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| number = ML17310A412
| number = ML17310A412
| issue date = 10/27/2017
| issue date = 10/27/2017
| title = Seabrook Station Revision 18 to Updated Final Safety Analysis Report, Chapter 6, Engineered Safety Features
| title = 8 to Updated Final Safety Analysis Report, Chapter 6, Engineered Safety Features
| author name =  
| author name =  
| author affiliation = NextEra Energy Seabrook, LLC
| author affiliation = NextEra Energy Seabrook, LLC
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{{#Wiki_filter:SEABROOK STATION UPDATED FINAL SAFETY ANALYSIS REPORT    CHAPTER 6 ENGINEERED SAFETY FEATURES  SECTIONS  6.1(B) Engineered Safety Feature Materials 6.1(N) Engineered Safety Feature Materials
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===6.2 Containment===
Systems
 
===6.3 Emergency===
Core Cooling System
 
===6.4 Habitability===
Systems
 
===6.5 Fission===
Product Removal and Control Systems 6.6 In-Service Inspection of Class 2 and 3 Components 6.7 Main Steam Isolation Valve Leakage Control System (BWR)
 
===6.8 Emergency===
Feedwater System APPENDICES  6 A Design Guidelines for In
-Service Inspection of Piping Systems
 
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES  Engineered Safety Feature Materials Revision 8 Section 6.1(B)
Page 1  6.1(B) ENGINEERED SAFETY FEATURE MATERIALS This section provides a discussion of the materi als used in Engineered Safety Feature (ESF) components and the material interactions that potentially could impair operation of ESF.
6.1(B).1 Metallic Materials 6.1(B).1.1 Materials Selection and Fabrication Typical material specifications applicable to component s in the ESF not covered by Subsection 6.1(N).1 are listed in Table 6.1(B)-1. In some cases this list of materials may not be totally inclusive; however, the listed specifications are representative of those materials used.
Materials utilized in ESF have been selected for their compatibility with core and containment spray solution, and conform with the requireme nts of the ASME Boiler and Pressure Vessel Code, Section III, Articles NC-2160 and NC-3120, plus all applicable addenda and code cases.
Typical specifications for principal pressure-retaining ferritic material s, austenitic stainless steels and nonferrous materials, including bolting and weld rod materials used in pressure-boundary welds in the Engineered Safety Features, are listed in Table 6.1(B)-1. These materials are qualified to the requirements of the ASME Code, Section III and Section IX, and are used in procedures which have been qualified to these same rules.
The following controls are placed on fabrication and assembly of austenitic stainless steel materials used in the ESF compone nts to insure a high degree of quality and reliability. These controls assure compliance with Regulatory Guides 1.31, 1.37, and 1.44: a. Significant sensitization of austenitic stainless steel is avoided by imposing the following controls: 1. Use of low-heat input welding procedures and processe s, as well as maximum interpass temperature control 2. Use of fully annealed material heat-treated in accordance with specific parameters (e.g., water quenching) 3. Prohibition of stress relief after welding 4. Engineering review of welding and heat-treatment procedures. b. Specific controls are impos ed during fabrication and installation to preclude contamination of stainless steel by chlorides and low melting point constituents, particularly during welding and heat treatment. These controls are monitored by checking chemical analysis certifications of materials that contact stainless steel (such as tapes, marking crayons and cleaning solutions), and engineering review of final cleaning procedures.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES  Engineered Safety Feature Materials Revision 8 Section 6.1(B)
Page 2  c. Cold working of stainless steel is prohibited after solution annealing except in mild environments where residual stresses from bending or forming are minimal. Piping for the containment spray system spray headers undergoes moderate bending during fabrication; however, this will have no dele terious affects on system performance since internal pressure during system operation is low and the chemical environment mild. In no case is cold-worked stainless steel with a yield strength of 90,000 psi or greater used in ESF constituents. d. Each heat or lot of filler material is required to be checked to assure the presence of 5 to 20 percent ferrite as calculated from the chemical composition and/or by a magnetic measuring check of a weld pad made with the subject filler material. Maximum interpass temperature control is also imposed during welding to minimize hot cracking. The thermal-insulation used on ESF piping and equipment inside containment is fiberglass blanket insulation of the type commercially known as Nukon, manufactured by Owens-Corning Fiberglass, with a stainless steel jacket over the outside surface of the insulation. Nukon is consistent with the recommendations of Regulatory Guide 1.36. Owens-Corning Fiberglass submitted Topical Reports OCF-1 on Nukon to the NRC for review in August 1977. The thermal insulation used on ESF piping and equipment outside containment shall be either fiberglass or calcium silicate molded sections with an aluminum jacket over the outside surface of the insulation.
6.1(B).1.2 Composition Compatibility and Stability of Containment Core Coolants The pH of the coolants for the ESF during a loss of coolant is dependent upon the boron concentration of the Reactor Coolant System, the accumulators and the refueling water storage tank, and the concentration of sodium hydroxide in the spray additive tank. The concentrations of boron in the accumulators, and refueling water storage tank, and the concentration of sodium hydroxide in the spray additive tank are fixed and are verified periodically by analyzing samples of the solution in each tank and adjusting the chemical compositions if necessary. A recirculation/sample system is installed for the Spray Additive Tank to verify its concentration.
The boron concentration in the Reactor Coolant System can vary from 0 to 4000 ppm depending upon the requirements for reactivity control. Depending on the various initial chemical compositions of the injected coolant and the Reactor Coolant System, the pH of the coolant can range from 8.8 to 9.5 at the beginning of the recirculation phase of ESF operation, and remains constant thereafter. The Ph of the spray solution during the injection phase will average between 9.1 and 10.3. The pH of the containment spray system sump water (and therefore the long-term ESF coolant) following a LOCA is monitored by withdrawing samples downstream of the RHR heat exchanger. Two sample points exist: the normal connection to the sample sink and a local sample point. Sodium hydroxide can be added, if necessary for pH adjustment, using the chemical and volume control system tanks and pumps. This assures the capability of maintaining a sump pH greater than 7.0 as recommended in Branch Technical Position MTEB
 
6-1. The solution would be prepared in the chemical mixing tanks and supplied to the suction of the charging pumps. The charging pump suction is fed from a cross-connect by the RHR system.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES  Engineered Safety Feature Materials Revision 8 Section 6.1(B)
Page 3  Information concerning hydrogen release by the corrosion of containment metals and the control of the hydrogen and combustible gas concentrations within the containment following a loss-of-coolant accident is contained in Subsection 6.2.5. Water for containment spray and emergency core cooling is stored in the refueling water storage tank and the spray additive tank.
Both tanks are constructed of stainless steel which has been demonstrated by test and experience to be compatible with solutions of borated water and sodium hydroxide.
6.1(B).2 Organic Materials Significant quantities of coated surfaces inside containment that would be exposed to the post-LOCA environment are listed in Table 6.1(B)-2. The coating systems for these surfaces, except PCCW piping, are epoxy-based Keeler & Long coating systems designed for a 40-year life and are in compliance with the applicable ANSI standards for coating systems inside containment (ANSI N45.2, ANSI N101.2, ANSI N101.4 and ANSI N512). Thus the coating systems meet Regulatory Guide 1.54.
Other significant quantities of organic materials inside containment are listed in Table 6.1(B)-4.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES  Engineered Safety Feature Materials Revision 8 Section 6.1(N)
Page 1  6.1(N) ENGINEERED SAFETY FEATURE MATERIALS 6.1(N).1 Metallic Materials 6.1(N).1.1 Materials Selection and Fabrication Typical materials specifications used for components in the Engineered Safety Features are listed in Table 6.1(N)-1. In some cases, this list of materials may not be totally inclusive. However, the listed specifications are representative of those materials used. Materials used conform with the requirements of the ASME Boiler and Pressure Vessel Code, Section III, plus applicable and appropriate addenda and code cases. The welding materials used for joining the ferritic base materials of the Engineered Safety Features conform to or are e quivalent to ASME Material Specifications SFA 5.1, 5.2, 5.5, 5.17, 5.18, and 5.20. The welding materials used for joining nickel-chromium-iron alloy in similar base material combination and in dissimilar ferritic or austenitic base material combination conform to ASME Material Specifications SFA 5.11 and 5.14. The welding materials used for joining the austenitic stainless steel base materials conform to ASME Material Specifications SFA 5.4 and 5.9. These materials are qualified to the requirements of the ASME Code, Section III and Section IX, and are used in procedures which have been qualified to these same rules. The methods utilized to control delta ferr ite content in austenitic stainless steel weldments are discussed in Subsection 5.2.3. All parts of components in contact with borated water are fabricated of or clad with austenitic stainless steel or equivalent corrosion-resistant material. The integrity of the safety-related components of the Engineered Safety Features is maintained during all stages of component manufacture. Austenitic stainless steel is utilized in the final heat-treated condition as required by the respective ASME Code, Section II, material sp ecification. Furthermore, it is required that austenitic stainless steel materials used in the engineered safety features components be handled, protected, stored, and cleaned according to recognized and accepted methods which are designed to minimize contamination that could lead to stress corrosion cracking. These controls are stipulated in Westinghouse specifications which are discussed in Subsection 5.2.3. Additional information concerning austenitic stainless steel, including the avoidance of sensitization and the prevention of intergranular attack, can be found in Subsection 5.2.3. No cold-worked austenitic
 
stainless steels having yield strengths greater than 90,000 psi are used for components of the Engineered Safety Features with in Westinghouse scope of supply.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES  Engineered Safety Feature Materials Revision 8 Section 6.1(N)
Page 2  Westinghouse-supplied engineered safety features components within the containment that would be exposed to core cooling water and containment sprays in the event of a loss-of-coolant accident used materials listed in Table 6.1(N)-1. These components are manufactured primarily of stainless steel or other corrosion-resistant material. The integrity of the materials of construction for engineered safety features equipment when exposed to post-design basis accident conditions has been evaluated. Post-design basis accident conditions were conservatively represented by test conditions. The test program (Reference 1) performed by Westinghouse considered spray and core-cooling solutions of the design chemical compositions, as well as the design chemical compositions contaminated with corrosion and deterioration products which may be transferred to the soluti on during recirculation. The effects of sodium (free caustic), chlorine (chloride), and fluorine (fluoride) on austenitic stainless steels were
 
considered. Based on the re sults of this investigation, as well as testing by ORNL and others, the behavior of austenitic stainless steels in the post-design basis accident environment will be acceptable. No cracking is anticipated on any equipment even in the presence of postulated levels of contaminants, provided the core cooling and spray solution pH is maintained at an adequate level. The inhibitive properties of alkalinity (hydroxyl ion) against chloride cracking and the inhibitive characteristic of boric acid on fluoride cracking have been demonstrated. Information concerning compliance with Regul atory Guides 1.31, 1.37, a nd 1.44 can be found in Section 1.8.
6.1(N).1.2 Composition, Compatibility, and Stability of Containment and Core Spray Coolants Westinghouse supplied the accumulator vessels used for storing ESF coolants. The accumulators are carbon steel clad with austenitic stainless steel. Because of the corrosion resistance of these materials, significant corrosive attack on the accumulator vessels is not expected. The accumulator vessels are filled with borated water and are pressurized with nitrogen gas. The boron concentration, as boric acid, is 2600-2900 parts per million (ppm). Samples of the solution in the accumulators are taken periodicall y for checks of boron concentration. Principal design parameters of the accumulators are listed in Table 6.3-1.
The method of establishing containment spray and recirculation sump pH following a loss-of-coolant accident is discussed in Subsection 6.2.2. Information concerning hydrogen release by the corrosion of containment metals and the control of the hydrogen and combustible gas concentrations within the containment followi ng a loss-of-coolant accident is discussed in Subsection 6.2.5.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES  Engineered Safety Feature Materials Revision 8 Section 6.1(N)
Page 3  6.1(N).2 Organic Materials Quantification of significant amounts of protective coa tings on Westinghouse-supplied components located inside the Containment Build ing is given in Table 6.1(N)-2; the painted surfaces of Westinghouse-supplied equipment comprise a small percentage of the total painted surfaces inside containment. For large equipment requiring pr otective coatings (specifically itemized in Table 6.1(N)-2, Westinghouse specifies or approves the type of coating systems utilized; requirements with which the coating system must comply are sti pulated in Westinghouse process specifications, which supplement the equipment specifications. For these components, the generic types of coatings used are zinc-rich silicate or epoxy-based primer with or without chemically cured epoxy or epoxy-modified phenolic top coat.
The remaining equipment requires protective coatings on much smaller surface areas and is procured from numerous vendors; for this equipment, Westinghous e specifications require that high quality coatings be applied using good commercial practices. Table 6.1(N)-2 includes identification of this equipment and total quantities of protective coatings on such equipment. Protective coatings for use in the reactor containment have been evaluated as to their suitability in post-design basis accident conditions. Tests have shown that certain epoxy and modified phenolic systems are satisfactory for in-containment use. This evaluation (Reference 2) considered resistance to high temperature and chemical conditions anticipated during a loss-of-coolant accident, as well as high radiation resistance. Information regarding assurance requirements for protective coatings is addressed in the discussion on conformance to Regulatory Guide 1.54 in Section 1.8. Further compliance information has been submitted to the NRC for review (via letter NS-CE-1352 dated February 1, 1977 to C. J. Heltemes, Jr., Quality Assurance Branch, NRC, from C. Eicheldinger, Westinghouse PWRSD, Nuclear Safety Dept.) and accepted (via letter dated April 27, 1977, to C. Eicheldinger from C. J. Heltemes, Jr.).
6.1(N).3 References
: 1. "Behavior of Austenitic Stainless Steel in Post Hypothetical Loss-of-Coolant Accident Environment," WCAP-7803, December 1971. 2. "Evaluation of Protective Coatings for Use in Reactor Containments," WCAP-7825, December 1971.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 1  6.2 CONTAINMENT SYSTEMS
 
====6.2.1 Containment====
Functional Design 6.2.1.1 Containment Structure
: a. Design Bases The containment design bases are established by the requirement that the system safely withstand the consequences of postulated accidents in conjunction with simultaneous occurrences of adverse environmental conditions. The containment structure and the containment enclosure, together with the exhaust system, are designed so that the offsite doses from radioactivity rele ased under accident conditions are less than the limits set forth in 10 CFR 100. The GOTHIC computer program (Reference
: 28) was used to develop a model of the Seabrook Station containment and associated safety systems. The model was employed to determine the containment response to various LOCAs and main steam line breaks at an analyzed core power level of 3659 MWt. The GOTHIC model is similar to the CONTRAST-S-MOD1 model that was originally used to determine the Seabrook Station containment response. The peak containment pressure and temperature predicted by GOTHIC for an analyzed core power level of 3659 MWt is bounded by the results of the original containment analysis described in the following subsections. Additionally, an evaluation of the short-term LOCA mass and energy releases presented in Table 6.2-30 determined that they are bounding for an analyzed core power level of 3659 MWt, without the need to be adjusted. Therefore, the containment design bases, evaluation, and results presented in the following subsections remain bounding and appl icable for an analyzed core power level of 3659 MWt and have not been revised. 1. Postulated Accident Conditions for Containment Design Accidents postulated to determine the containment internal design pressure and the containment design temperature include ruptures of the primary and secondary coolant system piping concurrent with a variety of single failures. The simultaneous loss of offsite power (LOOP) has also been assumed whenever it results in more restrictive design conditions. The detailed accident conditions for primary system pipe rupture are given in Subsection 6.2.1.3; those for secondary system pipe ruptures in Subsection 6.2.1.4.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 2  The single failures postulated for the primary system pipe ruptures include failure of a containment spray train and failure of a diesel generator. Those postulated for the main steam line breaks include failure of main feedwater pump to trip, a feedwater isolation or control valve, a main steam isolation valve, an emergency feedwater pump run out control, and a containment spray train. The calculated maximum internal containment pressure is 49.6 psig, resulting from a (full) double-ended guillotine rupture of the primary coolant system pipe at the pump suction, with one of the two containment spray pumps failed at time of containment spray actuation, maximum initial containment pressure of 1.5 psig and minimum flow rate of 2808 gpm for the Containment Building Spray System. The system design flow rate is 2930 gpm. This is the containment design basis (DB) accident. In accordance with General Design Criterion 50 of 10 CFR 50, this value was increased to 52.0 psig, thus providing a 4.8 percent margin between the design and maximum calculated values. Use of containment temperature responses following the main steam line breaks (MSLBs), which are more severe than those for
 
loss-of-coolant-accident (LOCAs), to obtain an envelope for equipment qualification, is disc ussed in Section 3.11. 2. Postulated Accident Conditions for Subcompartment Design Ruptures of appropriate high-energy lines at various locations within a subcompartment, concurrent with the SSE, have been postulated to determine the design requirements for the subcompartment structure. The maximum calculated pressure is not affected by LOOP or any postulated single failure because of the rapid occurrence of the peak pressure. The accidents postulated for each subcompartment are described in Subsection 6.2.1.2. This subsection also contains the maximum calculated pressures and the design pressures associated with each subcompartment. 3. Mass and Energy Releases Accidents involving ruptures in the primary or secondary coolant system pipes can result in the release of a significant amount of mass and energy into the containment atmosphere.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 3  (a) Loss-of-Coolant Accident The assumptions and details of calculational methods concerning the mass and energy releases into the containment following a rupture in the Primary Coolant System are treated in Subsection 6.2.1.3. The sources of water mass available for release include the initial reactor coolant, borated water from the accumulators and the
 
refueling water storage tank (RWS T), and the aqueous solution in the spray additive tank (SAT). The amount of water in each source is shown in Table 6.2-1.
During the blowdown phase, the reactor coolant energy is the principal source of energy released to the containment. The high-enthalpy, high-pressure water is rapidly discharged from the break at a critical-flow rate that depends upon the conditions at the break location. Some portion of it flashes into steam due to the comparatively lower pressure and lower temperature of the containment atmosphere. The discharge rate soon drops as the reactor coolant pressure is reli eved. The end of blowdown is defined as the time when the flow at the break reaches a minimum. Following the blowdown phase, additional heat is transferred via the coolant to the containment. This is comprised of decay heat, core internal energy, reactor vessel metal energy, steam generator energy, metal-water heat of reaction, and the energy of the coolant itself. The post-blowdown phase is characterized by a long, slow transient in which mass and en ergy discharge rates depend upon the flooding rate of the core. The flooding rate and the performance of the Emergency Core Cooling System (ECCS) are discussed in Section 6.3.  (b) Secondary System Pipe Rupture The details concerning assumptions and calculational methods dealing with mass and energy releases into the containment
 
following a rupture in the Secondary Coolant System are discussed in Subsection 6.2.1.4.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 4  For ruptures in the main steam line, the sources of mass available for release include the steam generator initial inventory, the main and emergency feedwater pumped into the steam generator before
 
isolation, the feedwater in the uni solated piping which eventually flashes into the steam generator, and the steam in the unisolated main steam piping. The first item varies from 124,000 lbm to 168,000 lm, depending on the initial operating power level. The rest depends on the time of is olation and the single failure postulated. The sources of energy for release into the containment include the energy of the initial steam generator inventory and the heat transferred from the primary system to the secondary system
 
during the transient period.
The high-enthalpy and high-pressure steam is discharged from the break at the critical-flow rate upon rupture of a main steam pipe at the exit of a steam generator flow restrictor. The discharge rate decreases as the affected steam generator is depressurized.
Depending on the break type, an isol ation signal is generated by either the Reactor Protection System or by the instrumentation system monitoring the containmen t pressure. The signal causes isolation of the main feedwater lines and the main steam. The blowdown drops considerably follow ing the isolation, but the flow from the affected steam generator continues until dryout time. 
 
Thereafter, the blowdown flow rate drops to a value equal to the emergency feedwater flow rate.
The effects of a postulated feedwater line rupture are not as severe as the main steam line break because the break effluent of a feedwater line rupture is at a lo wer specific enthalpy. Therefore, feedwater line break mass and energy releases to the containment are not addressed since they are bounded by steam line break
 
releases. 4. Effects of Engineered Safety Features on Energy Removal The energy released as a result of a LOCA or a secondary system pipe rupture is partially removed from the containment by the Containment Building Spray (CBS) System and the Residual Heat Removal (RHR)
System through the Station Service Water (SSW) System.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 5  Shortly following a coolant system pipe rupture, ECCS automatically starts. When the containment pressu re reaches a high-pr essure setpoint, borated water from the RWST mixed with spray additives is pumped through the spray nozzles to the containment atmosphere by the CBS system. When the RWST reaches a low-level setpoint, the source of water for the CBS system and the RHR portion of the ECCS is switched to the containment sump. In this recirculation mode the containment sump water is cooled by the CBS and RHR heat exchangers, and either pumped into the reactor vessel or sprayed into the containment atmosphere. The CBS and RHR heat exchangers transfer heat to the service water via the intermediate closed-loop prim ary component cooling water. All Engineered Safety Features which are available for containment heat removal are described in Subsection 6.2.2. Relevant system parameters are summarized in Table 6.2-2. 5. Capability of ESF for Post-Acci dent Pressure Reduction and Energy Removal All Engineered Safety Features are separated into two independent subsystems of equal capability to meet the single failure criteria. One
 
hundred percent redundancy is also pr ovided in the associated electrical actuation systems. Emergency powe r is supplied from redundant onsite power sources. The containment and its heat removal systems are designed so that operation of only one of the two CBS tr ains, in conjunction with the ECCS at any point in its range of capability, is sufficient to reduce the pressure of the containment atmosphere to within half of its calculat ed peak value in less than 24 hours following containment design basis LOCA.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 6  6. Containment Leakage Rate Bases The containment is isolated from the outside environment following major accidents by the Containment Isolation System. The presence of the containment enclosure and the use of exhaust fans to produce a slightly sub-atmospheric pressure in the space between the containment enclosure and the containment structure reduce the direct leakage from the structure to the environment to zero. The containment design is such that the maximum rate of leakage from the containment structure to the containment enclosure following a coolant pipe rupture is 0.15 percent of the containment air mass per day. The containment heat removal systems are capable of reducing the containment pressure, within 24 hours following the accident, to such a value that the volumetric leakage rate is less than one-half of the maximum valu
: e. The use of HEPA and charcoal filters in the exhaust line from the containment enclosure reduces the discharge of radioactive iodine into the environment to the extent that offsite doses following an accident are within the guidelines of 10 CFR 100. The exhaust system is discussed in Subsection 6.2.3.
The periodic testing and surveillan ce program to assure the above containment leakage rate is discussed in Subsection 6.2.1.6 and in the Technical Specifications. 7. Bases for Minimum Containment Pressure Analysis Assumptions in the minimum pressure analysis for ECCS confirmatory studies are based upon maximizing the ESF heat removal capability and other heat removal mechanisms. Th ey are discussed in Subsection 6.2.1.5. b. Design Features
: 1. Containment Structure The containment structure is a rein forced concrete cylinder with a hemispherical dome and a reinforced concrete foundation, keyed into the rock by the depression for the reacto r cavity pit and by continuous bearing around the periphery of the foundation mat. A welded steel liner plate is anchored to the inside face of the concrete as a leak-tight membrane. The liner plate on top of the foundation sl ab is protected by a 4-foot thick concrete slab which serves to carry internal equipment loads and forms the floor of the containment. A detailed description of the containment structure is given in Subsection 3.8.1.
Figures showing typical sections through the containment can be found in Section 1.2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 7  2. Containment Enclosure The containment enclosure is a reinforced concrete cylindrical structure with a hemispherical dome. Detailed descriptions of the structure are presented in Subsection 3.8.4. Figures s howing sections and elevations of the containment enclosure can be found in Section 1.2. Fans maintain the pressure in the space between the containment structure and the containment enclosure at a value slightly below the atmospheric pressure following a LOCA. All joints and penetrations are welded or sealed to ensure air tightness. 3. Protection Against Dynamic Effects Provision is made for protecting the containment structure, internal compartment and ESF systems against loss of function from effects that
 
could occur following postulated accidents. These provisions include physical barriers designed to minimize the dynamic effects of missiles and pipe whip, pipe whip restraints to limit damage from ruptured lines, physical separation by distance a nd redundancy of components and/or safety trains, as appropriate. A detailed discussion of provisions for protection against dynamic effects inside the containment is presented in Section 3.6 and Subsection 3.5.1.2. 4. Codes, Standards and Guides Codes, standards and guides applied to the design of the containment structure and internal structures are identified in Subsections 3.8.1 and
 
3.8.3. 5. Protection Against External Pressure The containment structure, including its steel components, is designed to withstand a maximum external pressure of 3.5 psi (differential pressure). The most limiting event for establishi ng the required external pressure is the inadvertent actuation of the Containment Spray System, which results in a negative pressure differential of 2.6 psi. The analysis follows: The containment is normally maintained at a slight positive pressure by the containment online purge subsystem (see Updated FSAR Subsection 9.4.5.2c.2). With this system in operation, the containment is maintained at a nominal positive pressure of 0.5 psig, with high and low pressure alarms at 0.65 psig and 0.35 psig, respectively.
Accordingly, if this system is in operation, the initial pressure inside containment is
 
always positive as an initial starting condition.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 8  However, because the above system is not redundant, the analysis for this event is based upon an initial containment ambient pressure of 14.6 psia. Initial temperatures and relative humidity are 120F and 90 percent. The analysis utilized the containment free volume, geomet rical data and passive heat sink data as provided in Updated FSAR Table 6.2-1, Table 6.2-3 and Table 6.2-4. Both spray trains are assumed to operate with a maximum flow rate of 3500 gpm per train. The analysis assumes the total capacity of the refueling water storage tank of 475,000 ga llons is available for spray at a minimum temperature of 50F (see Updated FSAR Subsection 6.2.2.3). Using the computer code CONTRAST-S-MOD-1, the minimum resultant containment pressure of 12.0 psia results from the inadvertent actuation of sprays. The above event produces a negative pressure differential of 2.7 psig. This is to be compared with the containment structure design differential pressure of negative 3.5 psig (see Updated FSAR Subsection 6.2.1.1b.5). This provides a nominal negative pressure margin of 0.8 psig. 6. Potential Water Traps Principal areas where water could be trapped and prevented from being circulated by the ECCS and CBS system during a major accident are the reactor cavity and refueling canal.
Figures showing plan and sectional views of these areas are given in Section 1.2. Other minor volumes not addressed here have been taken into account in calculating the NPSH available to the ECCS and CBS system pumps, as discussed in
 
Subsection 6.2.2. In the event of the LOCA, the reactor cavity, incore instrument sump, ECCS sump, and other minor volumes will be filled with water up to
 
elevation (-)26 feet during the operation of the ECCS and CBS systems.
A total of 17,070 cubic feet of water is required to fill these two volumes up to the (-)26 feet elevation, resulting in a reduction of water height of approximately 17 inches which would otherwise contribute to NPSH available to the ESF pumps. This has been factored into the NPSH
 
calculations for the pumps.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 9  Three four-inch diameter drain lin es connected to a common header permit a flow path between the reactor cavity, refueling canal, and the rest of the containment, thereby preventing water from being trapped in the reactor cavity and refueling canal to elevations above the water level in the rest of the containment. In the event that all the drain lines were blocked or all the normally open drain valves were left closed (see Subsection 6.2.2.3a.3 for further details) the loss of 5760 cubic feet of water above the (-)26 feet elevation woul d occur. This would result in a
 
further reduction of water height of approximately 6 inches. This has been factored into the NPSH calculation for pumps also.
: 7. Containment Cooling and Ventilation System During normal operation, fan coolers maintain the containment atmosphere below 120F. It uses the cooling water from the Primary Component Cooling Water System, while its humidity is permitted to vary. Five of six fan coolers opera te continuously, with the sixth fan cooler serving as an installed spare. Cooled air from the containment fan coolers is also directed into the reactor cavity at various locations to maintain equipment and concrete at or below design temperatures.
 
Three induced draft fans, two of which are normally operating, draw air past the control rod drive mechanism through a cooling shroud to maintain the equipment at or below its design operating temperature. The third fan serves as an installed spare.
An Online Purge System is provided to periodically purge the containment air to control airborne radioactivity. A separate system provides pre-entry purging of the containment and purging during refueling operations. The Containment Structure Heating and Cooling System and the Containment Online and Pre-entry Purge Systems are discussed in detail in Subsection 9.4.5. c. Design Evaluations
: 1. Containment Pressure-Temperature Response Following A LOCA or a Secondary Coolant System Pipe Rupture Containment pressure and temperatur e responses following a variety of
 
postulated ruptures in the primary and secondary coolant system pipes have been calculated by means of the computer program CONTRAST-S-MODl which is descri bed in detail in Reference 1.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 10  (a) System Parameters and Initial Conditions The system parameters and the initial conditions used in the pressure-temperature response analysis are presented in Table 6.2-1 and Table 6.2-2. The parameters and the initial conditions are chosen to maximize the containment pressure and temperature responses unless th eir effects are insignificant. (b) Actuation of Containment Sprays The Containment Spray System is initiated by a containment spray actuation signal which is generated by the containment Hi-3 ("P" signal). The analysis limit for this setpoint is 19.8 psig. The maximum delay in signal processing and the response time of the protection system instrumenta tion is one second. The stroking time of the spray system valves is 20 seconds to become fully open with the exception of CBS-V38 and CBS-V43 which have a maximum stroke time of 25 seconds. The maximum delay time to bring the pumps to full speed and to fill the feed lines and headers is 38 seconds following the receipt of this signal. The valve opening and fill-up of the line takes place concurrently. Therefore, the maximum delay of 39 seconds after generation of the "P" signal consists of 1 second plus 38 seconds for fill-up.
In the case of loss of offsite power concurrent with a coolant system pipe rupture, the emergenc y electric power from the onsite diesel generators will be available in 12 seconds. Receipt of the "P" signal by the actuation sequencer, which is discussed in detail in Subsection 8.3.1, will cause the spray valves to start opening immediately, or as soon as the emergency power is available. If the "P" signal is rece ived within 27 seconds or between 27 and 52 seconds following an accident, the spray pumps will be started at 27 seconds or 52 seconds respectivel
: y. If the signal is received after 52 seconds, the pumps will be started immediately. Thus, for any "P" signal received before 27 seconds following an accident, which is true for all LOCA cases analyzed, the spray time is always at 65 seconds after the accident. For MSLB cases analyzed, the time to generate the "P" signal varies over a wide
 
range. However, for all MSLB ca ses, a constant conservative spray delay time of 65 seconds after receipt of the signal has been assumed.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 11  (c) Containment Passive Heat Sinks The structure and equipment within the containment which have been modeled as passive heat sinks are listed in Table 6.2-3.
Table 6.2-4 gives the thermophysical properties of the materials for the heat sinks. In Table 6.2-3, some of the heat sinks of similar
 
structure have been lumped into various thickness groups. For internal heat sinks exposed to the containment on one side only, it is conservatively assumed that th e other side is insulated. For those exposed on both sides and having a plane of symmetry, only half of the heat sink is modeled but the effective surface area is doubled. For the steel-lined structur es, the steel-concrete interface resistance has been modeled as an equivalent air gap of inch. This is conservative when compared with experimental data in References 2 and 3. The interface conductance therein varies from 100,000 Btu/hr-ft 2- F for a very good contact to 10 Btu/hr-ft 2- F for a very poor contact. The lower limit corresponds to the equivalent conductance of an air gap approximately 20 mils thick. The heat transfer between the containment and the passive heat sinks is calculated in the CONTRAST-S code by combining the contribution from the condensation and convection. An effective heat transfer coefficient, h eff, based on the temperature difference between the containment atmosphere and the heat sink surface can
 
be defined as follows:
heff = f(h cond-h conv)(T sat-T wall)/(T con-Twall) + h conv where, h cond = condensing heat transfer coefficient h conv = convective heat transfer coefficient T sat  = containment atmosphere dew point T wall = surface temperature of heat sink T con  = containment atmosphere temperature f = 1, if T sat > T wall , 0 otherwise.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 12  In the estimation of the condensing heat transfer coefficient, the modified Tagami correlation (Reference 4) is used for the LOCA analysis while the Uchida correlation (Reference 5) is used for the MSLB analysis. The convective heat transfer coefficient has been assumed to have a constant value of 2 Btu/hr-ft 2-F in all accident cases analyzed. This corresponds to the minimum value given in the Tagami correlation. The heat transfer through a passive heat sink is computed in the CONTRAST-S code by solving the partial differential equation for unsteady, one-dimensional heat conduction using a fully implicit finite-difference scheme which is unconditionally stable. The accu racy of the solution can be improved by decreasing the grid spacing as well as the time increment. A sensitivity study (Reference 6) has been performed to establish the upper limit of the grid spacing in concrete for a reasonable degree of accuracy. It was found that grid spacings of 0.05 inch for concrete a few inches thick is adequate.
Considerably larger grid spacings have been found adequate for steel, due to the relatively smaller temperature gradients.
The selection of the grid spacings has also been guided by the following criteria, suggested in Reference 7, to avoid large, meaningless fluctuations in the solution: kxh 1tx where, x = grid spacing t = time increment  = thermal diffusivity of the material h  = heat transfer coefficient at the heat sink surface k  = thermal conductivity of the material For a typical heat transfer coefficient of 80 Btu/hr-ft 2- F during the containment pressure-temperature transient, this criterion suggests a grid spacing greater than 0.038 inch for concrete (k = 0.083 Btu/hr-ft- F and  = 0.028 ft 2/hr) and greater than 0.138 inch for carbon steel (k = 27.0 Btu/hr-ft- F and  = 0.459 ft 2/hr), when a time increment of 1 second is used.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 13  (d) Containment Pressure and Temperature Responses (1) Loss-of-Coolant Accident A spectrum of postulated reactor coolant system pipe ruptures has been considered. This includes three break locations listed in Table 6.2-5 and three break types and
 
sizes given in Table 6.2-6. The ESF systems relied upon to mitigate the consequences of a LOCA are the ECCS and CBS systems operating in conjunction with the Primary Component Cooling Water System and the SSW system.
Failure of one of the two CBS trains obviously would result in more severe containment conditions. The failure may be caused by failure of a pump, failure of a valve, or, assuming loss of
 
offsite power concurrent with the LOCA, failure of a diesel generator. As to the ECCS, any of its various components may fail, leading to partial loss of its cooling effect. In the extreme case, one of the two trains may fail entirely, due to failure of a diesel generator to start, assuming a concurrent loss of offsite power. The above two limiting single active failures (SAFs) are delineated in Table 6.2-7. The cooling water for ultimate heat disposal is available from two sources, namely, the circulating water tunnels and the cooling tower basin. Section 9.2 presents a detailed
 
description of these sources. The tunnel water temperature is not expected to exceed 65F. During the summer months, extended hot weather combined with ocean current changes can result in minor ocean temperature excursions above the 65°F design temperature threshold. System analysis has been performed to permit continued plant operation up to a maximum ocean temperature of 68.5°F. Concerning containment heat removal, the use of tunnel water following a LOCA is more restrictive than that of the cooling tower basin water. All LOCA analyses presented have been performed using the ocean as the ultimate
 
heat sink.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 14  The effects of the brea k type/size and the SAF have been fully investigated in the case of a rupture at the pump suction (Location No. 1). In the case of Locations No. 2 (cold leg) and No. 3 (hot leg), the results are expected to be less severe than those for Location No. 1 for the most limiting break type/size and SAF combination, namely Break No. 1 and SAF No. 2. Overall, a total of six cases have been analyzed. The details of the calculation of the mass and energy releases for the six cases analyzed are given in Subsection 6.2.1.3. Those results have been based on a temperature of 120F for the safety injection water. For Seabrook Station Unit 1, the maximum temperature of the injection water is 100 F. The time for switchover from the injection mode to the recirculation mode depends upon the injection flow (charging pumps, high pressure safety injection pumps, and low pressure safety injection pumps), spray flow and the
 
quantity of water available in the RWST. Since one of the two spray trains has been assumed to fail in all cases analyzed, the recirculation times are calculated for the maximum safety injection (two injection trains) and the minimum safety injection (one injection train) cases. With the injection and spray flow rates given in Table 6.2-2 and
 
the available quantity of water in the RWST provided in Table 6.2-1, the recirculation times for the maximum safety injection and the minimum safety injection cases are calculated to be 1688 seconds and 2755 seconds, respectively.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 15  The transient responses of the containment pressure, temperature and sump water temperature for the six cases analyzed are shown in Figure 6.2-1, Figure 6.2-2, Figure 6.2-3, Figure 6.2-4, Figure 6.2-5, Figure 6.2-6, Figure 6.2-7, Figure 6.2-8, Figure 6.2-9, Figure 6.2-10, Figure 6.2-11, Figure 6.2-12, Figure 6.2-13, Figure 6.2-14, Figure 6.2-15, Figure 6.2-16, Figure 6.2-17 and Figure 6.2-18. The transients show that, following blowdown, and prior to refill, the containment pressure and containment temperature drop because the mass and energy released through the breaks ceases completely at the end of
 
the blowdown period. Ho wever, the reflood and post-reflood mass and energy released from the break increase the containment pressure and temperature again.
The containment pressure and temperature eventually drop due to decreases in the mass and energy release rates, and due to energy removal by containment spray and passive heat sinks. After the switchover from the injection mode to the recirculation mode, the containment spray water is taken from the containment sump through the containment spray heat exchanger, and is at a higher temperature than that of the RWST. The spray heat removal rate thus drops. The containment pressure and temperature, therefore, increase after the start of recirculation. The mass and energy release data presented in Subsection 6.2.1.3 is based on the conservative assumption that the remainder of the energy (Reactor Coolant System, core-stored, primary and secondary metal, etc.) exits through the break within 3600 seconds. The reduced h eat removal capacity by the spray due to increased temperature, together with conservative energy release to the containment atmosphere, results in a recirculation pressu re peak which is observed to be the maximum peak for all pump suction breaks. The
 
energy release through the br eak drops at 3600 seconds, and only the decay heat is released thereafter. The containment pressure and temperature continue to drop monotonically from then on.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 16  The calculated peak containment pressure and peak containment temperature, along with the energy released to the containment up to the end of blowdown, are summarized in Table 6.2-8 for each of the six cases analyzed. The maximum peak containment pressure is seen to be 49.6 psig (see al so discussion presented in Subsection 6.2.1.1a.1). It occurs in Case 1.1.2, namely, a
 
full double-ended guillotine rupture at the pump suction
 
with the single active failure of one spray train, which corresponds to maximum safety injection and minimum
 
spray cooling. Case 1.1.2 is therefore taken as the containment DB LOCA. The transient responses have been calculated for up to 10 5 seconds after th e accidents, except for Case 3.1.1, (Hot-Leg Break). As can be seen from the
 
plots presented, the periods covered include the most important aspects of the tran sient and show the general trend of responses. The accident chronology is given in Table 6.2-9, one for each break location. It includes the time when the ESF system begins operation and time of occurrence of other important events. The distri bution of energy inventories prior to the accident, at the end of the blowdown phase and at the end of the core reflood phase, and also the steam generator energy releases during post-reflood phase, are provided in Subsection 6.2.1.3. The long-term recirculation operation causes reduction of the containment pressure to within a few psi above the atmospheric pressure in one day, which is well below one-half of the calculated peak pressure in all cases
 
analyzed. For the DB LOCA, the effective heat transfer coefficient based on the temperature difference between the containment atmosphere and the heat sink surface is plotted as a function of time in Figure 6.2-19.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 17  (2) Secondary System Pipe Ruptures Containment temperature and pr essure responses have been evaluated following a spectrum of breaks in the main steam line occurring at various plant operating power levels, each with a single failure in the safety systems postulated to concur with the accident. The details of calculating the mass and energy releases into the containment after a main steam line break (MSLB) are given in Subsection 6.2.1.4. The initial plant operating power levels and the spectrum of break types and sizes analyzed are summarized in
 
Table 6.2-10 and Table 6.2-11, respectively. The blowdown data for doubl e-ended breaks has been developed for breaks located downstream of the steam
 
generator flow restrictor and upstream of the main steam isolation valve (MSIV). The postulation of a break
 
downstream of the flow restrict or is conservative for large double-ended breaks since a break upstream of the flow restrictor would result in a smaller energy release because of severe water entrainment in the forward flow and the flow-limiting effect of the restrictor on the reverse flow.
For small double-ended breaks, there is no difference between the two locations. Since a break immediately downstream of the flow restrictor allows the steam in this section to blow down completely shortly after isolation, blowdown based on this brea k location is considered conservative.
For the split rupture, the blowdown data developed are valid for any break location in the steam piping and the
 
header. As presented in Table 6.2-10 a nd Table 6.2-11, a total of 17 operating power break type/size combinations for a MSLB have been considered. The c oncurrence of a single failure in the safety systems can result in more severe conditions in the containment. In this analysis, a loss of offsite power has been assumed to concur with the MSLB whenever it results in more severe containment conditions, in addition to a single failure in the safety systems.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 18  However, for the conservative estimation of mass and energy releases, offsite power is assumed to be available. The safety-related equipment relied upon to mitigate the consequences of an MSLB include those required for isolation of the main feedwater lines and the main steam lines, delivery of emergency feedwater into the steam
 
generators, and delivery of spray water into the containment. The Containment Spray System is the only active heat removal system for which credit has been taken in mitigating the consequences of a MSLB. The signal to isolate the main feedwater lines and the main steam lines and to start the Emergency Feedwater System is generated by the Reactor Protection System for all double-ended ruptures listed in Table 6.2-11. For the split ruptures, this isolation signal is conservatively assumed to be generated when the containment pressure reaches the Hi-1 setpoint of
 
6.8 psig for isolation of feedwa ter lines and Hi-2 setpoint of 7.4 psig for isolation of main steam lines, although the nominal values for the Hi-1 and Hi-2 setpoints are 4.3 psig. After the isolation setpoint is reached, a delay of 1.0 second for instrument response and signal processing has been allowed. The containment spray actuation signal ("P" signal) is generated when the containment pressure reaches
 
the Hi-3 setpoint, as discusse d in Subsection 6.2.1.1c.1(b).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 19  The isolation of the main feedwater lines is achieved by tripping the main feedwater pumps and closing the feedwater control valves (FCVs) and the feedwater isolation valves (FWIVs). The pump trip is immediate.
The FCVs are capable of closing completely within 10 seconds (5-second delay and 5-second stroke) after receipt of the isolation signal, and the FWIVs within 10 seconds after receipt of the isolation signal. Failure to trip a main feedwater pump would result in more main feedwater being pumped into the steam generators and an increase in the blowdown.
Failure of the broken-loop FWIV would allow additional feedwater, namely that in the piping between the FWIV and FCV to flash into the steam generator. Loss of offsite power, however, would result in less feedwater being fed to the steam generators because of the coastdown of the various pumps in the Condensate and Feedwater System. The isolation of the main steam lines is achieved by closing the MSIVs, the turbine control valves (TCVs) and the turbine stop valves (TSVs). The MSIVs are capable of closing completely within 5 seconds after receipt of the
 
isolation signal and the TCVs and the TSVs within 0.2
 
second. Failure of the broke n-loop MSIV to close would allow additional steam, namely that remaining in the steam
 
piping bounded by the MSIVs, the TSVs, the condenser, and the moisture separators/r eheaters, to blow down after isolation. Failure of a TCV or a TSV, however, would not cause more blowdown since cl osure of either one would effectively isolate that line.
Loss of offsite power has no effect on the isolation of the main steam lines.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 20  The emergency feedwater that enters the affected steam generator would eventually blow down. For conservatism, it is assumed that the emergency feedwater is pumped into the affected steam generator immediately after the MSLB. The flow isolation valves in each line are pre-set in the open position to provide a flow of at least 235 gpm to the intact steam generator, and ar e closed by a high flow signal in the event of a broken loop. In case of failure of a flow isolation valve in the open position, the flow is limited by
 
the flow restricting venturi to at most 750 gpm. In either case, the flow of emergency feedwater to the affected steam generator is assumed to be terminated manually 30 minutes after the MSLB. For the Containment Spray System, failure of one of the two trains would reduce its he at removal capacity by half and result in more severe containment conditions. The effect of loss of the offsite power on the actuation time of the spray system has been take n into account, as previously discussed by assuming a later spray time. The single failures which have been considered, as discussed above, are listed in Table 6.2-12. The first 5 failures would increase mass-energy release to the containment, while the last failure would result in a
 
reduction in the heat-removing capacity of the Containment
 
Spray System.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 21  Table 6.2-10, Table 6.2-11 and Table 6.2-12 show that there is a total of 102 combinations of plant operating powers, break type/sizes, and single failures. Mass and energy releases to the containment for these 102 cases have
 
been calculated. The effect of single failures No. 3 and No. 4 of Table 6.2-12 on the blowdown for any type of
 
ruptures, and single failure No. 2 for split ruptures, is to increase the steam generator dryout time only. This is essentially true for single failure No. 5 for all ruptures since a somewhat larger blowdown rate after the dryout time has been observed not to be sufficient to cause a second peak in the containment temperature or the containment pressure. 
 
Thus, the above single failure combinations can be eliminated. However, to clearly illustrate the effect of the single failures on the containment pressure-temperature transient, all 6 single failures have been analyzed for the 102 percent power level and for each of the three break types, No. 1, No. 2 and No. 3. This amounts to a total of 56 cases. After carefully examining the containment
 
pressure-temperature responses for these cases, the following general inferences are drawn: The containment pressure and temperature transients show that the peak containment pressure occurs either at the time the spray water enters the containment or at the steam generator dryout time. The latter occurs when the blowdown is severe enough to cause increases in the containment pressure even with the containment spray system operating, which is true for all double-ended ruptures. The peak containment temperature always occurs at the spray time.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 22  When considering accidents at the same power level and with the same single failure, the peak containment pressure increases with the size of the break in the case of
 
double-ended ruptures. The peak s for the split ruptures are always higher than those for the small double-ended
 
ruptures, but lower than those for the full double-ended
 
ruptures. The effect of power level, with the inherent break size change in the case of sm all double-ended breaks, is not always monotonic. The peak containment pressure varies
 
monotonically with power level only in small double-ended
 
breaks with a concurrent single failure of a MSIV or a containment spray train, a nd in full double-ended breaks with a concurrent single failure of a containment spray train. in the former, the peak pressure increases
 
monotonically as the power level increases, while in the latter, it decreases monotonically. The computed peak containment pressure and peak containment temperature, and the times of their occurrence are summarized in Table 6.2-13, Table 6.2-14 and Table 6.2-15 for full double-ended ruptures, small
 
double-ended ruptures, and split ruptures, respectively. Only those cases which effectively envelope containment pressure and temperature responses following MSLBs are shown. Also included in the tables is the total energy
 
released to the containment in each case. The transients for Cases 5.1.6 and 1.1.3 are shown in Figure 6.2-20, Figure 6.2-21, Figure 6.2-22 and Figure 6.2-23.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 23  The maximum peak containment pressure, as can be seen in these tables, is 36.1 psig and occurs in Case 5.1.6, namely a full double-ended guillotine rupture at hot shutdown with a concurrent failure of one containment spray train. This pressure is significantly lower than that for the DB LOCA. For this maximum peak containment pressure case, the mass-energy release is given in Subsection 6.2.1.4. The effective heat transfer coefficient, based on the temperature difference between the containment atmosphere and the heat sink surface, is plotted in Figure 6.2-24 as a function of time for four representati ve heat sinks, along with the Uchida correlation (Reference 5) which was used to calculate the condensing heat tran sfer coefficient. The heat sinks are the containment liner and typical thin-steel, thick-steel and concrete structures. All four effective heat
 
transfer coefficients are seen to be much lower than those predicted by the Uchida correla tion (taking into account the
 
===0.4 factor===
for concrete surfaces) when the containment atmosphere is superheated. When the containment atmosphere becomes saturated, the effective heat transfer coefficients become identical to that of the Uchida
 
correlation, as long as the heat sink surface temperature is lower than the containment atmosphere (which is approximately equal to the saturation temperature
 
corresponding to the partial pressure of steam in the containment). If the heat sink surface temperature exceeds the containment atmosphere temperature, condensation
 
ceases and convective heat transfer from the heat sink to the containment atmosphere takes place. The maximum peak containment temperature is 364F and occurs in Case 1.1.2, namely a full double-ended guillotine rupture at 102 percent power with a concurrent failure in the broken-loop MSIV. The ma ss-energy release for this case is presented in Subsection 6.2.1.4.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 24  A plot of the heat transfer coefficients similar to that presented for the maximum peak containment pressure is shown in Figure 6.2-25 for Case 1.1.2. Again, a similar behavior is observed. Th e effective heat transfer coefficients for the containment liner, the thin steel structure, the thick steel structure, and the concrete structure are considerably lower than those predicted by the
 
Uchida correlation when the containment atmosphere is superheated. They become identical only when the containment atmosphere becomes saturated and the heat sink surface temperature is lower than the containment atmosphere saturation temperature. (e) Post-Accident Containment Temperature/Pressure Monitoring Instrumentation provided in the containment can be used to monitor and record containment pressure in the event of an accident. This system is discussed in detail in Section 7.5.
Containment sump temperature is not monitored since it is not consequential to verification of proper operation of the Post-Accident Heat Removal Systems. 6.2.1.2 Containment Subcompartments
: a. Design Bases The major subcompartments within the containment are the reactor cavity, the steam generator compartments, the pressurizer compartment and the pressurizer skirt cavity. These subcompartments are designed to withstand the differential pressures and jet impingement forces resulting from a postulated pipe break. Reactor cavity and steam generator compartment overpressurization has been deleted from the design basis in accordance with the final NRC Rulemaking with respect to GDC-4, dated October 27, 1987. Sufficient openings for venting these subcompartments are provided to k eep differential pressures on the subcompartment walls and forces imposed on equipment supports within their structural limits. In addition, restrain ts and supports on the various equipment contained within these subcompartments are designed so that pipe whip and forces transmitted through component supports do not threaten the structural integrity of these subcompartments or the containment structure. The pipe breaks considered in all subcompartments are full double-ended ruptures.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 25  Of all the postulated break locations, the ones considered severe for the subcompartment pressurization analyses are listed in Table 6.2-16. The loads on the subcompartment walls and on the equipment supports for a given subcompartment are determined for various break locations. The subcompartment walls and the equipm ent supports are designed so that the maximum calculated load does not exceed the design load. b. Design Features
: 1. Reactor Cavity The reactor cavity is a cylindrical annulus around the reactor vessel. The surrounding structure, termed the primary shield wall, is a heavily reinforced concrete stru cture which provides support for the reactor vessel and its associated coolant system pi ping. The lower portion of the cavity, where the core lies, has an outside diameter of 17.08 ft. The inner diameter of the annulus, formed by the outer diameter of the reactor
 
vessel, is 16.76 ft. The upper region of the cavity where the hot and cold leg nozzles emanate from the vessel has a diameter of 25.5 ft. Contained within this upper region is the ring gird er which gives added rigidity to the vessel support. The reactor coolant loop pipe whip restraints were eliminated from the design bases by ECA 25/113665, Rev. A, in accordance with the final NRC Rulemaking with respect to GDC-4, dated October 27, 1987.
A neutron shield consisting of borated concrete, and which is integral to the permanent reactor cavity seal ring, is installed around the reactor vessel refueling flange to reduce neutron streaming and dose rates on the containment operating floor during power operation. The permanent seal ring is equipped with removable hatch covers and neutron shield plugs to allow for the required ventilation air flow rate and access to the reactor cavity annular space, respectively. Reactor cavity overpressurization has been deleted from the design basis in accordance with the final NRC Rulemaking with respect to GDC-4, dated October 27, 1987. A vertical section through the reactor cavity is shown in Figure 6.2-26. Figure 6.2-27 shows the plan view at the elevation of the nozzles. The total free volume of the reactor cavity plus the reactor pool region above is 48,978 cu. ft. The total vent area from the reactor cavity to the containment is 1,757 sq. ft, primarily through the reactor pool region. The free volume and the vent area are conservatively calculated with the insulation in place.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 26  2. Steam Generator Compartment A steam generator compartment is a reinforced concrete structure which encloses the steam generator, a reactor coolant pump and its associated primary and secondary coolant system piping. Figure 6.2-29 and Figure 6.2-30 show sections of the steam generator compartment. 
 
Horizontal sections at various elevations are provided in Figure 6.2-31, Figure 6.2-32, Figure 6.2-33 and Figure 6.2-34. Large vent paths from the steam generator compartment to the containment are available via the reactor coolant pump area, and the adjacent steam generator compartment. The top of the steam generator compartment is also open to the containment. The steam generator sta nds on four supports anchored to the floor at El. (-)26'-0."  The total free volume of a steam generator compartment is 23,040 cu. ft. Ther e are no blowout panels or other pressure-dependent areas considered in the analysis. Free volumes and vent areas have been calculated with the insulation in place. Steam generator compartment overpressurization has been deleted from the design basis in accordance with the final NRC Rulemaking with respect to GDC-4, dated October 27, 1987. 3. Pressurizer Compartment The pressurizer compartment is a reinforced concrete structure extending from El. 0'-0" to El. 63'-0" which encloses the pressurizer and its
 
associated piping. The pressurizer skirt, which is a cylindrical support extending from the bottom of the pressu rizer, anchors the pressurizer to the compartment floor. A ring support at El. 23'-63/4" provides lateral
 
support for the pressurizer. Section and plan drawings of the pressurizer compartment are shown in Figure 6.2-35, Figure 6.2-36, Figure 6.2-37, Figure 6.2-38, Figure 6.2-39 and Figure 6.2-40. Free volumes and vent areas have been calculated assuming the insulation remains intact during the transient. The HVAC ducting and sheet metal panels at elevation 16'-6" are designed to bl ow out in the event of a pressure buildup of 0.25 psig in the compartment to provide additional vent area. The total free volume of the compartment used in the analysis is 6638 cu. ft and the total vent area to the containment is 400 sq. ft.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 27  4. Pressurizer Skirt Cavity The pressurizer skirt cavity is formed by the bottom of the pressurizer and its supporting skirt. A 14" surge line which conn ects the reactor coolant system with the pressurizer passes through a 51/2 ft diameter opening in the pressurizer compartment floor. Figure 6.2-41 and Figure 6.2-42 show the plan and elevation drawings of the pressurizer skirt cavity. The volume below the pressurizer skirt has a large vent opening to the containment.
The total free volume of the skirt cavity is 1860 cu. ft and the total vent area to the containment is 238 sq. ft.
The insulation on the pressurizer and on the surge line is assumed intact in calculating the free volume and vent
 
openings. c. Design Evaluation
: 1. Mass and Energy Release Data The mass and energy release data for all the breaks considered for the subcompartment analyses has been generated by Westinghouse.
Discussions of the blowdown model are provided in Reference 8. 2. Computer Code for Subcompartment Pressurization Analysis The subcompartment pressure transients were calculated using COMPRESS - a digital computer program. A detailed description of the analytical method can be found in Appendix 15C and Reference 9. Some important aspects of the method are outlined below: (a) Mathematical Model The COMPRESS computer program calculates the pressure and temperature transient responses in a set of inter-connected volumes following a high energy line rupture. The subcompartment initial conditions, free volumes, vent path areas, inertias and loss coefficients, as well as the mass and energy release from the break, are input to the code. For vent pa ths covered by blowout panels or hinged doors, the flow area, inertia and loss coefficient are to be supplied as functions of the position of the vent cover. Information is also required on the physical prop erties of the vent cover related to the dynamics.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 28  Computations for the transient pressure and temperature are made by performing mass and energy balances in each volume during small time steps. The flow between inter-connected volumes is calculated assuming quasi-steady thermodynamic conditions in the volumes. It is assumed that the air-steam-water mixture in each volume is homogeneous and in thermodynamic equilibrium during the time step. The entrainment of the water from condensed vapor is conservatively assumed to be 100 percent. The flow is based on the vent path characteristics at the previous time step. The position of a blowout panel or hinged door, which determines the vent path characteristics, is obtained by solving the equation of motion at each time step, utilizing the calculated nodal pressures. (b) Vent Flow Calculations The COMPRESS code offers various options for calculating the flow between connected volumes. The flow can be sonic or subsonic depending upon the ratio of upstream and downstream
 
pressures. Subsonic flow is calculated using the incompressible flow equation or the ideal nozzl e equation. Sonic flow is calculated using the ideal nozzl e equation or M oody's two-phase flow correlation. However, in the present analyses, the vent flow calculations were made using the incompressible flow equation and the ideal nozzle equation, and the smaller of the two calculated flows is used as the actual vent flow for the conservative prediction of the subcompartment pressure. The complete entrainment of water with the air-steam mixture, the homogeneity of air-steam-water mixture and the thermodynamic equilibrium between the gas and liquid phases are assumed throughout the vent
 
path. (c) Vent Path Loss Coefficient The resistance to flow in a vent path is, in general, due to change in the flow area (expansion and contraction), friction, change in the flow direction and flow obstructions. Various components of the loss coefficient are calculated following the procedures outlined in Reference 10. All contractions an d expansions are considered to be sudden changes in the flow area. All the components are appropriately lumped together a nd used for a given vent path.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 29  3. Pressure Transient Analyses (a) Subcompartment Modeling Each subcompartment is subdivided into a number of nodes to determine a realistic subcompartment pressure response. Boundaries between the nodes are pl aced at locations where large pressure gradients can occur either due to physical flow restrictors or due to the fluid inertia. A particular nodalization is developed for a given subcompartment through a nodalization sensitivity study, to determine the minimum number of nodes required to adequately predict the pressure profiles in the subcompartment. This involves several nodalizations for each subcompartment with
 
various numbers of nodes, especi ally in regions where large pressure gradients exist. By comparing the results of various nodalization schemes for a given subcompartment, that scheme
 
which results in a converged solution so that the subcompartment pressure profiles are not appreciably changed by further nodal refinement is selected for the pressurization analysis. (b) Nodalization Sensitivity Study The transient pressure responses in various subcompartments are analyzed by subdividing into a set of inter-connecting nodes. The pressure at any particular location is usually sensitive to the nodalization scheme employed. As the nodalization is made more refined, the calculated subcompartment pressure responses are closer to the real situation. To ensure that a particular nodalization scheme is adequate to predict the subcompartment pressure responses, a nodalization sensitivity study is usually performed by employing various nodalization schemes. When it is established that by further refinement of the nodalization the subcompartment pressure responses or the equipment loads within the subcompartment do not change significantly, the nodalization scheme is considered adequate. (1) Reactor Cavity Reactor cavity overpressurization has been deleted from the design basis in accordance with the final NRC Rulemaking with respect to GDC-4, dated October 27, 1987.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 30  (2) Steam Generator Compartment Steam generator compartment overpressurization has been deleted from the design basis in accordance with the final NRC Rulemaking with respect to GDC-4, dated October 27, 1987. (3) Pressurizer Compartment The pressurizer compartment pressure transient analysis for the nodalization sensitivity study has been performed
 
following a double-ended rupture in the spray line at El. 34'-0". The three nodal schemes employed for the sensitivity study are given in Figure 6.2-54. In the 15-node scheme, the break region is divided into 2 nodes and the whole compartment is divided into 6 vertical regions. The 23-node scheme has 6 nodes in the break region and has 6 vertical regions. In the 29-node scheme, the whole compartment is divided in to 7 vertical regions.
The horizontal pressure prof iles around the pressurizer at the break elevation are shown in Figure 6.2-55, while Figure 6.2-56 gives the vertical profile. For the three nodalizations, Figure 6.2-57 depict s the sidewise forces and Figure 6.2-58 depicts the moments that act on the
 
pressurizer. The horizontal a nd vertical pressure profiles, as well as the forces and moments on the pressurizer, demonstrate that the 23-node scheme adequately predicts the compartment pressure response.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 31  (4) Pressurizer Skirt Cavity The pressure transients in the pressurizer skirt cavity are calculated following a double-ended guillotine rupture in the surge line. The two nodal schemes used for the sensitivity study are shown in Figure 6.2-59 and Figure 6.2-60. In the 5-node scheme the volume underneath the pressurizer down to El. (-)5'-11" is taken as the break node. In the 6-node scheme the break node consists of the volume down to El. 0'-0" only. The peak pressure in the two cases is within 3 percent, even though the break node volume in the 6-node model is about half of that in the 5-node model. Further refinement of the nodalization below El. (-)5'-11" will not affect the pressure in the skirt cavity because a choked flow condition is established from the break node to the adjacent volume in the skirt cavity. The 6-node scheme, therefore, is adequate for the skirt cavity pressurization analysis. (c) Results (1) Reactor Cavity Reactor cavity overpressurization has been deleted from the design basis in accordance with the final NRC Rulemaking
 
with respect to GDC-4, dated October 27, 1987. (2) Steam Generator Compartment Steam generator compartment overpressurization has been deleted from the design basis in accordance with the final NRC Rulemaking with respect to GDC-4, dated
 
October 27, 1987. (3) Pressurizer Compartment All postulated double-ended guillotine ruptures in the spray
 
line at various discontinuities have been considered to determine the design loads on the pressurizer supports and also on the compartment structures.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 32  The sensitivity study performed for a full double-ended rupture of the spray line (5.187 in I.D.) demonstrated the adequacy of the 23-nodal model. Using the model, the compartment pressurization was analyzed for a break at El. 34'-0". Figure 6.2-69 graphically shows the transient pressures in each node in the pressurizer compartment. The differential pressures for various compartment walls are given in Figure 6.2-70. The mass and energy releases
 
following a double-ended guillotine rupture used in the analysis are provided in Table 6.2-30. Table 6.2-31 presents the vent path characteristics used in the analysis.
The nodal data, the pressurizer compartment initial
 
conditions, calculated peak diffe rential pressures, design peak differential pressures and the design margins for a
 
spray line rupture are presented in Table 6.2-32. The calculated peak differential pressures listed in Table 6.2-32 are the maximum differential pressures considering all the postulated break locations listed in Table 6.2-16. (4) Pressurizer Skirt Cavity A double-ended guillotine rupture of the surge line (11.188 in. I.D.) at the pressurizer nozzle was analyzed using a 6-node model. The pressure response of the nodes is shown in Figure 6.2-71. The mass and energy release data used for this break is given in Table 6.2-33. The vent path characteristics are presented in Table 6.2-34. Table 6.2-35 presents the noda l data, skirt cavity initial conditions, calculated peak diffe rential pressures, design peak differential pressures and the design margins for this rupture. 6.2.1.3 Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents This analysis presents the mass and energy releases to the containment subsequent to a hypothetical loss-of-coolant accident (LOCA). The release rates are ca lculated for pipe failure at three distinct locations:  Hot leg (between vessel and steam generator)  Pump suction (between steam generator and pump)
Cold leg (between pump and vessel).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 33  During the reflood phase, these brea ks have the following different characteristics. For a break in the pump suction or cold leg piping, a portion of the accumulator or safety injection flow in the intact loop can bypass the downcom er and flow directly to the break. For a cold leg pipe break, all of the fluid which l eaves the core must vent through a steam generator and be vaporized by heat addition to the primary from th e secondary. However, relative to breaks at the other locations, the core flooding rate (and therefore the rate of fluid leaving the core) is low, because all the core vent paths include the resistance of the reactor coolant pump. For a hot leg pipe break, the vent path resistance is relatively lo w which results in a high core flooding rate but the majority of the fluid which exits the core bypasses the steam generators in venting to the containment. The pump suction break combines the effects of the relatively high core flooding rate, as in the hot leg break, and steam generator heat addition as in the cold leg break. As a result, the pump suction breaks yield the highest energy flow rates during the post-blowdown
 
period. The spectrum of breaks analyzed includes the largest cold and hot leg breaks, reactor inlet and outlet respectively, and a range of pump suction breaks from the largest to a 3.0 ft 2 break. Table 6.2-36 presents the specific cases analyzed and a list of tables which contain the results for each case. Because of the phenomena of reflood as discussed above, the pump suction break location is the worst case. For this reason a spectrum of break sizes has been used in this analysis for the pump suction location. Other break locations result in less severe containment pressure transients than the pump suction location. Smaller break sizes at these locations result in less severe transients than full double-ended guillotine breaks. Therefore the hot leg and cold leg locations have only been analyzed with an assumed double-end ed guillotine break. The LOCA transient is typically divided into four phases:  Blowdown - includes the period from accide nt occurrence (when the reactor is at steady-state operation) to the time when the total break flow stops. Refill - the period of time when the lower plenum is being filled by accumulator and safety injection water.  (This phase is conservatively neglected in computing mass and energy releases for containment evaluations.)  Reflood - begins when the water from the lower plenum enters the core and ends when the core is completely quenched. Post-Reflood - describes the period following the reflood transient. For the pump suction and cold leg breaks a two-phase mixture exits the core, passes through the hot legs and is superheated in the steam generators. After the broken loop steam generator cools, the break flow becomes two phase.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 34  a. Mass and Energy Release Data
: 1. Blowdown Mass and Energy Release Data Table 6.2-37, Table 6.2-38, Table 6.2-39, Table 6.2-40, Table 6.2-41 and Table 6.2-42 present the calculated mass and energy releases for the
 
blowdown phase of the various break s analyzed with the corresponding break size. 2. Reflood Mass and Energy Release Data Table 6.2-43, Table 6.2-44, Table 6.2-45, Table 6.2-46 and Table 6.2-47 present the calculated mass and energy releases for the reflood phase of the various breaks analyzed, along with the corresponding safeguards assumption (maximum or minimum). The reflood results have been omitted for the hot leg break since the blowdown releases are sufficient to determine the peak containment pressure for this break location. 3. Dry Steam Post-Reflood Mass and Energy Release Data The calculated mass and energy releas es for the post-reflood phase with dry steam are provided in the reflood mass and energy release tables (Table 6.2-45, Table 6.2-46, Table 6.2-47) after end of 10-foot entrainment occurs. 4. Two Phase Post-Reflood Mass and Energy Release Data Table 6.2-48 and Table 6.2-49 present the two phase (froth) mass and energy release data for a double-ended pump suction break using minimum and maximum safeguards assumptions, respectively. Table 6.2-50 presents the results for a 0.6 ft 2 double-ended pump suction break using minimum safeguards. The double-ended pump suction minimum safeguards case is normally limiting. The two phase results are provided for other cases to prove that an upper bound calculation has been performed. This information is not provided for the three-foot squared pump suction split or the double-ended cold leg or hot leg cases. The peak containment pressures for these cases will occur during the blowdown phase of the transient. 5. Equilibration and Depressurization Energy Release Data The equilibration and depressurization energy release has been incorporated in the post-reflood mass and energy data. This eliminates the need to determine additional releases due to the cooling of steam generator secondaries and primary metal.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 35  b. Mass and Energy Sources The sources of mass considered in the LOCA mass and energy release analysis are given in the mass balance tables (Table 6.2-51, Table 6.2-52, Table 6.2-53, Table 6.2-54, Table 6.2-55 and Ta ble 6.2-56). These sources are: 1. Reactor Coolant System 2. Accumulators
: 3. Pumped injection.
Likewise the sources of energy considered in the LOCA mass and energy release analysis are given in the energy balance tables (Table 6.2-57, Table 6.2-58, Table 6.2-59, Table 6.2-60, Table 6.2-61 and Table 6.2-62). These sources include: 1. Reactor Coolant System
: 2. Accumulator
: 3. Pumped injection
: 4. Decay heat
: 5. Core stored energy
: 6. Primary metal energy
: 7. Secondary metal energy
: 8. Steam generator secondary energy
: 9. Secondary transfer of energy (feedwater into the steam out of the steam generator secondary). The balances are presented at the following times:
: 1. Time zero (initial conditions)
: 2. End of blowdown time
: 3. End of refill time (The only difference that will be noted from the values at the end of blowdown is that some accumulator water will be transferred to the reactor coolant. Thus, the low plenum will be full at the beginning of the reflood transient.) 4. End of reflood time 5. The time when the broken loop steam generator reaches thermal equilibrium (for froth cases only)
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 36  6. The time when the intact loop steam generator reaches thermal equilibrium (for froth cases only) 7. Time of full depressurization (for froth cases only). The methods and assumptions used to release the various energy sources are given in Reference 11.
The following items ensure that the core energy release is conser vatively analyzed for maximum containment pressure.
: 1. Maximum expected operating temperature 2. Allowance in temperature of instrument error and dead band (+4 F) 3. Margin in volume (1.4 percent) 4. Allowance in volume for thermal expansion (1.6 percent) 5. Margin in core power associated with use of engineered safeguards design rating (ESDR) 6. Allowance for calorimetric error (2 percent of ESDR)
: 7. Conservatively modified coefficients of heat transfer
: 8. Allowance in core-stored energy for effect of fuel densification 9. Margin in core stored energy (+15 percent). c. Blowdown Model Description The model used for the blowdown transient (SATAN-VI) is the same as that used for the ECCS calculation. This model is described in Reference 12 and 13. Reference 11 provides the method by which this model is used. d. Refill Model Description At the end of blowdown, a large amount of water remains in the cold legs, downcomer and lower plenum. To conservatively model the refill period for the purpose of containment mass and energy releases, this water is instantaneously transferred to the lower plenum along with sufficient accumulator water to completely fill the lower plenum. Thus, the time required for refill is conservatively neglected.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 37  e. Reflood Model Description The model used for the reflood transient (WREFLOOD) is a slightly modified version of that used in the ECCS calculation. This model is described in Reference 12 and 14. Reference 11 describes the method by wh ich this model is used, and the modifications. Transients of the principal parameters during reflood are given in Table 6.2-63 and Table 6.2-64 for the double-ended pump suction break with minimum and maximum safeguards. f. Post-Reflood Model Description (FROTH)
The transient model (FROTH) along with its method of use is described in Reference 8. g. Single Failure Analysis The effect of single failures of various ECCS components on the mass and energy releases is included in these data. Two analyses bound this effect. No single failure is assumed in determining the mass and energy releases for the maximum safeguards case. For this case a failure must be assumed in the Containment Cooling Systems. Normally the limiting case is the loss of one spray pump. For the minimum safeguards case, the single failure assumed is the loss of one emergency diesel. This failure results in the loss of one pumped safety injection train. The analysis of both maximum and minimum safeguards cases assures that the effect of all credible single failure is bounded. h. Metal-Water Reaction In the mass and energy release data presented here, no Zr-H 20 reaction heat was considered because the clad temperature did not rise high enough for the rate of the Zr-H 2 0 reaction to be of any significance. i. Energy Inventories Energy inventories for primary and secondary systems are tabulated for hot leg, cold leg, and pump suction breaks in Table 6.2-57, Table 6.2-58, Table 6.2-59, Table 6.2-60, Table 6.2-61 and Table 6.2-62. j. Additional Information Required for Confirmatory Analysis System parameters needed to perform confirmatory analysis are provided in
 
Table 6.2-65.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 38  6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures inside Containment
: a. Mass and Energy Release Data The mass and energy releases into the containment following a postulated main steam line break (MSLB) have been calculated by using the model described in Subsection 6.2.1.4d and incorporating th e balance-of-plant parameters for Seabrook Station via the procedur e described in Reference 15. The effects of a postulated feedwater line rupture are not as severe as the main steam line break because the break effluent of a feedwater line rupture is at a lower specific enthalpy. Therefore, feedwater line break mass and energy releases to the containment are not a ddressed here since they are bounded by steam line break releases. 1. Break Type/Size and Operating Power The plant operating power levels at the time of the MSLB and the spectrum of break types and sizes an alyzed have been presented in Table 6.2-10 and Table 6.2-11, respectively. Full double-ended rupture (DER) area is determined by the integral flow restrictor area. This break represents the largest possible break. A small double-ended rupture has been considered for each power level. These break sizes have been chosen to be large enough to generate a steam line isolation signal from the Primary Protection System. For any ruptures smaller than these small
 
double-ended ruptures, an isolation signal is generated by containment pressure. Two such cases have been analyzed with approximately half the corresponding size of the small double-ended rupture. These breaks are expected to cover adequately the full spectrum of double-ended break sizes. For the split ruptures, the break sizes selected are the largest sizes which will not generate a steam line isolation signal from the Primary Protection System. An isolation signal is generated on containment pressure. Larger split ruptures will generate primary protection signals and are expected to be bounded by the double-ended ruptures. The breaks are assumed to be at the exit of a steam generator flow restrictor for
 
double-ended ruptures, and at any point on the piping between a steam generator and the first main steam pipe whip restraint inside the containment for split ruptures.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 39  For full-size double-ended guillotine ruptures, the model allows credit for the flow-limiting effect of the MSIV (one for each loop) on the reverse flow blowdown. The reverse flow cons ists of (1) an initial blowdown of the steam in the piping between the break (immediately downstream of a steam generator flow restrictor) and the nearest MSIV, which is controlled by the piping cross-sectional area (4.18 ft 2); (2) a subsequent flow from the intact steam generators, which is controlled by the MSIV seat area
 
(1.97 ft 2); and (3) a post-isolation piping steam blow down. The initial piping blowdown is assumed at a consta nt choked flow rate corresponding to the initial pressure in the line. The flow from the intact steam
 
generators, which is controlled by MSIV seat area, is conservatively calculated assuming a pressure decay curve for a break area smaller than the MSIV seat area for a given powe r level. Following isolation, the blowdown of the steam remaining in the piping (from the break up to the MSIVs of the intact steam generators for a single failure of a MSIV and up to the nearest MSIV for all other single failures) is calculated assuming the flow rate drops to zero linearly. The post-isolation blowdown was calculated by following the procedure given in Reference 15 for the small double-ended breaks.
The reverse flow consists of (1) a reverse flow from the intact steam generators, and (2) a post-isolation piping steam blowdown, both
 
controlled by the break size.
For split ruptures, isolation of the main feedwater and main steam lines is initiated when the containment pressure reaches 6.8 and 7.4 psig, which are the upper bounds of the isolation setpoints for the main feedwater and main steam. The time when this pressure is reached was found by computing the containment pressure-temperature response using blowdown data obtained by assuming no isolation. The analytical method and initial conditions for this calculation have been presented in Subsection 6.2.1.1. The correct blowdown in the period following
 
isolation was then calculated by applying the calculated isolation time.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 40  2. Fluid Inventory for Release The total inventory of fluid available for release is characterized by the steam generator dryout time, which is defined as the time when the blowdown rate of the affected steam generator is equal to the rate at which the feedwater is entering. The dryout time depends on four steam/water sources:  (1) initial steam generator mass, (2) mass added by feedwater flashing, (3) mass added by the Main Feedwater System, and (4) mass added by the Emergency Feedwater System. Item (1) is discussed in Subsection 6.2.1.4c. For Item (2), the actual water volume in the main feedwater piping between the affect ed steam generator and the valve effecting the isolation was used. Item (3) is obtained by integrating the transient flow rate over an appropriate time period. The flow rate was first obtained as a function of the pressure in the affected steam generator, considering the system resistance and characteristics of the various pumps. The pressure in the intact steam generators was assumed to remain
 
unchanged during the transient. The main feedwater pumps were assumed to be operating at the maximum speed until they were tripped and all feedwater control valves (FCV) were assumed to be at their initial
 
position, except the one in the broken loop which was assumed fully open, until their isolation is completed. It was assumed that the main feedwater pumps are immediately tripped upon rece ipt of the isolation signal, while all FCVs close instantaneously after a further delay equal to the maximum valve stroking time. The maximum stroking time of the feedwater isolation valves (FWIVs) and FCVs is discussed in Subsection 6.2.1.1. 
 
The pressure-flow relationship thus calculated was then converted to give the main feedwater flow rate as a function of time using the calculated pressure transient for the affected steam generator. Finally, Item (4) was obtained by multiplying the constant flow rate by the dryout time, or 30 minutes (time of isolation of the emergency feedwater line), as
 
appropriate.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 41  3. Mass-Energy Release Data for Most Severe Ruptures The mass and energy releases have been calculated for the 102 accident-single failure combinations, as can be deduced from Table 6.2-10, Table 6.2-11 and Table 6.2-12. The mass flow rate and the enthalpy for the forward flow and the reverse flow for the case of maximum containment peak pressure (a full DE rupture at hot shutdown with one containment spray train failed) are given in Table 6.2-66 and Table 6.2-67, respectively. Similar blowdown information for the case of maximum containment peak temperature (a full DE rupture at 102 percent
 
power with the broken loop MSIV failed) is given in Table 6.2-68 and Table 6.2-69. b. Single Failure Analysis The single failures postulated to concur with the MSLB have been presented in Table 6.2-12. Failure of a FWIV (Single Failue No. 1) increases the dryout time through Item (2) described in the previous subsection. The main feedwater available for flashing with and without this failure is 558.0 cu. ft and 160.5 cu. ft, respectively. Failure of a FCV (Single Failure No. 3) and failure of the main feedwater pump trip (Single Failure No. 4) increase the dryout time through Item (3) only, while failure of the emergency feedwater pump run out control (Single Failure No. 5) increases the dryout time through Item (4) only. Failure of the broken-loop MSIV (Single Failure No. 2) extends the steam volume available for the post-isolation piping blowdown to in clude the portion bounded by the MSIVs, the turbine stop valves (TSVs), the condensers, and the moisture separators/reheaters. The steam volum e with and without this failure is 11907.1 cu. ft and 969.9 cu. ft, respectively.
Failure of one of the two containment spray trains (Single Failure No. 6) results in a reduced containment heat removal rate, but has no eff ect on the blowdown. c. Initial Conditions A spectrum of power levels spanning the operating range, 102 percent, 75 percent, 50 percent, 25 percent, as well as the hot shutdown condition, has been considered. At each power level, plant initial conditions corresponding to the power level were assumed. Initial steam generator mass corresponding to the design mass limits was assumed.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 42  Offsite power was assumed to be available. Specifically, this means no credit was taken for tripping of the reactor coolant pumps in determining the mass and energy releases. Tripping of the pumps reduces heat transfer capability from the primary plant into the steam generator, which reduces the effects of core power generation decay heat and thick metal energy and energy from intact steam generators on break releases. Further details of the initial conditions may be found in Reference 15. d. Description of Blowdown Model The steam generator blowdown model assumes dry steam. Details are available in Reference 15. e. Energy Inventories Mass and energy balances are provided in Table 6.2-69a and Table 6.2-69b for the most severe secondary pipe ruptur es based on the highest peak calculated containment pressure and temperature. f. Additional Information Required for Confirmatory Analysis No additional information is deemed necessary for the performance of a confirmatory analysis.
6.2.1.5 Deleted 6.2.1.6 Testing and Inspection Information concerning preoperational leakage testing and periodic in-service leakage surveillance of the containment to ensure functional capability of the containment and associated structures is provided in Subsection 6.2.6. Testing and inspection requirements for safety systems which support the functiona l capability of the containment and asso ciated structures are discussed in the respective sections for the individual systems. 6.2.1.7 Instrumentation Requirements Instrumentation is provided to monitor containment pressure, temperature humidity, hydrogen concentration, radiation levels and sump and flood water level to assist normal plant operations. Instrumentation to monitor containment parameters for accident monitoring is discussed in detail in Section 7.5. Containment post-LOCA radiation monitoring, area and airborne radioactivity monitoring instrumentation is discussed in detail in Subsection 12.3.4. Containment post-LOCA hydrogen monitoring is discussed in Subsection 6.2.5.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 43  6.2.1.8 Containment Analysis at an Analyze Core Power Level of 3659 MWt As discussed in Section 6.2.1.1.a, the peak containment pressure and temperature predicted by the GOTHIC model for an analyzed core power level of 3659 MWt is bounded by the results of the original containment analysis performed using the CONTRAST-S-MOD1 program. This section provides a discussion of the results/data associated with containment response analyses of record, as shown on the applicable tables and figures in this section, as compared with the results/data associated with the containment re sponse at the analyzed co re power level of 3659 MWt. a. Long-Term LOCA Containment Pressure and Temperature Response The pressure and temperature time/histor ies shown in Tables 6.2-8, 6.2-9 and Figures 6.2-1 through 6.2-18 are based on the original and bounding containment analysis. A new GOTHIC code analysis was performed to baseline the original analysis and determine the impact of the analyzed core power level of 3659 MWt.
The peak temperature and pressure va lues shown for the bounding EQ profile curves remain bounding at the 3659 MWt an alyzed core power level. The actual calculated pressure and temperature values at the 3659 MWt analyzed power level vs. time differ slightly from the curves presented. The results show that the peak containment temperatures and pressures and the peak containment sump water temperature at an analyzed core power level of 3659 MWt are bounded by the peak values determined in the analysis of record.
: b. Long-term LOCA Mass and Energy Release Analysis The data presented in Tables 6.2-37 through 6.2-64 and Figures 6.2-87 through 6.2-90, applicable to the containment response analysis of record, remain bounding at the analyzed co re power level of 3659 MWt for the following reasons: (1) The LOCA mass and energy release data used in the analysis of record are generic, whereas the mass and energy release data used in the analyzed 3659 MWt core power analysis were generated based on a Seabrook Station plant-specific model, (2) The decay heat model used in the generic data was based on American Nuclear Society (ANS) 1971 + 20%, whereas the 3659 MWt analyzed core power analysis was performed with ANS 1979 + 2, and (3) In the 3659 MWt analyzed core power analysis, credit was taken for steam-water interaction in the Reactor Coolant System loop piping, which was not available when the generic data was calculated.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 44  c. MSLB Containment Pressure and Temperature Response The pressure and temperature time/hist ories shown in Tables 6.2-13 through 6.2-15 and Figures 6.2-20 through 6.2-23 are based on the original and bounding MSLB containment analysis. For the 3678 MWt NSSS analyzed thermal power level, a new GOTHIC based analysis was performed to baseline original analysis and determine the impact of the analyzed core power level of 3659 MWt. The peak temperature and pressure values shown for the bounding EQ profile curves remain bounding at the 3659 MWt analyzed core power level. The actual calculated pressure and temperature values at the 3659 MWt analyzed core power level vs. time differ slightly from the curves presented. The limiting peak temperature for MSLB under 3659 MWt analyzed core power conditions is 357.4°F for a double ended rupture at 100% power with a single train failure. This is bounded by the existing peak containment temperature during a MSLB of 364°F.
The existing containment peak pressu re following a MSLB is 36.1 psig determined for a double-ended rupture of the main steam line at hot standby power with failure of one spray train. This peak containment pressure determined for MSLB conditions is well below the peak containment pressure determined under LOCA following a full double-ended rupture of the reactor coolant pump suction piping with two trains of safety injection in operation.
The highest containment peak pressure based on 3659 MWt analyzed core power level conditions following a MSLB is cal culated to be 37.3 psig for a doubled ended rupture at near zero power with failure of one diesel generator (i.e., one train). Peak containment pressure of 30.8 psig is developed for MSLB under 3659 MWt analyzed core power conditions for the double-ended rupture at 100%
power with failure of one train. The current analysis of record was performed using mass and energy release data for a generic Westinghouse plant, whereas the results under the 3659 MWt analyzed core power level conditions are specific to Seabrook Station. Although the peak containment pressure resulting from MSLB under 3659 MWt analyzed core power conditions is slightly higher than that developed in the analysis of record, this value remains bounded by the peak pressure determined in the loss of coolant accident analysis.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 45  d. MSLB Mass and Energy Release Analysis The containment MSLB response analysis of record was performed using mass and energy release data for a generic Westinghouse plant (Tables 6.2-66 through 6.2-69b), whereas the 3659 MWt core power analysis uses the RETRAN computer code for determining the MSLB mass and energy releases, as well as inputs that are plant-specific for Seabrook Station. As indicated above, results of the 3659 MWt core power analysis show that the containment MSLB response analysis of record remains bounding for peak containment temperature; the peak containment pressure determined in the 3659 MWt core power analysis is bounded by the peak pressure determined in the containment LOCA response analysis of record. 6.2.2 Containment Heat Removal System The containment is maintained below design pre ssure following a primary or secondary system line rupture by the parallel action of the Emergency Core Cooling System (ECCS) and the Containment Building Spray (CBS) System as active heat removal systems and by the passive heat sinks such as structural components. The ECCS is discussed in Section 6.3; details of the CBS system and the inter-relationship between the ECCS and CBS systems for removing heat from the containment are discussed in this section. Passive heat sinks, such as the containment liner and other structures, are described in Subsection 6.2.1. 6.2.2.1 Design Bases
: a. The ECCS and the CBS system are each comprised of two identical trains, each train independent of the other and fully redundant. Failure of a single active component will not cause the loss of more than half of either system's 200 percent heat removal capacity. Sufficient capacity to mitigate the consequences of an accident is thus assured with one CBS train and one ECCS train available. b. The reactor unit has its own CBS system and ECCS. c. The CBS system is designed to remove the energy discharged to the containment following a loss-of-coolant accident (LOCA) or main steam line break (MSLB) to prevent the containment pressure from exceeding design pressure and to reduce and maintain containment temperature and pressure within acceptable limits. The
 
postulated accident conditions for which the CBS performance is evaluated are discussed in Subsection 6.2.1.1. d. The sources and amounts of energy released to the containment during accident conditions that determine the required cap acities of the containment heat removal systems are discussed in Subsections 6.2.1.3 and 6.2.1.4.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 46  e. Only minimum containment spray capacity (one train) is required for design heat removal. The containment energy removal rate is sufficient to reduce containment pressure so that leakage is reduced to one-half of the design leakage of 0.15 percent of the containment air mass per day within 24 hours after the DBA. f. Assuming a loss of offsite power, the most limiting single failure is the failure of an emergency diesel generator to function leading to the loss of one safety train at the time of actuation. A detailed discus sion of single active failures is presented in Subsection 6.2.1.3. g. Components in the CBS system that are required to function during and subsequent to an accident are designated seismic Category I, and are designed to withstand the SSE wit hout loss of function. h. The capability of mechanical, instrumentation and electrical components in the containment heat removal systems to withstand the post-accident containment environmental conditions is discussed in Section 3.11. i. Design of the CBS system to withstand the effects of wind and tornado loading, floods, and missiles is discussed in Secti ons 3.3, 3.4, and 3.5, respectively. j. Components in the CBS system that are required to function during and subsequent to an accident are protected against the dynamic effects of pipe rupture, as discussed in Section 3.6. k. Design of the CBS system to withsta nd the effects of floods is discussed in Section 3.4. The contents of the RWST and SAT are require d for the safe shutdown of the reactor and cooling of the containment following the rupture of primary or secondary coolant system piping in the containment. The water in the RWST and SAT is not required for reactor cooldown if there is no pipe rupture. The RWST and SAT are not protected from tornado wind loads and missiles since the simultaneous occurrence of a pipe rupture in the containment and a tornado is considered incredible.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 47  6.2.2.2 System Design The P&I diagrams of the ECCS and CBS system are shown in Figure 6.3-1 and Figure 6.2-74, respectively. Figure 6.2-75 is a complete flow diagram of all the Engineered Safety Feature Systems. The valves in Figure 6.2-75 are shown in their open position so that the same diagram can be used to represent all modes of operation. Design parameters for the CBS system components are listed in Table 6.2-75.
: a. Operation The CBS system is actuated by a containment spray actuation signal (CSAS), which is initiated by high pressure in the containment. The CSAS is discussed in Section 7.3. The CBS system pumps water from the refueling water storage tank (RWST) to the spray nozzles located high in the Containment Building. The RWST contains a minimum of 450,000 gallons of borated water at a maximum temperature of 98F, and provides cooling for a minimum of 26 minutes after an accident, based upon maximum pumps in operation at maximum flow rates. Upon a low-low level signal from the RWST (approximately 350,000 gallons removed) in conjunction with an "S" signal, the sucti ons of the residual heat removal (RHR) and CBS pumps automatically re-align to take suction from the containment recirculation sumps. The operator then manually re-aligns the centrifugal charging pumps to take suction from RHR pump P-8A discharge and the safety injection pumps to take suction from RHR pump P-8B discharge. All pumps continue to operate in the recirculation mode until no longer required. Heat tracing is not required for the piping in this system since no part of the system is exposed to temperatures below 40 F. b. Component Description The following are descriptions of the components in the CBS system. RHR pumps and heat exchangers are described in Section 6.3 and Subsection 5.4.7; ECCS component descriptions are found in Section 6.3. 1. Containment Spray Pumps The CBS pumps are horizontal centrif ugal pumps selected to supply the design spray flow rate at containment design pressure. The pumps are designed to take suction from the containment sump at the most limiting NPSH condition (atmospheric pressure and a temperature of 212F) and pump it back into the containment through the spray nozzles. Design pump discharge pressure takes into account containment pressure, elevation head to the highest nozzles, and piping frictional losses.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 48  Each CBS pump is designed to deliver 3010 gpm from the lowest level in the RHR equipment vault to the highest point in the Containment Building. The minimum calculated CBS system flow rate of 2808 gpm during injection has been shown to be adequate to maintain the containment pressure and temperatur e within the design envelope for worst case primary and secondary side ruptures. 2. Spray Additive Tank The spray additive tank (SAT) is mounted adjacent to the RWST, and drains by gravity into the RWST mixing chamber through a six inch diameter pipe which has redundant valving. This line connects the bottoms of the SAT with the RWST mi xing chamber. External heaters are provided to prevent freezing or chemical precipitation during cold weather. The mixing ratio of the spray additive tank volume to the RWST volume is such that the pH of the sp ray solution during the injection phase will average between 9.0 and 9.6 units. The tank is sized to provide the correct amount of sodium hydroxide so lution to insure that the final containment recirculation sump pH after injection will be between 8.5 and 11.0 units for the various reactor coolant conditions. No provision is made
 
in the design of the SAT to pr event the reaction of NaOH with atmospheric carbon dioxide during long-term storage. Proper concentration of sodium hydr oxide between 19 and 21% by weight will be verified periodically by chemical analysis. 3. Containment Spray Heat Exchangers The containment spray heat exchange rs are shell and tube-type heat exchangers with spray flow in the tube side and primary component cooling water (PCCW) on the shell side. They are sized such that one containment spray heat exchanger and one residual heat removal heat
 
exchanger provide 100 percent of design heat removal capacity. Heat exchanger parameters, including flow rates, were selected so that one RHR heat exchanger and one CBS heat exchanger satisfy containment cooling requirements. Table 6.2-76 contains the heat exchanger performance data used for the accident analyses. 4. Spray Headers and Nozzles The spray headers are positioned in the containment dome to maximize coverage of the containment volume. Four separate headers are used to obtain the distribution of the flow, two for each train. Each train contains 198 nozzles with each nozzle providing a design flow of 15.2 gpm (see Figure 6.2-76 and Figure 6.2-77).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 49  5. Refueling Water Storage Tank The refueling water storage tank (RWST) is designed to store 475,000 gallons of borated water. Th is tank is designed to supply water both for refueling operations and to the Containment Spray System and the Emergency Core Cooling System during accident operations. The RWST capacity is based on accident requirements and will supply the safety injection, the charging, residual heat removal and containment spray pumps for at least 26 minutes dur ing the injection phase of a design base accident. Margin is provided to allow time for transfer of the systems to the recirculation mode and to account for instrument errors. Analysis is based on a minimum of 350,000 gallons of water being injected. An external steam heating supply system is provided to protect against freezing. Tank temperature is indicated locally and alarmed in the main control room. c. Material Compatibility The components of the CBS system, including the spray nozzles, are fabricated of materials listed in Table 6.2-75. The pH of the sump water following an accident is monitored to ensure that the pH is maintained in the correct range (discu ssed in part f. of this section) as the hydroxide is consumed by chemical reaction with zinc and aluminum within the containment. Two sample points exist to withdraw samples downstream of the RHR heat exchangers: the normal connection to the sample sink, and a local sample point. Sodium hydroxide can be added for pH adjustment using the chemical and volume control system tanks and pumps. The solution is prepared in the chemical mixing tanks and supplied to the suction of the charging pumps. The charging pump suction is fed from the RHR system during recirculation. Neither the containment spray pumps or motors nor other e ngineered safeguard pumps or motors are exposed to the containment atmosphere; accordingly, no adverse effects are considered due to the post-accident containment environment.
: d. Redundancy Minimum allowable cooling capacity is assured by utilizing the "double train" concept in both the Emergency Core Cooling and the Containment Spray Systems. These trains are independent of each other, with no interconnection, so that a single active component failure will not cause loss of function of the system. An analysis of a failure of each component is presented in Table 6.2-77.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 50  e. Recirculation Piping There are two penetrations from the containment sump to the Primary Auxiliary Building (PAB), with each pipe encased in a sleeve. In each line, immediately inside the PAB, is a motor-operated ga te valve. After passing through the isolation valve, the flow in each lin e divides to supply one CBS and one RHR pump. Each isolation valve is enclosed within a housing de signed to withstand containment design pressure to prevent any leakage to the PAB atmosphere. f. Containment Spray System pH Values The pH of the system is dependent on the ratio of boric acid to caustic. The interaction of these two compounds has been investigated (References 17 and 18) and this data has been used to correlate containment spra y composition and pH. The 21 percent by weight sodium hydroxide in the spray additive tank, when mixed with the borated water in the RWST, nominally produces a pH of 9.6 or less during the injection phase. For maximum initial tank level mismatch, due to instrument uncertainties, the spray pH could exceed 10.3 for about 6 minutes. The calculations of the spray pH range included such considerations as the physical piping arrangement between the additive and water storage tanks, variation in relative liquid heights in the tanks due to instrument uncertainties and permissible variations in the concentra tions of sodium hydroxide in the spray additive and boron in the refueling wa ter storage tanks. Since the UE&C computer program (MIXCH) used in the above calculations has successfully simulated the test data gathered in the testing of a similar gravity feed chemical injection system (Arkansas Nuclear One), a full-scale pH test is not considered necessary for Seabrook. This analysis technique (MIXCH) is detailed in Reference 23. The maximum pH of the containment sump after a LOCA depends on the concentration of boron in the reactor coolant, i.e., the sump pH is 9.4 at zero ppm boron in the reactor coolant and 8.8 at a 4000 ppm boron concentration.
Corrosion products in the solution have a te ndency to slightly suppress the pH, generally to an extent less than 0.1 unit. In particular, for the composition of the containment sump water, the reduction in pH is expected to be even less (about 0.02 unit).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 51  g. Net Positive Suction Head Requirements Adequate net positive suction head is assured by locating the RHR and CBS pumps at the lowest level in the Aux iliary Building. Th e RHR and CBS pump available net positive suction heads from the containment sump were determined by assuming the limiting conditions in accordance with NRC Regulatory Guide 1.1 (pressure equal to atmospheric and temperature equal to 212ºF). The CBS pump available and required net positive suction head is shown in Table 6.2-75.
The RHR pump available and required net positive suction head is shown in Table 6.3-1.
: h. Heat Exchanger Surface Fouling The materials used for the CBS heat exchanger are listed in Table 6.1(B)-1. The shell side of the heat exchanger is cool ed by the PCCW system which contains a corrosion-inhibiting agent and operates as a closed system. The tube side, which is in contact with the emerge ncy core coolant, is corrosion resistant. The effect of corrosion fouling on heat exchanger surfaces will, therefore, be minimal;
 
however, fouling factors were included in the detailed design of the units to assure the required heat removal capabili ty through conservative design. i. Heat Exchanger Performance The heat exchanger (residual heat removal, containment spray, primary component cooling water) temperatures have been selected based upon maximum service water (ultimate heat sink) temperatures and the amount of heat removal required. The flows, geometry, and surface area were studied in the evaluation of the containment pressure-temperature analysis. Those parameters selected and tabulated in Table 6.2-76 are those which meet the design basis requirement for containment cooling. j. Containment Recirculation Sump and Strainer Design The containment recirculation sump collects and strains the water available for supplying the residual heat removal, containment spray, safety injection and high head charging pumps during the recircul ation mode of oper ation following an accident. The sump is designed to meet the intent of Regulatory Guide 1.82. Two completely independent sumps are located in the containment to maintain the "double train" concept as de scribed in Subsection 6.2.2.2d. One sump supplies water to Train A and the other sump supplies Train B. The arrangement of these sumps is shown in Figure 6.2-79. The minimum water level in containment during a loss-of-coolant accident is nominally Elevation (-)23.79 ft.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 52  A series of debris interceptors are provided on the containment floor within the recirculation flow paths. The debris interceptors reduce the quantity of debris transported to the sumps by trapping debris and allowing the remaining debris more time to settle prior to reaching the sumps. Heavy particles are prevented from reaching the sumps by sloping the surrounding floor away from the sumps. This facilitates settling of debris on the floor prior to reaching the sump area. A strainer is installed in each sump. Each strainer consists of rows of vertically oriented strainer panels, consisting of a framework sandwiched between two sets of wire cloth attached to perforated plates. The maximum hole size of the perforated plates and maximum width of a gap between bolted structures is 0.068 inches. The strainer would therefore pr event debris particles 0.068 inches or greater in diameter which may be genera ted following a large break LOCA from passing through or bypassing the strainer and entering the ECCS system. The minimum physical restricti on in the ECCS flow path consists of 0.073 inches, which is the effective opening of the fuel assembly debris filter bottom nozzle in combination with the P-grid. Therefore, the strainer will prev ent recirculation of debris particles of sufficient size to impede cooling flow to the core.
The strainer panels are mounted on a plenum structure within the sump. The plenum is sealed to the sump floor and at the sump wall adjacent to the ECCS pipe inlet to ensure that all water entering the sump passes through the strainer panels. Water is drawn through the strainer panels and plenum and into the lower portion of the sump. The strainer will also act as a vortex preventor to further preclude air intrusion into the ECCS piping. The strainers have been designed to accommodate the debris generated and transported to the sump during the recirc ulation phase of a LOCA. The head loss due to debris on the strainer is less than the available NPSH margins for operating ECCS pumps, thereby ensuring that cavitation of the ECCS pumps will not occur. Therefore, the design meets the in tent of Regulatory Guide 1.82.
The potential for clogging of the sump strainers by equipment and piping insulation or loose insulation in the containment is minimized by the type of insulation used.
 
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 53  The thermal insulation inside the containment for piping and equipment except the reactor pressure vessel is fiberglass blanket insulation of the type commercially known as Nukon, manufactured by Owen's-Corning Fiberglass. The outside surface of the insulation blankets is covered with a stainless-steel jacket or is encapsulated in stainless steel wire mesh. Nukon is consistent with the recommendations of Regulatory Guide 1.36. The reactor pressure vessel is insulated with stainless-steel reflective insulation or fiberglass blanket.
Clogging of the strainers by nonsafety-related equipment is unlikely due to the remote location of the sump s relative to the NNS equipment and physical barriers separating the sumps from other areas in the containment. The supplementary neutron shielding around the reactor vessel which could be displaced by blowdown forces during an accident is designed to remain anchored and intact; hence, it is not a potential source of strainer blockage during an accident. The design of the sump suction piping ensu res that adequate flow and net positive suction head are available to all pumps under the most limiting containment conditions, as required by Regulatory Guide 1.1. The two sumps and the pumps they service are designed so that any single active or passive failure will not cause the loss of both A and B Train components. The sumps are visually inspected on a periodic basis to assure that they are clean, free of debris and that all strainers are intact and in position. The containment sump line isolation valves are exercised periodically to assure operability within Technical Specification requirements. k. Periodic Testing The provisions for periodic testing and inspection of the containment spray system are discussed in Subsection 6.2.2.4. l. Applicable Codes, Standards and Guides The codes, standards and guides applicable to the containment spray system are summarized in Table 6.2-79. m. Remote Manual Operation of the Containment Building Spray System The CBS system is designed to function completely automatically under accident conditions, hence there are no operations which must be performed manually by the operator from the main control board to initiate the proper function of the system during an accident. After the suctions of the CBS pumps are automatically switched over from the RWST to the containment recirculation sumps, the isolation valves (CBS-V2, CBS-V5) in the discharge line from the RWST will be closed by the operator.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 54  n. Plant Protection System Signals and Setpoints Operation of the CBS system is initiated automatically upon receipt of a containment spray actuation signal (CSA S). A CSAS is generated when the containment pressure reaches 19.8 psig (see also Subsection 6.2.1.1c.1(b)). The analysis limit of 19.8 psig is established by the requirement to maintain the maximum containment pressure during an accident as low as practical while keeping the setpoint as high as practical to minimize the probability of spray actuation following a small high energy li ne break. The Engineered Safety Features Actuation System is fu rther described in Section 7.3. o. Equipment Qualification Components in the CBS system which are re quired to function during the accident are qualified (vendor certification) to verify the ability of the components to perform their intended functions under the conditions specified in the purchase documents and/or by test. Environmental qualification of safety-related equipment is discussed in Section 3.11.
Tests and inspections are discussed in Subsection 6.2.2.4. Seismic qualification is addressed in Section 3.10. Pump and valve operability assurance is di scussed in Subsection 3.9.3.2. p. Containment Spray System Response Time Containment spray system response time is discussed in Subsection 6.2.1.1c. 6.2.2.3 Design Evaluation The analyses of the post-accident containment pressure transients are discussed in Subsection 6.2.1. The double train concept insure s that sufficient heat removal capacity will exist, even with a single active failure. Containment design pressure is not exceeded and containment pressure reduction reduces containment leakage to 50 pe rcent of the design leak rate within 24 hours after the DBA. The refueling water storage tank and the spray additive tank are located within an enclosure building in the yard area. The RWST and SAT are fully enclosed with insulated siding and roof, as well as by two heated buildings (PAB and WPB). Included within these enclosures is the associated piping, vent lines, and instrument tubing.
During cold weather conditions, both the SAT and RWST are heated by steam heating panels mounted on the exterior surface of the tanks. Calculations demonstrate that the RWST heating panel can maintain a minimum water tank temperature of 50F, and concurrently provide sufficient heat into the enclosure area to maintain an enclosure temperature of 39F. No credit was taken for heat contributions from the SAT heaters. The site environmental condition for this design evaluation assumed -17F and 30 mph winds, and are more conservative than the minimum outdoor conditions listed in Updated FSAR Figure 3.11-1.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 55  For the above environmental conditions, the heat loss from the Enclosure Building, including
 
infiltration losses, is 158,000 Btu/hr. as compared to an RWST heating panel capacity of 674,000 Btu/hr. Accordingly, freeze protection is provided for all equipment. In addition, both tank low temperature and enclosure temperature alarms are provided in the main control room. The maximum temperature of the water in the tanks is calculated to be 86 F, using plant specific meteorological data, assuming maximum solar heat gain and failure of the ventilation fans. The heat of reaction upon the mixing of the boric acid and the sodium hydroxide would raise the temperature of the tank contents approximately 2F, thereby raising the RWST maximum supply temperature to 88 F. Following a refueling operation, the RWST water temperature may reach a higher value. The maximum temperature considered for all evaluations is 98 F. Neither tank is protected against tornado missiles, and a tornado and accident are not considered simultaneous events. In the event of tornado damage to eith er tank, the affected unit would be shut down. The SAT is connected to the RWST by two parallel lines each with an automatic motor-operated valve. The valves are actuated and powered from separate sources to insure that the NaOH solution can be added to the containment spray even in the event of a single active failure. The method of addition of 20 percent NaOH solution in required concentrations to the borated water drawn from the RWST immediately following a LOCA is primarily dependent on passive components, such as tanks, pipes and a baffled mixing chamber. The rate of addition is dependent on the drawdown rate of the RWST and is based on principl es of hydrostatics and hydrodynamics. A description of the system is contained in the following paragraphs. The system outlined here was chosen in preference to other available systems because it relies upon a minimum number of active components for adding and mixing the NaOH with the borated water and therefore is not dependent on the proper operation of active components (such as eductors and associated recirculation valves, etc.), to achieve the required mixing ratio.
Addition of NaOH to the borated water is accom plished by gravity feed through a 6" pipe connecting the SAT and RWST which together remain in hydrostatic equilibrium throughout the period of ECCS injection into the core and the containment atmosphere. A mixing chamber is provided inside the RWST to thoroughly mix the NaOH with the borated water supplied to the containment spray system by the RWST. The redundant motor-operated isolation valves between the SAT and the RWST are normally closed and are automatically opened by the containment spray actuation signal. Once these isolation valves are open, the SAT draws down simultaneously with the RWST as the residual heat removal, safety injection, charging and containment spray pumps withdraw water from the RWST. The operating levels of the SAT and RWST are maintained during normal operation to ensure that the two tanks are in hydraulic equilibrium. Sin ce the two tanks are in hydraulic conjunction with each other, they remain in hydrostatic equilibrium throughout the injection phase.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 56  The Containment Spray Actuation System and po wer supplies are independent and redundant to ensure actuation and power in the event of a single failure. An analysis of possible failures is presented in Table 6.2-77. a. Passage of Spray Water to the Recirculation Sumps The containment spray water is dist ributed over the operating floor at elevation 25' and is then directed to one of three di fferent flow paths to the containment sump. Water holdup is minimized by grated floors and adequate openings in compartment walls and shielding to allow passage of water to the recirculation sumps, as described below: 1. The operating floor has 3520 ft 2 of grating which will pass the spray directly to the lower floor at elevation 0'-0". Spray that impacts and collects on the solid floor at this elevation will freely drain to the grated floor areas. Elev ation 0'-0" has 1270 ft 2 of grating which will allow the water draining from above to drain to elevation -26' where the spray will drain to the recirculation sump. 2. The steam generator shielding extends above the operating floor to elevation 32 ft. The top of these shielded cubicles is open to collect spray. However, this path is open to elevation -26 ft and the spray will drain directly to the sump. 3. There are openings in the reactor operating floor for the refueling canal, reactor internals lay down and access to the reactor head region. 
 
Eventually, all spray im pacting this area drains to the annular region between the reactor vessel and the concrete primary shield wall. This area is not isolated during normal plant ope ration. Low elevation portions of the refueling canal (See Figure 1.2-6) will drain to the -26' elevation through three four-inch lines each having two valves in series, normally open, but both closed during refueling connected to a common drain path.
Each drain path is isolated by two valves in series during refueling.
Adequate openings have been provided in the missile shield walls to allow free passage of water at elevation -26' to the recirculation sump. The maximum total trapped volume of spray water is 22,830 ft 3 (170,780 gal). See Subsection 6.2.1.1b.6 for details. This is significantly less than the 46,788 ft 3 (350,000 gal) of water supplied from the RWST. b. Changeover from Injection to Recirculation The changeover from the injection mode to the recirculation mode during an accident is described in Subsection 6.3.2.8. The containment spray pumps function the same as the RHR pumps during the changeover.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 57  Upon reaching a low-low level in the RWST, the recirculation mode of operation is automatically initiated. The two containment sump isolation valves are automatically opened when the low-low level signal in the RWST is indicated in conjunction with an "S" signal. This va lve opening realigns the suction of the containment spray pumps to the sumps for the start of recirculation. The containment spray pumps continue to ope rate during this switchover. After the switchover is completed, the isolation valves in the discharge line from the RWST are closed by the operator. Each of the containment spray subsystems is capable of satisfying the system function and either can be removed from service. A single failure does not prevent the transfer to the recirculation mode since each active component is duplicated by the dual train concept. c. Spray Effectiveness Each spray train contains 198 SPRAYCO 1713A hollow-cone ramp bottom nozzles. Sixty-five nozzles were randomly selected from a quantity of 325 to evaluate the performance of the nozzles and verify the required flow of 15.2 gpm at 40 psi differential pressure. The average mean droplet diameter for the nozzles tested was 660 microns. This compares to a conservative value of 1250 microns used in the containment iodine removal analysis. The average mean drop diameter was arrived at by numerical averaging based on an instantaneous sampling of spray at design conditions. Table 6.2-80 lists the percentage of sprayed volume and Figure 6.2-80 and Figure 6.2-81 show the extent of overlap ping of the sprays in plan for spray loops A and B, respectively. These figures show virtually 100 percent coverage of the containment at the operating floor level. Figure 6.2-82 and Figure 6.2-83 show the spray loops A and B covera ge pattern in elevation views.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 58  Figure 6.2-84, Figure 6.2-85 and Figure 6.2-86 summarize the operating characteristics of the spray noz zles at 40 psi. It should be noted that 99.99 percent of the drops are below a diameter of 1500 microns, have a terminal velocity less than 17.88 fps and contain over 99.99 percent of the total liquid volume. It is assumed that, following a LOCA, the sprays are initiated at a time when the containment atmospheric temperature is 266F, so that the air steam ratio is 0.78 pounds air per pound steam and the initial spray temperature is 100 F. The drop is considered to be a rigid sphere of radius r o initially at T o in an air/steam atmosphere at T oo, and T oo does not change during the time the drop is falling. Steam in contact with the drop will condense, leaving a boundary layer of air around the drop. This is e quivalent to having an extremely large air-steam mass ratio, essentially 100 percent air and 0 percent steam. It is also conservatively assumed that this boundary layer is created instantaneously. Heat is transferred to the drop through th e boundary layer by convection, and water vapor diffuses through the boundary layer. Ranz and Marshall (Reference 19) provide correlations for both mass and heat transf er. After diffusing through the boundary layer, the steam will condense on the drop surface and the latent heat of condensation will act as a surface heat source. The assumption of a rigid drop implies the longest time for the drop to heat up, thus providing the most conservative case.
The following equations were used in the analysis: 121 2 2ScRe0.31PD VRTrMPk 1/3 1/2 o m BM g 3PrRe0.31 k hr 1/3 1/2 o S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 59  where T TTT o r t o 2 r r o aa r o V t2 Re 428028607140142171401530..a.ap.p.D g.V t and diameter drop D p dropofdensitya mixtureofdensityp ityviscos gravityofonaccelerati g tcoefficien transfer mass K g droptheofvelocity thermal V t mixturefortconsgasR tan mixtureofpressuretotalP mixtureforydiffusivit thermal timet layerboundaryininertofpressure average P BM layerboundaryofweight molecular average M m vaporwaterofydiffusivit D v tcoefficien transferheath tyconductivithermalk S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 60  Re, Pr, Sc =Reynolds, Prandtl and Schmidt numbers Equation (1) describes the temperature behavi or of the drop; equation (2) describes the mass transfer coefficient; and equation (3) the convective heat transfer coefficient at the drop boundary. Parsly (Reference 20) has solved this set of equations numerically using a finite difference method for a range of drop sizes from 500 to 4000 microns in diameter, containment temperatures of 212 F and 266F, and initial drop temperatures of 86 F, 122 F and 176 F. Using these results for an averag e spray nozzle height of 134 feet and initial drop temperature of 100F, equilibration time is much shorter than the time required for the drops of essentially all sizes to reach the containment sump. Table 6.2-81 presents the parametric results. It can be concluded from Table 6.2-81 that even the largest spray drops attain the containment temperature at times far shorter than the time required to reach the containment sump. As a result, the spray effectiveness value of 1.0 is fully justified in the case of a LOCA. The effectiveness of the sprays following an MSLB, when the containment atmosphere is superheat ed, is discussed in Appendix 15B. d. Net Positive Suction Head Available Adequate net positive suction head (NPSH) for the containment spray pumps is
 
assured under all postulated operating conditions by analys is of the suction head available and vendor testing of the completed pumps. Maximum calculated flow under the most limiting NPSH conditions, i.e., during recirculation, is 3660 gpm. NPSH available at this flow is 23.76 feet versus a maximum required NPSH of 23.6 feet. The CBS pump analys is of available NPSH conservatively assumes that each residual heat removal pump (which shares a common suction on a train basis with each CBS pump) is also operating at design cold leg recirculation flow of 4388 gpm and considers the suction flow path with the highest hydraulic resistance. The formulas and flow resistance data in Reference 21 were used, along with GE strainer test data, to compute NPSH
 
available.
Table 6.2-78 lists the values of containmen t pressure head, vapor pressure head of pumped fluid, suction head, and friction head used in the analysis.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 61  e. Integrated Energy, Content of the Containment Atmosphere and Recirculation Water Figure 6.2-87 and Figure 6.2-88 show the integrated energy content of the containment atmosphere and recirculation water, respectively, as functions of time following the postulated design basis loss-of-coolant accident. The integrated energy absorbed by the structural heat sinks and removed by the containment heat removal heat exchangers is shown in Figure 6.2-89 and Figure 6.2-90, respectively. f. Debris The major source of debris that coul d be generated during a loss-of-coolant accident is insulation. The thermal insulation being used inside the containment will be both stainless steel reflective insulation and fiberglass insulation of the
 
type commercially known as Nukon, manufactured by Owen's-Corning Fiberglass, with a stainless steel jacket over the outside surface of the insulation.
Nukon is consistent with the recomm endations of Regulatory Guide 1.36. 6.2.2.4 Testing and Inspection The preoperational testing of the containment heat removal system verified the functional capability of the individual systems under operational conditions. Testing and inspection of the ECCS systems is discussed in Section 6.3 and Chapter 14. The preoperational testing of the Containment Spray System verified the operational parameters of the spray pumps during recirculation to the RWST. This testing included a demonstration of system response to ESF signals and the ability of the sump to supply the containment spray and residual heat removal pumps. Flow testing of the nozzles was performed by the manufacturer and was not performed in the field. An air flow test was performed to verify that no nozzles are plugged. Operability of the gravity feed system was demonstr ated during preoperati onal testing of the ECCS Performance Test (Table 14.2-3, Item 8).
The preoperational test will demonstrate the draw-down characteristics of the RWST and SAT during the different flow conditions of the ECCS Performance Test. The Containment Spray System will be inspected and tested periodically in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section XI and the requirements of the Technical Specifications.
The containment recirculation sumps will be visually inspected periodically to insure that they are free of debris and all strainers are intact.
 
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 62  6.2.2.5 Instrumentation The Containment Heat Removal System is provided with instruments and controls to allow the operator to monitor the status and operation of the spray system and to allow the automatic or manual initiation of the injection and recirculation modes of operation. The manual spray actuation consists of four momentary controls (see Figure 7.2-1, sh.8). Actuation occurs only if two associated controls are operated simultaneously. This prevents inadvertent spray initiation as a result of operator error. The automatic initiation is by coincidence of 2 out of 4 prot ection set loops, monitoring the containment pressure. The spray actuation signal starts the containment spray pumps and positions all valves to their operating configuration. The design details of the Engineered Safety Features Actuation System are
 
presented in Section 7.3.
The details of the interlocks involved in the suction valve realignment from the RWST to the containment sump during the switchover from inj ection to recirculation mode are presented in Table 6.3-7. RWST instrumentation is discussed in Subsection 6.3.5.
Indications of pump operation are provided by pump status indication lamps and the pressure indications at the main control room. Alignment of automatic valves is indicated by the valve status indications. Additionally, a separate status monitoring indication system is provided at the control room for both modes of the spray system. This enables the operator to evaluate the extent to which the valves are open and if the system is operating effectively. Alarms are also provided to indicate that either train of the Containment Spray System is inoperative. The design features of the bypass and inoperable status alarm system which provide system level indication, in compliance with Regulatory Guide 1.47, are presented in Subsection 7.1.2.6.
Abnormal conditions of RWST level and temperature, RWST enclosure temperature, containment sump level, pump discharge pressure, pump motor temperatures, and heat exchanger outlet temperature are alarmed at the main control room to alert the operator. The design details of the Accident-Monitoring Instrumentation System are presented in Section 7.5. The Control and Display Instrumentation System is designed to operate under all normal and abnormal conditions, including loss-of-coolant accident and loss of power. Diversity, redundancy of the sensors, circuitry and actuating devices meet the requirements of IEEE-Standard 279 and ensure that minimum system function is provided under postulated abnormal conditions. No single failure of the control and instrumentation will prevent the spray system minimum safety function. The design details of the instrumentation system are presented in Section 7.1. A comparison of the containment water level instrumentation desi gn with each of the five clarification points of NUREG-0737, Position II.F.1.2, (Page II.F.1.16), is addressed below: a. The Seabrook design for containment water level complies with this requirement. Refer to clarification c. below for a discussion of the narrow range qualification.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 63  b. The wide-range level measurement is designed to monitor water levels that correspond to all the water from the primary and safety systems and one-half the condensate storage tank. This capability exceeds a liquid volume of 600,000 gallons. c. The narrow-range water level monitors are not required to operate after their respective sumps have been flooded as their function is to monitor operational leakage. They will only be exposed to a mild environment as any leakage that would cause a harsh environment woul d flood their sumps and would be detectable by the wide-range (recirculation) sump level indicators and instruments monitoring the containment atmosphere. The narrow-range containment sump level instrumentation will be covered by the maintenance/surveillance for equipment that is located in a mild environment. d. This requirement is not applicable to Seabrook. e. The functions of the wide-range level indication are: 1. Verify the existence of water in the containment as corroboration of detection of a LOCA 2. Verify that the water injected from the RWST is accumulating in the containment 3. Verify that there is adequate NPSH for the containment spray pumps 4. Verify that the containment water le vel is less than the design basis flood level. The accuracy of this indication has been determined, has been reviewed against the functions listed above, and has been found acceptable to support each function. 6.2.3 Secondary Containment Functional Design The function of the secondary containment (containment enclosure) is to collect any fission products which could leak from the primary containment structure into the containment enclosure and contiguous areas following a loss-of-coolant accident (LOCA). The containment enclosure provides a low leakage rate barrier between the containment and the environment to control all leakage from the containment boundary. The system is comprised of (a) a structural barrier surrounding the containment, adjacent vaults and penetration areas; and (b) a containment enclosure emergency cleanup system which mainta ins a pressure lower than ambient in the enclosure to prevent uncontrolled releases of radioactivity into the environment.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 64  6.2.3.1 Design Basis
: a. Containment Enclosure
: 1. The containment enclosure is design ed for 3 psig differential pressure. 2. The containment enclosure is designed to withstand the transient pressure and temperature conditions produced in the annulus between the containment and the enclosure as a result of either a LOCA within the containment or a high-energy pipe rupture within the containment
 
enclosure annulus. 3. The containment enclosure is capa ble of withstanding the external pressure conditions resulting from the maximum wind pressure postulated for the site, the external pressure drop resulting from a tornado, and tornado-generated missiles. 4. The containment enclosure is de signed to withstand a safe shutdown earthquake. b. Containment Enclosure Emergency Cleanup System
: 1. The system is capable of reducing the containment enclosure pressure to negative 0.25 inches w.g. (water gauge) following an accident and maintaining it at or below that level uniformly for up to one year. 2. The system is capable of processing the atmosphere of the containment enclosure space while maintaining the design negative pressure differential. 3. The system is designed to permit periodic inspection and monitoring of functional capability. 4. This system is designed in accordance with Regulatory Guide 1.52. 5. The system is designed to seismic Category I and Safety Class 2 requirements. 6. The system is designed to retain functional capability while experiencing a loss of offsite power concurrent wi th a LOCA and any single active component failure.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 65  6.2.3.2 System Design
: a. Containment Enclosure The containment enclosure is comprised of a right cylindrical structure with a hemispherical dome and other penetration and equipment areas as described in Table 6.2-82. These structures completely enclose the containment, forming a second barrier to the uncontro lled escape of radioactive s ources in the event of an accident. The inside diameter of the cylinder, constructed of reinforced concrete, is 158 feet. The vertical wall varies in thickness from 15 to 36 inches, and the dome is 15 inches. The inside of the dome is 5'-6" above the top of the containment structure. Design and performance data are listed in Table 6.2-82. The annular cylinder formed by the containment and the enclosure is shown on Figure 1.2-2, Figure 1.2-3, Figure 1.2-4, Figure 1.2-5 and Figure 1.2-6. Codes, standards and guides applied in the design of this structure are discussed in Subsection 3.8.4.2. b. Features in Support of the Containment Enclosure All piping penetrating the containment structure is sealed and anchored at the containment structure and at the containment enclosure so as not to be overstressed by thermal or seismic-induced motion. Electrical penetrations are sealed and anchored at the containment structure. The containment recirculation sump lines are enclosed in a sleeve which extends out to a vessel which encloses the first isolation valve outside the containment. This enclosure serves to contain any leakage from the sump line and first isolation valve.
The containment enclosure isolation features are discussed in Subsections 9.4.6.2 and 6.5.1. All personnel doors and equipment hatches in the containment enclosure are under administrative control. The doors are provided with position indicators and alarms having readout and alarm capability at the primary and secondary alarm stations. These doorways and hatches must be closed to insure a negative pressure in the containment enclosure.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 66  c. Containment Enclosure Emergency Cleanup System This system has two functions: (1) to pr oduce a negative pressure post accident in the annular, cylindrical volume between the containment and the containment enclosure, and (2) to collect any hazardous materials that might leak into these areas from the containment structure or equipment/systems located within the enclosure (ECCS) so that they may be disposed of in a controlled manner. Both these functions are performed by redundant filter trains, redundant fans, dampers and controls, and a common discharge ductwork system to the unit plant vent. Each exhaust fan has a sufficient capacity, with a clean and dirty filter train, to remove the in-leakage into the entire containment enclosure area calculated to occur at the design negative differential pressure, maintaining the required differential pressure. Subsection 6.2.3.3a discusses the performance of the fans. The presence of the containment enclosure and the use of the exhaust fans to
 
produce a slightly negative pressure betw een it and its external surroundings minimize the direct leakage from the containment structure to the environment.
The redundant filter trains contain moistu re separators, upstream HEPA filters, carbon adsorber bank and a downstream HEPA filter bank. The use of HEPA and charcoal filters in the exhaust from the containment enclosure reduces the discharges of radioactive iodine so that offsite doses following a LOCA are within the guidelines of 10 CFR 100. All components of the Containment Enclosure Emergency Cleanup System required to operate following an accident are Safety Class 2, seismic Category I. The system does not have provision for r ecirculation flow. Additional details are presented in Subsections 6.5.1 and 6.5.3. d. Containment Enclosure Bypass Leakage The maximum allowable leakage from the containment structure following an accident is 0.15 percent of the mass of its atmosphere per day. This would occur at maximum pressure. During the first 24 hours following a LOCA, the containment heat removal systems reduce the pressure, the driving force behind the leakage, to less than one-half the maximum value. As discussed in the preceding section, the direct leakage to the environs of radioactive contaminants from the containment is within the guidelines of 10 CFR 100.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 67  Although, as discussed in the preceding section, a containment enclosure emergency cleanup system has been provided to minimize leakage to the environs, a significant number of lines penetrate the containment and terminate in areas not treated by this cleanup system. Therefore, all leakage attributed to
 
penetrations and isolation valves, requiring Type B and Type C Test per 10 CFR 50, Appendix J, is conservatively assumed to bypass the cleanup system. The total allowable leakage for Type B and Type C Tests and for combined bypass leakage is discussed in Subsection 6.2.6.3. This is in accordance with
 
Appendix J acceptance criteria. 6.2.3.3 Design Evaluation The containment enclosure system design is evaluated from two viewpoints. Subsection 6.2.3.3a investigates the adequacy of the structure and associated equipment to achieve its functional goal, a negative pressure differential. Subsection 6.2.3.3b considers the vulnerability of the system to damage from a high-energy line rupture within the enclosure. a. Containment Enclosure Analyses One train of the Containment Enclosure Emergency Cleanup System is required
 
to be able to draw down the entire Containment Enclosure Area to a negative
 
differential pressure of 0.25 iwg. This differential pressure is required to be established between all areas that comprise the Containment Enclosure Area and their external surroundings. The areas that comprise the Containment Enclosure Area are listed in Table 6.2-82. This ne gative differential pressure has to be established within 8 minutes following a LOCA. The radiological dose analyses for a LOCA described in Section 15.6.5.4 begins to take cred it for filtration of radioactive contaminants that leak into the Containment Enclosure Areas at 8 minutes following an accident. Per Appendix 15B, no credit is taken for filtering out any of the radioactive contaminants released into the Containment Enclosure Areas for the first 8 minutes following an accident. The filter removal efficiencies are set at 0 during this time. Analysis has shown that one containment enclosure exhaust filter fan is capable of drawing down the entire containment enclosure area to the design negative differential pressure in less than 8 minut es after the initiation of a design basis LOCA. This analysis takes into account the engineered safe ty feature actuation system signal delay time, delay time for th e diesel generator to supply power in the event of a simultaneous loss of offsite power, and the time for the filter fan to come up to speed.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 68  In the event of a LOCA, the containmen t could experience an increase in volume on the order of 11,500 cubic feet because of thermal and pressure expansion. This is determined using the data listed in Table 6.2-82, and considers the swelling of the containment structure due to the design pressure of 52 psig. This would result in a decrease in the free volume of the Containment Enclosure Building of less than 1 percent with a similar corresponding rise in its pressure. The time for an exhaust filter fan to bring the Containment Enclosure Area back down to atmospheric pressure to compensate for the swelling of the containment structure is included in the analysis of the Containment Enclosure Area draw down time. The analysis also includes the time required for an exhaust filter fan to reduce the Containment Enclosure Area to the required negative differential pressure. A separate analysis has determined that it is necessary to establish a negative differential pressure of 0.685 iwg at the 21' -0" elevation of the Containment Enclosure Ventilation Area. This will ensure that a negative differential pressure
 
of 0.25 iwg exists at the top elevation of the Containment Enclosure for the full range of design basis outside ambient temperatures. The draw down analysis conservatively includes the time required for one filter exhaust fan to remove
 
enough air to draw down the entire Containment Enclosure Area to an internal pressure of negative 0.685 iwg. The analysis to verify this draw down time also takes into account a total calculated maximum in-leakage of 1025 cfm. This is the in-leakage through various air flow paths such as electrical, piping and duct penetrations, concrete structure, construction joints, doors, seal plates, metal partitions, ducts and floor drains. Air in-leakage was determined using data from the penetration sealant supplier, analytical calculation and experimental leakage data provided in "Conventional Buildings for Reactor Containment NAA-SR-10100 (1965)," issued by Atomics International, a Division of North American Aviation Incorporated. For conservatism, this in-leakage is calculated at the maximum
 
differential pressure of 0.685 iwg. The calculated maximum in-leakage also includes leakage from the primary containment at a rate of 0.2% of the primary containment volume for the first day fo llowing a design bases LOCA. This assumption is conservative because primary containment leakage is limited to 0.15 percent by weight. This analysis also assumes a low airflow of 1890 cfm from a single fan. This is the minimum value acceptable during system surveillance testing. When all of these factors are taken into account, one filter exhaust fan is still capable of achieving the design negative differential pressure in less than the required design basis draw down time of 8 minutes.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 69  Test or sampling connections in lines penetrating both the containment and containment enclosure are protected by either two isolation valves or by a locked-closed valve and one isolation valve so that no single failure can compromise the ability to achieve negative pressure by allowing a source of suction fluid to the exhaust fans other than the atmosphere of the containment enclosure. The analyses of the pressure/temperature response of the containment to a LOCA, performed for Subsection 6.2.1, have demonstrated that there is never any significant change in the temperature on the outside of the containment wall. Accordingly, the temperature in the containment enclosure is determined by the heat generated by the equipment present inside it and energy removal by the containment enclosure cooling units which function both during normal plant operations and in the event of a LOCA. The cooling coils have been sized to continuously maintain the temperatures, in the areas to be cooled, for normal, abnormal and accident conditions as discussed in Section 3.11(B). b. High Energy Line Rupture The main steam, feedwater, and steam ge nerator blowdown lines pass through the containment enclosure, but not directly. The enclosure bo undary terminates on the fluid portion of the penetration for these pipes on main steam tunnel sides. The residual heat removal line also passes through the containment enclosure, but is classified as a moderate energy line because of its short operational period.
Therefore, ruptures of these lines within the containment enclosure are not considered.
Failure of a high-energy line would result in pressurization of the containment enclosure due to the mass and energy release. The high-energy lines that penetrate the containment and traverse the Enclosure Buil ding without guard pipes are: 1. Sample lines from the pressurizer
: 2. Sample lines from reactor coolant loops
: 3. Excess letdown line
: 4. Letdown line. The sample lines (items 1 and 2 above) are normally isolated, a nd are only opened daily for short duration to purge the line and collect samples.
Since such lines are under the direct control of the operator taking samples, their failure could be immediately detected by lack of pressure at the sample sink, and isolation effected by the operator. Also, the sample line isolation valves will automatically close on a containment isolation "T" signal.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 70  The excess letdown line (item 3 above) is norma lly isolated. During operation, the liquid letdown is cooled to approximately 165F by primary component cooling water in the excess letdown heat exchanger. Accordingly, the failure of this line within the containment enclosure is less severe than for the normal letdown line. The flow rate in the excess letdown line, when in use, is approximately 2100 lb/min of subcooled water. Because of the mitigating factors associated with the rupture of other lines, a complete double-ended guillotine break of the 3-inch letdown line (item 4 above) becomes the design basis accident for the pressurization of the containment enclosure. This break is also the most severe small line rupture outside containment from the standpoint of radiological consequences, and is discussed in Subsection 15.6.2. As described there, this break leads to a flow of 140 gpm of water at 380F (having been cooled from 560F by the regenerative heat exchanger) for a period of less than 30 minutes, by which time the operator will have isolated the rupture. With the secondary containment at 104F, 10 percent relative humidity, and atmospheric pressure, the water flashes at the break into a steam-water mixture which would pressurize the containment enclosure to 1.5 psig within the first half hour. Whenever the flow is terminated, the pressure would begin to drop.
No credit is taken for ventilation and purging, and only the containment enclosure annulus free volume was available for pressurization for conservatism. The design pressure for the containment enclosure structure is 3 psig. 6.2.3.4 Tests and Inspections Preoperational testing of the containment enclosure and its associated exhaust system is discussed in Subsection 6.5.1.4 and Chapter 14. Periodic testing of the containment enclosure exhaust fans is discussed in Subsection 6.5.1.4. This periodic testing will also include a visual surveillance of containmen t enclosure penetration and other seals. 6.2.3.5 Instrumentation Requirements The system monitoring instrumentation and controls are provided in the main control room. Instrumentation associated with the Containment Enclosure Emergency Cleanup System is described in Subsection 6.5.1.5. The system is automatically initiated on a 'T' (containment isolation phase A) signal. The Engineered Safety Features Actuation System is described in detail in Section 7.3. The logic, controls and instrumentation of this engineered safety feature system function so that a single failure of any com ponent will not result in the loss of functional capability for the system. Further information on safety-related instrumentation is included in Chapter 7. Area radiation and airborne radioactivity monitoring instrumentation details are presented in Subsection 12.3.4.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 71  6.2.4 Containment Isolation System The Containment Isolation System is comprised of the valves, piping and actuators required to isolate the containment following a LOCA or steam line rupture. The systems establish and/or maintain isolation of the containment from the outside environment to prevent the release of fission products, and to ensure that the public is protected in accordance with 10 CFR 100 guidelines. Each piping penetration of the containment, except penetrations associated with engineered safety features equipment, is required to maintain or establish isolation of the containment under any loss-of-coolant accident or main steam pipe rupture that will initiate the containment isolation signals. 6.2.4.1 Design Bases The design bases for the containment isolation system are as follows: a. Minimize and limit the atmospheric release of radioactive materials in the event of a LOCA by isolating those lines penetrating the containment which are not required for the operation of the engineered safety feature systems. b. Avoid the reactivity effects that could result from excessive cooldown of the Reactor Coolant System in the event of a steam line break accident, and prevent the overpressurization of the containment during such an occurrence by isolating the containment as well as the steam generators, as may be required to fulfill these objectives. c. Provide double barrier protection for all lines that penetrate the containment, where a barrier may consist of a valve, a closed system, or a diaphragm depending upon its location and application. A closed system is one which satisfies all of the following requirements:
: 1. The system does not communicate with either the Reactor Coolant System or the containment atmosphere. 2. The system is protected against missiles and pipe whip.
: 3. The system is designated seismic Category I.
: 4. The system is classified Safety Class 2.
: 5. The system is designed to withstand temperatures at least equal to the containment design temperature. 6. The system is designed to withstand the external pressure from the containment structural acceptance test.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 72  7. The system is designed to withstand the environment and transient conditions resulting from either a loss-of-coolant accident or a main steam line break. Sealed-closed barriers which replace automatic isolation valves include blind flanges and locked-closed isolat ion valves. These barriers, which remain closed after a LOCA, will be managed through administrative controls. d. The containment isolation system design shall comply with the requirements of 10 CFR 50, Appendix A, General Design Criterion 54, and ot her applicable criteria as follows:
: 1. Lines that are part of the reacto r coolant pressure boundary and penetrate the containment, or are connected to the containment atmosphere, have their penetrations designated as Type I and are provided with valves as follows: (a) One locked-closed isolation valve inside and one locked-closed isolation valve outside containment; or (b) One automatic isolation valve inside and one locked-closed isolation valve outside containment; or (c) One locked-closed isolation valve inside and one automatic isolation valve outside containment; or (d) One automatic isolation valve inside and one automatic isolation valve outside containment. These provisions are in accordance with General Design Criteria 55 and
: 56. A simple check valve is considered an automatic isolation valve only on the inside of the containment on lines with flow coming into the containment. 2. Lines that penetrate the containment and are neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere have their penetrations designated as Type II. These lines are provided with valves as follows: (a) One automatic isolation valve outside containment; or (b) One locked-closed isolation valve outside containment; or (c) One isolation valve outside containment capable of remote manual operation.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 73  The second isolation barrier on thes e lines is provided by the closed system inside containment, as defined in Subsection 6.2.4.1c. These provisions are in accordance wi th General Design Criterion 57. A simple check valve is considered an automatic isolation valve only on the inside of the containment on the lines with flow coming into the containment. Therefore, a stop-check valve is provided on the emergency feedwater line outside the containment. 3. Connections/lines only for test purpose are provided with manual isolation valves that are opened only for testing. These valves are always closed when not testing. 4. Instrument lines which are connected to the RCS are designed so that a break within these boundarie s results in a relatively small flow that can be made up with the normal charging systems, which is in accordance with Regulatory Guide 1.11. The penetrations of these lines and lines connected to the containment atmosphere are designated as Type III, and are provided with isolation barriers in accordance with 10 CFR 50, Appendix A, General Design Criteria 55 and 56. See Subsection 6.2.4.2m(4) for exceptions to Regulatory Guide 1.141. The containment pressure, reactor coolant wide-range pressure, reactor vessel level and core differential pressure instrument lines are designed in accordance with the provisions of Regulatory Guides 1.141 and 1.151. For the containment pressure transmitters, isolation from the containment atmosphere is provided by a sealed bellows arrangement located immediately inside the containment wall, and is connected to the pressure transmitter outside containment by sealed, fluid-filled, tubing. Isolation outside containment is provided by the diaphragm in the pressure transmitter.
The RCS wide-range pressure, level and core differential pressure transmitters have a sealed bellows connected to the RCS and a second sealed bellows-type isolator outside the containment to provide two barriers in addition to the diaphragm in the transmitters. All components
 
are connected by sealed, fluid-filled, tubing. e. Relief valves are used as isolation valves and their relief set point is greater than 1.5 times the containment design pressure, in accordance with Standard Review Plan 6.2.4. f. The containment isolation systems are designed to remain functional following a safe shutdown earthquake.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 74  g. Containment isolation valve closure speeds and leak tightness will prevent radiological effects from exceeding the guidelines established by 10 CFR 100. h. Classification of essential and nonessential systems that penetrate containment is given in Table 6.2-83. Essential systems are defined as those piping systems penetrating containment which are necessary for mitigating the consequences of an accident; nonessential systems are classified as those piping systems penetrating containment which provide a uxiliary service functions for operation of the plant, and which are not required for mitigation of accidents. i. Although it is not specifically required to isolate containment in response to Station Blackout, the capability to establish containment integrity is provided (see Section 8.4.4.5). 6.2.4.2 System Design
: a. General Description Schematic representations of the isolation valving systems which define the fluid systems penetrating the containment wall, including instrument lines, are shown on Figure 6.2-91, sh.1, Figure 6.2-91, sh.2, Figure 6.2-91, sh.3, Figure 6.2-91, sh.4, Figure 6.2-91, sh.5, Figure 6.2-91, sh.6, Figure 6.2-91, sh.7, Figure 6.2-91, sh.8, Figure 6.2-91, sh.9, Figure 6.2-91, sh.10, Figure 6.2-91, sh.11. All valves and piping are fabricated of suitable stai nless and/or carbon steel to conform to the ASME Boiler and Pressure Vessel Code, Section III, Code Class NC. Special attention is given to materials to ensu re there are no radiol ytic or pyrolytic decomposition products to interfere with the safe operation of any engineered safety features. Section 6.1 includes further discussion of the materials.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 75  b. Component Description A summary of the fluid system lines penetrating containment and the valves and closed systems employed for containment isolation is presented in Table 6.2-83. Each process valve is described as to type, open or closed status during normal operation, shutdown and accident conditions, and closure/opening time. Information is also presented on valve preferential failure mode, position indication, isolation signal and location relative to containment. The test, vent and drain (TVD) valves associated with penetrations are not specified in Table 6.2-83, but are shown on Figure 6.2-91, sh.1, Figure 6.2-91, sh.2, Figure 6.2-91, sh.3, Figure 6.2-91, sh.4, Figure 6.2-91, sh.5, Figure 6.2-91, sh.6, Figure 6.2-91, sh.7, Figure 6.2-91, sh.8, Figure 6.2-91, sh.9, Figure 6.2-91, sh.10, Figure 6.2-91, sh.11. All TVD valves located between the containment isolation valves are identified. For the remaining TVD valves (located outside the containment isolation valves), only the test valves are schematically shown as arrows to identify all containment isolation valves, demonstrate the ability to perform the Type C test (if required), and yet provide a cl ear, unobstructed schematic representation of containmen t penetrations. Fo r further clarity, alphabetic suffixes were added to the individual lines of the multiple-line penetrations. These suffixes do not appear in other design documents. c. Valve Actuation Signals The design of the system providing the signals for containment isolation complies with the following general requirements: 1. The containment isolation signal overrides all signals for actuations of containment isolation valves for nonessential systems. 2. Phased isolation is used. With phased isolation, all systems except engineered safety features and non-engineered safety features-related systems are automatically isolated. Only those engineered safety feature-related systems that can be justified to remain operational shall not be automatically isolated during the initial phase. 3. Diverse parameters are used wherever possible for developing isolation signals. 4. Concurrent containment isolation occurs coincident with initiation of emergency core cooling. 5. All valves that receive a containment isolation signal cannot be reopened until the isolation signal is reset and manual action is taken to reopen the
 
valve. The controls are separated so that only one valve, or group of valves associated with a penetration, open for each manual action.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 76  Automatically tripped isolation valves are actuated to the cl osed position by one of two separate containment isolation signals. The first of these signals ("T" Signal) is derived in conjunction with automatic safety injection actuation on Hi-1 containment pressure, and trips the majority of the automatic isolation valves. The Hi-1 setpoint is the lowest practical and includes margin for containment pressurization, instrument error and operating margin. These are valves in the nonessential process lines wh ich do not increase the potential for damage to in-containment equipment when isolated. This is defined as "phase A" isolati on, and the valves are designated by the letter "T" in the isolation diagrams of Figure 6.2-91. The second, or "phase B," containment isolation signal ("P" Signal) is derived from Hi-Hi containment pressure and/or actuation of the Containment Spray System, and trips the automatic isolation valves in the other pr ocess lines (which do not include safety injection lines) penetrating the containment. These isolation valves are designated by the letter "P" in the isolation diagrams.
Containment air purge (CAP) and containment online purge (COP) system lines which provide an open path from the c ontainment environs are equipped with radiation monitors that are capable of isolating these lines upon receipt of a high radiation signal, in addition to automatic safety injection actuation, manual containment spray actuation and manual pha se "A" isolation signals. Further discussion of containment isolation signals is found in Section 7.3.  (Refer to Figure 7.2-8.) d. Valve Closure Time The objective in establishing valve closure time is to limit the release of radioactivity from the containment to as-low-as-is-reasonably-achievable. Consideration is given to the fluid system requirements (e.g., water hammer) in determining the valve closure time, the effect of closure time on valve reliability, as well as the containment isolation requirements. These considerations have been addressed in the design of the containment isolation system, within the context and requirements of the guidelines and applicable criteria presented in Subsection 6.2.4.1, Design Bases. Isolation valve closure times for the Containment Isolation System are presented in Table 6.2-83. The valves listed there reflect the maximum time required to isolate a system so that radioactive release to the environs during a design basis accident is within limits in 10 CFR 100. Refer to Subsection 9.4.5 for discussion of containment online purge line isolation.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 77  e. Operability of Valves Inside Containment Isolation valves located inside containmen t are subject to the high pressure, high temperature, steam-laden atmosphere resulting from an accident. Operability of these valves in the accident environment is ensured by proper design, construction and installation, as reflected by the following considerations: 1. All components in the valve installation, including valve bodies, trim and moving parts, actuators, instrument air control and power wiring are constructed of materials sufficiently temperature and humidity resistant to be unaffected by the accident environment. Special attention is given to electrical insulation, air operator diaphragms and stem packing material. Section 3.11 discusses the qualification of this equipment for operation in the containment atmosphere during an accident condition. 2. In addition to normal pressures, the valves are designed to withstand maximum pressure differentials in the reverse direction imposed by the accident conditions. Additionally, a review was performed to ensure that the containment isolation valves inside the containment were not overpressurized due to trapped fluid in adjacent piping in the post-LOCA environment. 3. The containment structure online purg e subsystem is designed to prevent debris from entering the exhaust and s upply lines to ensure the operability of the isolation valves. This is accomplished by debris screens installed in the ends of the lines. Each debris screen consists of heavy-bar stainless steel grating, banded and welded to the exhaust and supply ends of the lines. Both the exhaust and inlet piping have two 90 bends and a minimum of 14 feet. This design greatly reduces the possibility of direct impingement of debris on the valves. The pipe, screen s and supports are seismic Category I. The screens will be capable of withstanding the differential pressure resulting from a LOCA up to the point of containment isolation. Operability of valves and their operators within containment atmosphere is addressed by qualifying this equipment to IEEE Standard 382-1972, Guide for Type Test of Class I Electric Valve Operators for Nuclear Power Generating Stations, and NRC Regulatory Guide 1.73, Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants. Updated FSAR Section 3.9 provides additional information on valve operability.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 78  Containment isolation valve operators ha ve been provided to ensure adequate reliability for the operation of the valves. Lines penetrating the containment that serve the engineered safety features as well as their associated support systems are provided with motor operators. To verify that motor operators have sufficient torque to operate the valves, the valves are tested by opening and closing under full system pressure. Operability duri ng and after a LOCA is thus assured. Isolation valves outside containment, wh ich are not required to function during a
 
post-LOCA condition, are provided with air operators and spring return to the fail-safe position. All motor-operated valves have manual handw heel operators and can be closed or opened manually on loss of primary power. f. Location of Valves Outside Containment All isolation valves outside the containment are located as close to the containment as practical. All exterior isolation valves are located within 15 feet of containment, except for the main steam, feedwater and combustible gas control isolation valves. The main steam isolation valves are located within 75 feet of penetrations X-1 (westside) and X-2 (eastside), and 85 feet away from penetrations X-4 (westside) and X-3 (eastside). The feedwater isolation valves are located within 30 feet of the containment. The combustible gas control isolation valves are located within 50 feet of containment. The distances of the main steam isolation valves from the containment mentioned above are required to accommodate the main steam safety valves and power-operated relief valves. The distance of the feedwater and combustible gas control isolation valves from the containment accommodates the required physical piping arrangement and provides accessibility to the valves for maintenance. Two main steam and two feedwater lines emerge from the containment structure and enclosure on each side of the building, 180 from each other. On each side, the four lines turn and run parallel toward the Turbine-Generator Building. These lines and their isolation valves are enclosed in a seismic Category I structure and are shielded from tornado-generated missiles. The structure is designed with sufficient vent openings to the external atmosphere so that any possible pipe failure inside will not cause failure of the building by overpressurization.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 79  g. Actuation and Control Equipment Containment isolation valves are provided with actuation and control equipment appropriate to the valve type. For example, globe and diaphragm valves are generally fitted with air diaphragm operators which will fail in the safe position on loss of operating air. Gate valves are generally fitted with motor operators and are powered from emergency buses. On lo ss of offsite power, the power source is automatically switched to the diesel generators which feed the emergency buses. Motor-operated valves fail in the as-is position. No manual operation is required for immediate isolation. If actuating power is lost, the automatic air-operated isolation valves assume the position (open or closed) that provides gr eater safety. Motor-operated isolation valves fail "as is."  Manual control switches provide a secondary mode of actuation for the automatic isolation valves. The positions of the automatic isolation valves and remote manual valves are displayed in the main control room at both their control switch location and as part of the Post-Accident Monitoring System except for the main steam isolation valve (MSIV) bypass valves (MS-V-204, MS-V-205, MS-V-206, and MS-V-207). The circuit breakers for these valves are administratively controlled locked open (off) with the valves locked closed during power operation except for surveillance testing and to equalize pressure across the MSIVs before they are opened. With the circuit breakers open, the indicating lights at the control switch are off but the Post-Accident Monitoring System lights are still operable.
: h. Seismic Design Protection for containment isolation systems and components against loss of function due to seismic event forces is provided. Containment is olation valves and their operators are designated as seismic Category I. Containment isolation provisions are capable of maintaining th e isolation function during and after the Safe Shutdown Earthquake (SSE). The valves are capable of being realigned after the Design Basis Event and to withstand seismic aftershocks following the SSE. The containment isolation valves, their operators, and supports are designed to assure that they are capable of with standing the Safe Shutdown Earthquake as recommended by Regulatory Guide 1.29. To assure their adequacy in this respect:
: 1. Valves are located in a manner to reduce their accelerations. Valves suspended on piping spans are designed for the loads to which the span would be subjected. Valves are mounted in the position recommended by the manufacturer.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 80  2. Valve yokes are designed for adequacy and strengthened as required for the response of the valve operator to seismic loads. 3. Where valves are required to operate during seismic loading, the operator forces are factored in the design to assure that system function is preserved. 4. Control wires and piping to the valve operators are designed to assure that the flexure of the line does not endanger the control system. Appendages to the valve, such as position indicators and operators, are designed for structural adequacy. Additional information regarding the seismic design of systems and components is presented in Sections 3.7 and 3.9. i. Missile Protection All containment isolation valves, actuators, and controls are located so as to be
 
protected against accident-generated missiles including those caused by a loss-of-coolant accident. The isolation valve gallery outside the containment is partitioned into four areas by a vertical wall and platform. In addition to providing missile protection, this arrangement provides for separation of
 
redundant safety-related lines and serves as radiation shie lding and a work platform during maintenance periods. Inside the containment, missile protection is provided by a missile shield wall and accumulator tanks. No extraneous equipment is placed in the valve galleries. Only valves so protected are considered to qualify as containment isolation valves. Details regarding the probability of missile generation and design features to prevent the formation of missiles are given in Section 3.5. j. Potential Leakages All nonautomatic isolation valves which are left open to perform post-accident functions, except MS-V393 and MS-V394, ha ve potential leak paths into the containment enclosure only. The desi gn of the enclosure recirculation and filtration system is such that all possible leaks have been considered in determining the capacity of the system to maintain the potential radioactive discharge within the limits of 10 CFR 100 subsequent to a LOCA. Provisions are made to detect and minimize the leakage from the engineered safety features and auxiliary systems located outside containment. Leakage
 
monitors are installed to provide the plant staff with the current knowledge of the system leakage rates. Multiple monitors for noble gas effluents are installed with an extended range designed to function during accident as well as normal
 
operating conditions.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 81  The potential sources of bypass leakage past the containment en closure are listed in Table 6.2-83. Details of leakage acceptance limits and testing are given in Subsection 6.2.6. k. Physical Separation Physical separation of containment isolation systems and components is provided where required in accordance with design guidance contained in "Physical Protection for Systems and Comp onents Important to Safety, "ANS-58.3/N182 - 1977, "Standard Crit eria for Separation of Class 1E Equipment and Circuits," Trial-Use, IEEE-384-1974, and in NRC Regulatory Guide 1.75, "Physical Independence of Electric Systems." l. Fire Protection Fire protection for containment isolation provisions against loss of function from fire is provided in accordance with "Generic Requirements for Light Water Nuclear Power Plant Fire Protection," ANSI/ANS-59.4-1979, "Standard for Fire
 
Protection for Nuclear Power Plants, NFPA report 803-1978, and "Seabrook Station Fire Protection System Evaluation and Comparison to APCSB 9.5-1 Appendix A." Additional information on fire protecti on is presented in Subsection 9.5.1. m. Clarifications All characteristics of the containment isolation system design conform to the criteria, regulatory guidelines, and other standards mentioned above, with the following clarification: 1. Only one isolation valve outside the reactor containment is provided on each of the two lines between the containment building recirculation sump and the suction of the residual heat removal and containment spray pumps. This reduces the probability of the valve failing to open when called upon to function. The pipe between the containment sump and the isolation valve is jacketed, and the isolation valve is enclosed in protective chambers so that failure of the pipe or the valve body will not result in release to the environment of radioactive fluid or gases. 2. Closed system (as defined in Subsection 6.2.4.1c) is used as a second isolation barrier for the containment penetrations of the following systems: residual heat removal, safety injection, chemical volume and control (SI portion only), containment building spray, main steam, feedwater, steam generator blowdown, and primary component cooling water (thermal barrier portion only).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 82  (a) The containment penetrations for the main steam, feedwater, steam generator blowdown and primary component cooling water (thermal barrier portion only) are designated as Type II penetrations. The definition of the Type II penetrations is provided in Subsection 6.2.4.1d.2. (b) The remaining penetrations which use a closed system as a second isolation barrier are designated as Type E. The following provides the details for the individual penetrations: (1) The residual heat removal system utilizes two normally closed, pressure-interlocked valves in series for each suction line inside the containment. Further details on interlocks are discussed in Chapter 7. This arrangement
 
decreases the probability of release to the environment of radioactive fluid or gases by eliminating a potential leakage
 
point, and retains redundant isolation capability should a residual heat removal system pipe rupture occur outside the containment. The valve which is located closer to the RCS is not considered a containment isolation valve. The second valve defines the containment isolation barrier inside containment and is considered to be sealed closed. This containment isolation arrangement is as described in ANS 56.2/ANSI N271-1976 and endorsed by Regulatory
 
Guide 1.141. (2) The arrangement of CBS/ECCS suction penetrations is described in Item 1 above. (3) The discharge of all engineering safety features systems utilizes the check valves located inside containment as automatic isolation barriers. 3. Each supply line to the hydrogen anal yzer portion of the combustible gas control system is provided with two closed manual isolation valves outside containment. The isolation valve inside containment is locked open to allow for post-LOCA operation of the an alyzers. The first isolation valve outside containment is locked closed except when the hydrogen analyzers are in operation. An additional isolation barrier is provided by a closed system outside containment.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 83  The locked closed containment isolat ion valves are not required to be Type C tested as per Table 6.2-83. These lines form a closed, seismic Category I system outside containment. The integrity of the closed system is maintained by performing periodic surveillance testing as described in
 
Subsections 6.2.5.1.g and 6.2.6. 4. The sealed, fluid-filled, instrument systems which penetrate the containment boundary are designed to ANSI B31.1, Seismic Category I, not ASME Section III. The penetration is designed to ASME Section III, Code Class 2. See Subsection 7.1.2.12 for a discussion of compliance to Regulatory Guide 1.151. These sealed sensing lines, shown on Figure 6.2-91, sh.9, have no isolation valves, but because of the isolation barriers provided as part of the sealed, fluid-filled, system, a postulated severing of the line during either normal operation or accident conditions will not result in any radioactive release from the containment. This containment isolation arrangement is as described in ANS 56.2/ANSI N271-1976 and endorsed by Regulatory Guide 1.141. See
 
Subsection 6.2.4.1d for additional discussion. 5. Not used. 6. Some of the pressure indicators and transmitters on the main steam line outside the containment (such as PT 3001 to 3004, 3173, 3174, 3178, 3179, and PI 3051 to 3054) are not qualif ied since they are not required for engineered safety systems, are on secondary (noncontaminated fluid) loops, and are on small lines whose breaks will not result in any significant steam leaks. This is consistent with the requirements of Regulatory Guide 1.11. 7. The CAP System penetrations each have a blind flange using a resilient double o-ring design installed on the penetration outside containment during Modes 1, 2, 3, and 4. The design includes a test groove between the o-rings for Type "B" testing in accordance with 10 CFR 50, Appendix J. The penetrations terminate in the containment enclosure during plant
 
Modes 1, 2, 3, and 4, thereby allowing any leakage to be collected and processed by the Containment Enclosure Ventilation System. 6.2.4.3 Design Evaluation The Containment Isolation System has been designed in accordance with Regulatory Guide 1.11, Design Criteria 54, 55, 56, 57. Accordingly, it has been specifically designed to: a. Isolate lines penetrating the containment, which are not required for the operation of the engineered safety feature systems, in the event of a LOCA; S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 84  b. Isolate lines penetrating the containment, which are not required for the operation of the engineered safety feature systems, in the event of a main steam line break; c. Shut the isolation valves in the CAP and COP systems upon detection of high radioactivity. Isolation of the containment in the event of a LOCA or a steam line rupture is initiated upon receipt of isolation signals as discussed in Subsection 6.2.4.2c and Section 7.3. 6.2.4.4 Tests and Inspections During preoperational testing, tests are performed on the containment isolation system to verify valve response to containment isolation signals, and to determine valve leakage rates. Chapter 14 contains a further discussi on of the preoperational tests performed on this system. Periodic testing of the containment isolation system is performed in accordance with Technical Specification requirements. Subsection 3.9.6.2 describes the in-service inspection program for these valves. 6.2.5 Combustible Gas Control in Containment Following a loss-of-coolant accident (LOCA) hydrogen gas may be generated inside the containment by reactions such as zirconium fuel cladding with reactor coolant, corrosion of metals of construction by solutions used for em ergency core cooling or containment spray, and by radiolysis of aqueous solution in the core and sump. To ensure that the containment integrity is not compromised by burning or explosion of this hydrogen, a combustible gas control system (CGCS) has been provided to mix the containment atmosphere, monitor combustible gas concentrations within the containment regions, and reduce the combustible gas concentrations within the containment by recombination of the free hydrogen with the oxygen in the containment air. 6.2.5.1 Design Bases
: a. The CGCS is capable of continuously monitoring the hydrogen concentration in the containment during and after a desi gn basis LOCA. The CGCS is also capable of monitoring hydrogen concentration in containment during a beyond-design-basis accident for accident management, including emergency planning. The operator in the main control room is alerted of the need to activate systems to reduce combustible gas concentrations, when required, by an alarm from the operating hydrogen monitor. b. The containment mixing portion of the CGCS is designed to uniformly mix the containment atmosphere for as long as is necessary during and following an accident which generates hydrogen. Mixing of the containment atmosphere prevents high concentrations of hydrogen from accumulating locally.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 85  c. The CGCS, containment mixing subsystem, meets the redundancy and power source requirements for engineered safety features. It is de signed to withstand a single active mechanical component failure and a passive elec trical failure; no single failure will incapacitate the containment mixing system. d. Components of the CGCS located within the containment ar e protected against damage from internally generated missiles or jet impingement in the post-LOCA environment. Moreover, such component s have been subjected to qualification tests to demonstrate their capability to remain operable in the LOCA environment for as long as may be required. e. The CGCS, including its foundations and supports, is designed to withstand the effects of an SSE without loss of function. f. The CGCS design will permit periodic in-service inspection, operability testing and leak rate testing of the system or its components. g. The unit is provided with its own permanently installed combustible gas control equipment. h. In the absence of containment isolat ion "T" signal, the CGCS is capable of purging the containment in the event that more than a single failure of active elements of the system occurs. i. The CGCS is classified as a seismic Category I system, and is designed, fabricated, erected and tested to Safety Class 2 quality standards, except for the containment structure recirculating filter system which is Safety Class 3. The hydrogen analyzers, which are normally isolated from the containment, are comprised of Class 1E, seismic Category I components. j. The design of the combustible gas control system is in accordance with NRC Regulatory Guides 1.7, 1.22, 1.26 and 1.29, General Design Criteria 5, 41, 42, 43, and 50 and SECY 03-0127, "Final Rulemaking Risk-Informed 10 CFR 50.44, "Combustible Gas Control in Containment."
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 86  6.2.5.2 System Design The CGCS consists of subsystems which monito r the combustible gas concentrations in the containment, and which possess the capability for maintaining a mixed containment atmosphere to ensure that hydrogen concentrations remain below flammable levels following a loss-of-coolant accident. The overall system is depicted on Figure 6.2-
: 92. A portion of the CGCS, the fans installed to mix the containment atmosphere, are not shown on this figure, but are discussed in Subsection 6.2.5.2b. a. Hydrogen Monitoring Subsystem The containment atmosphere is mon itored by two completely independent hydrogen sampling and analysis systems whic h are started after an accident. The hydrogen analyzer design parameters are summarized in Table 6.2-84.
The suction intakes are located at elevation 183'-6", at opposite sides of the containment dome, terminating in 90 elbows pointing downward to minimize entry of spray into the sample lines. To prevent condensation of moisture in the suction lines to the analyzers, the lines are heat-traced and maintained at a temperature of 278 or greater, after penetrating the containment wall. This ensures that the gas sample is maintained above the steam saturation temperatures postulated to occur during design basis accidents (ref. Figure 3.11-1). The analyzers are located outside containment at the 22'-0" level of the Main Steam and Feedwater Pipe Chase Building, and take suction through a heavy-walled tube of approximately 1/4" bore, with lengths varying from 150 feet to 300 feet inside the containment. To maintain the requirement for separation of the two redundant monitoring subsystems, each analyzer is powered from a different electrical train, and has its own distinct discharge piping to return the sampled gas to the containment.
 
The hydrogen analyzers are normally isolated from the containment-related piping by Safety Class 2 manual valves at the analyzers. Since the analyzers are located outside the containment, periodic inspection and testing are facilitated.
The less strenuous conditions ex pected in the Main Steam and Feedwater Pipe Chase Building, rather than the post-accident environment inside the containment, are all the analyzers need be qualified for.
The analyzers work on the principle of thermal conductivity of hydrogen at various concentrations. Grab sampling provision is available in the vicinity of the hydrogen analyzers on a per train basis.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 87  b. Containment Atmosphere Mixing Mixing of the containment atmosphere to prevent localized buildup of hydrogen concentrations is provided by the Containment Spray System, described in Subsection 6.2.2. This aspect of mixing is discussed further in Subsection 6.2.5.3. Subcompartments not exposed to the sp rays have been designed so that no pocketing of hydrogen in their upper levels is possible. The top of the entrance to these subcompartments is always flush with their ceiling, and the elevation of the ceiling immediately outside of them is never lower than the entrance way.
Compartments located beneath the operating floor connect with the free containment volume by their doorways and via piping penetrations through which they also communicate with each other.
Hydrogen vent ports have also been included in the ceiling of such subcompartments at the end opposite from their entrances to prevent any dead-ending wh ich could inhibit natural dispersion by convection and diffusion. The associated Containment Structure Recirculation Filter System (CSRFS) is described in Subsection 9.4.5. Following a LOCA, the two 4000 scfm fans are started by an engineered safety feature actuation signal, as discussed in Chapter 7. The fans take suction from the apex of the dome and discharge below the operating floor. The fans, the dampers, and the suction and discharge ductwork are capable of withstanding the physical, chemical and radiological environment to which they will be subjected in the event of a LOCA. These elements of the
 
CSRFS are redundant, separate, and built to Safety Class 2 and seismic Category I standards.
: c. Hydrogen Recombiners One means of combustible gas control in the containment is through the use of electric hydrogen recombiners. The c odes, standards and regulatory guides employed in their design are listed in Table 6.2-85. The unit has a pair of recombiners, located at the perimeter of the operating floor inside the containment. Thus, there is no need to protect personnel from radiation in the vicinity of an operating recombiner. The separateness extends beyond the physical distinction to the independence of instrumentation, control circuits and power supply so that no single failure can impede the operation of more than one recombiner. Table 6.2-86 summarizes the recombiner design parameters.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 88  The recombiner, Figure 6.2-94, consists of an inlet preheater section, a heater-recombiner section, and a discharge mixing chamber. The inlet preheater section is a thermally insulated vertical metal duct positioned around a central heater section to take advantage of heat losses from the heater section. The heater section consists of four vertically stacked assemblies of electric heaters, each assembly containing individual heating elements. An outer enclosure provides protection against containment spray water. The overall assembly is mounted on structural steel framing which provides a substantial foundation free of normal operating vibration. The recombiner is fabricated of a corrosion-resistant high temperature material, except for the base which is of carbon steel. The heaters are commercial-type electric resistant heat ers sheathed with Incoloy-800, an excellent corrosion-resistant material for this purpose. These heaters are operated at significantly lower power densities than in commercial service. Air is first drawn into the preheater section by natural convection, where it is warmed. It then passes through an orifice, plate, and enters the electric heater section where it is heated to approximately 1150 to 1400 F, thus causing recombination between the oxygen a nd hydrogen. The efficiency of recombination is 99.9 percent, minimum, at all hydrogen concentrations between 2 and 4 percent. Tests have verified that the recombina tion is not a catalytic surface effect associated with the heaters, but occurs as a result of the increased temperature of the process gases. Since the phenomenon is not a catalytic effect, saturation of the unit does not occur. Operation of the recombiner is manually controlled from a panel located in the main control room (the recombiner, power supply panel and control panel are shown schematically in Figure 6.2-95). The recombiner power panel contains an isolation transformer plus an SCR controller to regulate power into the recombiner. This equipment is mounted outside the containment and thus, is not exposed to the post-LOCA environment. To control the recombination process, the correct power input for bringing the recombiner above the threshold temperature for recombination is set on the controller. The correct power required for recombination depends upon containment atmosphere conditions, and is determined when recombiner operation is required. For equipment tests and periodic checkouts, a thermocouple readout instrument is also provided in the control panel for monitoring temperatures in the recombiner. Reference 22 further describes these recombiners and their qualification testing.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 89  d. Backup Purge System The capability for purging of the containment at a controlled rate is also provided. Purging is accomplished by replacing the purged gas with clean compressed air. The purge system, Figure 6.2-92, consists of a flow meter, a throttle valve and associated piping. After the flow meter, the piping terminates adjacent to the inlet screens of the emergency cleanup filters located within the containment enclosure. From the filters, there is a direct path to the unit plant vent. The purge line separates close to the emergency filters, and each of the two lines ends in a pipe reducer bell close to the inlet screen of one of the filter units. Each of these two lines includes an isolation valve, so that th e purge gas flow can be directed to whichever filter unit is operating. The purge system is aligned and the purge flow is initiated manually. The Containment Enclosure Emergency Cleanup System is discussed in Subsection 6.2.3; the filter system's design requirements are detailed
 
in Subsection 6.5.1. Components and ductwork associated with the filters are classified as Safety Class 3 and seismic Category I. All piping inside the containment and the containment penetration connections associated with the purge system are duplicated to provide independent and redundant capability and to prevent a single failure from stopping the containm ent purge/flow vent. These are Safety Class 2. External to the containment, a single line runs from the penetration area to the containment enclosure emergency filters. 6.2.5.3 Design Evaluation A beyond design bases accident could generate hydroge
: n. Turbulence created by the rupture of a coolant line serves to thoroughly mix the atmosphere of the containment in the early phases of the post-LOCA period. The release to the containment of the hydrogen produced by radiolysis of cooling water in the core would be through th e break, and would be associated with the local mixing caused by the break flow. The hydrogen generated by corrosion of metals by spray water would, in turn, be mixed with the free containment volume by that spray. In addition, the aluminum and zinc in the containment are wi dely distributed. The inventories of these corrodible metals are shown in Table 6.2-87 and Table 6.2-88, respectively.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 90  In the long term, after the temperature in the containment has been reduced by the containment heat removal system, and the rate of hydrogen production by corrosion is diminished, the principal source of combustible gas buildup is the radiolysis of the water injected into the reactor vessel and sprayed into the containm ent. Hydrogen originating from core radiolysis is deemed to be distributed in the region of the break by the dispersion caused by the mass flow from it. The radiolysis of sump water by fission products released from the core produ ces hydrogen at a lower rate than radiolysis in the core a day or two after the accident. Hydrogen generated by sump water radiolysis is distributed by two means, the second of which serves to insure that combustible gas concentrations, in general, are homogeneous within the containment, including recombiner or sampling locations: a. The minimum post-LOCA spray rate is 3000 gpm. With a maximum possible sump water volume of 475,000 gallons the entire amount of sump water is sprayed through the containment atmosphere approximately every three hours. 
 
The fission products are uniformly distributed in the sump water, being mixed by turbulence caused by injection, break flow and spraying, but the means of release to the containment atmosphere must be across the water-ga s interface. Hydrogen atoms, not bubbles, are formed by radiolysis. They diffuse through the sump water forming molecules, which themselves diffuse and escape to the atmosphere at the surface once the relatively low solubility of hydrogen in water is passed, saturating the sump water. Ultimately , the release rate of hydrogen to the containment free space equals the production rate from radiolysis of sump water. 
 
Using the data concerning spray droplets, fall height, etc., detailed in Subsection 6.5.2, it is calculated that the surface-to-mass ratio for the water in the sprays exceeds that of the water in the sump by a factor of more than 5000. Thus, virtually all of the hydrogen generated by radiolysis cause d by fission products outside the core is released by the water while it is being sprayed. The spray
 
pattern is designed so that, even if only one train of sprays is available, the volume of the containment above the operating floor is uniformly exposed to the
 
sprays. b. To insure that the atmosphere of the containment is mixed, so that hydrogen concentrations are virtually identical, the fans in the containment structure recirculation filter system discussed above recirculate the gases in the containment more rapidly than twice a day, even with only one fan available. The discharge of the fan flow below the operating floor promotes a flow from those regions upward to the suction of the fans at the apex of the dome, with the containment atmosphere passing through the sprayed region. Thus, no
 
stratification of hydrogen is possible. The system meets the requirements of Regulatory Guide 1.7.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 91  The pertinent data for evaluation of the control aspect of the combustible gas control system is summarized in Table 6.2-89. Although Subsection 15.6.5.3c.1 indicates that only < 1.0 percent of the zirconium would react, for conservatism, 5 percent is assumed to participate in a reaction with th e coolant to produce hydrogen coincident with the accident. To prevent the concentration of hydrogen from reaching 4 volume percent in the containment, the recombiners would be turned on at any time before, or when, the concentration reached 3.5 volume percent. If only one recombiner is turned on 278 hours after a LOCA when the 3.5 volume percent setpoint is reached, the hydrogen concentration wi ll begin to drop immediately. The combustible gas concentration in the containment as a function of time is shown on Figure 6.2-96, both with and without the operation of a single recombiner, at an efficiency of 99.9 percent. The backup purge system, as described in Subsection 6.2.5.2d, could be operated at rates up to 1000 scfm. This far exceeds that of a single recombiner which processes only 100 scfm. Although actuation of the recombiners could be delayed, the 3.5 volume percent is selected to have marg in for detection of the failure of both recombiners to function. In this event, the backup purge system would be started at the 3.8 volume percent level. This would leave approximately 41 hours to start purging. Figure 6.2-97 shows the containment hydrogen concentration with
 
neither recombiner nor purge, and with purging at a rate of only 3 percent of the containment volume per day starting at the 3.8 volume percent mark, 319 hours after the LOCA. The actual purge rate used will be based on an analysis of the containment atmosphere following a LOCA. Projected offsite doses resulting from containment purging, if required, are described in Subsection 15.6.5.4. 6.2.5.4 Testing and Inspection
: a. Hydrogen Analyzer The hydrogen analyzer is shop-tested using a gas mixture closely simulating the containment post-LOCA atmosphere expected at the time the units would be placed into service, with the temperature, pressure, humidity and hydrogen concentration conditions approximated.
During preoperational testing, the hydrogen analyzers are calibrated and checked for proper operation. System integrity will be maintained by performing periodic surveillance testing as described in Subsection 6.2.6. Periodic calibration tests will be performed in accordance with Technical Requirements.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 92  b. Hydrogen Recombiner The electric hydrogen recombiners have undergone extensive testing in the Westinghouse development program. These tests encompassed the initial analytical studies, laborator y proof-of-principal tests, and full-scale prototype testing. The full-scale protot ype tests included the effects of: 1. Varying hydrogen concentrations 2. Alkaline spray atmosphere
: 3. Steam
: 4. Convection currents
: 5. Seismic activity
 
A detailed discussion of these te sts is given in Reference 22.
During preoperational testing, the functional operability of the recombiner and its control will be demonstrated. Periodic testing of the recombiners will be performed in accordance with plant procedures. c. Backup Purge System The containment purge function of the CGCS will be demonstrated during preoperational testing. Initial and periodic tests of the containment enclosure emergency exhaust filters is described in Subsection 6.5.1. d. Containment Structure Recirculation Filter System The operability of containment atmospheric mixing fans and dampers will be
 
verified during preoperational testing. Pe riodic tests of the fans and dampers will be conducted in accordance with Technical Specification requirements. 6.2.5.5 Instrumentation Requirements With the occurrence of a LOCA, the ESF actuation signal will start the Containment
 
Recirculation Fan System. Two independent hydrogen analyzers which monitor containment hydrogen concentrations after an accident, are located outside the containment. Off-normal conditions, such as low temperature, low sample flow and pressure, and cell failure are alarmed at the control room on a system level as a common alarm, and individually indicated at the local panel. The output of either ch annel of the hydrogen analyzers is available to the operator, both locally and at the MCB. When either channel indicates a hydrogen concentration at or prior to 0.5 volume percent below the limiting hydrogen concentration, this fact is alarmed at the main control board (MCB). The operator would then start the recombiners, if he had not done so already.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 93  The recombiners are thermal electric types. Temp erature sensors located at the recombiner input to the MCB temperature indication to maintain the recombiner temperature at an optimum value for efficient recombination of H2 and O2 into water. The power input to the recombiner is manually set from the MCB. Recombiner temp erature and power input are indicated at the MCB. The temperature indication is used for equipment tests and periodic checkouts and is not required for the safety-related func tion of the hydrogen recombiners. In the event that neither recombiner starts, the hydrogen concentration in the containment would continue to rise. If it has rise n significantly above the first alarm point but is still below the limiting hydrogen concentration, this fact would be alarmed at the MCB as a signal to the operator to initiate the purging of the containment. The Containment Purge System is normally closed, and is isolated from the containment by four valves:  CGC-V14 and CGC-V28 inside the containment are remotely controlled, motor-operated valves that close on a "T" signal; valves CGC-V15 and CGC-V36 located outside the containment are manually operated and normally locked closed. Initiating purge flow is a manually controlled operation, with a combination of both local and remote control from the MCB. Valve status for the remotely operated valves CGC-V14 and CGC-V28 is provided near the associated control switches. Pressure and fl ow instruments are provided in the purge line. Purge flow indication is available at the MCB. Air for the purge system is supplied from the Compressed Air System. 6.2.6 Containment Leak Rate Testing The reactor containment structure, the containment penetrations and the containment isolation barriers are designed to permit periodic Type A integrated leakage rate testing. The reactor containment and its leakage limiting barriers are also designed to permit periodic inspection of important areas such as penetrations. Penetrations with resilient seals or expansion bellows are designed to permit periodic leakage testing at pressures up to the containment design pressure. Piping systems penetrating the reactor containment are provided with the capability of redundant isolation, as dictated by their importance to safety functions. These systems are designed so that their isolation capabilities can be periodically tested for operability and leakage to ensure compliance with the established leakage rate lim its. The foregoing are intended to be in full compliance with General Design Criteria 52, 53 and 54 of 10 CFR 50, Appendix A. All portions of the above systems (RHR, CBS, SI and CS) are located within the containment enclosure boundary except piping associated with the injection phase of ECCS and a minor amount of charging pump piping used during the recirculation mode. Any leakage from these systems following a LOCA is therefore filtered by the containment enclosure emergency exhaust filters prior to release to the environment.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 94  The piping that lies outside the containment enclosure boundaries includes the pump suction lines from the refueling water storage tank to the RHR, SI and CBS pumps (lines 1201-1-151-14" and 1202-1-151-14", see Figure 6.2-74) and the centrifugal charging pumps (lines 1205-1-151-8" and 1206-1-151-8", including valves LCV-112D and LCV-112E). These lines are used only during the in jection phase of post-accident operation. They are isolated within the containment enclosure, will not be contaminated during recirculation, and will not present a release path. A portion of line CS-374-1-2501-4" is run outside the containment
 
enclosure. This line is used during the recirculation phase of post-accident operation; however, there are no valves or equipment in this pipe segment. The leakage potential for this line is therefore small. The RHR pumps are periodically tested via a recirculation path. This will function as a leak test because operating in the recirculation mode exposes the discharge pipe and pump to a pressure higher than that experienced during a LOCA. The RHR pumps are tested at least once every 18 months according to the Technical Specifications. CBS equipment is periodically tested via a recirculation path. This will function as a leak test, as pressure during the recirculation mode exposes the discharge pipe and pump to a pressure higher than that experienced in a LOCA. The containment spray pumps are tested at least once every 18 months per Technical Specification 4.6.2.1.
The safety injection pumps are tested via a recirculation miniflow test line to the refueling water storage tank. The pressure in the pump and pump discharge line is greater than that seen in a LOCA. This test is performed at least ever y 18 months per Technica l Specification 4.5.2. Portions of the RHR, CBS, and SI systems ar e periodically inspected for leakage under the Leakage Reduction Program in accordance with Technical Specifications 6.7.6.a. The leakage from ECCS and containment spray system equipment has been evaluated in Subsection 15.6.5.4d and the impact of this leakage is factored into the offsite dose calculations. The Containment Leakage Rate Testing (CLRT)
Program is described in the Leakage Test Reference (SLTR). Periodic Type A, B, and C leakage tests will be performed and reported in compliance with the intent of 10 CFR 50, Appendix J. Deviations from the wording of Appendix J are included in the Technical Specifications. The testing program included a complete series of Type B and Type C tests as presented in Subsections 6.2.6.2 and 6.2.6.3, prior to fuel loading. During plant operation, each penetration requiring testing will be tested periodically to ensure continued compliance with leakage limits.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 95  6.2.6.1 Containment Integrated Leakage Rate Test - Type A Test The initial containment integrated leakage rate test (Type A test) was performed after completion of construction of the containment structure and prior to initial fuel loading. Periodic integrated leakage rate tests shall be conducted in accordance with Subsection 6.2.6.4 and Technical Specifications 4.6.1.2. The maximum allowable integrated leakage rate, La, at the calculated peak accident pressure, Pa, is 0.15 weight percent per 24 hours. The calcula ted peak accident pressure, Pa, is 49.6 psig. Prior to conducting the initial Type A test all penetrations were installed and all systems penetrating the containment were complete, up to and including all automatic isolation valves external to the containment. Deviations from this schedule were documented and properly considered when reporting final leakage rate test results.
The structural integrity test (SIT) preceded the initial Type A test. A minimum of 24 hours elapsed from the time the containment was in excess of 85 percent Pa for the SIT and the commencement of the Type A test, to assure sufficient time for outga ssing from the internal structure. The structural integrity of the containment vessel and of the containment enclosure building shall be determined in accordance with the Containment Leakage Rate Testing Program. Any abnormal degradation detected during these inspect ions will be reported as part of a special report as required by Te chnical Specifications. Systems will be aligned for the Type A test based on the requirements of the Containment Leakage Rate Testing Program. Table 6.2-90 lists systems typically vent ed prior to and during the conduct of the Type A test. Table 6.2-91 lists those systems typically not vented and drained, and the justification thereof.
The required tests including applicable pre-test requirements, data analysis methods, test acceptance criteria, test schedule requirements, and reporting requirements are discussed in the Containment Leakage Rate Testing Program. 6.2.6.2 Containment Penetration Leakage Rate Test - Type B Test Type B tests are required on all containment penetrations with resilient seals, gaskets, or expansion bellows. These include, but are not limited to, air locks, air lock door seals, piping penetrations with expansion bellows and blind fla nges, and electrical seals.
Those penetrations which are seal-welded are exempt from this testing requirement. Table 6.2-92 lists all containment penetrations falling into this category.
 
All penetrations requiring Type B testing will be tested in accordance with the Containment Leakage Rate Testing Program. This is in accordance with Appendix J of 10 CFR 50.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 96  6.2.6.3 Containment Isolation Valve Le akage Rate Test - Type C Test Type C tests are required on all lines that penetrate the primary containment and present a potential leakage path between the inside and outside atmospheres of the primary containment under postulated accident conditions. These include lines: (1) that provid e a direct connection between containment atmosphere and the outside, like purge and vent lines; (2) whose isolation valves are required to close automatically upon receipt of a containment isolation signal to isolate containment atmosphere or the Reactor Coolant System; or, (3) whose isolation valves are required to operate intermittently under post-accident conditions to isolate containment atmosphere or the Reactor Coolant System. Table 6.2-83 lists all lines penetrating the containment and, where applicable, the containment isolation valves associated with those lines. Those lines not considered as requiring testing are noted. Containment isolation valves which are not Type C tested, and the reasons thereof, can be categorized as follows: a. Valves that isolate lines which form a closed system inside containment satisfying the criteria of Updated FSAR Subsection 6.2.4.1c are not Type C tested. These systems, and therefore their respective containment isolation valves, will not communicate with containment atmosphere or the reactor coolant system under post-accident conditions. These systems include main steam, feedwater, and steam generator blowdown, and component cooling water supply and return for the thermal barrier heat exchangers. b. Certain ECCS containment isolation valves are not Type C tested. The primary function of many of these valves is not to isolate containment following an accident, but rather to direct emergency core cooling water as desired. In fact, most of the valves will be open during one or more of the three ECCS post-accident modes. In addition, the valves are part of Safety Class 2/seismic Category I systems that are closed outside containment and liquid-filled, with an assured post-accident 30-day water supply.
A water seal at a pressure greater than 1.10 Pa will be maintained at the containment penetrations associated with these isolation valves for the 30-day post-accident period. This seal precludes leakage of containment atmosphere. c. The containment isolation valves on the CGC hydrogen analyzer lines are not required to be Type C tested. These lines form a closed, seismic Category I system outside containment. The integrity or the closed system will be maintained by performing periodic surv eillance testing as described in Subsections 6.2.4.2.m.3 and 6.2.5.1.g. Type C tests will be performed in accordance with the Containment Leakage Rate Testing Program. For valves tested in this manner, a radiological assessment will be made to establish the leakage limits. This form of testing meets the intent of 10 CFR 50, Appendix J (III.C.2), and no exemption is noted.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 97  As discussed in 6.2.3.2.d, a containment enclosure emergency cleanup system has been provided to minimize leakage to the environs. A significant number of lines penetrate the containment and terminate in areas not treated by this cleanup system. All leakage attr ibuted to penetrations and isolation valves, requiring Type B and Type C leakage rate tests per 10 CFR 50, Appendix J, is conservatively assumed to bypass the cleanup system. 6.2.6.4 Scheduling and Reporting of Periodic Tests The Type A test schedule and reporting requirements will be in accordance with the Containment Leakage Rate Testing Program. 6.2.6.5 Special Testing Requirements This section addresses the special requirements associated with the secondary containment surrounding the primary containment. The maximum allowable leakag e rate and in-leakage limits are discussed in Subsection 6.2.3.3(a) and Technical Specification 3/4.6.5.2.
 
====6.2.7 References====
: 1. "CONTRAST-S MOD1 - A Digital Computer Program to Predict Containment Pressure-Temperature Response," UEC-TR-006-SUP, June 1979, A Supplement to UE&C Topi cal Report UEC-TR-006-0, March
 
1976. 2. Roshenow, W.M. and Hartnett, J.P., "Handbook of Heat Transfer," McGraw-Hill, New York (1973) 3. Brazlay, M.E., "Range of Interface Thermal Conductance for Aircraft Joints," NASA TND-426, May 1960. 4. Tagami, T., et al., "Studies for Safety Analysis of Loss-of-Coolant Accidents in Light Water Power Reactors," NSJ-TR-223, March 1968, Japan Atomic Energy Research Institute, Japan. 5. Uchida, H., et al., "Evaluation of Post-Accident Cooling Systems of Light Water Power Reactors," Third International Conference on the Peaceful Uses of Atomic Energy, New York, 1965. 6. "Preliminary Safety Analysis Report," Public Service of New Hampshire, Seabrook Station Units 1 & 2, Docket-Nos. 50-443 and 50-444. 7. Clausing, A.M., "Practical Techniques for Estimating the Accuracy of Finite-Difference Solutions to Para bolic Equations," J. Appl. Mech., March 1973. 8. Shepard, R. M., et al., "Westinghouse Mass and Energy Release Data for Containment Design," WCAP-8312-A, Rev. 2 and WCAP-8264-P-A (Proprietary) February 1974.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 98  9. Niyogi, K.K., Bhatnager S.K. and Rathi, S.S., "COMPRESS - A Code for Calculating Subcompartment Pressure Response," UEC-TR-004-1, July 1976. 10. Idel' Chik, I.E., "Handbook of Hydraulic Resistance, Coefficients of Local Resistance and of Friction," Israel Program for Scientific Translations, Jerusalem, 1966. 11. NS-TMA-2075 letter from T. M. Anderson, Westinghouse to J. F. Stolz, NRC, April 25, 1979, Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version. 12. Westinghouse ECCS Evaluation Model, February 1978 Version, WCAP-9220, February 1978. 13. F. M. Bordelon, et al., "SATAN-VI Program:  Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-6174, June
 
1974. 14. G. Collier, et al., "Calculati onal Model for Core Reflooding After a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Nonproprietary), June 1974. 15. Land, R.E., "Mass and Energy Releases Following a Steam Line Rupture," WCAP-8822, September 1976.
: 16. Deleted. 17. Kirk-Othmer, Encyclopedia of Chemical Technology, 2nd Edition Volume 3, Page 6.9, Figure 4. 18. Transactions of the Electrochemical Society, 1935, Volume 47, Page 88-100. 19. Ranz, W.E. and Marshall, W. R. Jr., Chemical Engineering Progress, 48, 4p. 173 (1952). 20. Parsly, L. F., Design Considerations of Reactor Containment Spray Systems - Part VI, ORNL-TM-2412. 21. Flow of Fluids through Valves, Fitt ings, and Pipe; Crane Technical Paper No. 410, 13th Printing - 1973. 22. J. F. Wilson, "Electric Hydrogen Recombiner for PWR Containments," Westinghouse WCAP-7709-L (Proprietary) and WCAP-7820 (Non-Proprietary), Supplements 1, 2, 3, 4 and 5. 23. Niyogi, K. K. and Lin, S. D., "An Analysis of Two Stream Gravity Feed Mixing Process," ASME 77-DE-44, February 14, 1977.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Containment Systems Revision 15 Section 6.2 Page 99  24. Deleted.
: 25. Deleted.
: 26. Deleted.
: 27. SECY 03-0127, "Final Rulemaking Risk-Informed 10 CFR 50.44, "Combustible Gas Control in Containment," dated July 24, 2003. 28. Calculation No. 4.3.16-13F, Revision 3, "Free Hydrogen in Containment." 29. NAI 8907-2, Revision 13, "GOTHIC Containment Analysis Package User Manual," Version 7.0, July 2001.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 1  6.3 EMERGENCY CORE COOLING SYSTEM
 
====6.3.1 Design====
Bases The Emergency Core Cooling System (ECCS) consists of the centrifugal charging pumps, safety injection pumps, a refueling water storage tank, the residual heat removal pumps, the residual heat removal heat exchanger, the safety injection accumulators, and the associated valves and piping. Plants listed in Subsection 1.3.1 have similar Emergency Core Cooling Systems to that of Seabrook.
The primary function of the ECCS following an accident is to remove the stored and fission product decay heat from the reactor core so that fuel rod damage, to the extent that it would impair effective cooling of the core, is prevented. The ECCS is designed to cool the reactor core as well as to prov ide additional shutdown capability following initiation of the following accident conditions: a. Loss-of-Coolant Accident (LOCA) including a pipe break or a spurious relief or safety valve opening in the Reactor Coolant System (RCS) which would result in a discharge larger than that which could be made up by normal makeup system. b. Rupture of a control rod drive mechanism causing a Rod Cluster Control Assembly (RCCA) ejection accident. c. Steam or feedwater system break accident including a pipe break or a spurious relief or safety valve opening in the secondary steam system which would result in an uncontrolled steam rele ase or a loss of feedwater. d. A steam generator tube rupture.
The acceptance criteria for the consequences of each of these accidents are described in Chapter 15 in the respective accident analysis sections.
 
The bases used in design and for selection of ECCS functional requirements are derived from Appendix K Limits for fuel cladding temperature, etc., following any of the above accidents as delineated in 10 CFR 50.46. The subsystem functional parameters are selected so that, when integrated, the Appendix K requirements are me t over the range of anticipated accidents and single failure assumptions.
NOTE:  All drawings referenced that are not included were provided separately to the NRC as required in Updated FSAR Section 1.7.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 2  The reliability of the ECCS has been considered in selection of the functional requirements, selection of the particular components and location of components and connected piping. Redundant components are provided where the loss of one component would impair reliability. Valves are provided in series wher e isolation is desired, and in parallel when flow paths are to be established for ECCS performance. Redundant s ources of the safety injection actuation signal are available so that the proper and timely operation of the ECCS will be ensured. Sufficient instrumentation is available so that a failure of an instrument will not impair readiness of the system. The active components of the ECCS are powered from separate buses which are energized from offsite power supplies. In addition, redundant sources of auxiliary onsite power are available through the use of the emergency diesel generators to ensure adequate power for all ECCS requirements. Each generator is capable of driving all pumps, valves, and necessary instruments associated with one train of the ECCS.
All valves required to be actuated during ECCS ope ration are located to pr event vulnerability to flooding. Repositioning of valves due to spurio us actuation coincident with a LOCA has been analyzed and is not considered credible for a design basis. The environmental qualification of active ECCS equipment is discussed in Section 3.11. Protection of the ECCS from missiles is discussed in Section 3.5. Protection of the ECCS against dynamic effects associated with ruptures of piping is described in Section 3.6. Protection from flooding is also discussed in Section 3.4.
The elevated temperature of the sump solution during recirculation is well within the design temperature of all ECCS components. In additi on, consideration has been given to the potential for corrosion of various types of metals expos ed to the fluid condition prevalent immediately after the accident or during long-term recirculation operations. 6.3.2 System Design 6.3.2.1 Schematic Piping and Instrumentation Diagrams Flow diagrams of the ECCS are shown in Figure 6.3-1, Figure 6.3-2, Figure 6.3-3, Figure 6.3-4, Figure 6.3-5 and Figure 6.3-6. a. System Operation Upon the initiation of a safety injection "S" signal, the following automatic actions are initiated to commence the injection phase of emergency core cooling: 1. Centrifugal charging pump s start (see Dwg. NHY-503335). 2. Refueling water storage tank suction valves to chargi ng pumps open (see Dwg. NHY-503335). 3. Normal charging path valves close (see Dwg. NHY-503337).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 3  4. Charging pump miniflow valves close (see Dwg. NHY-503380 & NHY 503398). 5. Safety injection pumps start (see Dwg. NHY-503900). 6. Residual heat removal pumps start (see Dwg. NHY-503761). 7. Any closed accumulator isolation valves open. These valves will open only if power is available to the normally de-energized motor control centers E 522 and E 622 (see Dwg. NHY-503907). 8. Volume control tank outlet isolation valves close. Valves are interlocked with the RWST suction valves to the charging pumps (see Dwg. NHY-503341). 9. High head safety injection valves open (see Dwg. 1-NHY-503903).
During the injection phase, the two centrifugal charging pumps (CCPs) operate to inject into the cold legs of all four loops. The source water to the CCPs is the refueling water storage tank (RWST). Once the Reactor Coolant System (RCS) pressure is below shutoff head of the two safety injection pumps (SIPs), they begin to take borated water from the RWST and deliver it to the cold legs of the four loops. This is done through the residual heat removal (RHR) injection/accumulator discharge lines. In the case of a steam line break or small RCS break, the system pressure remains high for a long period of time, and the CCPs and SIPs supply core cooling.
When the RCS pressure drops below the pressure of the four safety injection accumulator tanks, they discharge their conten ts in the four RCS cold legs. These accumulators contain borated water and are pressurized with nitrogen. This portion of the ECCS is most effective in the case of large RCS breaks where system pressure drops rapidly to the accumulator pressure. The two residual heat removal pumps (RHRPs) take water from the RWST and inject it into the cold legs of all four RCS loops via the accumulator discharge lines once system pressure drops below the shutoff head of the pumps.
Therefore, upon the initiation of the safety injection "S" signal, borated water is injected into the RCS via the CCPs, SIPs, accumulator tanks and RHRPs. The point at which these various injection modes commence operating is controlled by the rate at which the reactor coolant is lost and system pressure drops. The RWST supplies the borated water used for the injection phase of the ECCS. When the RWST water level drops to the low-low-1 level alarm point, the
 
injection phase is discontinued and the co ld leg recirculation phase is initiated.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 4  The changeover from the injection mode to recirculation mode is initiated automatically and completed manually by operator action from the main control room. Protection logic is provided to automatically open the two containment recirculating sump isolation valves when two out of four refueling water storage tank level channels indicate a refueling water storage tank level less than a low-low-1 level setpoint in conjunction with the initiation of the engineered
 
safeguards actuation signal ("S" signal). This automatic action would align the two residual heat removal pumps to take suction from the containment sump and to deliver directly to the RCS. It should be noted that the residual heat removal pumps would continue to operate during this changeover from injection mode to recirculation mode.
The two charging pumps and the two safety injection pumps would continue to take suction from the refueling water storage tank following the above automatic action, until manual operator action is taken to align these pumps in series with the residual heat removal pumps. Valves SI-V138 and SI-V139, in the injection path, are normally closed (but not de-energized). A safety injection ("S") signal then opens the valves in the injection line.
The refueling water storage tank level pr otection logic consists of four level channels, with each level channel assigned to a separate process control protection set. Four refueling water storage tank level transmitters provide level signals to corresponding normally de-energized level ch annel bistables. Each level channel bistable would be energized on receipt of a refueling water storage tank level signal less than the low-low-1 level setpoint.
The two-out-of-four coincident logic is utilized in both protection cabinets A and B to ensure a trip signal in the event that two out of the four level channel bistables are energized. This trip signal, in conjunction with the "S" signal, provides the actuation signal to automatically open the corresponding containment sump isolation valves.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 5  The low-low-1 refueling water storage tank level signal is also alarmed to inform the operator to initiate the manual acti on required to realign the charging and safety injection pumps for the recirculation mode. The manual switchover sequence that must be performed by the operator is delineated in Table 6.3-7. Following the automatic and manual switchover sequence, the two residual heat removal pumps would take suction from the containment sump and deliver
 
borated water directly to two RCS cold legs. A portion of the Number 1 residual heat removal pump discharge flow would be used to provide suction to the two charging pumps which would also deliver directly to the RCS cold legs. A portion of the discharge flow from the Number 2 residual heat removal pump would be directed to the RCS cold legs. As part of the manual switchover procedure (see Table 6.3-7, Step 5), the suctions of the safety injection and charging pumps are cross connected so that one residual heat removal pump can deliver flow to the RCS and both safety injection and charging pumps, in the
 
event of the failure of the second residual heat removal pump. After approximately 5 to 6 hours, cold leg recirculation is terminated and hot leg
 
recirculation is initiated. This is done to terminate any boiling in the core should
 
the break be in one of the RCS cold legs.
During this phase of recirculation, the SIPs discharge is aligned to supply wate r to all four RCS hot legs and the RHRPs discharge is aligned to supply water to RCS hot legs 1 and 4. The CCPs do not have the capability to feed the hot legs and continue to supply the cold legs. 6.3.2.2 Equipment and Component Descriptions The component design and operating conditions listed in Table 6.3-1 are specified as the most severe conditions to which each respective component is exposed during either normal plant operation, or during operation of the ECCS. For each component, these conditions are considered in relation to the code to which it is designed. By designing the components in accordance with applicable codes, and with due consideration for the design and operating conditions, the fundamental assu rance of structural integrity of the ECCS components is maintained. Components of the ECCS are de signed to withstand the appropriate seismic loadings in accordance with their safety class as given in Table 3.2-2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 6  Descriptions of the major mechanical components of the ECCS follow: a. Accumulators The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. During normal operation, each accumulator is isolated from the RCS by two check valves in series. Shoul d the RCS pressure fall below the accumulator pressure, the check valves open and borated water is forced into the RCS. One accumulator is attached to each of the cold legs of the RCS. Mechanical operation of the swing disc check valves is the only action required to open the injection path from the accumulators to the core via the cold leg. Connections are provided for remotely adjusting the level and boron concentration of the borated water in each accumulator during normal plant operation as required. Accumulator water level may be adjusted either by draining to the primary drain tank or by pumping borated water from the refueling water storage tank to the accumulator. Alternatively, the boron concentration may be adjusted by a feed and bleed process. This would involve pumping borated water into the accumulator from the refueling water storage tank, as described above, with simultaneous draining of the accumulator through the sample connection to the Containment Building drainage sump. Technical Specification actions related to out-of-service equipment during this evolution would be applied. Samples of the solution in the accumulators are taken
 
periodically for checks of boron concentration. Accumulator pressure is provi ded by a supply of nitrogen gas, and can be adjusted as required during normal plant operation; however, the accumulators are normally isolated from this nitrogen supply. Gas relief valves on the accumulators protect them from pressures in excess of design pressure. The accumulators are located within the containment but outside of the secondary shield wall which protects them from missiles. Accumulator gas pressure is monitored by indicators and alarms. The operator can take action as required to maintain plant operation within the requirements of the Technical Specification covering accumulator operability. b. Residual Heat Removal Pumps In the event of a LOCA, the residual heat removal pumps are started automatically on receipt of an "S" si gnal (see Dwg. NHY-5 03761). The residual heat removal pumps take suction from th e refueling water storage tank during the injection phase and from the containmen t sump during the recirculation phase. Each residual heat removal pump is a si ngle-stage vertical-position centrifugal pump.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 7  A minimum flow bypass line is provided for the pumps to reci rculate and return the pump discharge fluid to the pump suction should these pumps be started with the RCS pressure above thei r shutoff head. Once flow is established to the RCS, the bypass line is automatically clos ed (see Dwg. NHY-503764). This line prevents deadheading of the pumps and permits pump testing during normal operation.
The safety intent of Regulatory Guide 1.1 is met by the desi gn of the ECCS so that adequate net positive suction head is provided to system pumps. In addition to considering the static head and sucti on line pressure drop, the calculation of available net positive suction head in the recirculation mode assumes that the vapor pressure of the liquid in the sump is equal to the containment ambient pressure. This assures that the actual available net positive suction head is always greater than the calculated net positive suction head. The residual heat removal pumps are discussed further in Subsection 5.4.7. A pump performance curve is given in Figure 6.3-7. Available and required net positive suction head are shown in Table 6.3-1. c. Centrifugal Charging Pumps In the event of an accident, the charging pumps are started automatically on receipt of an "S" signal (see Dwg. NHY-503330), and are automatically aligned to take suction from the refueling water storage tank during injection (see Dwg. NHY-503335). During recirculation, sucti on is provided from the residual heat removal pump discharge. These high head pumps deliver flow to the RCS at the prevailing RCS pressure. Each centrifugal charging pump is a multi-stage diffuser design, barrel-type casing with vertical suct ion and discharge nozzles. A minimum flow bypass line is provided on each pump discharge to recirculate flow to the pump suction after cooling via the seal water heat exchanger during normal plant operation. Each minimum flow bypass line contains a motor-operated valve which closes on receipt of the safety injection signal. The miniflow isolation valve for each charging pump will re-open if the flow through its respective pump drops below the low flow setpoint. The valve will reclose in the presence of an "S" signal when the flow increases beyond the high flow setpoint (see Dwg. NHY-503398 & 503380). This signal also closes the valves to isolate the normal charging line and volum e control tank and ope ns the charging pump/refueling water storage tank suction valves to align the high head portion of the ECCS for injection (see Dwg. NHY-503335). The charging pumps may be
 
tested during power operation via the minimum flow bypass line.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 8  A pump performance curve for the centrifugal charging pumps is presented in Figure 6.3-8. Available and required ne t positive suction h ead are shown in Table 6.3-1. d. Safety Injection Pumps In the event of an accident, the safety injection pumps are started automatically on receipt of an "S" si gnal (see Dwg. NHY-503900). The pumps deliver water to the RCS from the refueling water storage tank during the injection phase, and from the containment sump via the residual heat removal pumps during the recirculation phase. A minimum flow bypass line is provided on each pump discharge to recirculate flow to the refueling water storage tank in the event that the pumps are started with the RCS pressure above pump shutoff head. This line also permits pump testing during normal plant operation. Two parallel valves in series with a third, downstream of a common header, are provided in this line. These valves are manually closed from the control room as part of the ECCS realignment from the injection to the recirculation mode. The common recirculation header for the safety injection pumps is seismically analyzed and is seismically supported.
A pump performance curve is presented in Figure 6.3-9. Available and required net positive suction head are shown in Table 6.3-1. e. Residual Heat Exchangers The residual heat exchangers are conventional shell and U-tube type units. During normal cooldown operation, the residual heat removal pumps recirculate reactor coolant through the tube side while component cooling water flows through the shell side. During emergency core cooling recirculation operation, water from the containment sump flows th rough the tube side.
The tubes are seal welded to the tube sheet.
 
A further discussion of the residual he at exchangers is found in Subsection 5.4.7. f. Valves Design features employed to minimize valve leakage include:  Where possible, packless valves are used. Other valves which are normally open, except check valves and those which perform a control function, are provided with backseats to limit stem leakage.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 9  Normally closed globe valves are installed with recirculation fluid pressure under the seat to prevent stem leakage of recirculated (radioactive) water. Relief valves are enclosed, i.e., they are provided with a closed bonnet. 1. Motor-Operated Valves The seating design of all motor-operated valves is of the Crane flexible wedge design. This design releases the mechanical holding force during the first increment of travel so that the motor operator works only against the frictional component of the hydr aulic unbalance on the disc and the packing box friction. The disc is guided throughout the full disc travel to
 
prevent chattering and to provide ease of gate moveme nt. The seating surfaces are hard faced to prevent galling and to reduce wear. Where a gasket is employed for the body to bonnet joint, it is either a fully
 
trapped, controlled compression, spir al-wound asbestos gasket with provisions for seal welding, or it is of the pressure seal design with provisions for seal welding. Many of the valves stuffing boxes were
 
originally designed with a lantern ring leakoff connection with packing configurations which minimize stem seal leakage to the full extent
 
possible by the design. Exceptions to th is criterion are gate valves that have been determined to be susceptible to pressure locking, which have been modified to utilize the valve stem leakoff connection as a vent path for the bonnet cavity. These valves us e only a single packing set. Based on industry recommendations, many of the double packed stuffing boxes have been modified to a single packing configuration. The motor operator incorporates a "hammer blow" feature that allows the motor to impact the discs away from the backseat upon opening or closing. This "hammer blow" feature not only impacts the disc but allows the motor to attain its operational speed prior to impact. Valves which must function against system pressure are designed so that they function with a pressure drop equal to full system pressure across the valve disc. 2. Manual Globe, Gate and Check Valves Gate valves employ a wedge design a nd are straight through. The wedge is either split or solid. All gate va lves have backseat and outside screw and yoke.
Globe valves, "T" and "Y" style, are full-ported with outside screw and yoke construction.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 10  Check valves are spring-loaded lift piston types for sizes 2 inches and smaller, swing type for sizes 21/2 inches to 4 inches and tilting-disc type for sizes 4 inches and larger. Stainless steel check valves have no penetration welds other than the inlet, outlet and bonnet. The check hinge is serviced through the bonnet. The stem packing and gasket of the stainless steel manual globe and gate valves are similar to those described above for motor-operated valves. Carbon steel manual valves are employed to pass nonradioactive fluids only and therefore do not contain the double packing and seal weld provisions. 3. Accumulator Check Valves (Swing-Disc)
The accumulator check valve is designed with a low pressure drop
 
configuration with al l operating parts contained within the body.
Design considerations and analyses which ensure that leakage across the check valves located in each accumulator injection line will not impair accumulator availability are as follows: (a) During normal operation, the chec k valves are in the closed position with a nominal differential pressure across the disc of approximately 1650 psi. Since the valves remain in this position
 
except for testing or when called upon to open following an accident, and are therefore not subject to the abuse of flowing operation or impact loads caused by sudden flow reversal and
 
seating, they do not experience significant wear of the moving parts, and are expected to function with minimal back leakage.
The back leakage can be checked via the test connection as described in Subsection 6.3.4. (b) When the RCS is being pressurized during the normal plant heatup operation, the check valves are tested for leakage as soon as there is a stable differential pressure of about 100 psi or more across the valve. This test confirms the seat ing of the disc and whether or not there has been an increase in the leakage since the last test. When this test is completed, the accumulator discharge line motor-operated isolation valves are opened and the RCS pressure increase is continued. There shoul d be no increase in leakage from this point on since increasing reactor coolant pressure increases the seating force and decreases the probability of leakage.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 11  (c) The experience derived from the check valves employed in the emergency injection systems indicates that the system is reliable and workable; check valve leakage has not been a problem. This is substantiated by the satisfactor y experience obtained from operation of the Robert Emmett Ginna and subsequent plants where the usage of check valv es is identical to Seabrook. (d) The accumulators can accept some in-leakage from the RCS without affecting availability.
Continuous in-leakage would require, however, that the accumulator volume be adjusted
 
periodically to Technical Specification requirements. 4. Relief Valves Relief valves are installed in various s ections of the ECCS to protect lines which have a lower design pressure than the RCS. The valve stem and spring adjustment assembly are isolated from the system fluids by a bellows seal between the valve disc and spindle. The closed bonnet provides an additional barrier for enclosure of the relief valves.
Table 6.3-2 lists the system's relief valves with their capacities and setpoints. 5. Butterfly Valves Each main residual heat removal line has an air-operated butterfly valve (RH-HCV-606 and 607) which is normally open and is designed to fail in the open position. The actuator is arranged so that air pressure on the diaphragm overcomes the spring force, causing the linkage to move the butterfly to the closed position. Upon loss of air pressure, the spring returns the butterfly to the open position. These valves are left in the full-open position during normal operation to maximize flow from this system to the RCS during the injection mode of the ECCS operation.
These valves are used during normal residual heat removal system (RHRS) operation to control cooldown flow rate. Each residual heat removal heat ex changer bypass line has an air-operated butterfly valve which is normally closed and is designed to fail closed. Those valves (RH-FCV-618 and 619) are used during normal cooldown to
 
avoid thermal shock to the residual heat exchanger.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 12  6. Accumulator Motor-Operated Valve Controls As part of the plant shutdown admini strative procedures, the operator is required to close these valves. This prevents a loss of accumulator water inventory to the RCS, and is performed shortly after the RCS has been depressurized below the safety injection unblock setpoint. The redundant pressure and level alarms on each accumulator would remind the operator to close these valves, if any were inadvertently left open. Power is disconnected after the valves are closed.
During plant startup, the ope rator is instructed via procedures to energize and open these valves when the RCS pressure reaches the safety injection unblock setpoint. Monitor lights in conjunction with an audible alarm will alert the operator should any of these valves be left inadvertently closed once the RCS pressure increases beyond the safety injection unblock
 
setpoint (see Dwg. NHY-503907). Once open, power is disconnected and remains off except during valve testing. The accumulator isolation valves are not required to move during power operation or in a post-accident situa tion, except for valve testing. For a discussion of limiting conditions for operation and surveillance requirements of these valves, refe r to the Technical Specifications. For further discussion of the instrumentation associated with these valves, refer to Subsections 6.3.5, 7.3.1b and 7.6.4. 7. Motor-Operated Valves and Controls Remotely operated valves for the injection mode which are under manual control (i.e., valves which normally are in their ready position and do not require a safety injection signal) have their position indicated on the control board. If a com ponent is out of its proper position, its monitor light will indicate this on the control panel. At any time during operation, when one of these valves is not in the ready position for injection, this condition is shown visually on the board, and an audible alarm is sounded in the control room on a system basis, as part of the Bypass and Inoperable Status alarms. The ECCS delivery lag times are given in Chapter 15. The accumulator injection time varies as the size of the assumed break varies since the RCS pressure drop will vary proportionately to the break size. Inadvertent mis-positioning of a motor-operated valve due to a malfunction in the control circuitry in conjunction with an accident has been analyzed and found not to be a credible event for use in design.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 13  Table 6.3-3 is a listing of motor-opera ted isolation valves in the ECCS showing interlocks, automatic f eatures, and position indications. 8. Motor-Operated SI Isola tion Valves and Controls SI-V93, the combined recirculation isolation valve from both safety injection pumps, is a normally open, motor-operated valve. This valve is closed by the operators, from the control room, during the switch over to the recirculation mode of safety injection.
Red/green valve position indication and valve full-closed monitor light is provided on the main control board. Additionally, any time SI-V93 leaves the full open position, an annunciator alarms for both the "SI Train A Inoperative" and the "SI Train B Inoperative" status monitoring alarms.
To prevent spurious operation or operator error, the control circuit for the motor operator is equipped with a dual contactor arrangement (see Updated FSAR Figure 8.3-45). This circ uit requires two separate operator actions, involving the normal valve control switch plus a separate
 
key-operated switch, to reposition the valve. 6.3.2.3 Applicable Codes and Classifications Applicable industry codes and classifications for the ECCS are discussed in Section 3.2. 6.3.2.4 Material Specifications and Compatibility Materials employed for components of the ECCS are given in Table 6.3-4. Materials are selected to meet the applicable material requirements of the codes in Table 3.2-2 and the following additional requirements for compatibility with the reactor coolant during the
 
recirculation phase following a LOCA: a. All parts of components in contact with borated water are fabricated of or clad with austenitic stainless steel or equivalent corrosion- resistant material. b. All parts of components in contact (internal) with sump solution during recirculation are fabricated of austenitic stainless steel or equivalent corrosion-resistant material. c. Valve seating surfaces are hard faced with Stellite Number 6 or equivalent to prevent galling and to reduce wear. d. Valve stem materials are selected for their corrosion resistance, high tensile properties, and resistance to surface scoring by the  packing.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 14  6.3.2.5 System Reliability Reliability of the ECCS is cons idered for all aspects of the system, from initial design through periodic testing of the components during plant operation.
The ECCS is a two train, fully redundant standby safeguard feature. The system has been designed and pr oven by analysis as having the ability to withstand any single credible active failure during injection, or any active or passive failure during recirculation, while still meeting the performance objectives described in Subsection 6.3.1. Two trains of pumps, heat ex changers, and flow paths are pr ovided for redundancy, as only one train is required to satisfy the performance requirements. The initiating signals for the ECCS are derived from independent sources, as measured from process (e.g., low pr essurizer pressure) or environmental variables (e.g., containment pre ssure). Redundant as well as functionally independent variables are measured to initiate the safeguards signals. E ach train is physically separated and protected where necessary so that a single event cannot initiate a common failure. Power sources for the ECCS are divided into two independent trains supplied from the emergency buses which can receive power either from onsite or offsite power sources. Sufficient diesel generating capacity is maintained onsite to provide power to each train. The diesel generators and their auxiliary systems are completely independent, and each supplies power to one of the two ECCS trains. The reliability program extends to the procurement of the E CCS components such that only designs which have been proven by past use in similar applications are acceptable for use. The quality assurance program, as described in Chapter 17, assures receipt of components only after manufacture and test to the a pplicable codes and standards. The preoperational testing program assures that the ECCS as designed and constructed will meet the functional requirements calculated in its design.
The ECCS is designed with the ability for online testing of most components so the availability and operational status can be readily determined. In addition to the above, the integrity and operability of the ECCS is assured through examination of critical components dur ing the routine in-service inspection. a. Active Failure Criteria The ECCS is designed to accept a single failure following the incident without loss of its protective function. The system design will tolerate the failure of any single active component in the ECCS itself or in the necessary associated service systems at any time during the period of required system operations following the incident.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 15  A Failure Mode and Effect Analysis is presented in Table 6.3-5, which demonstrates that the ECCS can sustain the failure of any si ngle active component in either the short or long term and still meet the level of performance for core cooling. Since the operation of the active compone nts of the ECCS following steam line rupture is identical to that following a LOCA, the same analysis is applicable.
The ECCS can sustain the failure of any single active component and still meet the level of performance for the addition of shutdown reactivity. b. Passive Failure Criteria The following design philosophy assures the necessary redundancy in component and system arrangement to meet the intent of the General Design Criteria on single failure as it specifically applies to failure of passive components in the ECCS. Thus, for the long term, the system design is based on accepting either a
 
passive or an active failure. 1. Redundancy of Flow Paths and Components for Long-Term Emergency Core Cooling In design of the ECCS, Westinghouse utilizes the fo llowing criteria: (a) During the long-term cooling peri od following a loss of coolant, the emergency core cooling flow paths shall be separable into two subsystems, either of which can provide minimum core cooling
 
functions and return spilled water from the floor of the containment back to the RCS. (b) Either of the two subsystems can be isolated and removed from service in the event of a leak outside the containment. Redundant
 
motor-operated valves arranged in series are provided for this isolation function. (c) Adequate redundancy of check va lves is provided to tolerate failure of a check valve duri ng the long term as a passive component. (d) Should one of these two subsystems be isolated in this long-term period, the other subsystem remains operable. (e) Provisions are also made in the design to detect leakage from components outside the containment, to collect this leakage, and to provide for maintenance of the affected equipment.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 16  A single passive failure analysis is presented in Table 6.3.6. It demonstrates that the ECCS can sustain a single passive failure during the long-term phase and still retain an inta ct flow path to th e core to supply sufficient flow to maintain the core covered and affect the removal of
 
decay heat. The procedure followed to establish the alternate flow path also isolates the component which failed. Thus, for the long-term emergency co re cooling functi on, adequate core cooling capacity exists with one flow path removed from service. 2. Subsequent Leakage from Components in Safeguards Systems With respect to piping and mechanical equipment outside the containment, considering the provisions for visual inspection and leak detection, leaks will be detected before they propagate to major proportions. A review of the equipment in the system indicates that the largest sudden leak potential would be the sudden failure of a pump sh aft seal. Evaluation of leak rate, assuming only the presence of a seal retention ring around the pump shaft, showed that flows less than 50 gpm would result. Piping leaks, valve packing leaks, or flange gasket leaks tend to build up slowly with time and are considered less severe than the pump seal failure. Larger leaks in the ECCS are prevented by the following:
(a) The piping is classified in accordance with ANS Safety Class 2 and receives the ASME Class 2 quality assurance program associated with this safety class. (b) The piping, equipment and suppor ts are designed to ANS Safety Class 2 seismic classification, permitting no loss of function for the design basis earthquake. (c) The system piping is located within a controlled area on the plant site. (d) The piping system receives periodic pressure tests, and is accessible for periodic visual inspection. (e) The piping is austenitic stainless steel which, due to its ductility, can withstand severe distortion without failure. (f) Instrument tubing is designed to the requirements of Regulatory Guide 1.151 as discussed in Subsection 7.1.2.12.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 17  Based on this review, the design of the Primary Auxiliary Building and related equipment was verified for its ability to handle leaks up to a maximum of 50 gpm. Leakage would dr ain to and collect in the primary auxiliary building sump. Automatic initiation of the sump pumps at a predetermined setpoint would be indicated at the main control board and would alert the operator to an abnormal condition. Corrective action would include determining the location of the leak by visual inspection, and remote or manual isolation of the leak point from the rest of the system within 30 minutes. c. Potential Boron Precipitation Boron precipitation in the reactor vessel can be prevented by a backflush of
 
cooling water through the core to reduce boil-off and resulting concentration of boric acid in the water remaining in the reactor vessel. This is accomplished by a switch from cold to hot leg recirculati on about 5 to 6 hours following an accident. The minimum Hot Leg Recirculation flow meets decay heat removal requirements at this time. Three flow paths are available for hot leg recirculation of sump water. Each safety injection pump can discharge to two hot legs with suction taken from residual heat removal pump discharge either directly or indirectly via the charging pump cross connect. The residual heat removal pump(s) will also be aligned to deliver flow to the hot leg injection header. Loss of one pump or one flow path will not prevent hot leg recirculation since redundant methods are available for use. d. Submerged Valve Motors All electrically operated valves in the ECCS required to be functional during and following a LOCA are located outside containment. All other electrical equipment in the ECCS that is require d during post-LOCA is either located outside containment or above the maximum calculated water level inside containment. 6.3.2.6 Protection Provisions The provisions taken to protect the system from damage that might result from dynamic effects are discussed in Section 3.6. The provisions to protect the system from missiles are discussed in Section 3.5. The provisions to protect the system from seismic damage are discussed in Sections 3.7, 3.9 and 3.10. Thermal stresses on the RCS are discussed in Section 5.2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 18  6.3.2.7 Provisions for Performance Testing Test lines are provided for performance testing of the ECCS as well as individual components. These test lines and instrumentation are shown in Figure 6.3-1. All pumps have miniflow lines for use in testing operability. Additional information on testing can be found in Subsection 6.3.4.2. 6.3.2.8 Manual Actions No manual operator actions are re quired for proper operation of the ECCS during the injection mode of operation. Only limited manual actions are required by the ope rator to realign the system for the cold leg recircul ation mode of operation, and, after approximat ely 5 to 6 hours, for the hot leg recirculation mode of operation. These actions are delineated in Table 6.3-7. The transfer from the injection mode to recirculation mode is initiated automatically and completed manually by operator action from the main control room. Protec tion logic is provided to automatically open the two containment recircul ation sump isolation valves when two out of four refueling water storage tank level channels indicate a refueling water storage tank level less than a low-low level setpoint in conjunction with the initiati on of the engin eered safeguard actuation signal ("S" signal - see Dwg. NHY-503258). The automatic action would align the two residual heat removal pumps to take suction from the containment sump and deliver directly to the Reactor Coolant System. The automatic action also aligns the two containment building spray pumps to take suction from the containment sump and deliver to the containment building spray headers. It should be noted that the residual heat removal and containment building spray pumps would continue to operate during this transfer from injection mode to recirculation mode. The two charging pumps and the two safety injection pumps would continue to take suction from the refueling water storage tank following the above automatic action, until manual operator action is taken to align these pumps in series with the residual heat removal pumps. The consequences of the operator failing to act altogether will be loss of high head safety injection pumps and charging pumps. The refueling water storage tank level protection logic consists of four level channels, with each level channel assigned to a separate process contro l protection set. Four refueling water storage tank level transmitters provide level signals to corresponding normally de-energized level channel bistables. Each level channel bistable would be energized on receipt of a signal that the refueling water storage tank level is less than the low-low level setpoint.
The two-out-of-four coincident logic is utilized in both protection cabinets A and B to ensure a trip signal in the event that two out of the four level channel bistables are energized. This trip signal, in conjunction with the "S" signal, provides the actuation signal to automatically open the corresponding containment sump isolation valves.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 19  The first manual actions are required of the operator after the "Lo-Lo" level signal actuates the opening of the sump valves. The "Lo-Lo" signal also alarms at the MCB to alert the operators to the need for action. Figure 6.3-10 shows a schematic of the tank, the level instrumentation setpoints and the various water volume allowances.
All level setpoints were selected to ensure a minimum required injection volume, adequate transfer volumes and, at the same time, to avoid any spurious alarm actuations.
The "high level" alarm setpoint was selected to assure that, during filling of the RWST, the operator is alerted prior to a possible spillage of water from refueling water storage tank via overflow.
The "tech spec" level alarm setpoi nt was selected to assure that the minimum required injection volume remains above the "Lo-Lo" level (transfer) setpoint. If the water level drops below the "tech spec" setpoint, the alarm sounds and the plant will be placed in the mode mandated by the plant technical specifications. A "tech spec approach" alarm is provided to alert the operator to a need for makeup prior to reaching a water level requiring plant shutdown. Spurious alarm actuation is prevented by an adequate separation between the instrument error bands associated with each setpoint. In addition to this, temperature compensation is employed for "high level," "tech spec" and "tech spec approach" alarms to account for level changes due to temperature fluctuations. Instrument error bands were calculated accounting for uncertainties such as measurement accuracy, calibration accuracy, signal drift, environment changes, etc. The time from accident initiation to the first required manual actions is dependent on the initial tank water level, the flow rate out of the tank and the "Lo-Lo" level setpoint. The minimum time is calculated based on injecting the minimum required injection volume of 350,000 gallons. As shown on Figure 6.3-10, the minimum injection allowance is contained between extreme low range of the "tech spec" alarm error band and the extreme upper range of the "Lo-Lo" signal error band. The maximum flow rate out of the RWST is based on the "Maximum Safeguards" flow rate of 13,200 gpm out of the tank. Based on injecting the minimum injection allowance at a maximum rate of 13,200 gpm, the first manua l operator actions are not required until approximately 26 minutes after accident initiation.
It should be noted that the entire injection volume is assumed to come from the RWST, neglecting the additional volume ava ilable in the spray additive tank.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 20  Figure 6.3-10 depicts two additiona l allowances of RWST inventor
: y. The transfer allowance is the volume of RWST inventory set aside to allow the operators sufficient time to complete the transfer from the injection mode to the recirculation mode of ECCS operation. This transfer allowance is the volume of water contained between the low range of the RWST Lo-Lo level setpoint and the high range of the RWST EMPTY alarm setpoint. The RWST EMPTY alarm setpoint is based on the postulated single failure discussed below. The shutoff or single failure allowance is the volume of RWST inventory between the RWST EMPTY alarm setpoint and the calculated level at which vortexing is possible for the flow rate from the RWST associated with the postulated worst case single failure th at could occur during this transfer. In the event of the design bases LOCA, the containment sump isolation valves open automatically upon actuation of the RWST LO-LO level signal. The combination of containment pressure and elevation head from the sump seats the check valves in the lines between the RWST, CBS, and RHR pumps (CBS-V 3, -V7, -V55, -V56), reducing the flow rate out of the RWST to that of the safety injection and charging pumps; or approximately 1500 gpm. At this flow rate, the operators would have over 27 minutes to complete the transfer before the RWST EMPTY alarm could actuate. It should be noted that the RWST EMPTY alarm setpoint is based on the higher flow rate associated with the single fail ure discussed below, and only the safety injection and charging pumps would be susceptible to vortexing at this RWST level. 
 
However, operators would still secure these pumps if the transfer was not completed before the RWST EMPTY alarm was actuated. In an accident for which the RWST water is at the minimum allowed temperature, the containment heat sinks are at a low temperature and the heat transfer rate in the containment is high, the containment pressure may be high eno ugh to actuate spray, but not high enough to seat the check valves in the suction lines from the RWST to the RHR and CBS pumps. For this case, the high flow rate out of the RWST would continue until the tank suction isolation valves (CBS-V2, -V5) are closed by the operators. The operators would have over 3 minutes to close these valves before vortexing is calculated to occur. Assuming the operators take the full time allotted to close these two valves, this would still leave an additional 5 minutes to complete the transfer.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 21  The limiting single failure for this design is the failure of one of the RWST isolation valves to close (CBS-V2, -V5). Note that the failure of one of the containment sump isolation valves to open has the same affect with respect to flow ra te out of the tank, however , it occurs earlier in the transfer and the operator has additional time to respond. If one of the RWST isolation valves does not close, the flow rate out of the RWST is reduced from 13,200 gpm to 7400 gpm, as opposed to approximately 1500 gpm when both va lves close. The RWST EMPTY alarm is designed for this single failure. If the operators are not able to get the affected valve closed, the RWST EMPTY alarm will sound, alerting them to immediately shut off any pumps still taking suction from the RWST. The shutoff allowance is established to give the operators 1 minute of operation between the EMPTY alarm and the level at which vortexing could potentially occur for shutting off the pumps. Note that though all of the pumps still taking a suction from the RWST are shut off when this alarm sounds, the operato rs only have to stop th e affected RHR and CBS pumps in order to prevent vortexing. The safety injection and charging pumps could operate for over 11 additional minutes before reaching their calculated vortex level. After the pumps are stopped, the transfer is completed and all available pumps are restarted. Following the automatic and manual transfer sequence, the two residual heat removal pumps would take suction from the containment sump and deliver borated water directly to two RCS cold legs. A portion of the Number 1 residual heat removal pump discharge flow would be used to provide suction to the two charging pumps which would also deliver directly to the RCS cold legs. A portion of the discharge flow from the Number 2 residual heat removal pump would be used to provide suction to the two safety injection pumps which would also deliver directly to the RCS cold legs. As part of the manual transfer, the suctions of the safe ty injection and charging pumps are cross connected so that one residual heat removal pump can deliver flow to the RCS and both safety injection and charging pumps, in the event of the failure of the second residual heat removal pump. See Section 7.5 for process information available to the operator in the control room following and accident. 6.3.3 Performance Evaluation Chapter 15 accidents that result in ECCS operation are as follows: a. Inadvertent opening of a steam generator relief or safety valve (see Subsection 15.1.4) b. Small break LOCA (see Subsection 15.6.5) c. Large break LOCA (see Subsection 15.6.5)
: d. Major secondary system pipe failure (see Subsection 15.1.5) e. Steam generator tube failure (see Subsection 15.6.3).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 22  Safety injection is actuated from any of the following: a. Low pressurizer pressure
: b. Low steamline pressure
: c. High containment pressure
: d. Manual initiation. A safety injection signal will rapidly trip the main turbine, close all feedwater control valves, trip the main feedwater pumps, and close the feedwater isolation valves.
Following the actuation signal, the suction of the centrifugal charging pumps is diverted from the volume control tank to the refueling water storage tank. Simultaneou sly, the valves isolating the charging pumps from the injection header automatically open. The safety injection pumps also start automatically but operate at shut off head when the RCS is at normal pressure. The passive injection system (accumulators) and the low head system (residual heat removal pumps) also provide no flow at normal RCS pressure. Figure 6.3-6 is a simplified illustration of the ECCS. The notes provided with Figure 6.3-6 contain information relative to the operation of the ECCS in its various modes. The modes of operation illustrated are full operation of all ECCS components, cold leg recirculation with residual heat removal pump Number 2 operating, and hot leg recirculation with residual heat removal pump Number 1 operating.
Lag times for initiation and operation of the ECCS are limited by pump startup time and consequential loading sequence of these motors onto the safeguard buses. Most valves are normally in the position conducive to safety; therefore, valve opening time is not considered for these valves. If there is no power blackout, all pump motors and valve motors are started immediately upon receipt of the "S" signal. In the case of a blackout, a 10 second delay is assumed for diesel startup, then pumps and valves are loaded acc ording to the sequencer. The charging pumps will be applied to the buses in 10 seconds, the safety injection pumps will start in 15 seconds, and the residual heat removal pumps in 20 seconds. These times refer to time after receipt of the "S" signal, and are in addition to the time for attainment of speed of the diesel generators. See Subsection 8.3.1 for details of diesel generator operation. The time sequence of ECCS components is also discussed in Chapter 15 with the a ppropriate accident analysis.
 
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 23  6.3.3.1 Inadvertent Opening of a Steam Generator Relief or Safety Valve The assumed steam release is typical of the capacity of any single steam dump, relief or safety valve. The boron solution provides sufficient negative reactivity to maintain the reactor well below criticality. The cooldown for this case is more rapid than the actual case of steam release from all steam generators through one steam dump, re lief, or safety valve. The transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the steam generators.
Since the transient occurs over a period of about 5 minutes, the negl ected stored energy is likely to have a significant effect in slowing the cooldown. The analysis shows that there will be no return to criticality after reactor trip, assuming a stuck Rod Cluster Control Assembly, with offsite power available, and assuming a single failure in the engineered safe ty features. Since the reactor does not return to criticality, a Departure from Nucleate Boiling Ratio (DNBR) less than the safety analysis limit value does not exist. 6.3.3.2 Small Break LOCA A LOCA is defined as a rupture of the RCS pi ping or of any line connected to the system. Ruptures of small cross section will cause expul sion of the coolant at a rate which can be accommodated by the charging pumps which would maintain an operational water level in the pressurizer permitting the operator to execute an orderly shutdown. The maximum break size for which the normal makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the RCS thr ough the postulated break against the charging pump makeup flow at normal RCS pressure, i.e., 2250 psia. A makeup flow rate from one centrifugal charging pump is adequate to sustain pressurizer level and pressure for a break through a 0.375 inch diameter hole. This break results in a loss of approximately 17.5 lb/sec (127 gpm at 130 F and 2250 psia). The safety injection signal stops normal feedwa ter flow by closing the main feedwater line isolation valves and initiates emergency feedwater flow by starting emergency feedwater pumps. The small break analyses deal with breaks of up to 1.0 ft 2 in area, where the safety injection pumps play an important role in the initial core recovery because of th e slower depressurization of the RCS. The RCS depressurization water level transients show that for a break of approximately 3.0 inch equivalent diameter, the tran sient is turned around and the core is recove ring prior to accumulator injection. For a 3.5 inch equivalent diameter break, the core remains uncovered with a decreasing level until accumulator action. Thus, the maximum break size showing core recovery prior to accumulator injection will be approximately 3.0 inch equivalent diameter. Accumulator injection commences when pressure reaches 600 psia, i.e., approximately 1200 seconds for the 3.0 inch break size.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 24  The analysis of this break has shown that th e high head portion of the ECCS, together with accumulators, provide sufficient core flooding to keep the calculated peak clad temperature below required limits of 10 CFR 50.46. Hence, ad equate protection is afforded by the ECCS in the event of a small break LOCA. 6.3.3.3 Large Break LOCA A major LOCA is defined as a rupture 1.0 ft 2 or larger of the RCS piping including the double-ended rupture of the largest pipe in the RCS or of any line connected to that system. The boundary considered for LOCA as related to c onnection piping is de fined in Section 3.6. Should a major break occur, depr essurization of the RCS results in a pressure decrease in the pressurizer. Reactor trip occurs and the Safety Injection System is actuated when the pressurizer low pressure trip setpoint is reached. Reactor trip and safety injection system actuation are also provided by a high containmen t pressure signal. These countermeasures will limit the consequences of the accident in two ways: a. Reactor trip and borated water injection provide additional negative reactivity insertion to supplement void formation in causing rapid reduc tion of power to a residual level corresponding to fission product decay heat. b. Injection of borated water ensures su fficient flooding of the core to prevent excessive clad temperatures. When the pressure falls below approximately 600 psia, the accumulators begin to inject borated water. The conservative assumption is made that injected accumulato r water bypasses the core and goes out through the break until the termination of the blowdown phase. This conservatism is again consistent with the Final Acceptance Criteria. The pressure transient in the reactor containment during a LOCA affects ECCS performance in the following ways. The time at which end of blowdown occurs is determined by zero break flow which is a result of achieving pressure equilibrium between the RCS and the containment. In this way, the amount of accumulator water bypass is also affected by the containment pressure, since the amount of accumulator water discharged during blowdown is dependent on the length of the blowdown phase and RCS pressure at end of blowdown. During the reflood phase of the transient, the density of the steam ge nerated in the core is dependent on the existing containment pressure. The density of this steam affects the amount of steam which can be vented from the core to the break for a given downcomer head, the core reflooding process, and thus, the ECCS performance. It is through these effects that containment pressure affects ECCS performance.
For breaks up to and including th e double-ended severance of a r eactor coolant pipe, the ECCS will limit the clad temperature to well below the melting point and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved. See Table 15.6-1 for ECCS sequence of events.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 25  For breaks up to and including th e double-ended severance of a r eactor coolant pipe, the ECCS will meet the Acceptance Criteria as presented in 10 CFR 50.46. That is: a. The calculated peak fuel element clad temperature provides margin to the requirement of 2200 F. b. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor. c. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17 percent are not exceeded during or after quenching. d. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core. 6.3.3.4 Major Secondary System Pipe Failure The steam release arising from a rupture of a main steam pipe would result in energy removal from the RCS causing a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin. There is an increased possibility that the core will become critical and return to power. A return to power following a steam pipe rupture is a potential problem. The core is ultimately shut down by the borated water injection delivered by the Safety Injection System. Minimum capability for injecti on of the high concentration boric acid solution is assumed corresponding to the most restric tive single active failure in the Safety Injection System. The calculated transient delivery times for the borated water are listed in Table 15.1-1. In all cases, injection of the refueling water is preceded by the low concentration borated water, which is swept from the lines. For the cases where offsite power is available, the sequence of events in the Safety Injection System is the following. After the generation of th e safety injection signal (appropriate delays for instrumentation, logic, and signal transport in cluded), the appropriate va lves begin to operate and the high head safety injection pump starts. In 30 seconds, the valves are assumed to be in their final position and the pump is assumed to be at full speed. The volume containing the low concentration borated water is swep t, of course, before the refueli ng water reaches the core. This delay, described above, is inherently included in the modeling. In cases where the offsite power is not available, and additional 10 second delay is assumed to start the diesels and to load the necessary safety injection equipment onto them. The analysis has shown that even assuming a stuck Rod Cluster Control Assembly with or without offsite power, and assuming a single active failure in the engineered safeguards, the core remains in place and intact. Radiation doses will not exceed 10 CFR 100 guidelines.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 26  Although Departure from Nucleate Boiling (DNB) (with possible clad perforation) following a steam pipe rupture is not necessarily unacceptable and not precluded in the criterion, the above analysis, in fact, shows that no Departure from Nucleate Boiling occurs for any rupture assuming the most reactive Rod Cluster Control Assembly stuck in its fully withdrawn position. 6.3.3.5 Steam Generator Tube Failure The accident examined is the complete severance of a single steam generator tube, assuming it to
 
take place at power. Assuming normal operation of the various plant control systems, the following sequence of events is initiated by a tube failure: a. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side there is a steam flow/feedwater flow mismatch before the trip as feedwater flow to the affected steam generator is reduced due to the additional break flow which is now being supplied to that unit. b. Continued loss of reactor coolant inventory leads to a reactor trip and safety injection signals generated by an OTDT signal. The resultant plant cooldown also leads to a rapid change in pressurizer level. The safety injection signal automatically terminates normal feedwater supply and initiates emergency
 
feedwater supply and initiates emergency feedwater addition. c. The four steam generator blowdown and flash tank concentrates liquid monitors and the condenser off-gas radiation monitor will alarm indicating a sharp increase in radioactivity in the secondary system. A high radioactivity signal from any one of the four steam generator blowdown radiation monitors or the flash tank
 
concentrates radiation monitor will automatically isolate the concentrates discharge from the blowdown flash tank. d. The reactor trip automatical ly trips the turbin e and, if offsite pow er is available, the steam dump valves open permitting steam dump to the condenser. In the
 
event of a coincident station blackout, the steam dump valves would automatically close to protect the condenser. The steam generator pressure would rapidly increase resulting in steam discharge to the atmosphere through the steam generator safety and/or power-operated relief valves. e. Following reactor trip, the continued action of emergency feedwater supply and borated safety injection flow (supplied from the refueling water storage tank) provide a heat sink which absorbs some of the decay heat. Thus, steam bypass to the condenser, or in the cas e of loss of offsite power, steam relief to atmosphere, is attenuated during the recovery procedure leading to isolation.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 27  f. Safety injection flow results in increasing pressurizer water level. The time after trip at which the operator can clearly see returning level in the pressurizer is dependent upon the amount of operating auxiliary equipment. A steam generator tube rupture will cause no subsequent damage to the RCS or the reactor core. An orderly recovery from the accident can be completed even assuming simultaneous loss of offsite power. 6.3.3.6 Existing Criteria Used to Judge the Adequacy of the ECCS Criteria from 10 CFR 50.46: a. Peak clad temperature ca lculated shall not exceed 2200 F. b. The calculated total oxidation of the clad shall nowhere exceed 17% of the total clad thickness before oxidation. c. The calculated total amount of hydrogen generated from the chemical reaction of the clad with water or steam shall not exceed 1 percent of the hypothetical amount that would be generated if all of the metal in the cl ad cylinders surrounding the fuel, excluding the clad around the plenum volume, were to react. d. Calculated changes in core geometry shall be such that the core remains amenable to cooling. e. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptable low value and decay heat shall be removed for the extended period of time required by long-lived radioactivity remaining in the core. In addition to and as an extension of the Final Acceptance Criteria, two accidents have more specific criteria as shown below. In the case of the inadvertent opening of a steam generator relief or safety valve, an additional criteria for adequacy of the ECCS is: assuming a stuck Rod Cluster Control Assembly, offsite power unavailable, and a single failure in the engineered safety features, there will be no return to criticality after reactor trip for a steam release equivale nt to the spurious opening, with failure to close, of the larger of a single steam dump, relief, or safety valve. For a major secondary system pipe failure, the added criteria is:  assuming a stuck RCCA with or without offsite power, and assuming a single failu re in the engineered safeguards, the core remains in place and intact.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 28  6.3.3.7 Use of Dual Function Components The ECCS contains components which have no other operating function, as well as components which are shared with other systems. Components in each category are as follows: a. Components of the ECCS which perform no other function are: 1. One accumulator for each loop which discharges borated water into its respective cold leg of the reactor coolant loop piping. 2. Two safety injection pumps, which supply borated water for core cooling to the RCS.  (May be used dur ing check valve testing also.) 3. Associated piping, valves and instrumentation. b. Components which also have a normal operating function are: 1. Residual Heat Removal Pumps and the Residual Heat Exchangers These components are normally used during the latter stages of normal reactor cooldown and when the reactor is held at cold shutdown for core decay heat removal or for flooding the refueling cavity. However, during
 
all other plant operating periods, they are aligned to perform the low head injection function. 2. Centrifugal Charging Pumps These pumps are normally aligned for char ging service. As a part of the Chemical and Volume Control System, the normal operation of these pumps is discussed in Subsection 9.3.4. 3. Refueling Water Storage Tank This tank is used to fill the refuelin g canal for refueling operations and to provide makeup to the spent fuel pool. However, during all other plant operating periods, it is aligned to the suction of the safety injection pumps, and the residual heat removal pumps. The charging pumps are automatically aligned to the suction of the refueling water storage tank upon receipt of the safety injection signals or a volume control tank low level signal. During normal operation they take suction from the volume control tank.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 29  An evaluation of all components required for operation of the ECCS demonstrates that either: a. The component is not shared with other systems, or b. If the component is shared with other systems, it is either aligned during normal plant operation to perform its accident function or, if not aligned for its accident function, two valves in parall el are provided to align the system not utilized for injection and two valves in series are provided to isolate portions of the system not utilized for injection. These valves are automatically actuated by the safety injection signal. Table 6.3-8 indicates the alignment of components during normal operation, and the realignment required to perform the accident function.
In all cases of component operation, safety injection has the prior ity usage so that an "S" signal will override all other signals and start or align systems for injection. 6.3.3.8 Limits on System Parameters The analyses show that the desi gn basis performance characteristic of the ECCS is adequate to meet the requirements for core cooling following a LOCA with the minimum engineered safety features equipment operating. In order to ensure this capability in the event of the simultaneous failure to operate any single active component, Technical Specifications are established for reactor operation.
Normal operating status of ECCS components is given in Table 6.3-9. The ECCS components are available whenever th e coolant energy is high and the reactor is critical. During low temperature operation there is a negligible amount of stored energy in the coolant and low decay heat; therefore, an accident comparable in severity to accidents occurring at operating conditions is less probable and fewer ECCS components are required. The principal system parameters and the number of components which may be out of operation or in test, quantities and concentrations of coolant available, and allowable time in a degraded status are illustrated in the Tec hnical Specifications. If efforts to repair the faulty component are not successful, the plant is placed into a lower operational status, i.e., hot standby to hot shutdown, hot shutdown to cold shutdown, etc.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 30  6.3.4 Tests and Inspections 6.3.4.1 Preoperational Performance Tests Preoperational performance testing of the ECCS is discussed in Chapter 14. 6.3.4.2 Reliability Tests and Inspections Capability is provided for routine periodic testing of the ECCS components and all necessary support systems at power. Valves which operate after a LOCA are operated through a complete cycle, and pumps are operated individually in this test on their miniflow lines. The charging pumps also can be tested by their normal charging function. If such testing indicates a need for corrective maintenance, the redundancy of equipment in these systems permits such maintenance to be performed without shutting down or re ducing load under certai n conditions. These conditions include considerations such as the period within which the component should be restored to service and the capability of the remaining equipment to provide the minimum required level of performance during such a period.
The operation of the remote stop valve and the check valve in each accumulator tank discharge line may be tested by opening the remote test line valves just downstream of the stop valve and check valve. Flow through the test line can be observed on instruments, and the opening and closing of the discharge line stop valve can be sensed on this instrumentation. Where series pairs of check valves form the high pressure to low pressure isolation barrier between the RCS and safety injection system pipi ng, periodic testing of these check valves must be performed to provide assurance that certain postulated failure modes will not result in a loss of coolant from the low pressure system outside containment with a simultaneous loss of safety injection pumping capacity. The safety injection system test line subsystem provides the capability for determination of the integrity of the pressure boundary formed by series check valves. The tests performed verify that each of the series check valves can independently sustain differential pressure across its disc, and also verify that the valve is in its closed positi on. The required periodic tests are to be performed after each refueling just prior to plant startup, after the RCS has been pressurized. Lines in which the series check valves are to be tested are the safety injection pump cold and hot leg injection lines and the residual heat removal pump cold and hot leg injection lines. To implement the periodic component testing requirements, Technical Sp ecifications have been established. During periodic system testing, a visual inspection of pump seals, valve packings, flanged connections, and relief valves is made to detect leakage. In-s ervice inspection provides further confirmation that no signi ficant deterioration is occurr ing in the ECCS fluid boundary.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 31  Design measures have been taken to assure that the following testing can be performed: a. Active components may be tested periodically for operability (e.g., pumps on miniflow, certain valves, etc). b. An integrated system actuation test
* can be performed when the plant is cooled down and the RHRS is in operation. c. An initial flow test of the fu ll operational sequence can be performed.
The design features which assure this test capability are specifically: a. Power sources are provided to permit individual actuation of each active component of the ECCS. b. The safety injection pumps can be tested periodically during plant operation using the minimum flow recirculation lines provided. c. The residual heat removal pumps are used every time the RHRS is put into operation. They can also be tested period ically when the plant is at power using the miniflow recirculation lines. d. The centrifugal charging pumps are either normally in use for charging service or can be tested periodically on miniflow. e. Remote operated valves can be exercised during routine plant maintenance. f. Level and pressure instrumentation is provided for each accumulator tank for continuous monitoring of these parameters during plant operation. g. Flow from each accumulator tank can be directed through a test line to determine valve operability. h. A flow indicator is provided in the charging pump, safety injection pump, and residual heat removal pump headers. Pressure instrumentation is also provided in these lines. i. An integrated system test can be performed when the plant is cooled down and the RHRS is in operation. This test demonstrates the operation of the valves, pump circuit breakers, and automatic circu itry including diesel starting and the automatic loading of ECCS components of the diesels (by simulating a loss of
 
offsite power in conjunction with an SI actuation test signal).
* Details of the testing of the sensors and logic circuits associated with the generation of the safety injection signal, together with the application of this signal to the operation of each active component, are given in Section 7.2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 32  See the Technical Specifications for the selection of te st frequency, acceptab ility of testing and measured parameters. ECCS components and systems are designed to meet the intent of the ASME Code, Section XI for in-service inspection. 6.3.5 Instrumentation Requirements Instrumentation and associated analog and logic channels employed for initiation of ECCS operation are discussed in Section 7.3. This section describes the instrumentation employed for monitoring ECCS components during normal plant operation and ECCS post accident operation. All alarms are annunciated in the control room. 6.3.5.1 Temperature Indication: Residual Heat Exchanger Temperature The fluid temperature at both the inlet and the outlet of each residual heat exchanger is recorded in the control room.
6.3.5.2 Pressure Indication
: a. Charging Pump Inlet and Discharge Pressure There is local pressure indication at the suction and discharge of each centrifugal charging pump. b. Safety Injection Pump Suction Pressure There is a locally mounted pressure indicator at the suction of each safety injection pump. c. Safety Injection Pump Discharge Pressure Safety injection pump discharge pressure is indicated in the control room for both pumps. d. Accumulator Pressure Duplicate pressure channels are installed on each accumulator. Pressure indication in the control room and high and low pressure alarms are provided by each channel. e. Test Line Pressure A local pressure indicator used to check for proper seating of the system check valves between the injection lines and the RCS is installed on the leakage test line. The pressure is also indicated on the control board. f. Residual Heat Removal Pump Suction Pressure Local pressure indication is provided at the inlet to each residual heat removal pump.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 33  g. Residual Heat Removal Pump Discharge Pressure Residual heat removal pump discharge pressure for each pump is indicated in the control room. A high pressure alarm is actuated by each channel. h. Containment Building Spray Pumps Discharge Pressure Containment building spray pump discharge pressure is indicated in the control room for both pumps. 6.3.5.3 Flow Indication
: a. Charging Pump and Injection Header Flow Flow for each centrifugal charging pump and the total centrifugal charging pump injection flow are indicated in the control room. A low-flow alarm for each pump is provided on the main control board. b. Safety Injection Pump Flow Injection flow for each of the safety injection pumps is indicat ed in the control room. A low-flow alarm for each pump is provided on the main control board. c. Safety Injection Pump Minimum Flow A flow indicator is installed in the safety injection pump minimum flow line. d. Test Line Flow Local and main control board indication of the leakage test line flow is provided to check for proper seating of the system check valves between the injection lines and the RCS. e. Residual Heat Removal Pump Cold Leg Injection Flow The flow from each residual heat removal subsystem to the RCS cold legs is indicated in the control room. These instruments also control the residual heat removal bypass valves, maintaining constant return flow to the RCS during normal cooldown. f. Residual Heat Removal Pump Minimum Flow A flowmeter installed in each residual heat removal pump discharge header provides control for the valve located in the pump minimum flow line. A low-flow alarm is provided on the main control board.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 34  6.3.5.4 Level Indication
: a. Refueling Water Storage Tank Level Three separate level indicators are available at the main control board, from three separate level transmitters. Two channels are recorded. Apart from this, four channels of RWST level are provided to the ESF portion of the protection system.
A low-low level, 2/4 channels, in combination with the "S" signa l will generate an "ECCS/CONTN SPRAY RECIRC" signal.
This signal is used to open the containment sump isolation valves. The following alarms are available through the computer: 1. RWST Level High
: 2. RWST Level Approachi ng Technical Specification 3. RWST Level Technical Specification Limit
: 4. RWST Level Lo-Lo at Recirc Setpoint
: 5. RWST Empty Recirculation actuation is indicated on the control board by a monitoring light grouped with the appropriate valve position lights. b. Accumulator Water Level Duplicate water level channels are provided for each accumulator. Both channels provide indication in the control room and actuate high and low water level alarms. c. Spray Additive Tank Level Two level indication channels ar e provided in the control room. d. Containment Flood Level Level indication is provided in the control room on a per train basis. e. Encapsulated Tank for Containment Recirculation Sump Valves Water accumulation inside this tank is alarmed in the control room. This indicates leakage from the containment sump recirculation valves.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Core Cooling System Revision 15 Section 6.3 Page 35  6.3.5.5 Valve Position Status Monitoring Indication In addition to red (open) and green (closed) position indicating li ghts located above the associated control switch, the positions of the valves in Table 6.3-3 are also indicated by arrays of monitoring status lights at the main control board. These accident DBE status lights monitor the status of various engineered safeguard equipment following an accident and/or other Design Basis Events (DBE). The grouping of these lights prov ides the opera tor with a qui ck and easy identification of the required status of the important equipment for the various operational modes following an accident. The lights are grouped and are designed to go on upon these events occurring. Deviations are quickly identified by the light(s) failing to go on. Isolation valves are provided with red (open) and green (closed) position-indicating lights located above the associated control switch. These lights are actuated by valve motor-operator limit switches or stem-mounted switche s, as applicable.
A monitor light for each isolation valve that is on when the valve is in isolation position is provided in an array of monitor lights located on the MCB. For valves whose position indication lights are powered independently from the control circuit, control power availability is indicated by an indicating light at the MCB.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 1  6.4 HABITABILITY SYSTEMS The habitability systems include the control room complex which houses the controls to operate the plant safely under normal conditions and main tain it in a safe c ondition under all postulated accident conditions, as well as the supporting equipment, supplies and procedures for an emergency team of 10 operating and 25 technical support personnel. The Technical Support Center is in the control room complex and has the capacity to accommodate 25 technical support
 
personnel. 6.4.1 Design Bases The control room complex design provides sp ecific areas for continuous occupancy by the station operating personnel and the technica l support center personnel during postulated emergency conditions, including equipment and materials to which the operators may require access. The structural design of the control room complex together with its supporting systems will ensure access and occupancy under accident conditions without occupants receiving radiation exposures in excess of 5 rem total effective dos e equivalent (TEDE) for the duration of the accident. The control room complex has sanitary facilities and a first aid kit. The Control Room Ventilation System, which includes redundant emergency makeup air/filtration subsystems, will prevent the buildup of airborne particulates and radioactive iodines within the control room complex during an accident. Equipment, including necessary instrumentati on, controls and proce dures are provided at appropriate locations outside the control room with the design capability for (1) initiating prompt hot shutdown of the reactor and maintenance of the unit in a safe hot shutdown condition and (2) with a capability for subsequent cold shutdown of the reactor. The ventilation design will allow unit operation from the isolated control room without makeup or exhaust air for a period of time of at leas t thirty hours before th e carbon dioxide buildup would reach the concentration of 0.6 percent by volume.  (Carbon dioxide concentration would not exceed 3 percent within 290 hours. At 1.5 percent, basic performance and physiological functions are not affected, but concentrations of 3 percent by volume should not be exceeded. 
 
Per ASHRAE Handbook, 1978 Applications, Chapter 12.) Both control room makeup air intakes are located at a distance sufficiently away from each other and from major potential accident release points to minimize control room contamination in the event of a release of airborne radi oactivity or other toxic substances.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 2  The control room complex is maintained at a positive pressure of at least
  " w.g. with respect to outside and the adjacent cable spreading room. This positive pressure preven ts the infiltration of hazardous contaminants. Self-contained breathing apparatus is supplied within the complex to provide breathing protection in the presence of hazardous contaminants. The control room envelope boundary is designed and maintained so that unfiltered air inleakage is limited to  150 cfm during the emergency mode of operation. Redundant air conditioning systems are provided to ensure that the control room atmosphere is maintained within acceptable temperature and humidity limits for equipment operability and
 
personnel comfort. Meteorological information instruments are provided. These instruments aid the operator in deciding to isolate a contaminated makeup air intake if necessary. The habitability of the control room complex will not be compromised by the simultaneous occurrence of an SSE, a loss of offsite power and a loss-of-coolant accident. The active and passive components of the Control Room Ventilation System provide an environment that is consistent with the requirements of General Desi gn Criterion 19, which are to provide adequate ventilation air, maintain positive pressure to prevent infiltration of contaminants, temper the air for operator comfort, remove equipment heat, and maintain the control room below the guideline radiation dose of 5 rem TEDE. 6.4.2 System Design The habitability systems include: a. A concrete radiation shielded control room (Subsection 12.3.2) b. Heating, Ventilating and Air Conditioning Systems (Subsection 9.4.1)
: c. Redundant normal makeup air fans and associated discharge dampers d. Redundant emergency makeup air fans and associated HEPA/carbon/HEPA filtering units for airborne particulate and iodine removal from all makeup air and a portion of the total control room r ecirculation air under accident conditions e. Radiation and smoke monitored dual remote air inta kes (Subsection 12.3.4) f. Meteorological Information System (Section 2.3) g. Fire Protection System (Subsection 9.5.1)
: h. Full-face emergency breathing apparatus
: i. Communications (Subsection 9.5.2)
: j. Normal and emergency lighting (Subsection 9.5.3) k. Toilet facilities S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 3  l. Medical supplies m. Kitchen area. 6.4.2.1 Definition of Control Room Envelope The control room occupies the entire 75'-0" level of the Control Building, and includes the main control room area, computer room, Technical Support Center, office, conference room and library, emergency storage room, HVAC equipment room, kitchen and sanitary facilities, as shown in Figure 1.2-32. All controls, equipment and materials to which the control room operator would require access during an emergency are contained within this envelope, except for the makeup air intakes' manual isolation valves which are located on elevation 51'-6" of the Diesel Generator Building. 6.4.2.2 Ventilation System Design Details of the control room complex ventila tion and filtration systems are described in Subsections 9.4.1 and 6.5.1. An air flow diagram of the Control Room Ventilation System which identifies equipment, ducting, dampers, instrumentation and air flow rates for both normal and emergency modes is shown in Figure 9.4-1, Figure 9.4-2, Figure 9.4-3, Figure 9.4-25; major components and their major design parameters are included in Table 9.4-1 and Table 6.5-6. Two remote air intakes (east and west) are provided to furnish makeup air to the control room complex. The locations were selected to precl ude both intakes from being susceptible to accident-generated airborne radioactivity or toxic gases at the same time. The east makeup air intake is located approximately 380 feet northeast from the center of the Unit 1 containment structure. The west intake is located approximately 500 feet southwest of the Unit 1 containment structure (see Figure 1.2-1).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 4  The east air intake consists of a vertical 12-inch diameter carbon-steel pipe terminating in a tee-section. Both openings are pr otected by 1/2-inch square stainless steel wire mesh welded to the inside diameter of the respective opening. Prot ection of the intake against tornado missiles is provided by a reinforced concrete slab (see Figure 6.4-1). The we st air intake consists of a vertical 12-inch diameter carbon-steel pipe terminating in a 180 bend. The opening is protected by 1/2-inch square screen and security grating welded to the inside diameter. A portion of the west intake pipe above grade adjacent to the cooling tower wall and approximately a 2-foot vertical section just below grad e is not protected against tornado missiles. The unprotected 12-inch pipe, the vertical section above and below grade, and the 180 bend (see Figure 6.4-2) have an equivalent length of about 15 feet. This low effective target area results in a low mean value probability, calculated in the range of 2 x 10
-9 to 3 x 10
-7 per year, for tornado missile impact. Smoke monitoring equipment for the east intake, which alarms in the control room, is installed in a vault beneath the slab and within the fence enclos ed area. Radiation monitoring equipment for the east air intake, which alarms in the control room, is located within the pipe in the Diesel Generator Building. Radiation and smoke monitoring equipment for the west intake, which also alarms in the control room, is located within the pipe in the Diesel Generator Building. The Diesel Generator Building is located within the protect ed area security fence, and access to the building is controlled by security doors which are part of the station access control system. Environmental conditions within the east intake vault is maintained by convection heaters and a sump pump. The makeup air is transported via heavy wall carbon steel pipes from the remote air intakes to the control room HVAC equipment room (see Figure 9.4-1, Figure 9.4-2, Figure 9.4-3). The pipe enters the control room complex through the floor of one of the red undant filter units. The makeup air enters a tee located in a compartment of the filter unit upstream of the various filter components. The makeup air divides so that a por tion of the air discharges through an isolation damper then enters this compartment, while the remainder of the air enters the branch leg of the tee to the interconnecting ductwork of the redundant filter unit. The filter units and connecting ductwork are located within the control room envelope.
During normal operations, makeup air is drawn from both remote intakes and delivered to the control room complex by one of the two redundant normal makeup air fans. The normal makeup air fans and associated discharge dampers are lo cated outside the control room envelope on the 50'-0" elevation of the Diesel Generator Building. The air passes through medium efficiency prefilter(s) and electric heater(s) in both emergency filter units prior to discharging through an orifice into the control room HVAC equipment room. The prefilters remove dust and other airborne particulates and the heaters operate continuously to ma intain the carbon filter relative humidity at or below 70 percent thereby optimizing carbon adsorber efficiency and life.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 5  Under emergency conditions, makeup air is drawn from both remote air intakes and delivered to the control room complex by two fully redundant emergency filtration system fans. One hundred percent of the makeup air passes th rough the prefilter and heater and a HEPA-Carbon-HEPA filter configuration in either or both emergency filter units prior to discharging into the control room HVAC equipment room. In addition, approximately 2 percent of the total control room complex recirculation air flow, (i.e., including the air conditioning system flow rate) is drawn through the HEPA-Carbon-HEPA filter configuration in either or both emergency filter units. The HEPA filter(s) and carbon adsorber(s) are designed to remove radioactive airborne particulates and iodines (see Subsection 6.5.1 for filter design specifications). Under emergency conditions, the normal makeup air fans are automatically tripped off and their associated discharge dampers closed. The makeup air is transported to the control room via piping and backdraft dampers c onfigured in parallel which bypass the normal makeup air fans and dampers. The backdraft dampers preclude short cycling of air during normal operations. The exhaust air and supply air registers are adequately separated to precl ude recycling stale air and other noxious gases. The outside makeup air intake and the point of discharge for the control room exhaust are also adequately sepa rated to preclude recyc ling stale air and other noxious gases.
During normal operations, the modulating damper in the exhaust ductwork controls the amount of air being exhausted thereby maintaining a positive pressure within the control room complex. The damper is under the control of three static pressure sensing devices. The first pressure sensing point for the complex is in the HVAC equipment room, which is slightly lower in pressure than the remainder of the control room complex. The HVAC equipment room is maintained at least " w.g. above the outside atmospheric pressure, the second sensing point, and at least " w.g. above the cable spreading room, the third pressure sensing point. Under emergency conditions, the exhaust system isolates by automatic trip of the exhaust fan and closure of the modulating damper and redundant isolation damper. The emergency makeup air is adequate to maintain the complex at a pressure at least " w.g. greater than the outside atmospheric and cable spreading room pressures. Air is exhausted from the complex by exfiltration.
The following system components are powered or controlled from the Emergency Electrical Distribution System, to ensure operating power during all modes of operation:  The normal makeup air fans and associated discharge dampers  Emergency makeup air fans and associated discharge dampers Filter system air heaters Radiation Monitoring Instrumentation System  Exhaust system isolation dampers (Isolation Control System only).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 6  The normal makeup air fans are electrically "cros s-trained" with their associated discharge dampers, that is, the damper configured immediately downstream of the Train A fan is powered from the Train B vital bus and is controlled by the Train B control system. The damper configured immediately downstream of the Train B fan is powere d from the Train A vital bus and is controlled by the Train A control system. This design ensures isolation of the normal makeup air system (i.e., trip of each makeup ai r and/or closure of its associated discharge damper) under emergency conditions rega rdless of any single active failure. All automatic system dampers are pneumatically actuated and are designed to fail in the safe position (emergency mode configuration) on loss of instrument air. Controls for habitability system components are located in the control room complex. The normal makeup air fans and discharge dampers are controlled from the main control board (MCB). The components are manually actuated. Detection of high radiation in either remote intake will automatically isolate the normal makeup air subsystem. Actuation of the emergency makeup air and filtration subsystem fans will automatically isolate the normal makeup air subsystem. Failure of a vital bus or loss of instrument air will also isolate the system. The emergency makeup air and filtration subsystem fans and discharge dampers are automatically actuated upon detection of high ra diation in either remote intake or upon generation of a safety injection 'S' signal. The filtration subsystem can also be normally actuated from the MCB. The exhaust subsystem fan and discharge dampers are controlled from the MCB. The subsystem will automatically isolate upon loss of control room pressurization, detection of high radiation in either remote intake, or actuation of the filtration subsystem fans. The safety-related active components of the system are designed to seismic Category I requirements, and satisfy the design criteria of IEEE Standard 279 and other industry standards for electrical equipment, as defined in Subsect ion 8.1.4. No single failure of any of the active components will degrade the system's performance, as shown in Table 6.4-1. The safety-related passive components (i.e., pressure boundary) of the makeup air system, ducts and filters, are also designed to seismic Category I requirements.
All safety-related active and passive components of the system are contained in missile-protected buildings, are underground, or in the case of some piping, consist of such a small exposed area that the possibility of being struck by a tornado missile is negligibly small. Specifically, the mean value probability of a missile impacting the unprotected portion of the west air intake is in the range of 2 x 10
-9 to 3 x 10
-7 per year. No internally generated missiles which could impair the system's ability to perform its safety-related functions are credible. The system is designed to meet the intent of 10 CFR 50, Appendix A, Ge neral Design Criterion (GDC) 19, NUREG-0800 (Standard Review Pl an) Section 6.4 and Subsection 6.5-1, and Regulatory Guides 1.52, 1.78, and 1.196.
 
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 7  The system is monitored and regulated by a Control Room Envelope Habitability Program which ensures that the control room and its Operators are in a condition to assure public safety during all modes of operation. It provides measures to allow the control room habitability systems (CRHS) to maintain a habitable environment for operators under normal and abnormal conditions, and identifies compensatory measures if habitability is in question. 6.4.2.3 Leak Tightness The only openings in the control room envelope boundaries are th e sealed cable penetrations, two personnel accesses, the exhaust air duct with isolation dampers, and building construction joints. The primary personnel access way is a double-door configuration. The total complex outleakage with the exhaust subsystem isolated is calculated to be 165 cfm at a pressure of
(+) " w.g. The control room envelope (CRE) boundary may be opened intermittently under administrative control. This applies only to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings that ex ceed the allowable opening  size, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the ope rators in the CRE. This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when a need for CRE isolation is indicate
: d. Unfiltered air inleakage testing (tracer gas testing) will be performed, as part of the control room habitability program, at a frequency of 6 years to ensure that the CRE boundary is meeting the design requirements. 6.4.2.4 Interaction with Other Zones and Pressure-Containing Equipment
: a. Interaction with Other Zones There are no potential adverse interact ions between the control room complex ventilation zones and adjacent zones that could transfer toxic or radioactive contaminants into the control room. No ducts or ventilation piping from any other zone penetrates the control room envelope. Only the remote air intakes, makeup air piping and associated isolation valves, and normal makeup air fans and discharge dampers are located outside the control room envelope. This portion of the system is heavy carbon steel piping with primarily welded construction to minimize infiltration. The emergency makeup air and filtration subsystem fans are located within the control room envelope and
 
downstream of their associated filter components. This configuration maintains the filter units at a negative pressure precluding makeup air from bypassing the filter components under emergency conditions.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 8  The exhaust and static pressure control subsystem functions to maintain a positive pressure in the control room with respect to the outside and cable spreading room during normal operations. The pneumatically operated modulating damper in the exhaust ductwork controls the amount of air being exhausted and, thereby, maintains a positive pressure within the control room complex. The damper is under the control of th ree static pressure sensing de vices. The pressure sensing point for the complex is in the HVAC equipment room, which is slightly lower in pressure than the remainder of the control room envelope. The mechanical equipment room is kept at least " w.g. above the outside atmospheric pressure, the second pressure sensing point, and at least " w.g. above the cable spreading room at all times, the third pressure sensing point. Under emergency conditions, the exhaust duct and associat ed control room envelope penetration are isolated by redundant dampers configured in series. During normal operation, 1000 cfm of makeup air will be delivered to the control room complex. Approximately 145 cfm will be exfiltrated and the remaining 855 cfm will be exhausted. Under emergency conditions, approximately 600 cfm of makeup air will be delivered to the control room complex all of which will be exfiltrated. An SSE will not provide an exfiltration path for control room complex air that will negate the ability to maintain a positive pressure since the Diesel Generator Building and Control Building are seismic Category I structures. The intake air equipment and welded piping to the control room complex are also seismic Category I. b. Pressure-Containing Equipment The pressure-containing equipment in the Control Room Complex consists of refrigerant lines, Computer Room Halon system, fire extinguishers, instrument air operated equipment, and self-c ontained breathin g apparatus. The Computer Room Air Conditioning Refrigerant System is nonsafety-related, nonseismic Category I and contai ns 10 pounds of Refrigerant 22.
The Uniform Mechanical Code, 1976 Editi on, Section 1505 a llows 22 pounds of Refrigerant 22 per 1,000 cubic feet of occupied space. The control room complex envelope is 246,000 cubic feet. The refr igerant charge of 10 pounds in the Computer Room Air Conditioning System is well below the allowable 22 pounds per 1,000 cubic feet.
 
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 9  The computer room Halon 1301 system is capable of a total concentration of 5 percent of the computer room volume of 13,340 cubic feet. The release of the entire volume of Halon into the control room complex envelope would result in a concentration of 0.30 percent. The National Fire Protection Association Code, Section 12A, lists 5 to 7 percent concentration as producing minimal, if any, central nervous system effects for exposures of approximately five minutes' duration.
Fire extinguishers of various types (Halon, dry chemical, and CO
: 2) are installed in the Control Room complex. All of the fire extinguishers are Underwriters Laboratories listed and/or Factory Mutual approved. If all of the Halon fire extinguishers we re discharged in the Control Room, the total amount of Halon released would be less than the total amount of Halon that would be released by the Computer Room Halon system. The release of the Computer Room Halon system will not adversely affect Control Room habitability. Therefore, the release of the lesser amount of Halon contained within the fire extinguishers would also not affect the Control Room complex habitability. The operation of the dry chemical and/or CO 2 fire extinguishers would not affect Control Room habitability due to low volume of gas released from the fire extinguishers as compared to the total volume of the complex.
Control Room self-contained breathing apparatus (SCBA) - The release of the contents of any or all of the SCBA air packs or spare bottles into the Control Room atmosphere would not pose a detriment to Control Room habitability.
Instrument air operated components in the Control Room, primarily air operated dampers, have a normal air consumption rate. This air is released into the Control Room area. A review of these components has determined that the amount of air released is approximately 1 CFM. This is a contributor to the unfiltered inleakage assumed in Control Room dose analysis, [See Appendix 15C.] The makeup air and exhaust air systems will serve to dilute any gaseous concentration of refrigerants or Ha lon below the already safe levels. 6.4.2.5 Shielding Design The design basis loss-of-coolant accident (LOCA) establishes the shielding requirements for the Control Building. The control room shielding design is discussed in Subsection 12.3.2, and is evaluated from design bases LO CA source terms and doses which are presented in Subsection 15.6.5.4.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 10  The external walls and roof of the control room are 2-foot thick reinforced concrete. These shield thicknesses in conjunction with the habitability systems will limit the integrated dose to the operators to less than 5 rem TEDE, in conformance with General Design Criterion 19 of Appendix A of 10 CFR, Part 50. 6.4.3 System Operational Procedures 6.4.3.1 Normal Mode
: a. Normal Makeup Air Subsystem During normal plant operation, the control room normal makeup air subsystem is aligned to deliver approximately 1000 cfm of outside air from both remote intakes (500 cfm per intake). With one normal makeup air fan operating and its associated discharge damper open, the inta ke isolation valves are positioned to allow equal amounts of air to be drawn from the east and west intakes. The normal makeup air flows through the prefilter and heater for each emergency filter unit and discharges th rough an orifice into the HVAC equipment room. The heater for each unit operates continuously to limit humidity to less than or equal to 70 percent. The prefilters are periodically replaced when particulate buildup causes the differential pressure across the filters to increase to a predetermined
 
value. In the event normal makeup air fails or is isolated for reasons other than those delineated in Subsections 6.4.3.2 and 6.4.3.3, appropriate operator action will be taken to re-establish makeup air. If make up air is lost because of fan failure, the redundant normal makeup air fan and its discharge damper will be manually actuated. If makeup air is lost because of a vital bus outage or failure, or a loss of instrument air supply to the dampers, the emergency makeup air and filtration subsystem will be manually actuated. b. Emergency Makeup Air and Filtration Subsystem During normal plant operation, the emergency makeup air and filtration subsystem fans are idle and their associated discharge dampers are closed. Normal makeup air flows through each filter unit's prefilter and heater as discussed above. In the event this subsystem must be manually actuated during normal operation, it functions similarly to emergency mode operation. c. Exhaust and Static Pressure Control Subsystem During normal plant operation, the control room exhaust fan is operating and its discharge control damper modulates to maintain the control room complex at a pressure of at least (+) " w.g. with respect to the outside and adjacent areas. The redundant exhaust isolation damper remains fully open.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 11  d. Cooling/Recirculation Subsystem The non-safety related control room air conditioning subsystem will normally operate. However the, the safety related Trains may be placed in operation during normal plant operation. In the event of a malfunction in the non-safety related subsystem, or during a loss of offs ite power, one of two 100% capacity safety-related trains of control room ai r conditioning will be placed in service manually. Following a loss of off-site power with the non-safety related subsystem de-energized, one of the two redundant safety related Trains will automatically start via the emergenc y diesel generator load sequence. Subsection 9.4.1 provides a more detailed description of this subsystem and its operation. The unit heaters are not required to maintain the operation of the control room.
Redundant unit heaters are not provided. If the unit heaters should fail during operation of the control room in the wintertime, the space temperature may drop below the normal comfort temperature. However, heat loads generated internally by electrical equipment would preclude excessively low temperatures. 6.4.3.2 Emergency Mode
: a. Normal Makeup Air Subsystem Following an accident, when high radiation is detected in either remote air intake or when the emergency makeup air and filtration subsystem fans are actuated, the normal makeup air fans automatically trip off and their associated discharge dampers automatically close. The control systems for these fans and dampers are "cross-trained," that is, the discharge damper associated with the Train A fan is controlled by the Train B control loop and vice versa. This c onfiguration ensures isolation of the normal makeup air subsystem by fan trip and/or damper closure regardless of any single active failure. Detection of smoke in either remote intake is alarmed only. Operator action is required to initiate the filter recirc. mode. Operations may, at their discretion, manually isolate the smoke-contaminated intake and re-establish makeup air from the unaffected intake to the control room complex via the emergency makeup air and filtration subsystem (see Subsection 6.4.3.3).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 12  b. Emergency Makeup Air and Filtration Subsystem Following an accident, when high radiation is detected at either remote intake or upon generation of an 'S' signal, both redundant emergency makeup air fans and their associated discharge damper are automatically actuated. Although the redundant filter system fans are designed to operate coincidentally and stabley in their parallel configuration, Operations may, at their discretion, shut down one of the systems during the course of the accident. Each filter system may also be initiated manually from the control room at anytime. The subsystem may be initiated manually upon detection of smoke in either remote intake (see
 
Subsection 6.4.3.2). Each emergency makeup air and filtration subsystem has a nominal capacity of 1100 cfm. This capacity is comprised of 600 cfm makeup air and 500 cfm recirculation air. These system flow rates have been calculated assuming both remote intake isolation valves are open to a throttle position allowing for 300 cfm makeup air from each intake. Following an accident, a contaminated remote intake does not have to be manually isolated. Design base analyses indicate that the makeup air dilution factor (i.e., 50 percent makeup air from "clean" intake, 50 percent makeup air from contaminated intake) and the radi oactive particulate and iodine removal capacity of the filters together are adequate to maintain control room doses below allowable limits for the 30-day accident mitigation
 
period. c. Exhaust and Static Pressure Control Subsystem Detection of high radiation in either remote makeup air intake or operation of either emergency makeup air and filtration subsystem fan will automatically isolate the exhaust and static pressure control subsystem. Under emergency conditions, the exhaust subsystem remains isolated at all times. 6.4.3.3 Smoke Removal Mode The control room ventilation intake is provided with smoke detection capability to automatically alarm. Upon receipt of a smoke alarm from either remote intake, the operator will manually initiate the filter recirc. mode. The control room filter recirc. mode signal (CRFRM) will activate the emergency makeup air and filtration subsystem and isolate the normal control room makeup air and the exhaust and static pressure control subsystem. The HEPA filters associated with this system will remove smoke from the incoming air. Manual isolation of the smoke-laden intake can be accomplished by closing the appropriate 1-CBA-V9 or 2-CBA-V9 valve, locally. Additional venting of the control room could be accomplished by opening the doors and using
 
portable exhausters.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 13  6.4.3.4 Isolation Mode The system operational procedure will be the same as in emergency mode, discussed in Subsection 6.4.3.2. 6.4.4 Design Evaluations 6.4.4.1 Radiological Protection Radiological protection for the control room opera tors during accident cond itions is provided by the control room shield walls and habitability system. An evaluation of the protection offered by the shielding and habitability systems is presented in Chapter 15 (see Subsection 15.6.5.4, Radiological Consequences of a LOCA). 6.4.4.2 Toxic Gas Protection As stated in Subsection 2.2.3.1, no si gnificant quantity of toxic gase s is stored at any industrial facility in the vicinity of the site. No t oxic gases are transported by the Boston and Maine Railroad. The distance to the nearest highway is one mile, so it is unlikely that any toxic chemical spill along a road will endanger control room habitability and, in addition, the control room design and operator training offer inherent protection against the consequences of a spill in the site vicinity.
An evaluation of chemical hazards identified six chemicals onsite that had the potential to provide a hazard to control room habitability. The six chemicals are propane, hydrogen (trailers), sodium hypochlorite (N aOCl), carbon dioxide (refrigerat ed liquid), nitrogen (gas) and sulfur hexafluoride (SF6). These chemicals were evaluated and it was determined that none of the chemicals would affect maintaining the control room habitable if released under the postulated accident scenarios (Reference 4). This evaluation also addressed potential off-site chemical hazards. No off-site chemical hazards were identified that would affect control room habitability. As part of the Control Room Envelope Habitability Program, an updated list of potential chemical hazards both on and off-site is available.
Toxic gas protection for control room operators in not required since no potential chemical hazards have been identified either on or off-site that could potentially create a hazardous chemical environment in the control room. 6.4.5 Testing and Inspection During preoperational system testing, air systems ar e balanced to achieve design flow rates. In addition, operability of all equipment and control functions are verified.
Periodic verification of equipment operability and certain system parameters will be performed in accordance with plant Technical Specifications.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 14  Preoperational testing of the emergency filter units and their associated components will be performed in accordance with ANSI-N510 (1980). Subsequent periodic testing of the filter units will be performed in accordance with Technical Specifications which invoke certain ANSI-N510 (1980) in-place testing requirements and acceptance criteria. Air system pressure boundaries are leak tested following installation. A Control Room Envelope Habitability Program has been implemented. Four (4) types of tests are required to ensure Seabrook's Control Room is maintained habitable:  an integrated tracer gas test, an administrative evaluation, a Control Room Envelope pressure test and equipment TS surveillances. 6.4.6 Instrumentation Requirements 6.4.6.1 Normal Makeup Air Subsystem The controls for the normal makeup air fans and discharge dampers are located on the main control board (MCB). During normal operations, one of the two normal makeup air fans is running and its associated discharge damper is open. Isolation of the normal makeup air subsystem is automatically initiated upon detection of high radiation in either remote air intake or upon actuation of the emergency makeup air fans. Each remote air intake is provided with fully redundant radiation monitoring systems. The east intake radiation monitors are located at the intake. The west intake radiation monitors are
 
located on the intake piping in the Diesel Generator Building. Details of the RDMS system are provided in Subsection 12.3.4. The control scheme of the normal makeup air fans and dampers is "cross-trained" to ensure automatic isolation regardless of any single active failure, that is, detection of high radiation in either intake by Train A monitor or actuation of the Train A emergency makeup air fan will trip the Train A normal makeup air fan and close the discharge damper downstream of the Train B normal makeup air fan. Detection of high radiation in either intake by either Train B monitor or actuation of the Train B emergency makeup air fan will trip the Train B normal makeup air fan and close the discharge damper downstream of the Train A normal makeup air fan. This
 
"cross-trained" design also ensures that the normal makeup air subsystem automatically isolates on an outage or failure of either vital electrical bus.
Status lights are provided on the MCB for the normal makeup air fans and discharge dampers. 
 
Position indication is provided for the remote intake manual isolation valves. The following alarms are provided at the MCB:  Control room normal makeup air isolation  Loss of control room makeup air East air intake contaminated - smoke, radiation S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 15  West air intake contaminated - smoke, radiation. 6.4.6.2 Emergency Makeup Air and Filtration Subsystem The controls for the emergency makeup air fans and dampers are located on the MCB. The fans and dampers are automatically actuated upon detection of high radiation in either remote air intake or upon generation of an 'S' signal. The fans and dampers may also be manually actuated from the MCB. The filter unit heaters operate continuously.
 
Status lights for the emergency makeup air fans and discharge dampers are provided on the MCB. Grouped status lights ar e also provided for the fans.
Differential pressure across each filter component is indicated locally in the control room HVAC equipment room. The temperature for each filter unit is indicated locally. Relative humidity, differential pressure and air flow for each filter unit are indicated and recorded on the station computer.
The following alarms are provided at the MCB:  High filter unit differential pressure  High filter temperature High filter relative humidity High filter carbon monoxide levels (early fire detection)  High and low filter flow Filter isolation damper closed. 6.4.6.3 Exhaust and Static Pressure Control Subsystem The exhaust and static pressure control subsystem fan and discharge control damper are controlled from the MCB. During normal operations, the fan is running and the control/isolation damper modulates to maintain a positive control room pressure. This modulating feature is controlled by a differential pressure control loop. This control system senses pressure in the Control Room HVAC Equipment Room, Cable Spreading Room and the outside atmosphere. The damper is modulated automatically to maintain the HVAC Room at greater than or equal to " w.g. positive pressure with re spect to atmosphere and the Cable Spreading Room. The redundant exhaust isolation damper (CBA-DP-1058) is fully open during normal plant conditions. Loss of normal makeup air and/or loss of control room pressurization will close CBA-DP-28 and trip CBA-FN-15. Interlocks are provided so that isolation dampers CBA-DP-28, -1058 and exhaust fan CBA-FN-15 are is olated whenever a high radiation signal is present or fans 16A or 16B are running.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Habitability Systems Revision 13 Section 6.4 Page 16  Status indication for CBA-FN-15, CBA-DP-28, and CBA-DP-1058 is provided on the MCB.
Indication of differential pressure between the control room HVAC room and outside atmosphere is provided in the HVAC room. This differential pressure, as well as the differential pressure between the Cable Spreading Room and the HVAC room, is recorded on the station computer. The following alarms are provided at the MCB:  Low control room/outside atmo sphere differential pressure  Low control room/cable spreading room differential pressure  CBA-FN-15 tripped. 6.4.6.4 Cooling/Recirculation Subsystem All principal components of the safety related chilled water system (namely water chillers, chilled water pumps and air conditioning unit fans and dampers), except the chiller condenser exhaust fans are controlled from the MCB. The chiller condenser exhaust fans are controled from the control room air conditioning panel located in the control room air conditioning equipment room on elevation 75' within the control room pressure envelope. Additional instrumentation and control featur es are discussed in Subsection 9.4.1. The following alarms are provided at the MCB:  Control room high and low temperature  Computer room high and low temperature.
Safety related chiller trouble Condenser exhaust fan trip Condenser exhaust fan bypassed
 
====6.4.7 References====
: 1. N. Irving Sax, "Dangerous Properties of Industrial Chemicals," 5th Edition, Van Nostrand Reinhold, 1979. 2. "Handbook of Compressed Gases," Compressed Gas Association, Inc., Van Nostrand Reinhold, 1966. 3. "Hydrogen Chloride," Safety Data Sheet SD-39, Manufacturing Chemists Association, Washington, D.
C., Quoted in Reference 2; also extracts supplied by personal communication w ith Miss Mott of AICE. 4. "Evaluation of On-Site and Off-Site Toxic Chemicals - Control Room Habitability 2008 Update (SBC-1069)," Document No. 32-9098365-000 December 18, 2008. Areva NP Inc.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 1  6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS
 
====6.5.1 Engineered====
Safety Feature (ESF) Filter Systems The following engineered safety feature filter systems are atmospheric cleanup systems provided to remove fission products and to retain the radioactive material following a design basis accident. a. Containment Enclosure Emergency Air Cleaning System b. Fuel Storage Building Emergency Air Cleaning System c. Control room emergency make up air and filtration subsystem. These systems are secondary systems, as defined in NRC Regulatory Guide 1.52.
The following non-engineered safety-feature filter systems are described in other sections of the UFSAR: a. Primary Auxiliary Building Air Cleaning System (Subsection 9.4.3) b. Waste Processing Building Air Cleaning System (Subsection 9.4.4) c. Containment Structure Recirculating Air Cleaning System (Subsection 9.4.5) d. Containment Structure Purge Ai r Cleaning System (Subsection 9.4.5) The fans, ducting and dampers of the containmen t structure recirculating air cleaning system perform a post-accident H 2 mixing function (see Subsection 6.2.5). 6.5.1.1 Design Bases
: a. Containment Enclosure Emergency Air Cleaning System (CEEACS)
The Containment Enclosure Emergency Air Cleaning System is designed to maintain a negative pressure within the containment enclosure following an accident, to remove and retain airborne particulates and radioactive iodine, and to exhaust filtered air to the unit plant vent.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 2  1. The CEEACS is designed to maintain a negative pressure of greater than or equal to 0.25 inches of water, fo llowing a design basis accident, in the annular region defined by the containm ent structure and the containment enclosure, as well as in the additional building volumes associated with the electrical penetration areas, mech anical piping penetration area and engineered safeguard equipment cubi cles, so that any fission products leaking from these systems and from the primary containment will be
 
retained in these areas a nd eventually processed th rough the filt ers. In order to ensure a negative pressure of greater than or equal to 0.25 inches of water is maintained at the top of the containment enclosure for the entire range of design outside ambient temperatures, a negative pressure
 
greater than or equal to 0.685 inches of water has to be maintained at the 21' -0" elevation of the containment enclosure. The filter unit also accepts the discharge from the post-LOCA containment hydrogen purging duct, as discussed in Subsection 6.2.5.2. 2. The exhaust capacity is based on a conservative leak rate of 0.20 percent/day of the containment air mass at maximum internal pressure following a design basis LOCA as given in Table 6.5-7. Each containment enclosure exhaust fan is designed to exhaust at the rate of 2100 SCFM, which is equivalent to a volumetric inleakage rate of 325 percent/day from the containment structure to the containment enclosure annulus. 3. Sizing of the high efficiency particulate air filters (HEPA) and carbon adsorbers is based on the volumetric flow rate required to maintain the negative pressure in the containment enclosure annulus and connected
 
penetration and engineered safegua rd areas, and for fission product removal capability employing the cons ervative inventories given in Subsection 15.6.5. 4. The Containment Enclosure Emergency Air Cleaning System is a seismic Category I, Safety Class 2, system. b. Fuel Storage Building Em ergency Air Cleaning System
: 1. The Fuel Storage Building Emergency Air Cleaning System is designed to maintain a negative pressure of 0.25 inches of water within the Fuel Storage Building while in the irradiated fuel handling mode, to remove
 
and retain airborne particulates a nd radioactive iodine, and to exhaust filtered air to the unit plant vent following a fuel handling accident.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 3  2. The Exhaust Filter System is designed to remove and retain airborne particulate and radioactive iodine, and to exhaust the filtered air to the unit plant vent following a fuel-handling accident while either or both filters
 
are operating. 3. Sizing of the HEPA filter and carbon adsorbers is based on the volumetric flow rates required to maintain the required negative pressure in the Fuel Storage Building for both the normal fuel handling mode and the fuel handling accident mode, and for fission product removal capability employing the conservative invent ories presented in Subsection 15.6.5. 4. The Fuel Storage Building Emergency Air Cleaning System is a seismic Category I, Safety Class 3 system. c. Control Room Emergency Makeup Air and Filtration Subsystem The control room emergency makeup air and filtration subsystem is designed to supply makeup air from two remotely located intakes to the control room complex following an accident and/or a release of radiological contaminants or smoke, to maintain a positive pressure within the control room complex, and to remove and
 
retain airborne particulates and radioactive iodines from all makeup air and a portion of recirculated air. 1. The control room emergency makeup air and filtration subsystem is designed to maintain a positive pressure of greater than or equal to 0.125 inches of water in the control room complex relativ e to the outdoors and to the cable spreading room. The system will operate following a design basis accident or other abnormal operating scenarios such as smoke contamination of a remote air intake (see Section 6.4 for additional details). The control room complex occupies the entire 75'-0" elevation of the Control Building and includes the main control room area, computer room, technical support center, offices, conference room and library, emergency storage room, HVAC equipment room, kitchen and sanitary facilities, as shown in Figure 1.2-32. The positive pressure will preclude infiltration of radiological and other hazardous contaminants to maintain habitability of the complex for a safe plant operation and shutdown, as
 
necessary. 2. The makeup air portion of the nominal filter capacity (600 cfm) is conservatively greater than the minimum required ventilation rate for occupancy/breathing purposes (175 cfm) and the calculated complex outleakage at +0.125" W.G. (165 cfm).
The recirculation air portion of the nominal filter capacity (500 cfm) is based on a desired filter system decontamination factor (see Subsection 15.6.5).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 4  3. Sizing of the high efficiency particulate air (HEPA) filters and carbon adsorbers is based on the volumetric flow rate required to maintain the positive pressure within the control room complex, to satisfy ventilation requirements for occupancy/breathing purposes, and for fission-product removal capability employing the cons ervative inventories given in Subsection 15.6.5. 4. The control room emergency makeup air and filtration subsystem is a seismic Category I, Safety Class 3 system, except for some instrumentation which does not provide vital control or monitoring. 6.5.1.2 System Design
: a. Containment Enclosure Emergency Air Cleaning System The filter system consists of redundant filter trains, fans, dampers and controls and a common ductwork system. The air flow required to maintain a negative pressure in the Containment Enclosure Building is passed through demisters, which also function as prefilters, and through HEPA filters located both upstream and downstream of the carbon filter prio r to exhausting through the plant vent (see Figure 9.4-7 for an air flow diagram).
A ductwork cross-connection is provided be tween the two filter trains at a point between the downstream HEPA filter and th e fan inlet. Should the operating fan fail, this cross-connection will insure a continued air flow by manual startup of the redundant fan. Each redundant filter train is complete, separate and independent from both electrical and control standpoints. Each filter train fan is supplied power from an independent ESF power train source, whic h will furnish power to its fan during abnormal and post-accident conditions. The operation of mechanical equipment is controlled and monitored in the plant unit control room, as discussed in
 
Section 7.3.
The HEPA filters have a certified test efficiency of 99.97 percent based on DOP smoke test. For impregnated carbon filter efficiencies see Table 6.5-4. The evaluation of offsite effects due to potential accidents has been made in accordance with Appendix 15B, assuming minimum carbon filter efficiencies of 85 percent for organic iodines and 95 percent for elemental iodines for the conservative case. The carbon filters us e a deep bed design which provides a gas residence time of approximately 0.5 second.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 5  b. Fuel Storage Building Em ergency Air Cleaning System Separate redundant carbon filter exhaust systems are provided for filtering the building air prior to its exha ust to the unit plant vent.
One carbon filter train will be operated whenever irradiated fuel not in a cask is being handled. The filter unit, together with dampers and controls, will maintain the Fuel Storage Building at a negative pressure with respect to atmosphere so that any airborne particulate or radioactive iodine will be retained in the building and eventually processed through the filters (see Figure 9.4-4 for an air flow diagram). The Filter Cleanup System consists of redundant filter trains, redundant fans, ductwork, dampers and controls. Each filter train consists of demisters, which
 
also function as prefilters, heaters, medium efficiency filter, HEPA filters both upstream and downstream of the guard bed and carbon filters.
A ductwork cross-connection is provided to connect the two filter train systems at a point between the downstream HEPA filter and the fan inlet. Should the operating fan fail, this cross-connection will ensure continued air flow through a partially loaded or fully loaded filter by manual startup of the redundant fan. The cross-connection is sized so the temperature of a fully loaded filter bed will not rise above 200F (see Subsection 6.5.1.3c). Each redundant filter train is complete, separate and independent, from both electrical and control standpoints. Each filter train is supplied power to its fan during abnormal and post-accident conditions. The operation of mechanical equipment is controlled and monitored in the plant unit control room, as discussed in Section 7.3. The HEPA filter manufacturer's minimum test efficiency of 99.97 percent is based on the DOP smoke tests. The impregnated carbon filters have a manufacturer's rated minimum test efficiency of 97 percent for methyl iodide (CH 3I-131) and 99.9 percent for elemental iodine as delineated in Table 6.5-5. The carbon filter has a deep bed design having a gas residence time of
 
0.5 seconds. Prior to the start of handli ng of irradiated fuel not in a cask, one exhaust filter system is placed in operation, the normal exhaust system is shutdown, and the normal exhaust damper is closed. Therefore, the system will be in operation should a fuel handling accident occur.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 6  c. Control Room Emergency Makeup Air and Filtration Subsystem The filter system consists of redundant filter trains, fans, dampers, and controls. The redundant filter trains share a common suction makeup air piping system. The makeup air pipe penetrates the floor of a plenum integral to the Train A filter unit and upstream of the internal filter components. This plenum is connected to a systematically similar suction plenum integral to the Train B filter unit via ductwork. Both filter units and the connecting ductwork are located entirely within the control room complex. Only the makeup air pipe penetrates and enters the complex envelope. Each filter train consists of an isolation damper, a medium efficiency prefilter, an electric air heater, an upstream HEPA filter bank, a carbon adsorber section, a downstream HEPA filter bank, a backdraft damper, a fan, and a discharge control damper. Each filter train is also designed with an orifice plate downstream of the air heater and upstream of the first HE PA filter bank for bypass/recirculation air flow. During normal operations, when the filter system is idle, makeup air passes into the suction plenum of each filter, flows through the prefilter and heater, and discharges into the control room HVAC equipment room via th e orifice. Under emergency conditions when the filter system is operating, makeup air is drawn into each filter suction plenum, through the prefilter and heater, and then mixes with a portion of the recirculation air which enters via the orifice plate. This air mixture then flows through the HEPA-carbon-HEPA configuration, and is discharged into the HVAC equipment room vi a the filter train's associated fan and dampers. The electric heater for each filter operates continuously to ensure that the relative humidity within the carbon adso rber banks is equal to or less than 70 percent at all times. A pipe with a manual isolation valve is provided to cross-connect the two filter units at a point downstream of the carbon ad sorber sections.
The isolation valve is normally open / throttled to a preset position. During single train operation, this alignment will provide a small amount of air flow through the carbon adsorber section of the inoperable train. This air flow will remove decay heat satisfying fire protection concerns. The isolation valve will be closed for train isolation in the event a fire is detected in one of the units.
Each redundant filter train for both electrical and control requirements is complete, separate, and independent. Each filter train fan, air heater, and essential instrumentation and control components are powered from an independent ESF power train source, which will furnish power to the essential components during abnormal and accident conditions. Operation of the various system components is controlled and monitored in the plant unit control room, as discussed in Subsection 6.5.1.5c and Section 7.3.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 7  The HEPA filters have a certified test efficiency of 99.97 percent based on DOP smoke test. The carbon adsorber efficiencies are provided in Table 6.5-6. Filter efficiencies utilized to evaluate post-accident control room doses are provided in Appendix 15B. The conservative analysis assumes minimum filter efficiencies of 95 percent for organic and elemental i odines and 99 percent for particulate iodines. The carbon adsorber design provides an average gas residence time of 0.25 seconds. 6.5.1.3 Design Evaluation
: a. Containment Enclosure Emergency Air Cleaning System (CEEACS)
The containment enclosure exhaust filter trains are redundant, to insure the maintenance of a negative pressure in the containment enclosure and related areas and to insure cleanup of the exhaust air following an accident. All safety-related equipment and ductwork supports have been designed and seismically analyzed to
 
withstand and function through a Safe Shutdown Earthquake (SSE). The system is designed to limit offsite post-accident doses to values below those specified in 10 CFR 100 (see Subsection 15.6.5 for evaluation of system performance). A single component failure will not result in loss of function of this ESF system. In the unlikely event that an accident requiring filter operation occurs, both of the redundant filter train fans will be automatically started on the "T" signal (see
 
Drawing NHY-503515) to provide an air flow velocity of approximately 45 fpm through their associated carbon filter beds. In the further unlikely event of failure of one operating fan, the ductwork cross-connection will provi de redundant air flow from the redundant fan across the partia lly loaded or fully loaded filter bed.
See Section 6.2.3.3.a for a description of the analyses performed to demonstrate the capability of the system to draw down the Containment Enclosure Building to the design negative pressure within the design basis draw down time. HEPA filters and carbon adsorbers were tested at the expected accident environmental conditions for this secondary system. Results indicated no degradation of filtering efficiency. Subsection 15.6.5 analysis conservatively assumes lower efficiencies.
The systems are designed to meet the intent of Regulatory Guides 1.4 and 1.52. See Table 6.5-1 for a discussion relative to conformance with Regulatory Guide 1.52, Rev. 2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 8  b. Fuel Storage Building Emergency Air Cleaning System (FSBEACS)
The fuel storage building exhaust filter trains are redundant to insure cleanup capability and the ability to maintain a negative pressure following a fuel handling accident. All safety-related air handling equipment, equipment support and ductwork supports are designed to operate during and following an SSE. The system is designed to limit offsite post-accident doses to values not exceeding the requirements of 10 CFR 100 (see Chapter 15). Loss of one emergency exhaust filter train will not prevent the safety function from being performed. During fuel handling, only one set of filters and fan will normally be operating. In the unlikely event of an accident, the second set of filters and fan can be manually started to provide redundancy. The operating filter will provide an air flow velocity of approximately 40 fpm through its associated carbon filter bed. In the further unlikely event of failure of the operating fan, the ductwork cross-connection will provide redundant ai r flow across the partially loaded or fully loaded filter bed.
Note HEPA filters and carbon adsorbers have been tested at the expected accident environmental conditions for this secondary system. Results indicated no degradation of filtering efficiency; however, conservative parameters based on Regulatory Guide 1.25 were used in the conservative analysis in Subsection 15.7.4. The systems are designed to meet the intent of Regulatory Guides 1.25 and 1.52. See Table 6.5-2 for a discussion of conformance with Regulatory Guide 1.52, Rev. 2. c. Control Room Emergency Makeup Air and Filtration Subsystem The control room emergency filter trains are fully redundant to ensure the maintenance of a positive pressure within the control room complex, the provision of adequate supply makeup air for br eathing/occupancy purposes, and the filtration of this makeup air and a percentage of recirculation air following an accident or smoke contamination of an intake given any single active failure. All safety-related passive an d active components have b een designed and supported to withstand and function during and following a Safe Shutdown Earthquake (SSE). The system is designed to limit post-accident control room doses to values below those specified in 10 CFR 50, Appendix A, GDC 19 and Section 6.4 of the Standard Review Plan (see Subsection 15.6.5 for accident dose analyses).
 
Note All drawings referenced in this section were provided under a separate submittal to the NRC (see Section 1.7).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 9  In the unlikely event that an accident requiring filter system operation occurs, both of the redundant filter train fans and discharge control dampers will be automatically actuated upon generation of an "S" signal or detection of high radiation in either remote ai r intake (see Subsection 6.5.1.5c). The following analysis has been performed to demonstrate the capability of the system to maintain the control room complex at a positive pressure of 0.125 inches of water gauge: 1. Air Out-Leakage Analysis A calculation was performed to determine the control room complex envelope air out-leakage th rough various air flow path s such as electrical penetrations, concrete structure, construction joints, doors, and the worst-case exhaust isolation damp er. The analysis was developed utilizing vendor information, ASHRAE data and methodology, and analytical and experimental leak age data provided in "Conventional Buildings for Reactor Containment-NAA-SR-10100(1965)," issued by Atomics International, a division of North American Aviation Incorporated. The calculated maximum out-leakage is approximately 165 cfm at a control room complex positive pressure of 0.125" W.G. The nominal makeup air capacity of the emergency makeup air and filtration subsystem is 600 cfm. The Control Room Emergency Air and Makeup Filtration System is controlled by the Control Room Envelope (CRE) Integrity Program. This program ensures the control room and its Operators are in a condition to assure public safety during all modes of operation. It provides measures to allow the control room habitability systems (CRHS) to maintain a habitable environment for operators under normal and abnormal conditions, and identifies compensatory measures if hab itability is in question. d. Analysis of Heat Generation in Charcoal Beds A detailed analysis was performed to determine the highest attainable temperatures which could be reached by the charcoal beds in the Containment Enclosure System, Fuel Storage Build ing Filter System, and control room emergency makeup air subsystem due to decay heat from accumulated fission products resulting from a design basis accident. The filter loadings and other parameters used in the analysis for each filter system, as well as the results, are given below.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 10  1. Pertinent Parameters for Analysis Item CEEACS FSBEACS CREMAFS Initial Filter Temperature 100 F 100 F 105 F Thermal 0.55 Btu/ 0.55 Btu/ 0.55 Btu/
Conductivity hr ft F hr ft  F hr ft  F Specific Heat 0.242 Btu/lb F 0.242 Btu/ lb F 0.242 Btu/ lb  F Energy Absorption Coefficient 3.05/ft - 3.05/ft Charcoal Density 0.38 gm/cc 0.38 gm/cc 0.38 gm/cc Element Length 61 inches 61 inches *30 inches/  21 inches Width 66 inches 120 inches *24 inches/  38 inches Thickness 4 inches 2 inches
* 2 inches/  4 inches Number of Filter Elements 4 7
* 6/4
 
*1-CBA-F-38/1-CBA-F-8038 2. Filter Loading Isotope CEEACS FSBEACS CREMAFS I-131 3.3x10 3 6.3x10 2 1.89x10-1 I-132 4.1x10 0 nil 2.04x10
-1 I-133 3.4x10 3 6.8x10 1 3.66x10-1 I-135 6.0x10 2 5.0x10-2 3.16x10-1 S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 11  3. Results This analysis indicates that desorption temperatures will not be reached for the Containment Enclosure Emergency Air Cleaning System, the control room emergency makeup air subsystem and the fuel storage building emergency air cleaning system charcoal filter beds, even for the case of no air flow due to fan failure and unavailability of air flow vi a the ductwork cross-connections between the two redundant filter trains of each system. 6.5.1.4 Tests and Inspections
: a. Containment Enclosure Emergency Air Cleaning System Fan discharge ductwork was leak-tested during installation. Air system balancing and adjustment to design air flow was completed prior to system preoperational testing. The entire system underwent preopera tional testing as described in Subsection 14.2.11.
 
Periodic testing will verify the ability of the system to maintain the containment annulus at the required negative pressure. Pe riodic filter testing will be performed to meet the intent of Regulatory Guide 1.52, as clarified in Table 6.5-1. b. Fuel Storage Building Em ergency Air Cleaning System Fan discharge ductwork was leak-tested during installation. Air system balancing and adjustment to design air flow was completed during plant preoperational testing.
The installed systems underwent preope rational testing, as described in Subsection 14.2.11, prior to storage of fu el in the Fuel Storage Building. Prior to refueling operations, the ability of each fan to maintain a fuel storage building negative pressure greater than or equal to 0.25" H 2O will be verified. Periodic filter testing will be performed to meet the intent of Regulatory Guide 1.52, as clarified in Table 6.5-2. c. Control Room Emergency Makeup Air and Filtration Subsystems Fan discharge ductwork was leak-tested duri ng installation. Local leak testing is performed at potential leak paths on the makeup air piping (e.g., damper frame seals, cover plate gaskets, bolted flange seals, etc.). Air system balancing and adjustment to design air flow was completed prior to system preoperational testing.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 12  Acceptance and periodic filter testing will be performed in accordance with Regulatory Guide 1.52 except that the 1980 issues of ANSI N509 and ANSI N510 will be utilized (see Table 6.5-3).
Testing will be done following the Control Room Envelope Habitability Program.
Four (4) types of tests are required to ensure Seabrook's Control Room is maintained habitable:  an Integrated Tracer Gas Test (performed every six years), an Administrative Evaluation (performed every 18 months), a Control Room Envelope Pressure Test (performed every 18 months), and equipment TS surveillances (performed as speci fied by Technical Specifications).
6.5.1.5 Instrumentation Requirements
: a. Containment Enclosure Emergency Air Cleaning System This post-accident cleanup system is designed to function automatically upon receipt of an engineered safety featur e actuation "T" signal (containment isolation phase - A). Manual controls for the fans and dampers are also provided at the MCB (see Drawing NHY-503515). The containment enclosure (CE) pressure differential deviation is alarmed at the MCB. The discharge flow of the system is indicated locally, as well as recorded at the MCB. High and low flow alarms are also provided at the MCB. A high pressure drop condition across each filter train is recorded and alarmed at the MCB. In addition, the pressure differential of each filter component is indicated locally. Carbon adsorber discharge air high temperature and inlet air high humidity are alarmed at the MCB. Air temperature upstream of carbon adsorber and downstream of second HEPA filter is indicated
 
locally.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 13  b. Fuel Storage Building Em ergency Air Cleaning System The operation of the Fuel Storage Building (FSB) Emergency Air Cleaning System is controlled and monitored in the plant control room (see Drawing NHY-503543). FSB supply air is controlled through DP 13A/B (see Drawing NHY-503541). During fuel handling operations, the FSB is maintained at a slightly negative pressure with respect to the atmosphere through the FSB exhaust system (see Drawing NHY-503542). FSB hi gh and low differential pressure is alarmed at the MCB. Also, the FSB temperature and relative humidity are monitored, and high deviations are alarmed at the MCB. Each filter train is
 
monitored for differential pressure and high deviations are alarmed at the MCB.
Each cleaning unit is provided with a temperature switch which alarms high temperature on the MCB. Temperature switches for automatic control of cleaning unit heaters are also provided. Independent low-flow instrumentation is provided to alarm at MCB the discharge air flow through each of the redundant air cleanup filter units (see Drawing NHY-506452). In addition, the differential pressure across each individual filter unit is indicated locally. Local temperature indicators are also provided both upstream and downstream of the carbon filters. c. Control Room Emergency Makeup Air and Filtration Subsystem The emergency makeup air and filtration subsystem fans and discharge dampers are also controlled from the MCB. In the auto mode, CBA-DP-27A and 27B will open upon receipt of a remote intake high radiation signal or a safety injection ("S") signal. Opening of these dampers will automatically start associated fans CBA-FN-16A and 16B. A high radiation signal or starting fans 16A and 16B will trip CBA-FN-27A, 27B, and 15 and isol ate dampers CBA-DP-53A, 53B, 28 and 1058. Each filter air heater operates continuously. High and high-high temperature controls are provided. The high-high temperature control requires local/manual reset. Status lights are provided at the MCB for the fans and discharge dampers. Status lamps are also provided for the fans in the Accident Monitoring Instrumentation arrays. Differential pressure indication across each filter component is provided locally. High differential pressure across each filter unit generates a VAS alarm at the MCB. The temperature for each filter unit is indicated locally. High temperature generates a VAS alarm. Relative humidity for each filter unit is recorded on the station computer. High and low air flow generate VAS alarms. Two carbon monoxide detectors per filter (one each at the inlet and outlet of the carbon adsorber banks) provide fo r early fire detection. These devices monitor the filtration systems and generate an alarm in the control room.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 14  6.5.1.6 Materials Listed in Table 6.5-4, Table 6.5-5 and Table 6.5-6 by commercial name, quantity and chemical composition are the materials used in or on the filter systems. Extremes in temperature or radiation that could promote radiolytic or pyrolytic decomposition of materials are not applicable to the Containment Enclosure Building or the fuel storage for normal or postulated accident conditions. 6.5.2 Containment Spray System This section presents a deta iled description of the fission product removal function of the Containment Building Spray (CBS) System.
6.5.2.1 Design Bases The spray system provides for iodine removal to limit the consequences of a LOCA to the limits of 10 CFR 100 by providing a rapid reduction in containment elemental iodine concentration following a LOCA. This greatly reduces the amount of iodine that can leak from the primary containment into the containment enclosure structure. The combination of iodine removal by the sprays and the containment enclosure exhaust filters limits the offsite ra diological consequences resulting from the design basis accident. These consequences and the system performance are discussed in Subsection 15.6.5. Table 15.6-17, prepared in accordance with Regulatory Guide 1.4, lists the fission products that the CBS system is designed to remove during a LOCA. A discussion of the extent to which credit is taken for fission product removal by the CBS system is included in Subsection 15.6.5. The system uses the principles of hydrodynamic equilibrium to provide a reliable means for buffering the spray solution with an iodine rem oval agent. The spray rings are designed to provide a high degree of spray coverage to maximize the spray effectiveness for iodine removal. 6.5.2.2 System Design (for Fission Product Removal)
Iodine removal inside the containment following a LOCA is accomplished by 3010 gpm of spray (assuming only one train available) with a bor ic acid-sodium hydroxide spray solution. The design details of the spray system are presented in Subsection 6.2.2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 15  The spray system initially takes suction from the refueling water storage tank (RWST) and continues to add a fresh spray additive solution into the containm ent until the inventory of the RWST is depleted. Upon a low level in the RWST, the suctions of the spray pumps are automatically transferred to the containment sump. The time of transfer is a function of the number of emergency core cooling pumps and containment spray pumps in operation. At design flow rates and minimum pumps in operation (3575 gpm injection and 3010 gpm spray) the transfer to the recirculation mode takes place in 53.7 minutes, assuming a 10-second delay for injection and a 62 second delay in spray flow. With maximum safeguards in operation (6600 gpm injection and 6600 gpm spray), the transfer to the containment sump occurs at approximately 26 minutes, assuming a 10-second delay for injection and a 62 second delay in spray flow. The spray pumps remain in operati on for as long as is necessary to control the containment pressure within the required limits. The maximum delay in delivery of the NaOH solution to the spray nozzles is 2.47 minutes. The chemical additive for the spray is stored in the spray additive tank (SAT) located adjacent to the RWST. When a containment spray actuation signal occurs, two valves in parallel lines provide redundant flow paths for supplying the chemical additive to the RWST and spray pump suction. The chemical additive is stored as a 20 wt percent NaOH solution. The chemical spray additive flows by gravity into a mixing chamber in the RWST. The design is such that the pH of the solution leaving the mixing chamber is averaging between 9.0 and 9.6. The total amount of chemical supplied will result in a containment sump liquid having a pH range between 8.8 and
 
===9.4 during===
the recirculation phase.
The spray is delivered to the containment through 198 SPRACO 1713A nozzles per flow train, each having a flow rate of 15.2 gpm at containment design pressure and 40 psi differential operating pressure. The nozzles produce a drop size spectrum with a conservatively estimated volumetric drop diameter of 1000 microns.
The nozzles are spaced to provide a uniform spray pattern across the containment cross section. The nozzles of the redundant spray trains are uniformly spaced between the nozzles of the other spray train, so that either subsystem will provide uniform coverage. The SPRACO 1713A
 
nozzles have been used extensively in other nuclear plants and various iodine removal experiments. The location of the spray nozzles in the dome of the containment is shown on Figure 6.2-76 and Figure 6.2-77. The flow weighted average fall height of the spray drops is a minimum of 134 ft.
for headers 1 and 2. For headers 3 and 4, the average fall height is 144 ft. The operating floor is at an elevation of 25 ft.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 16  The following is a tabulation of important spray iodine removal parameters: Spray fall height 134 ft Total containment free volume 2.715x10 6 ft 3 Sprayed containment free volume 2.310x10 6 ft 3 Containment pressure, design 52.0 psig Containment temperature, design 296 F Spray flow rate per train design/minimum 3010/2808 gpm Spray solution pH, minimum
 
===9.0 Average===
9.0 to 9.6 Spray additive flow rate Maximum safety injection 240 gpm  Minimum safety injection 150 gpm These parameters are chosen to minimize credit taken for iodine removal by the spray system and provide a conservative basi s for calculating offsite doses. Two headers are provided for each 3010 gpm train. Headers No. 1 and 3 are supplied by one of the CBS pumps while Headers No. 2 and 4 are supplied by the other pump. Header No. 1 has 118 nozzles spaced approximately 3apart. Fifty-nine of these nozzles are oriented vertically downward. The remaining 59 are oriented at 45 on the inside of the header pointing inward and downward. Header No. 3, which is the companion header to header No. 1, has 80 nozzles spaced approximately 4 30' apart. Forty of these nozzles are at a 45 angle pointing downward toward the containment wall, 16 are at a 45 angle pointing downward and inward, 20 nozzles point directly downward and 4 point horizontally toward the containment vertical centerline. Header No. 2 contains 162 nozzles spaced approximately 2 15' apart. Eighty are mounted at a 45 angle pointing downward toward the containm ent wall, 41 point directly downward and 41 are at a 45 angle pointing inward. Header No. 4, which is the companion header to header No. 2, contai ns 36 nozzles spaced approximately 10 apart. Eighteen are mounted at a 45 angle pointing downward toward the containment wall. The remaining 18, point directly downward.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 17  Figure 6.2-84, Figure 6.2-85 and Figure 6.2-86 show the spray drop distribution, accumulated number percentage, and accumulated volume percentage for the SPRACO No. 1713A nozzle. These distributions are the basis for evaluation of iodine removal effectiveness. Approximately 85 percent of the containment net free volume is covered by spray. Figure 6.2-80, Figure 6.2-81, Figure 6.2-82, and Figure 6.2-83 show plan and elevation views of the expected spray patterns. Table 6.2-80 lists the regions of sprayed volumes in the containment and the volume of each sprayed region. It can be seen from Figure 6.2-80 and Figure 6.2-81 that more than 95 percent of the cross section of the containm ent at the operating floor (elevation 25') is covered by spray. 6.5.2.3 Design Evaluation The Containment Spray System has been evaluated by both conservative and realistic models. Subsection 15.6.5 presents the parameters used in both the conservative and realistic analyses and the results of these analyses.
The initial spray solution will be boric acid buffered with sodium hydroxide to a pH of approximately 9.0 to 9.6. Sump pH after recirculation commences will be above 8.5 to assure high iodine partition factors a nd prevent re-evolution of iodine from the spray solution.
Using the method described by Parsly (Reference
: 1) a fresh spray drop is calculated to be less than 10 percent saturated during its fall from the spray nozzles. The removal of iodine should therefore be controlled by gas phase resistance. Steam condens ation and drop coalescence have been discussed by Pasedag and Gallagher (Reference 2) and by Parsly (Reference 1). Pasedag shows that the total reduction in drop mass transfer surface area over a 100-foot fall due to these phenomena is only about 10 percent. Parsly recommends for calculational purposes that a mean drop diameter of 25 percent larger than the actual mean be used to conservatively compensate for coalescence and size distributi on. The calculations in Subsection 15.6.5 assume a mean drop diameter of 1250 microns, which conservatively estimates the iodine removal half-life. Table 6.2-80 lists the sprayed regions in the containment. The spray headers are designed to directly spray approximately 85 percent of the containment free volume. In addition the unsprayed areas are designed to allow, as much as possible, the free exchange of air with the sprayed area to prevent the pocketing of fission products.
Due to the relatively low surface to volume ratio in the reactor containment it is not expected that wall effects will play a large part in the removal of iodine by spray solution. No wall effects are included in the calculation in Subsection 15.6.5.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 18  The mathematical models used in calculatin g iodine removal by spray are presented in Subsection 15.6.5. These models assume that the spray removal function is effective throughout the containment sprayed volume and that the effectiveness is constant. Credit for mixing between the containment sprayed and unsprayed volumes is described in Subsection 15.6.5, in conjunction with the discussion of the multi-compartment spray model used to calculate iodine removal rate constants. 6.5.2.4 Tests and Inspections Subsection 6.2.2.4 discusses the tests and inspections performed to verify the functional capability of the spray system and components, including active valves, pumps, and the spray nozzles, to deliver the required flow for containment heat removal. Demonstration of those drawdown characteristics of the SAT and RWST which provide flow at the proper pH for effective iodine rem oval is also discussed in Subsection 6.2.2.4. The spray additives are sampled on a periodic basis to verify their continued state of readiness.
 
6.5.2.5 Instrumentation The system is provided with instrumentation and control to allow the operator to monitor the status and operation of the Containment Spray System from the control room to allow the automatic or manual initiation of the injection mode of operation. The instrumentation details are presented in Subsection 6.2.2.5. The SAT is steam heated, with the steam flow controlled by a temperature controller, which modulates a pneumatic-operated flow control valve. Low temperature of the tank is alarmed at the main control board (MCB). The tank level is indicated at the MCB. Refueling water storage tank instrumentation is de scribed in Subsection 6.3.5. Each of the SAT discharge valves may be opened either manually by the operator or automatically upon receipt of the containment spray signal. The valve positions are indicated at the MCB, above the control switches and also as a part of a separate status monitoring indication system.
6.5.2.6 Materials The spray solution is stored in the spray additive tank (SAT), and the RWST holds a boron solution. Both tanks are constructed of stainless steel which has been shown to be resistant to chemical attack by the respective stored solutions of 2400 to 2600 ppm bor on (as boric acid) and 19 to 21 percent by weight of sodium hydroxide. The spray solution is not susceptible to radiolytic or pyrolytic decomposition under the conditions anticipated in the post-accident environment. The corrosion properties of the spray solution are discussed in detail in Subsection 6.2.2.2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 19  6.5.3 Fission Product Control Systems This section provides a discus sion of the operation of all fission product control systems following the design basis loss-o f-coolant accident (LOCA). 6.5.3.1 Primary Containment The primary containment is a reinforced concrete structure with a hemispherical dome. The concrete thicknesses of the vertical wall and the dome are 41/2 feet and 31/2 feet respectively. A welded steel liner plate is anchored to the inside face of the concrete to function as a leaktight membrane. Figure 1.2-2, Figure 1.2-3, Figure 1.2-4, Figure 1.2-5, Figure 1.2-6 show typical sections through the containment and the locati on of the hydrogen purge system equipment and ductwork. Table 6.5-7 summarizes the operation of the containment following a LOCA. 
 
Calculation of offsite doses following a LOCA is discussed in Subsection 15.6.5. For the conservative case, 0.20 percent of the containment air mass is assumed to leak to the containment enclosure during the first 24 hours following an accident, and 0.10 percent/day thereafter. For the realistic case, 0.10 percent of the containment air mass is assumed to leak during the first 24 hours and 0.05 percent/day thereafter. The offsite doses presented in Subsection 15.6.5 are based on a containment free volume of 2.715x10 6 ft 3. The iodine removal function and effectiveness of the Containment Building Spray System is discussed in Subsection 6.5.2. This system will begin operation within 62 seconds after receipt of a LOCA-generated "P" signal, as described in Subsection 6.2.1.1. The function of the containment isolation systems is discussed in Subsection 6.2.4. No credit is taken for iodine removal by the Containment Online Purge System since it is only operated intermittently during normal plant operation and will isolate on a containment isolation signal if operating at the onset of an accident. Radiological consequences of this occurrence are addressed in Subsection 15.6.5. The combustible gas control system hydrogen recombiners, permanently located inside the containment, are designed to be operational with in seven days following a DBA as described in Subsection 6.2.5. Should both recombiners be inoperable for 50 days after the DBA, hydrogen concentration in the containment will be controlled by use of th e hydrogen purge line to the plant vent via the Containment Enclosure Emergency Cleanup System described in Subsection 6.5.1.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Fission Product Removal And Control Systems Revision 15 Section 6.5 Page 20  6.5.3.2 Secondary Containment The secondary containment is comprised of a reinforced concrete cylindrical structure with a concrete hemispherical dome, the Engineered Safety Features (ESF) equipment cubicles, and the pipe and electrical penetration areas. The release of airborne contamination following an accident due to leakage into the containment enclosure (the secondary containment) is controlled by a filtered exhaust system which maintains a subatmospheric pressure in each subcompartment. The Containment Enclosure Emergency Cleanup System (CEECS) is th e only fission product control system in the secondary containment. This system directs a nominal 2000 cfm of charcoal filtered exhaust air to the plant vent. Actual exhaust flow rate will be the sum of the primary containment leakage (see Subsection 6.5.3.1) and the inleakage from the surrounding environment. See Section 6.2.3.3.a for a description of the analyses performed to demonstrate the capability of the system to draw down the Containment Enclosure Building to the design negative pressure within the design bases draw down time. Iodine removal efficiency of the charcoal beds is assumed to be 85 percent for organic iodide and 95 percent for inorganic iodine for the conservative case; for the realistic case, the iodine removal efficiency is assumed to be 95 percent for organic iodines and 99 percent for inorganic iodines.
The air recirculation system for the ESF cubicles serves only for coolin g and provides no fission product control function; therefore, no mixing fraction was assumed. The gross volume of each compartment in the containment en closure is listed in Table 6.5-7. Anticipated and conservative assumptions concerning the fission pr oduct control functions of the secondary containment and the time sequence of events assumed in offsite dose estimates are
 
discussed in Chapter 15. Figure 9.4-5, Figure 9.4-6, Figure 9.4-7 and Figure 9.4-8, show each secondary containment area, the ventilation system associated with that area and the locations of exhaust intakes. There is no non-ESF filter systems used to cont rol pressure and fission product release in these areas. 6.5.4 References
: 1. Parsly, L.F., "Design Considerations of Reactor Containment Spray Systems -
Part VII, A Method for Calculating Iodine Removal by Sprays," ORNL-TM-2412 - Part VII, Feb. 1970. 2. Pasedag, W.F. and Gallagher, J.L., "Dr op Distribution and Sp ray Effectiveness" Nuclear Technology , Vol. 10, April 1971.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES In-Service Inspection Of Class 2 And 3 Components Revision 13 Section 6.6 Page 1  6.6 IN-SERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS The in-service inspection program for Class 2 and 3 component s is defined in the Technical Specifications. In general, this program meets all the requiremen ts of the ASME Code, Section XI, Edition and Addenda as requ ired in 10 CFR 50.55a. The references to ASME Section XI contained in paragraphs 6.6.6 are from the 1983 Edition Summer 1983 Addenda. 6.6.1 Components Subject to Examination The welds selected for examination in Class 2 piping systems have been determined in accordance with the requirements of the 1974 Edition of ASME Section XI with Addenda through Summer of 1975 as allowed or re quired by 10 CFR 50.55a, Paragraph (b)(2). All Class 2 components that do not meet the exemption requirements of IWC-1220, except the containment spray piping beyond the last downstream valve, will be examined in accordance with the requirements of IWC-2000, using the methods listed in Table IWC-2500-1 or the alternatives allowed by the Code Case entitled "Alternative Rules for Examination of Class 2 Piping." Containment spray piping beyond the last downstream valve is exempted from in-service inspection for the following reasons: a. These lines are normally empty and are required to function only in the unlikely event of a major LOCA or a main steam line break. b. These lines are not subject to fluctuating stresses, so that propagation of cracks due to defects in the pipe is unlikely. c. Spray piping welds are examined by ra diography during fabrica tion, so that large defects that could cause catastrophic failure during system operation will be
 
detected and repaired during fabric ation and installation of the piping. d. This piping contains hundreds of open nozzles so that splits or cracks which do not directly cause catastr ophic failure would not have a marked effect on the function of the system. e. System pressure is low, so that pressu re stresses in the piping would be unlikely to cause failure in the presence of small defects. All Class 3 components shall be examined per the requirements of Subsection IWD.
Safety classifications of all com ponents are presented in Subsection 3.2.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES In-Service Inspection Of Class 2 And 3 Components Revision 13 Section 6.6 Page 2  6.6.2 Accessibility The design and arrangement of Class 2 and Class 3 components provides adequate clearances to conduct the required examinations at the Code-required inspection interval. In general, access has been provided for examination of welds by pr oper layout of piping and equipment. Where surface or volumetric examination is required, insulation has been designed to be quickly and easily removed and replaced. Inaccessible Class 3 lines have been provided with taps to permit pressure decay tests in lieu of visual examination, as permitted in IWD-2000. Provisions for pump tests, as required by IWP-2000, have been made by incor porating suction and discharge pressure-measuring devices or flow measurement orifices adjacent to each safety class pump. Clearances around the pumps have been maintained to permit access for rotation speed and vibration measurements. Provisions for valve tests, as required by IWV-2000, have been incorporated by including valve position indicators in the design, as required, and by maintaining clearances for access to the valves. 6.6.3 Examination Techniques and Procedures In general, where volumetric examination is required to be performed, manual ultrasonic techniques are employed rather than radiography. Where equipment conf iguration, layout, or other considerations, make it undesirable to perform a manual ultrasonic examination, an automatic technique is employed. Radiography is employed if ultrasonic examination fails to give meaningful results. For components of the reactor coolant pressure boundary, see Subsection 5.2.4. Where surface examination is required, carbon steel components are typically examined by the magnetic particle method, and stainless steel components by the liquid penetrant method. The liquid penetrant method may be used when component size, geometry or space restriction prevents magnetic particle examination of carbon steel components. The steam generator tubing is examined by th e eddy current method (see Subsection 5.4.2.5). All examination techniques are qualified to Sec tion V or Section XI of the Code, using qualified personnel. 6.6.4 Inspection Intervals The inspection schedule for Class 2 and Class 3 components is in accordance with subarticles IWC-2400 and IWD-2400, and is included in the Technical Specifications. 6.6.5 Examination Categories and Requirements The examination categories and requirements for Class 2 components are in agreement with Section XI, Article IWC-2000. For Class 3 components, the requirements of Section XI, Article IW D-2000, are complied with, as applicable.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES In-Service Inspection Of Class 2 And 3 Components Revision 13 Section 6.6 Page 3  6.6.6 Evaluation of Examination Results
: a. Evaluation of Class 2 examination results will comply with the requirements of IWC-3000 of Section XI which applies the rules of IWB-3000. In general, indications detected during in-service examination that exceed the acceptance standards of IWB-3000 of the 1983 Edition, with addenda up to and including summer 1983, will require repair in accordance with IWC-4000 of Section XI. b. Repair procedures for Class 2 components comply with the requirements of IWC-4000 of Section XI. c. Evaluation of Class 3 components is consistent with IWD-3000 of Section XI which also applies the rules of IWB-3000. Visual examinations of Class 3 components consist of evaluations for leakage and integrity of structural attachments. Defects will be evaluated per IWB-3000 and repairs performed in accordance with the requirements of IWD-4000 of Section XI. In general, evaluation of leakage in Class 3 piping will be consistent with the
 
intent of IWA-3000, since defects resul ting in perceptible leakage are not acceptable. d. Defects in Class 3 pressure boundary components will be removed or reduced to acceptable size by grinding, cutting, or drilling.
If the defect and the repair do not encroach on the minimum wall, then no repair welding will be required. Repaired surfaces will be smoothly blended into the surrounding material with no discontinuities. If the defect or its removal encroaches upon the minimum wall thickness, the area will be built up to the minimum wall thickness by welding, and the surfaces smoothly blended into the surrounding surface. If the built-up material thickness of carbon st eel components that have be en repaired by welding is greater than 3/4", post-weld heat treatment is required. 6.6.7 System Pressure Tests The system pressure testing program for Class 2 components complies with the requirements of IWC-5000 of Section XI. The system pressure testing program for Class 3 components complies with the requirements of IWD-5000 of Section XI.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES In-Service Inspection Of Class 2 And 3 Components Revision 13 Section 6.6 Page 4  6.6.8 Augmented In-Service Inspection to Protect Against Postulated Piping Failures As stated in Section 3.6(B).2.1.a.4, for main steam and feedwater piping penetrating containment, no breaks were postulated between the first pipe whip restraint inside the containment and the five-degree restraint outside containment. To protect against postulated piping failures, this piping is subject to augmented inservice inspection as defined in the Risk-Informed ISI Break Exclusion Region (BER) evaluation (EE-07-035). The augmented lines are: MS-4000-02-30" FW-4606-03-18" MS-4000-41-30" FW-4606-04-16" MS-4001-02-30" FW-4607-03-18" MS-4001-41-30" FW-4607-04-16" MS-4002-02-30" FW-4608-03-18" MS-4002-37-30" FW-4608-04-16" MS-4003-02-30" FW-4609-03-18" MS-4003-37-30" FW-4609-04-16" S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES Main Steam Isolation Valve Leakage Control System (BWR)
Revision 8 Section 6.7 Page 1  6.7 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM (BWR) Not applicable to Seabrook.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 1  6.8 EMERGENCY FEEDWATER SYSTEM
 
====6.8.1 Design====
Bases The Emergency Feedwater (EFW) System provides the capability to remove heat from the Reactor Coolant System during emergency conditions when the Main Feedwater System is not available, including small LOCA cases. The EFW System operates over a time period sufficient to cool down the Reactor Coolant System to temperature and pressure levels at which the Residual Heat Removal System can operate. Unde r certain design basis scenarios, condensate storage tank inventory may be depleted prior to affecting the transition to RHR cooling. However, sufficient steam generator inventory will exist to maintain cooling until the RHR system can be placed in operation. For Station Blackout, the EFW turbine-driven pump will operate during the four-hour coping duration to cool down and maintain the secondary side pressure at about 250 psig (see Section 8.4.1).
During all other modes of plant operation, including startup, hot standby, and normal operation up to full power load, the EFW system has zero flow. The EFW system is designed to meet the following safety-related functional requirements: a. A malfunction or single active failure of a system component or nonessential equipment does not reduce the performance capabilities of the system. b. The functional performance of system components is not affected by adverse environmental occurrences, abnormal operational requirements, and off-normal conditions such as small breaks in the Reactor Coolant System or the loss of offsite power. c. System components and piping have sufficient physical separation and shielding to protect against the effects of internally and externally
 
generated missiles. d. The functional performance of the system is not affected by pipe whip and jet impingement that may result from high or moderate energy piping breaks or cracks. e. The system possesses diversity in motive power sources so that the system performance requirements are met with either power source. f. The system design precludes the occurrence of fluid flow instability during normal plant operation and during upset or accident conditions.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 2  g. Provisions are included to verify co rrect system operation, to detect and control system leakage, and to isolat e portions of the system in case of excessive leakage or component malfunctions. h. The system is capable of automatica lly initiating flow upon receipt of a system actuation signal. The system is also capable of manual actuation to provide protective action and for operational testing independent of the automatic signal. Single failure of the manual circuit will not result in loss of the system function. i. The system design possesses the capability to automatically terminate flow to a depressurized steam generator, while providing flow to intact steam generators. j. For Station Blackout analysis, no other active or passive failures, other than those causing the Station Blackout , nor other design basis events such as seismic or line breaks are assumed to occur (see Section 8.4.1). The Emergency Feedwater System is designed in accordance with ASME Code Section III, Class 3; IEEE Standards 323-1974 and 344-1975, Class 1E; and Seismic Category I requirements. System components are located within Seismic Category I structures and are thereby protected against effects of natural phenomena. The design of the EFW system wa s reviewed subsequent to the issuance of the NRC's March 10, 1980 letter to near-term operating license app licants (TMI-2 Task Action Plan, NUREG-0737, Item II.E.1.1). The review addressed the following areas: a. Subsection 10.4.9 of the Standard Review Plan b. Branch Technical Position ASB 10-1
: c. Generic short-term and long-term requirements applicable to the EFW system design, and operating procedures d. A reliability evaluation of the EFW system as outlined in NUREG-0611. An item-by-item discussion of the EFW system's compliance with each requirement was provided by applicants letter to NRC, dated July 27, 1982.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 3  6.8.2 System Description Upon loss of normal feedwater flow, the reactor is tripped, and the decay and sensible heat is transferred to the steam generators by the Reactor Coolant System via the reactor coolant pumps or by natural circulation when the pumps are not operational. Heat is removed from the steam generators via the main condensers or the main steam safety and/or steam generator atmospheric relief valves. Steam generator water inventory is maintained by water makeup from the Emergency Feedwater System. The system will supply feedwater to the steam generators to remove sufficient heat to prevent the over-pressurization of the Reactor Coolant System, and to allow for eventual system cooldown. The Emergency Feedwater System is comprised of two full-sized pumps (one motor- and one turbine-driven) whose water sour ce is the Condensate Storage Ta nk (CST). Su ction lines are individually run from the CST to each pump. A common EFW pump recirculation line discharges back to the Condensat e Storage Tank. This return line functions for recirculation pump testing and ensures minimum flow to prevent pump damage for any system low-flow operating condition. Both pumps feed a common di scharge header, which in turn supplies the four emergency feed lines. The common discharge header includes normally open gate valves between each branch connection to provide isolation in the ev ent of a pipe break or for maintenance. Each emergency feed line is connected to one of the main feedwater lines downstream of the feedwater isolation valve. Each main feedwater line enters the containment through a single penetration and feeds a single steam generator. For a diagram of the Emergency Feedwater System see Figure 6.8-1 and Figure 6.8-2.
Additional redundant pumping capability is provided by the startup feed pump in the feedwater system. The head and capacity curves for these pumps, which are plotted on the same sheet, show that the startup pump has sufficient capacity to serve as backup for the emergency feedwater pumps (see Figure 6.8-3). The startup feed pump (and steam generator wet-lay-up pumps) discharge line inside the Emergency Feedwater Pump Building and Main Steam and Feedwater Pipe Chase are seismically supported. All connections from thes e lines to other plant piping include normally closed valves. Valve and pump operation is administratively controlled. During normal plant operation, these lines are not pressurized. Valves V156 and V163 are normally closed, and are furnished with a motor operator.
 
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 4  Figure 10.4-6 and Figure 10.4-7 also shows a va lve in the branch line from the EFW turbine-driven pump suction line for use as a connection for emergency makeup from the condensate storage tank to the spent fuel pool or from the fire protection system to the condensate storage tank. This valve is normally closed, and is administratively opened. Additional makeup to the CST is available from the demineralized water storage tank (DM-TK-259) using the portable cooling tower makeup pump (SW-P-329) via demineralized water system valves DM-V-677, V-800, and V-490 (see Figure 9.2-15). A dedicated 194,000 gallons of demineralized water is maintained in th e lower half of the condensate storage tank for the exclusive use of the Emergency Feedwater System. For a description of the condensate storage facility see Subsection 9.2.6. Makeup to the tank is provided by the Demineralized Water Makeup System (see Subsection 9.2.3).
The motor-driven pump and pump controls are powered from an emergency bus. The startup feed pump is also capable of being powered from an emergency bus (see Subsection 8.3.1.1.b.9.(a)), and diesel generator cap acity is available to start this pump while carrying the maximum load listed in Table 8.3-1 (See Subsection 8.3.1.1.e.6). Steam fo r the turbine-driven pump is supplied from either of two main steam headers via branch lines connected upstream of the main steam isolation valves. Each branch lin e includes an air-operated, fail-open valve. The common EFW steam header contai ns one air-operated fail open valve (see Subsection 10.3.2.5). A summary of pump data is provided in Table 6.8-1. The branch lines to each steam generator include a manual gate isolation valve, two motor-operated flow control valves, a flow venturi, and a flow orifice. The flow control valves are normally in the open position when the system is not operating and are automatically closed during system operation in the event of a pipe break. These valves can be operated remotely as described in Subsection 6.8.5 to control steam generator water level. Two valves in series are provided for redundancy and are powered from different trains. Each valve is also provided with a handwheel to permit manual operation. The ope n position of the flow control valves for system limiting conditions will be set to insure the minimum required flow of 470 gpm to three steam generators and a minimum total flow of 650 gpm to four steam generators with one EFW pump operational. 6.8.3 Safety Evaluation The Emergency Feedwater (EFW) System components, instrumentation, and power supplies are sized and designed with sufficient redundancy to maintain the system's safety-related functions under all credible accident cond itions. The combination of one turbine-driven pump and one motor-driven pump provides a diversity of power sources to assure delivery of feedwater under emergency conditions.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 5  The system has been designed to provide the requ ired flow following a single active failure and a loss of offsite power and has been evaluated for a Station Blackout event (see Section 8.4.1). The common discharge header is not normally pressurized during normal plant operation. Figure 10.4-8 and Figure 10.4-9 shows the Main Feedwater System, the tie-in of the EFW and four individual stop-check valv es. These check valves (FW-V76, -V82, -V88 and V94) prevent backflow of feedwater from the main feedwater system or the steam generator (SG). Steam supply for the turbine-driven emergency feedwater pump is from either of two main steam headers via branch lines connected upstream of the main steam isolation valves. The branch lines contain normally closed, fail-open ai r-operated EFW isolation valves MS-V393 and MS-V394. These valves also serve a containment isolation function. These valves are controlled by Class 1E 125V DC solenoids. Branch isolation valve MS-V393 is controlled by a
 
Train A solenoid valve and the alternate branch isolation valve, MS-V394, is controlled by redundant Train A and Train B solenoid valve. The common EFW steam supply line contains a normally closed, fail-open air-operated valve (MS-V 395). This valve is controlled by both Train A and Train B Class 1E solenoi ds. Isolation valves MS-V393 and MS-V394 open in response to an EFW initiation signal to admit steam to MS-V395. Valve MS-V395 is timed to sequentially
 
open following opening of either MS-V393 or MS-V394. The sequential opening of MS-V395 allows for pressurizing the EFW steam header to discharge accumulated condensate via system drains prior to introducing steam to the turbine governor valve. Approximately 45 seconds
 
following an EFW initiation signal, MS-V395 is full open and supplying steam to the mechanical hydraulic turbine governor valve. The east and west EFW steam supply branches and the common header each contains a condensate drain pot and steam trap arrangement. The common header drain pot contains a constant steam vent line. The composite drain assemblies provide for minimizing accumulated steam header condensate during an EFW System standby, startup, and operational condition. The outside containment Steam Generator Blow down isolation valves close automatically whenever an EFW pump is in operation. Blow down isolation during EFW pump operation is a required support function that preserves Steam Generator and Condensate Storage Tank inventory. The outboard SB isolation valves are fail closed, air operated valves which are controlled by independent A-train and B-train pilot solenoid valves. The close signals for EFW pump operation derive from either position limit switches on MS-V393 (Train A) and MS-V394 (Train A and Train B), or motor-driven EFW pump breaker position.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 6  The EFW supply has two safety-grade flow control valves for each steam generator. One valve in each supply is powered by the A train emergency power source and the other valve is powered from the B train. The primary (or normal) flow control valves for the A and C steam generator will be powered by the A train with B and D steam generators' valves powered by the B train.
Backup flow control valves will be powered from the opposite emergency power train. These valves can be controlled from either the main control board or the remote shutdown panel using safety-grade controls. The five valves in the EFW pump discharge header are furnished with gear operators so that a concern for power diversity is not applicable.
The design and operation of the EFW system has been reviewed regarding the occurrence of hydraulic instabilities, characterized as water hammer. The EFW system is connected to the main feedwater system through stop-check valves outside the containment. The flow regulating valves in each EFW line are normally open, and are sized to pass the required flow under accident conditions. The only action required to establish EFW flow is to start the pumps. One pump has sufficient capacity to furnish the required flow to the Nuclear Steam Supply System. An analysis of the EFW system has established that its function and performance are not affected by the common causes for loss of flow resulting in water hammer, such as pump trip, or rapid valve closure. A pressure transient in the main feedwater system resulting from a pipe break, pump trip, and/or valve closure wo uld be dissipated before flow is established in lines to each SG. A trip of one pump will not affect the capability of the other pump to provide flow to the intact SGs. The only automatic valve closure in the EFW system would occur in the line to a faulted SG. During operation of the EFW system, the plant operators can initiate any changes in flow to each SG, as required. The EFW pumps' supply and recirculation line piping runs from nozzles on the Condensate Storage Tank (CST) to the EFW pumphouse. The CST nozzles and adjacent piping are protected by a seismic Category I structure which is part of the CST enclosure and tornado-missile shield.
The piping is routed underground and runs below grade into the EFW pumphouse, also a seismic Category I structure.
The EFW pump recirculation line to the CST, is designed to ASME Code Section III, Class 3 seismic Category I requirements. Valves FW-V346 and FW-V347 are administratively opened for EFW pump-surveillance testing and as required ensuring minimum EFW pump flow during system operation.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 7  The water lines to the oil cooler are designed to seismic Category I requirements. Water lines from the oil cooler are designed to ASME Code Section III, Class 3, seismic Category I requirements. The breakdown orifice in the line to the oil cooler limits the flow to 2-3 gpm. This flow was considered in sizing the pump capacity. In the unlikely event of pipe failure, this flow will easily be handled by the pump room floor drains. An accident analysis for this system in conjunc tion with the loss of the Main Feedwater System is provided in Chapter 15. A failure analysis of the Emergency Feedwater System following a feedwater pipe break is provided in Table 6.8-2. 6.8.4 Tests and Inspections Prior to initial plant startup, the Emergency Feedwater System is hydrostatically tested in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section III, Class 3, and pre-operationally test ed as described in Chapter 14. Periodic testing in accordance with Technical Specifications will be performed during normal plant operation. During periodic surveillance testing of the EFW pumps, manual valve alignments will be required. Only one EFW pump will be tested at a time. Because each EFW pump is capable of providing 100 percent of required flow, full system flow requirements will be available at all times. Automatic indication of EFW pump inoperable st atus is provided as discussed in subsection 7.1.2.6. 6.8.5 Instrumentation Requirements and Controls The Emergency Feedwater System will be actuated automatically on loss of offsite power, low-low level in any of the steam generators, safety injection signals or an ATWS mitigation system actuation signal. The engineered safety feature actuation system details are presented in Section 7.3. The ATWS mitigation system is disc ussed in Section 7.6. Manual controls for the turbine-driven pump steam supply valves are located at the main control board (MCB), as well as at the remote safe shutdown (RSS) panel. For the motor-driven pump, the controls are located at the MCB and in the switchgear room. The suc tion and discharge pressures of both pumps are indicated locally. Pump discharge pressure and CST level indication are provided on the MCB. Low suction pressures are alarmed at the MCB.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 8  Flow indications for all four individual emergency feedwater lines are provided. Safety grade flow orifice instrumentation readouts are displayed at the MCB. The instruments are powered from the safety grade inverters - A and C steam generators on the Train A inverter, and B and D steam generators on the Train B inverter. These instrumentation channels meet or exceed the requirements for Design Category 2 instrumentati on as provided in Subsection 7.5.5. The design details of the accident monitoring instrumentation are presented in Table 7.
5-1. Flow venturi are also provided in each emergency feedwater line.
Two of the four flow venturi instrumentation readouts are displayed at a RSS panel, and the remaining two flow venturi instrumentation readouts are displayed at a second RSS panel.
A sustained high-flow c ondition in any of the lin es is indicative of a line break. A break isolation is incorporated so that the affected line will be isolated by automatically closing the motor-operated valves on high flow signals from redundant flow instrumentation. Break isolation is required to conserve CST inventory and to provide the required flow to the intact steam generators. High flow alarms are also prov ided to alert the operator to this condition. The break isolation system is designed so that a single failure will not prevent break isolation or prevent emergency feedwater flow to at least two steam generators. Manual override provisions are also incorporated at the MCB as well as at the RSS panels, along with the open/close valve position indication. Each of the motor-operated control valves in each branch line is provided with fully independent power supplies, instrumentati on, and controls to ensure that at least one of the valves in each branch line can be closed when needed. All eight valves can be operated from the MCB. Four of the valves, one in each branch, can also be operated from an RSS panel and the remaining four valves, one in each branch, can be operated from a second RSS panel. To transfer control to the RSS pane l, remote/local selector switches on the RSS panels (CP-108A/B) and RSS disabling panels (CP-450A/B) must be transferred to the local position. The associated control switches are located on the MCB and the R SS panel so that the valv es normally used to control flow have the same train assignment as the flow indication for the steam generator (i.e., A and C SG are normally controlled with the A tr ain control valves and B and D SG with B train valves). Thus, complete redundancy is provided to control flow or to isolate any steam generator in the event of pipe breaks. The review of the emergency feedwater flow indication was performed as part of the detailed control room design review (DCRDR). As part of this effort, a determination of the needed characteristics of this display to support the emergency operating procedures (EOPs) was made. A comparison of the needed characteristics against the available instrumentation was made, and no deficiencies were found.
Each EFW pump recirculation line contains a normally closed motor-operated valve powered from its respective train.
SEABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Emergency Feedwater System Revision 15 Section 6.8 Page 9  Associated remote-manual control switches are lo cated at the MCB and at RSS panels CP-108A and CP-108B for EFW pumps P-37A and P-37B respectively. A flow orifice and associated instrumentation are provided in the common pump discharge recirculation path to the CST. This instrumentation is provided to permit periodic testing of the pumps to verify proper head-flow characteristics during periodic pump testing and to monitor recirculation flow during normal operation. Alarms ar e also provided to indica te that either train of the Emergency Feedwater System is inopera tive. The design features of the bypass and operable status alarm system which provide system level indications in compliance with Regulatory Guide 1.47 are presented in Subsection 7.1.2.6.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Design Guidelines For  In-Service Inspection Of Piping Systems Revision 8 Appendix 6A Page 6A-1  APPENDIX 6A DESIGN GUIDELINES FOR IN-SERVICE INSPECTION OF PIPING SYSTEMS The information in this Appendi x was not revised, but has been extracted from the original FSAR and is provided for historical information. Current desi gn guidelines for in-service inspection of piping systems are controlled by the Design Control Program.
 
==1.0 INTRODUCTION==
 
To verify the safety of a nuclear power plant on a continuing basis, the Owner must establish a program of in-service examination and testing such as is defined by Section XI of the ASME Boiler and Pressure Vessel Code. This program will include volumetric, surface and visual examination of welded joints in safety-related equipment and piping and operability assurance testing of safety-related pumps and valves.
It will be necessary for the nuclear power plant designer to provide access to the items to be examined, and for this reason, he must be cognizant of the clearance requirements for in-service inspection. It is intended that this general design guide provide the designer with the information required to guarantee the necessary access. 1.1 WELD PREPARATION FOR INSPECTION In order to provide assurance that the results of volumetric examination, especially ultrasonic data, are interpretable, piping welds should be prepared in such a way as to minimize random reflections and provide standardized configurations. Figure 7 describes weld preparation for In-service Inspection. 2.0 ACCESS REQUIREMENTS FO R IN-SERVI CE INSPECTION
 
===2.1 GENERAL===
REQUIREMENTS Examination of welds can be conducted either manually or by using automatic devices. The clearance requirements vary according to the method used. By means of special probes, the mechanized inspection can be conducted in much tighter areas than the manual inspection where much larger personnel access is required. The method used is also dependent on the environmental conditions in the area. Provision shall be made to ensure adequate access and work space for personnel and equipment in order to meet code requirements for in-service inspection. Unless otherwise specified, manual inspection will be assumed where there are no physical limitations preventing manual access; mechanized inspection will be assumed where there are physical limitations preventing manual access. Where direct manual operations are required, clearance shall be provided for the head and shoulders of a man within arm's length of the surface of the object to be inspected.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Design Guidelines For  In-Service Inspection Of Piping Systems Revision 8 Appendix 6A Page 6A-2  2.1.1 Access Requirements for Volumetric Examination
: a. Mechanical Inspections The standard butt weld inspection device and its associated tracks, used for mechanized inspection require a longitudinal clearance of 261/2 inches and a radial clearance of 81/2 inches around the entire longitudinal access (Figure 1). All insulation and obstructions shall be removable from the clearance area to perform the examinati ons. Personnel access is needed on one side to place the device on the tracks. Ideally, the weld should be inspec table from both si des, but certain situations may restrict the access to enable inspection from one side only. Figure 2 shows some such situations and calls for the minimum clearances needed to perform inspections meeting code requirements. The clearances specified are for mechanized inspections. b. Manual Inspection For pipe-to-pipe circumferential welds, the minimum longitudinal clear access on either side of the weld center line shall be 2T + 4 inches (6 inches minimum), where 'T' is the pipe wall thickness (Figure 3). For pipes less than 12 inches nominal size, a radial clearance of at least 24 inches shall be provided on one side and a minimum of 8 inches on the other sides. Radial clearance of 8 inches is required to operate the inspection equipment. For pipes greater than 12 inches nominal size, personnel radial clearance of 24 inch es shall be provided on two opposite sides of the pipe. Access on the two sides is needed since a technician is unable to reach all weld surfaces around large diameter pipes from one side. Such clearance requirements should be an important consideration for parallel piping runs. The clearance envelope needed for such runs shall be as shown in Figure 4.
Pipe-to-nozzle and pipe-to-fitting welds requiring clearances shall be as shown on Figure 5. Branch pipe or fitting connections located near circumferential butt welds requiring ultrasonic inspection must be so located that a clear distance equivalent to 2T + 4 (6 inches minimum) inches exists between the toe of the branch weld and center line of the circumferential butt welds. 2.1.2 Access Requirements for Surface Examination For surface inspection the longitudinal access shall be sufficient for placement of magnetic particle probes or dye penetrant brushed on the weld surface. A practical minimum length of access is 6 to 8 inches.
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Design Guidelines For  In-Service Inspection Of Piping Systems Revision 8 Appendix 6A Page 6A-3  2.1.3 Access Requirements for Visual Examination Piping systems subject to visual in-service inspections shall be designed with access to welds which shall allow direct visual access with or without mirrors.
For direct observations, the eye shall not be more than 24 inches away from the weld surface with the line of vision making at least a 30 degree angle with the surface to be examined for remote observations. Clearances shall be provided for the use of optical devices. The examination shall be carried out in a light that shall be sufficient to resolve a
  " line on 18% neutral gray card. 2.1.4 Access Requirements for Removable Insulation Piping systems with insulation shall have clearances for removal and storage of insulation materials around the welds to be examined. Insulation normally comes in two sections for pipes up to 25 inches O.D. and in three sections for sizes over 25 inches. For insulation comprised of
 
two sections, the clear area around the insulation should be at least one-half the diameter of the pipe, plus two times the insulation thickness plus an allowance for any lifting devices which might be placed on the insulation. This clearance must be provided on at least one section of the insulation since the second half ma y be slid around the pipe for it s removal. For three section insulation, the clearance area should be at least one-quarter the diameter of the pipe plus two times the wall thickness of the insulation plus an allowance for any required lifting devices. For insulated pipes with longitudinal welds on either sides of the circumferential welds, pipe insulation shall be removable for a minimum distance of 15 inches on either side of the circumferential weld edge. These requirements are illustrated in Figure 6. 3.0 OTHER LAYOUT CONSIDERATIONS
 
===3.1 PROHIBITED===
CONFIGURATIONS Certain system configurations which place fittings adjacent to each other are undesirable and prohibited from the in-service inspection point of view. Piping design should preclude the following fitting configurations: a. Valve to valve b. Valve to tee
: c. Valve to reducer
: d. Tee to tee
: e. Elbow to elbow
: f. Reducer to elbow
: g. Valve to elbow
: h. Nozzle to elbow 3.2 WELD SPACING The spacing between circumferential welds shall be 2T + 2 inches (6 inches minimum). Straight sections or spool pieces between the fittings shall be provided. The minimum spool piece length shall be 2T + 4 inches (6 inches minimum).
S EABROOK STATION UFSAR ENGINEERED S AFETY FEATURES Design Guidelines For  In-Service Inspection Of Piping Systems Revision 8 Appendix 6A Page 6A-4  Branch pipe or fitting connections located near circumferential butt we lds requiring ultrasonic inspection must be located so the distance equivalent to 2T + 4 inches exists between the toe of the branch weld and center line of the circumferential butt weld. This distance, however, shall not be less than 6 inches.
The designer shall review the pipe hanger, supports , seismic and pipe whip restraints to assure that this hardware does not interfere with the access already provided. No permanently installed part of any of these attachments shall be closer than 15" to the nearest circumferential weld. Removable parts shall not be installed closer than 2" from any circumferential weld. 3.3 PIPE CHASES AND PIPE RACKS Where the piping is run in a pipe chase or groups of pi pes banked together, care should be taken to assure that the piping to be inspected is located in the bank such that access can be acquired without having to remove other piping. An exam ple would be a group of piping several tiers deep suspended from above. The piping requiring inspection should be in the bottom tier to permit access and inspection from the floor or platforms below already designed into the plant. This would keep the amount of miscellaneous platforms and temporary scaffolding to a minimum. 4.0 IN-SERVICE TESTING REQUIREMENTS FOR PUMPS AND VALVES
 
===4.1 GENERAL===
REQUIREMENTS The applicable edition of the ASME Code for pump and valve testing is the 1983 edition with addenda up to and including Summer, 1983.
SEABROOK STATION UPDATED FINAL SAFETY ANALYSIS REPORT C HAPTER 6 ENGINEERED SAFETY FEATURES  TABLES  6.1(B)-1 Principal Engineered Safety Features Metallic Materials 6.1(B)-2 Coated Areas on Major Equipment and Structures inside Containment 6.1(B)-4 Other Organic Materials in Containment 6.1(N)-1 Engineered Safety Feature Materials 6.1(N)-2 Protective Coatings on Westinghouse Supplied Equipment Inside Containment 6.2-1 General Information and Initial Conditions 6.2-2 Engineered Safety Feature Systems Information 6.2-3 Containment Passive Heat Sink Data 6.2-4 Passive Heat Sink Material Properties 6.2-5 Break Locations Postulated for Loss
-of-Coolant Accident Analysis 6.2-6 Spectrum of Breaks Postulated for Loss
-of-Coolant Accident Analysis 6.2-7 Single Active Failures Postulated for Loss
-of-Coolant Accident Analysis 6.2-8 Calculated Peak Containment Pressure and Peak Containment Temperature Following Loss
-of-Coolant Accidents 6.2-9 Loss-of-Coolant Accident Chronology for Various Break Locations 6.2-10 Initial Plant Operating Power Levels Postulated for Main Steam Line Break Analysis 6.2-11 Spectrum of Ruptures Postulated for Main Steam Line Break Analysis 6.2-12 Single Active Failures Postulated for Main Steam Line Break Analysis 6.2-13 Calculated Peak Containment Pressure and Peak Containment Temperature Following Full Double
-Ended Guillotine Rupture in Main Steam Line
 
6.2-14 Calculated Peak Containment Pressure and Peak Containment Temperature Following Small Double
-Ended Guillotine Ruptures in Main Steam Line 6.2-15 Calculated Peak Containment Pressure and Peak Containment Temperature Following Ruptures in Main Steam Line 6.2-16 Postulated Breaks for Subcompartment Analysis 6.2-30 Mass and Energy Release Rates Following Spray Line Rupture 6.2-31 Pressurizer Compartment Vent Path Description
- Full Double
-Ended Spray Line Break 6.2-32 Pressurizer Compartment Nodal Description and Peak Nodal Differential Pressure - Full Double
-Ended Spray Line Break 6.2-33 Mass and Energy Release Rates - Surge Line Double
-Ended Guillotine 6.2-34 Pressurizer Skirt Cavity Vent Path Description
- Full Double
-Ended Surge Line Break 6.2-35 Pressurizer Skirt Cavity Nodal Description
- Full Double
-Ended Surge Line Break 6.2-36 Cases Analyzed and Tables with Results 6.2-37 Double-Ended Pump Suction Guillotine (Maximum Safety Injection)
- Slowdown Mass and Energy Releases 6.2-38 Double-Ended Pump Suction Guillotine (Minimum Safety Injection)
- Blowdown Mass and Energy Releases 6.2-39 0.6 Double
-Ended Pump Suction Guillotine
- Blowdown Mass and Energy Releases 6.2-40 Three-Foot Squared Pump Suction Split
- Slowdown Mass and Energy Releases 6.2-41 Double-Ended Hot Leg Guillotine
- Blowdown Mass and Energy Releases 6.2-42 Double-Ended Cold Leg Guillotine
- Slowdown Mass and Energy Releases 6.2-43 Double-Ended Pump Suction Guillotine (Minimum Safety Injection)
- Reflood Mass and Energy Releases 6.2-44 Double-Ended Pump Suction Guillotine (Maximum Safety Injection)
- Reflood Mass and Energy Releases 6.2-4 5 0.6 Double
-Ended Pump Suction Guillotine
- Reflood Mass and Energy Releases 6.2-46 Three-Foot Squared Pump Suction Split
- Reflood Mass and Energy Releases 6.2-47 Double-Ended Cold Leg Guillotine
- Blowdown Mass and Energy Releases 6.2-48 Double-Ended Pump Suction Guillotine (Minimum Safety Injection)
- Post-Reflood Mass and Energy Releases 6.2-49 Double-Ended Pump Suction Guillotine (Maximum Safety Injection)
- Post-Reflood Mass and Energy Releases 6.2-50 0.6 Double
-Ended Pump Suction Guillotine
- Post-Reflood Mass and Energy Releases 6.2-51 Double-Ended Pump Suction Guillotine (Minimum Safety Injection)
- Mass Balance 6.2-52 Double-Ended Pump Suction Guillotine (Maximum Safety Injection)
- Mass Balance 6.2-53 0.6 Double-Ended Pump Suction Guillotine - Mass Balance 6.2-54 Three-Foot Squared Pump Suction Split
- Mass Balance 6.2-55 Double-Ended Hot Leg Guillotine
- Mass Balance 6.2-56 Double-Ended Cold Leg Guillotine
- Mass Balance 6.2-57 Double-Ended Pump Suction Guillotine (Minimum Safety Injection)
- Energy Balance 6.2-58 Double-Ended Pump Suction Guillotine (Maximum Safety Injection)
- Energy Balance 6.2-59 0.6 Double-Ended Pump Suction Guillotine
- Energy Balance 6.2-60 Three-Foot Squared Pump Suction Split
- Energy Balance 6.2-61 Doub le-Ended Hot Leg Guillotine
- Energy Balance 6.2-62 Double-Ended Cold Leg Guillotine
- Energy Balance 6.2-63 Double-Ended Pump Suction Guillotine (Minimum Safety Injection)
- Principal Parameter Transients 6.2-64 Double-Ended Pump Suction Guillotine (Maximum Safety Injection)
- Principal Parameter Transients 6.2-65 System Parameters 6.2-66 Full Double
-Ended MSLB at Hot Shutdown (with SAF in Containment Spray System)
- Mass and Energy Releases (Forward Flow) 6.2-67 Full Double
-Ended MSLB at Hot Shutdown (with SAF in Containment Spray System)
- Mass and Energy Releases (Reverse Flow) 6.2-68 Full Double
-Ended MSLB at 102% Power (with SAF of Broken Loop MSIV) - Mass and Energy Releases (Forward Flow)
 
6.2-69 Full Double
-Ended MSLB at 102% Power (with SAF of Broken Loop MSIV) - Mass and Energy Releases (Reverse Flow) 6.2-69A Mass and Energy Balance Peak Calculated Containment Pressure for MSLB 6.2-69B Mass and Energy Balance Peak Calculated Containment Temperature for MSLB 6.2-75 Containment Spray System Parameters 6.2-76 Heat Exchanger Performance Data 6.2-77 Containment Cooling
- Active Component Failure Analysis 6.2-78 Deleted 6.2-79 Codes, Standards and Guides Applicable to the Containment Spray System 6.2-80 Containment Sprayed Volume 6.2-81 Effect of Drop Size on Spray Efficiency 6.2-82 Containment Enclosure Design and Performance Data 6.2-83 Containment Isolation System Design Information 6.2-84 Hydrogen Gas Analyzer Design Parameters 6.2-85 Applicable Codes, Standards, and Regulatory Guides Used in the Design of the Electric Hydrogen Recombiner 6.2-86 Electric Hydrogen Recombiner Typical Parameters 6.2-87 Containment Building Aluminum Inventory 6.2-88 Containment Building Zinc Inventory 6.2-89 Parameters Used to Evaluate CGCS Performance 6.2-90 Systems Open to Containment Atmosphere and/or Vented Prior to and During Type "A" Testing 6.2-91 Systems Not Vented During Type "A" Testing 6.2-92 Containment Liner Penetrations 6.3-1 Emergency Core Cooling System Component Design Parameters 6.3-2 Emergency Core Cooling System Relief Valve Data 6.3-3 Motor-Operated Isolation Valves in the Emergency Core Cooling System 6.3-4 Materials Employed for Emergency Core Cooling System Components 6.3-5 Failure Mode and Effects Analysis
- Emergency Core Cooling System
- Active Components
 
6.3-6 Emergency Core Cooling System Recirculation Piping Passive Failure Analysis - Long-term Phase 6.3-7 Manual Actions Required for ECCS Operation 6.3-8 Emergency Core Cooling System Shared Functions Evaluation 6.3-9 Normal Operating Status of Emergency Core Cooling System Components for Core Cooling 6.4-1 Control Room Complex Safety
-Related Ventilation System Malfunction Analysis 6.5-1 Compliance of Containment Enclosure Air Cleaning Units to Regulatory Guide 1.52, Rev. 2, March 1978 6.5-2 Compliance of Fuel Storage Building Air Cleaning Units to Regulatory Guide 1.52, Rev. 2, March 1978 6.5-3 Compliance of Control Room Emergency Filtration Subsystem to Regulatory Guide 1.52, Rev. 2, March 1978 6.5-4 Containment Enclosure Emergency Air Cleaning System Materials 6.5-5 Fuel Storage Building Emergency Air Cleaning System Materials 6.5-6 Control Room Emergency Makeup Air and Filtration Subsystem Materials 6.5-7 Volumes of Containment Enclosure Areas 6.5-8 Containment Operation Following a Design Basis Accident 6.8-1 Emergency Feedwater Pump Data 6.8-2 Emergency Feedwater System Failure Analysis
 
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.1(B)-1 Revision:
Sheet: 8 1 of 2  TABLE 6.1(B)-1 PRINCIPAL ENGINEERED SAFETY FEATURES METALLIC MATERIALS
* Valves  Bodies SA-182, Grade F304 or F316 SA-351, Grade CF3M or CF8M Bonnets SA-182, Grade F304 or F316 SA-240, TP 304 SA-351, Grade CF3M or CF8M Discs SA-182, Grade F316 SA-479, TP 316 Closure bolting/studs SA-453, Grade 660 SA-564, Grade 630 Nuts SA-193, Grade B6 SA-194, Grade 6, 8F, or 8M
 
SA-453, Grade 660 Tanks  Shells and heads SA-240, TP 304; SA-479, TP 304 Flanges and nozzles SA-276, TP 304; SA-182, TP 304 SA-312, TP 304 Closure bolting and nuts SA-193, Grade B8; SA-194, Grade 6 Pumps Pump casing SA-351, TP 316 Piping SA-213, TP 316 Pipe fittings Pipe plug, cross, bushing, & male connector SA-182, TP 316 Nipple SA-312, TP 316 Orifice SA-479, TP 316 Closure studs and nuts SA-193, Grade B8M; SA-194, Grade 8M or B8M
* Excluding NSSS-supplied S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.1(B)-1 Revision:
Sheet: 8 2 of 2  Heat Exchangers Shells and shell heads SA-285, Grade C Shell nozzles SA-106, Grade B Shell flanges SA-105 (normalized) Shell couplings SA-105 Channel SA-240, TP 304 Channel facing SA-182, Grade F 304 Channel nozzles SA-312, TP 304 Channel cover SA-516, Grade 70 lined with SA-240, TP 304 Channel couplings SA-479, TP 304 Tubes SA-249, TP 304 (.05% max. carbon) Tubesheets SA-182, Grade F 304 Closure bolting and nuts SA-193, Grade B7; SA-194, Grade 2H Piping SA-312, TP 304 or SA-358, TP 304 Fittings SA-182, TP 304 or SA-403, Grade WP 304 or WP 304W Flanges SA-182, TP F304 Closure Bolting and Nuts SA-193, Grade B8M or B8; 
 
SA-194, Grade 6 Weld Materials SFA-5.1, 5.4, 5.9, 5.17, 5.18, 5.20
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.1(B)-2 Revision:
Sheet: 8 1 of 1  Table 6.1(B)-2 COATED AREAS ON MAJOR EQUIPMENT AND STRUCTURES INSIDE CONTAINMENT
* Component/Structures Painted Surface Area (ft
: 2)                                                Containment Cylinder 63,774 Containment Dome 30,778 Miscellaneous Concrete 96,514 Structural Steel 64,688 Polar Gantry Crane 19,920 Floor and Sump 11,640 Equipment Hatch 3,250 Equipment Steel 2,345 Personnel Hatch 360 Containment Recirc. Unit 705
* Excluding NSSS-supplied S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.1(B)-4 Revision:
Sheet: 8 1 of 1  Table 6.1(B)-4 OTHER ORGANIC MATERIALS IN CONTAINMENT Item Material Amount Reactor coolant pump lubricant Petroleum base oil 1,060 gal.
Cable insulation 50,000 lbs.
Power cable Ethylene propylene rubber with Hypalon jacket Silicone Rubber  Control and instrument cable Cross-linked polyethylene insulation with Hypalon jacket
 
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.1(N)-1 Revision:
Sheet: 8 1 of 2 Table 6.1(N)-1 ENGINEERED SAFETY FEATURE MATERIALS Valves  Bodies SA-182, Grade F316; SA-351, Grade CF8 or CF8M Bonnets SA-182, Grade F316; SA-351, Grade CF8 or CF8M Discs SA-182, Grade F316; SA-564, Grade 630; SA-351, Grade CF8 or CF8M Pressure retaining bolting SA-453, Grade 660 Pressure retaining nuts SA-453, Grade 660; SA-194, Grade 6 Auxiliary Heat Exchangers Heads SA-240, Type 304 Nozzle necks SA-182, Grade F304; SA-312, Grade TP304; SA-240, Type 304 Tubes SA-213, Grade TP304; SA-249, Grade TP304 Tube Sheets SA-182, Grade F304; SA-240, Type 304; SA-516, Grade 70 with Stainless Steel Cladding A-8 Analysis Shells SA-240 and SA-312, Grade TP304 Auxiliary Pressure Vessels, Tanks, Filters, etc. Shells and heads SA-351, Grade CF8A; SA-240, Type 304; SA-264 Clad Plate of SA-537, Class 1 with SA-240, Type 304 Clad and Stainless Steel Weld Overlay A-8
 
Analysis Flanges and nozzles SA-182, Grade F304; SA-350, Grade LF2 with SA-240, Type 304 and Stainless Steel Weld Overlay A-8
 
Analysis Piping SA-312 and SA-240, TP304 or TP316 Seamless Pipe fittings SA-403, Grade WP304 Seamless Closure bolting and nuts SA-193, Grade B7; SA-194, Grade 2H S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.1(N)-1 Revision:
Sheet: 8 2 of 2 Auxiliary Pumps Pump casing and heads SA-351, Grade CF8 or CF8M; SA-182, Grade F304 or F316 Flanges and nozzles SA-182, Grade F304 or F316; SA-403, Grade WP316L Seamless Piping SA-312, Grade TP304 or TP316 Seamless Stuffing or packing box cover SA-351, Grade CF8 or CF8M; SA-240, Type 304 or Type 316; SA-182, Grade F304 or F316 Pipe fittings SA-403, Grade WP316L Seamless; SA-213, Grade TP304, TP304L, TP316 or TP316L Closure bolting and nuts SA-193 Grade B6, B7 or B8M; SA-194, Grade 2H or 8M; SA-453, Grade 660; and Nuts SA-194, Grade 2H; 6, 7 and 8M
 
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.1(N)-2 Revision:
Sheet: 8 1 of 1  Table 6.1(N)-2 PROTECTIVE COAT INGS ON WESTINGHOUSE SUPPLIED EQUIPMENT INSIDE CONTAINMENT Component Painted Surface Area (ft
: 2) RCS Component Supports 11,230 Reactor coolant pump assemblies 4,048 Accumulator tanks 5,400 Manipulator crane 2,600 Other refueling equipment 2,125 Remaining equipment (such as valves, auxiliary tanks and heat exchanger supports, transmitters, alarm horns, small instruments)
<1,300 S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-1 Revision:
Sheet: 8 1 of 1                                                  Table 6.2-1 GENERAL INFORMATION AND INITIAL CONDITIONS I. General Information, Containment Design Pressure, psig 52.0/-3.5  Design Temperature,  F 296  C. Free Volume, ft 3 2.704 x 10 6  D. Design Leak Rate, % mass/day 0.15 II. Initial Conditions A. Reactor Coolant System
: 1. Reactor Power Level, Mw(t) 3652.8  2. Average Coolant Temperature,  F 592.7  3. Mass of Reactor Coolant System Liquid, lbm 492,269  4. Mass of Reactor Coolant System Steam, lbm 12,370  5. Liquid Plus Steam Energy, Btu 302.3 x 10 6  B. Containment
: 1. Pressure, psig 1.5  2. Inside Temperature,  F 120  3. Outside Temperature,  F 110  4. Relative Humidity, %
1  5. Service Water Temperature,  F **65  6. Refueling Water Temperature,  F 98  7. Spray Water Temperature,  F 100  C. Stored Liquid
: 1. Refueling Water Storage Tank
*, gal 350,000 min. 2. Total Free Volume, Four Accumulators 3,800 nominal  3. Spray Additive Tank, gal 10,700  ** System analysis has been performed to permit plant operation up to a maximum ocean temperature of 68.5F to accommodate occasional summer ocean temperature excursions.
* 475,000 gal maximum capacity S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-2 Revision:
Sheet: 8 1 of 3  Table 6.2-2 ENGINEERED SAFETY FEATURE SYSTEMS INFORMATION Design Capacity Values Used for Containment Analysis A. Passive Safety Injection System
: 1. Number of Accumulators (Safety Injection Tanks) 4 4  2. Pressure Set Point, psig 600 600 B. Active Safety Injection Systems
: 1. Charging Pumps
: a. Number of Lines 2 1 (Min. SI), 2 (Max. SI)  b. Number of Pumps 2 1 (Min. SI), 2 (Max. SI)  c. Flow Rate per Pump, gpm 150 @ 2,518 psi 450 (Injection) 0 (Recirculation)  2. High Pressure Safety Injection
: a. Number of Lines 2 1 (Min. SI), 2 (Max. SI)  b. Number of Pumps 2 1 (Min. SI), 2 (Max. SI)  c. Flow Rate per Pump, gpm 440 @ 2,680 ft TDH 450 (Injection) 0 (Recirculation)  3. Low Pressure Safety Injection
: a. Number of Lines 2 1 (Min. SI), 2 (Max. SI)  b. Number of Pumps 2 1 (Min. SI), 2 (Max. SI)  c. Flow Rate per Pump, gpm 3,000 @ 375 ft TDH 4,000 (Injection) 3,000 (Recirculation)  4. Containment Spray System
: a. Number of Lines 2 1  b. Number of Pumps 2 1  c. Number of Headers 2 1  d. Flow Rate per pump, gpm 3,010 @ 550 ft TDH 2,808  5. Heat Exchangers
: a. Containment Spray Heat Exchanger (1) Type Shell and Tube
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-2 Revision:
Sheet: 8 2 of 3  Design Capacity Values Used for Containment Analysis    (2) Number of Units 2 1    (3) Heat Transfer Area/Unit, ft 2 3,468 3,468    (4) Overall Heat Transfer Coefficient, Btu/hr - ft 2 - &deg;F  (431 service)    (5) Flow Rates/Unit, gpm (a) Recirculation (Tube) Side 2,808    (b) Exterior (Shell) Side 4,800    (6) Source of Cooling Water Component Cooling Water    (7) Time of Recirculation, sec. 2755 (Min. SI) 1688 (Max. SI)  b. Residual Heat Removal Heat Exchanger (1) Type Shell and Tube (2) Number of Units 2 1 (Min. SI)    (3) Heat Transfer Area/Unit, ft 2 5444 5444    (4) Overall Heat Transfer Coefficient, Btu/hr - ft 2-&deg;F  385 (Service)    (5) Flow Rates/Units, gpm (a) Recirculation (Tube) Side 3,000    (b) Exterior (Shell) Side 5,000    (6) Source of Cooling Water Component Cooling Water    (7) Time of Recirculation, sec. 2755 (Min. SI) 1688 (Max. SI)  c. Component Cooling Water Heat Exchanger (1) Type Shell and Tube
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-2 Revision:
Sheet: 8 3 of 3  Design Capacity Values Used for Containment Analysis    (2) Number of Units 2 1 (Min. SI)    (3) Heat Transfer Area/Unit, ft 2 24,444 24,444    (4) Overall Heat Transfer Coefficient Btu/hr - ft 2 - &deg;F  539    (5) Flow Rates/Unit, gpm (a) Component Cooling Water(Shell) Side 11,727    (b) Station Service Water (Tube) Side 8,000 (Ocean) 11,000 (Tower) 7,500 (ocean) 9,851 (Tower)    (6) Source of Cooling Water Station Service Water    (7) Time of Recirculation, sec 2755 (Min. SI) 1688 (Max. SI)
 
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-3 Revision:
Sheet: 8 1 of 2 Table 6.2-3 CONTAINMENT PASSIVE HEAT SINK DATA Heat Sink Material One Side Exposed Area(ft 2) Thickness (in) Volume (ft3) Sides Exposed 1. Concrete Cylinder Steel-lined
 
Concrete Paint Carbon Steel Air Gap Concrete 63,774  0.018 0.375 0.625 54.0  1,993 286,983 one 2. Containment Dome Steel-lined
 
Concrete Paint Carbon Steel Air Gap Concrete 30,788  0.018 0.50 0.0625 42.0 1,282  107,723 one 3. Miscellaneous Painted Concrete
 
Paint Concrete
 
Paint 48,257  0.059 53.52 0.059 215,226 both 4. Refueling Canal Floor  SS-lined Concrete Stainless Steel Air Gap Concrete 1,152  0.25 0.0625 18.0  24  1,728 both 5. Refueling Canal SS-lined Concrete Stainless Steel Air Gap Concrete 6,005  0.1875 0.0625 36.0  94  18,014 both 6. Conduit Galvanized Steel 11,728 0.154 150.5 one 7. Ducts & Trays Galvanized Steel 55,927 0.0729 340 one 8. Structural Steel Painted Concrete Paint Steel Paint 32,344  0.018 0.558 0.018 1,503 both 9. Polar Crane, Equipment
: Hatch, Personnel Hatch Painted Steel
 
Paint Steel 23,529  0.018 0.736 1,443 one S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-3 Revision:
Sheet: 8 2 of 2 Heat Sink Material One Side Exposed Area(ft 2) Thickness (in) Volume (ft3) Sides Exposed 10. Equipment Steel Painted Steel Paint Steel 2,345  0.018 1.55 303 one 11. Containment Floor and Sump Painted Steel
 
Paint Concrete
 
Steel Air Gap Concrete 11,639  0.009 47.43 0.25 0.0625 120.0 46,003 243  116,390 one S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2_4 Revision:
Sheet: 10 1 of 1  TABLE 6.2-4 PASSIVE HEAT SINK MATERIAL PROPERTIES Material Thermal Conductivity (Btu/hr-ft- F) Volumetric Heat Capacity (Btu/ft 3- F) Paint 0.88 24.12 Carbon Steel 27.0 58.8 Stainless Steel 10.0 56.0 Air Gap 0.0184 0.0173 Concrete 0.92 22.62
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-5 Revision:
Sheet: 8 1 of 1  TABLE 6.2-5 BREAK LOCATIONS POSTUL ATED FOR LOSS-OF-COOLANT ACCIDENT ANALYSIS Code No. Break Location 1 Pump Suction 2 Cold Leg 3 Hot Leg S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-6 Revision:
Sheet: 8 1 of 1  TABLE 6.2-6 SPECTRUM OF BREAKS POST ULATED FOR LOSS-OF-COOLANT ACCIDENT ANALYSIS  Code No. Break Type and Size 1  Full Double-Ended Guillotine Rupture 2  Small Double-Ended Guillotine Rupture (0.6 of Full Size for Break at Pump Suction) 3  Split Rupture (3 ft 2 for Break at Pump Suction)
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-7 Revision:
Sheet: 8 1 of 1  TABLE 6.2-7 SINGLE ACTIVE FAILURES POSTULATED FOR LOSS-OF-COOLANT ACCIDENT ANALYSIS Code No. SAF 1  Failure of One Diesel Generator (Minimum Safety Injection) 2  Failure of One Diesel Generator at Time of Containment Spray Actuation (Maximum Safety Injection)
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-8 Revision:
Sheet: 10 1 of 1 TABLE 6.2-8 CALCULATED PEAK CONTAINMENT PRESSURE AND PEAK CONTAINMENT TEMPERATURE FOLLOWING LOSS-OF-COOLANT ACCIDENTS Case Accident Conditions Peak Pressure (psig) Time of Peak Pressure (sec)
Peak Temperature
( F) Time of Peak Temperature (sec) Energy Released to Containment Up To End of Blowdown (10 6 Btu) 1.1.1 Full DE Pump Suction Guillotine (Min. SI) 49.2 3,601 272 3,613 311.0 1.1.2 Full DE Pump Suction Guillotine (Max. SI) 49.6 3,603 273 3,614 311.0 1.2.1 0.6 DE Pump Suction Guillotine (Min. SI) 49.2 3,604 272 3,616 307.6 1.3.1 3 ft 2 Pump Suction Split (Min. SI) 35.6 33 251 32 306.9 2.1.1 Full DE Cold-Leg Guillotine (Min. SI) 36.6 18 254 18 305.9 3.1.1 Full DE Hot-Leg Guillotine
* (Min. SI) 37.5 20 256 22 311.8
* This case has been analyzed for the blowdown period only. Note: This table presents results related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 conta ins a discussion of the results of this analysis of record, as compared with the long-term LOCA containment response results at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-9 Revision:
Sheet: 10 1 of 1  TABLE 6.2-9 LOSS-OF-COOLANT ACCIDENT CHRONOLOGY FOR VARIOUS BREAK LOCATIONS Time (sec)
Event Case 1.1.2 Case 2.1.1 Case 3.1.1
* Pipe Rupture 0.0 0.0 0.0 Assumed Initiation of ECCS 10.0 10.0 10.0 Peak Containment Pressure During Blowdown 21.0 17.0 19.5 End of Blowdown 24.2 22.0 21.0 Beginning of Containment Spray 65.0 65.0 65.0 End of Reflood 155.1 212.1 - End of Broken-Loop Steam Generator Energy Release 678.0 - - Peak Containment Pressure After Blowdown and Before Circulation 695.0 215.0 - End of Injection and Beginning of Recirculation (350,000 gal. RWST Water Injected) 1688.0 2755.0 2755.0 Peak Containment Pressure During Recirculation 3603.0 6000.0 - Containment Pressure at 50% of Design Value 28000.0 19000.0 -
* This case has been analyzed for the blowdown period only. Note:  This table presents results related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the results of this analysis of record, as compared with the long-term LOCA containment response results at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-10 Revision:
Sheet: 8 1 of 1  6.2-10 INITIAL PLANT OPERATING PO WER LEVELS POSTULATED FOR MAIN STEAM LINE BREAK ANALYSIS Code No. Power Level 1 102% 2 75%
3 50%
4 25%
5 0% (Hot Shutdown)
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-11 Revision:
Sheet: 8 1 of 1  6.2-11 SPECTRUM OF RUPTURES POSTUL ATED FOR MAIN STEAM LINE BREAK ANALYSIS  Code No. Break Type and Size 1 Full Double-Ended Guillotine Rupture 2 Small Double-Ended Guillotine Rupture*
Power Level Break Size (sq..ft.)
102% 0.60  75% 0.55  50% 0.45  25% 0.33  0% 0.20 3 Split Rupture  Power Level Break Size (sq. ft.)
 
102% 0.80  75% 0.84  50% 0.80  25% 0.660  0% 0.40 4 Additional Small DE Guillotine Rupture
* (0.25 ft 2 for 75% and 0.10 ft 2 for 0% only.)
* Area given is for each of fo rward flow and reverse flow.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-12 Revision:
Sheet: 9 1 of 1  TABLE 6.2-12 SINGLE FAILURES POSTULATED FOR MAIN STEAM LINE BREAK ANALYSIS Code No. Single Failure 1  Failure of one Feedwater Isolation Valve (FWIV) to Close 2  Failure of Broken-Loop Main Steam Isolation Valve (MSIV) to Close 3  Failure of One Feedwater Control Valve (FCV) to Close 4  Failure of One Main Feedwater Pump Trip (MFPT) 5  Failure of Broken-Loop Emergency Feedwater Pump Runout Control (EFPRC) 6  Failure of One Containment Spray Train
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-13 Revision:
Sheet: 10 1 of 1  TABLE 6.2-13 CALCULATED PEAK CONTAINM ENT PRESSURE AND PEAK CONTAINMENT TEMPERATURE FOLLOWING FULL DOUBLE-ENDED GUILLOTINE RUPTURE IN MAIN STEAM LINE Case + Peak Pressure (psig) Time of Peak Pressure * (sec) Peak Temperature
( F) Time of  Peak Temperature (sec) Total Energy Release to Containment (10 6 Btu) 1.1.2 33.6 173 364 82 324.72      1.1.6 32.9 173 360 98 301.77      5.1.6 36.1 281 349 112 345.13        + The first digit refers to the power levels listed in Table 6.2-10, the second digit is the break spectrum in Table 6.2-11, and the third digit is the single failure in Table 6.2-12.
* Same as steam generator dryout time in these cases. Note:  This table presents results related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the results of this analysis of record, as compared with the MSLB containment response results at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-14 Revision:
Sheet: 10 1 of 1  TABLE 6.2-14 CALCULATED PEAK CONTAINM ENT PRESSURE AND PEAK CONTAINMENT TEMPERATURE FOLLOWING SMALL DOUBLE-ENDED GUILLOTINE RUPTURES IN MAIN STEAM LINE Case + Peak Pressure (psig) Time of Peak Pressure  (sec) Peak Temperature
(&deg;F) Time of Peak Temperature (sec) Total Energy Release to Containment (10 6 Btu) 5.2.6 28.8 1460*317 612 332.92 5.4.5 20.6 1165 306 1165 445.50 5.4.6 23.0 2435* 305 1260 291.62  + The first digit refers to the power levels listed in Table 6.2-10, the second digit is the break spectrum in Table 6.2-11, and the third digit is the single active failure in Table 6.2-12.
* Steam generator dryout time. Note:  This table presents results related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the results of this analysis of record, as compared with the MSLB containment response results at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-15 Revision:
Sheet: 10 1 of 1  TABLE 6.2-15 CALCULATED PEAK CONTAINM ENT PRESSURE AND PEAK CONTAINMENT TEMPERATURE FOLLOWING RUPTURES IN MAIN STEAM LINE Case + Peak Pressure (psig) Time of  Peak Pressure *  (sec) Peak Temperature
( F) Time of Peak Temperature  (sec) Total Energy Release to Containment (10 6 Btu) 4.3.6 34.7 353* 349 144 340.57  + The first digit refers to the power levels listed in Table 6.2-10, the second digit is the break spectrum in Table 6.2-11, and the third digit is the single active failure in Table 6.2-12.
* Steam generator dryout time. Note:  This table presents results related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the results of this analysis of record, as compared with the MSLB containment response results at an analyzed core power level of 3659 MWt.
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-16 Revision:
Sheet: 8 1 of 1  Table 6.2-16 POSTULATED BREAKS FOR SUBCOMPARTMENT ANALYSIS Containment Subcompartment Postulated Break Break Location Pressurizer Compartment Full double-ended guillotine spray line rupture EL 12'-0"; EL 24-6" EL 34'-0"; EL 45'-0" EL 54'-0 5/8"    Pressurizer Skirt Cavity Full double-ended guillotine surge line
 
rupture Pressurizer surge line nozzle
 
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-30 Revision:
Sheet: 8 1 of 2 Table 6.2-30 MASS AND ENERGY RELEASE RATES FOLLOWING SPRAY LINE RUPTURE Time (sec)
Mass Rate (lb/sec)
Enthalpy (Btu/lb) 0.00000 0.0 611.71 .00502 5756.7 611.71 .01002 5615.6 611.65 .02003 6046.9 607.30 .03004 6377.2 604.57 .04002 6514.4 603.09 .05005 6255.2 603.74 .06003 6325.0 602.99 .07006 6345.5 602.60 .08003 6287.2 602.63 .09001 6040.9 603.77 .10002 6273.6 602.47 .11510 6520.5 601.18 .20010 6057.1 603.29 .30004 5925.2 604.01 .34000 5884.1 604.24 .35027 5901.4 604.12 .37001 5858.6 604.35 .38001 5828.3 604.53 S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-30 Revision:
Sheet: 8 2 of 2 Time (sec)
Mass Rate (lb/sec)
Enthalpy (Btu/lb)
.40002 5902.6 604.07 .41010 5928.7 603.90 .43002 5845.8 604.38 .45755 5785.9 604.73 .47751 5795.2 604.65 .51009 5749.2 604.88 .55000 5797.0 604.52 .65003 5717.6 604.86 .70010 5736.6 604.64 .81021 5684.6 604.78 .85000 5699.4 604.60 .90008 5683.0 604.62 .94003 5662.9 604.68 1.00005 5671.7 604.51 .00000 5535.8 604.06 S EABROOK ENGINEERED S AFETY FEATURES Revision:
8 S TATION TABLE 6.2-31 Sheet: 1 of 4 UFSAR  Break Type : Full Double-Ended Spray Line Break Location: See Notes VENT PATH NO. HEAD LOSS, K Vol. Node No. From To AREA (ft 2) Ratio A L (1/ft) FRICTION K, fL/d TURNING AND OBSTRUCTION LOSS, K EXPANSION K CONTRACTION K TOTAL 1 1 2 139.40 .05 .150  .010 .020 .180 2 1 23 92.00 .05 .057  1.0 .300 1.357 3 2 3 21.50 .20 .110  .130 .310 .550      (.110)  (.390) (.180) (.680) 4 2 4 8.20 .35 .030  .110 .310 .450      (.030)  (.390) (.180) (.600) 5 2 5 10.67 .46 .060  .110 .310 .480      (.060)  (.390) (.180) (.630) 6 2 6 4.92 1.20 .130  .110 .310 .550      (.130)  (.390) (.180) (.700) 7 2 7 4.81 1.20 .130  .110 .310 .550      (.130)  (.390) (.180) (.700) 8 2 8 4.92 1.21 .130  .110 .310 .550      (.130)  (.390) (.180) (.700) 9 2 23 18.90 .50 .050  1.0 .500 1.550 S EABROOK ENGINEERED S AFETY FEATURES Revision:
8 S TATION TABLE 6.2-31 Sheet: 2 of 4 UFSAR  Break Type : Full Double-Ended Spray Line Break Location: See Notes VENT PATH NO. Vol. Node No. AREA (ft 2) Ratio A L (1/ft) HEAD LOSS, K 10 3 4 70.70 .12 .600  7.730 .190 8.520 11 3 8 13.58 .38 .080 .110  .300 .490 12 3 9 37.00 .34 .160    .160 13 4 5 69.75 .10 .210  1.710 1.920 14 4 10 19.60 .64 .050    .050 15 5 6 12.36 .42 .080 .110  .300 .490 16 5 11 14.70 .86 .080    .080 17 6 7 12.36 .39 .080 .110  .190 18 6 12 5.32 2.35 .170    .170 19 7 8 13.69 .35 .080 .110  .190 20 7 13 5.32 2.35 .170    .170 21 8 14 5.32 2.37 .170    .170 22 9 10 71.22 .12 .600  7.73 .190 8.520 23 9 14 13.56 .38 .080 .110  .300 .490 24 9 15 9.43 .33 .110  .370 .480      (.110)  (.550) (0.0) (.660)
S EABROOK ENGINEERED S AFETY FEATURES Revision:
8 S TATION TABLE 6.2-31 Sheet: 3 of 4 UFSAR  Break Type : Full Double-Ended Spray Line Break Location: See Notes VENT PATH NO. Vol. Node No. AREA (ft 2) Ratio A L (1/ft) HEAD LOSS, K 25 10 11 71.22 .10 .210  1.710 1.920 26 10 15 2.90 .84 .050  .430 .480      (.050)  (.740) (0.0) (.790) 27 11 12 13.56 .38 .080 .110  .300 .490 28 11 16 8.13 .51 .060  .220 .280      (.060)  (.200) (0.0) (.260) 29 12 13 13.56 .35 .080 .110  .19      (.110)  (0.0) (.050) (.160) 30 12 16 5.32 1.25 .110 .010  .120 31 13 14 13.56 .35 .080 .110  .190 32 13 16 5.32 1.25 .110  .010  .120      (.110)  (0.0) (.050) (.160) 33 14 15 5.32 1.27 .110  .010  .120      (.110)  (0.0) (.050) (.160) 34 15 16 8.66 1.66 .636 .424 .880 .770 2.710 35 15 17 18.02 .20 .070  .540 .310 .920 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-31 Revision:
Sheet: 8 4 of 4 Break Type : Full Double-Ended Spray Line Break Location: See Notes VENT PATH NO. Vol. Node No. AREA (ft 2) Ratio A L (1/ft) HEAD LOSS, K 36 16 18 19.43 .15 .050  1.320 .270 1.640 37 17 18 29.80 .37 2.088 1.392 4.120 3.760 11.360 38**17 19 1.20 .50  .050 .620 .170 .840 39* 18 20 12.30 .50 .050  .800 .410 1.260 40* 18 23 10.80 1.40 1.910  1.0 .400 3.310 41 19 20 37.90 .33 .120  .420 .470 1.010 42 19 21 91.60 .09 .040    .040 43 20 22 47.70 .17 .100  .060 .120 .280 44 20 23 114.50 .17 .040  1.0  1.040 45 21 22 45.80 .33 .120  .420 .470 1.010 46 22 23 119.10 .17 .040  1.0  1.040
* PRESSURE DEPENDENT AREAS.
S EABROOK ENGINEERED S AFETY FEATURES Revision:
8 S TATION TABLE 6.2-32 Sheet: 1 of 4 UFSAR  TABLE 6.2-32 PRESSURIZER COMPARTMENT NODAL DESCRIPTION AND PEAK NODAL DIFFERENTIAL PRESSURE Break Type : Full Double-Ended Spray Line Break Location: See Notes Initial Conditions Design Node Volume No. ( ) Description Height (ft) Volume (ft 3) Temperature
( F) Pressure (lb/in 2)(a) Relative Humidity % Calculated Peak Pressure Differential (lb/in 2 (g)) Peak Pressure Differential (lb/in 2 (g)) Margin % 1 Bounded by El.56'-6"-6 16 11" and El.63'-0" 6.44 902.1 120 15.2 90 5.37 1 6.5 21.0 2 Bounded by El.50'-1 16 7" and El.56'-6 16 11" 6.44 719.0 120 15.2 90 7.48 1 7.5 0.3 REGION BETWEEN EL.37'-63/4" AND EL.50'-1 Bounded by        3 Azimuth 60 and 150 12.56 457.7 120 15.2 90 16.18 2 18.5 14.3 4 Azimuth 150 and 180 12.56 243.1 120 15.2 90 13.96 3 18.5 32.5 5 Azimuth 180 and 240 12.56 184.4 120 15.2 90 13.82 3 18.5 33.9 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-32 Revision:
Sheet: 8 2 of 4 Break Type : Full Double-Ended Spray Line Break Location: See Notes Initial Conditions Design 6 Azimuth 240 and 300 12.56 67.3 120 15.2 90 13.48 3 18.5 37.2 7 Azimuth 300 and 360 12.56 67.3 120 15.2 90 13.48 3 18.5 37.2 8 Azimuth 0 and 60 12.56 66.6 120 15.2 90 13.65 3 18.5 35.5 REGION BETWEEN EL.25'-0" AND 37'-63/4"        Bounded by:        9 Azimuth 60 and 150 12.56 465.3 120 15.2 90 18.47 3 21.5 16.4 10 Azimuth 150 and 180 12.56 246.5 120 15.2 90 14.98 3 21.5 43.5 11 Azimuth 180 and 240 12.56 184.6 120 15.2 90 14.57 3 21.5 47.6 12 Azimuth 240 and 300 12.56 66.8 120 15.2 90 14.23 3 21.5 51.1 13 Azimuth 300 and 360 12.56 66.8 120 15.2 90 14.42 3 21.5 49.1 14 Azimuth 0 and 60 12.46 66.8 120 15.2 90 15.14 3 21.5 42 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-32 Revision:
Sheet: 8 3 of 4  Break Type : Full Double-Ended Spray Line Break Location: See Notes Initial Conditions Design REGION BETWEEN EL.22'-0" AND EL.25'-0"        Bounded by:        15 Azimuth 0 and 180 3.00 54.1 120 15.2 90 47.74 4 50 5 16 Azimuth 180 and 360 3.00 58.3 120 15.2 90 13.44 3 25 86 REGION BETWEEN EL.16'-6" AND EL.22'-0"        Bounded by:        17 Azimuth 0 and 180 5.50 141.8 120 15.2 90 25.19 4 25 0 18 Azimuth 180 and 360 5.50 245.2 120 15.2 90 11.53 4 25 117 REGION BETWEEN EL.8'-3" AND EL.16'-6"        Bounded by:        19 Azimuth 0 and 180 8.2 5 670.0 120 15.2 90 16.08 5 25 55 20 Azimuth 180 and 360 8.2 5 400.0 120 15.2 90 5.32 5 25 370 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-32 Revision:
Sheet: 8 4 of 4 Break Type : Full Double-Ended Spray Line Break Location: See Notes Initial Conditions Design REGION BETWEEN EL.0'-0" AND EL.8'-3"        B      ounded by:
21 Azimuth 0 and 180 8.25 746.0 120 15.2 90 64 15.25 5 25 22 Azimuth 180 and 360 8.25 518.0 120 15.2 90 224 7.71 5 25 23 C  15.2 90 -  - ontainment 2.71x10 6 120 NOTES: 1 .Peak differential pressure occurs following a postulated rupture of the spray line at elevation 54'-0". 2. Peak differential pressure occurs following a postulated rupture of the spray line at elevation 45'-0".
: 3. Peak differential pressure occurs following a postulated rupture of the spray line at elevation 34'-0".
: 4. Peak differential pressure occurs following a postulated rupture of the spray line at elevation 24'-6".
: 5. Peak differential pressure occurs following a postulated rupture of the spray line at elevation 12'-0".
 
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-33 Revision:
Sheet: 8 1 of 1 TABLE 6.2-33 MASS AND ENERGY RELEASE RATES SURGE LINE DOUBLE-ENDED GUILLOTINE Time (sec)
Mass Rate (lb/sec)
Enthalpy (Btu/lb) 0.0 0.0 705.37 .001 8.66lE3 705.37 .01011 1.365E4 705.19 .01501 1.468E4 705.07 .02011 1.522E4 704.89 .02511 1.544E4 704.68 .03007 1.546E4 704.43 .03513 1.536E4 704.15 .04012 1.519E4 703.88 .045 1.498E4 703.64 .05012 1.475E4 703.44 .05513 1.453E4 703.3 .06011 1.433E4 703.24 .0651 1.417E4 703.24 .07009 1.403E4 703.29 .07506 1.394E4 703.37 .08008 1.389E4 703.46 .08525 1.386E4 703.54 .09013 1.385E4 703.6 .0951 1.386E4 703.63 .10009 1.388E4 703.63 .15005 1.382E4 703.04 .20014 1.348E4 701.91 S EABROOK ENGINEERED S AFETY FEATURES Revision:
8 TABLE 6.2-34 Sheet: 1 of 1 S TATION  TABLE 6.2-34 PRESSURIZER SKIRT CAVITY VENT PATH DESCRIPTION Break Type
:  Full Double-Ended Surge Line Break Location:  Node #6 HEAD LOSS, K VOL. NODE NO. VENT PATH NO. From To AREA (ft 2) Ratio A L (1/ft) FRICTION K , d fL TURNING AND OBSTRUCTIO N LOSS , K EXPANSION K CONTRACTION K TOTAL  1 2 21.75 .323  .421  .421  1 6 22.86 .213    .127 .129  2 3 13.93 .504 .064  .402 .466  2 4 12.17 .209 .017  .56 .41 .987  2 5 193.7 .033  .528  .528  3 5 13.93 .47 .065  1.0  1.065  4 5 30.4 .263  1.0  1.000  5 6 .26 .2    1.770*
* Corresponds to an orifice coefficient = 0.6
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-35 Revision:
Sheet: 8 1 of 1 TABLE 6.2-35 PRESSURIZER SKIRT CAVITY NODAL DESCRIPTION Break Type:  Full Double-Ended Surge Line Break Location:  Node #6 Initial Conditions Design Node Volume No.( ) Description Height (ft) Volume (ft 3) Temperature
(&deg;F) Pressure Relative Humidity % Calculated Peak Pressure Differential (lb/in 2 (g)) Peak Pressure Differential Margin % 1 Volume from El.(-) 5'-11" to 0'-0"  128.7 120 15.2 90 186.4 N/A N/A 2 Volume from El.26'-0" to (-)5'-11"  1262.8 120 15.2 90 83 N/A N/A 3 Volume northeast  of the skirt cavity at El.(-) 12'-0"  181.1 120 15.2 90 6.5 N/A N/A 4 Volume southeast of the skirt cavity at El.(-) 12'-0"  121.6 120 15.2 90 5.8 N/A N/A 5 Containment 2.7x10 6 120 15.2 90 - N/A N/A 6 Volume under pressurizer vessel up to El.0'-0"  166.0 120 15.2 90 190.8 N/A N/A S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-36 Revision:
Sheet: 8 1 of 1  CASES ANALYZED AND TABLES WITH RESULTS Balance Case Blowdown Reflood Post-Reflood Mass Energy A Double-Ended Pump Suction Min. S.I.
6.2-38 6.2-43 6.2-48 6.2-51 6.2-57 B Double-Ended Pump Suction Max. S.I.
6.2-37 6.2-44 6.2-49 6.2-52 6.2-58 C 0.6 Doubled-Ended Pump Suction Min. S.I.
6.2-39 6.2-45 6.2-50 6.2-53 6.2-59 D Three Square Foot Pump Suction Split Min. S.I. 6.2-40 6.2-46 6.2-46 6.2-54 6.2-60 E Double-Ended Hot Leg Min. S.I.
6.2 - 6.2-55 6.2-61 F Double-Ended Cold Leg Min. S.I.
6.2-42 6.2-47 6.2-47 6.2-56 6.2-62 NOTES: 1 Table 6.2-38 is a duplicate of 6.2-37. It is provided for continuity only. Safety injection has no effect on the blowdown portion of the transient. 2 Double Ended refers to the size of break. A "0.6" Double Ende d Break is a guillotine break with a break equal to 60 percent of the double pipe area.
 
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-37 Revision:
Sheet: 10 1 of 2 TABLE 6.2-37 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MAXIMUM SAFETY INJECTION) BLOWDOWN MASS AND ENERGY RELEASES Note: BREAK PATH NO. 1 FLOW BREAK PATH NO. 2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 0.00 0.0 0.0 0.0 0.0 .100 41585.6 23222.0 22323.9 12416.9 .201 42362.9 23834.2 24356.8 13556.8 .400 44284.6 25474.7 23645.2 13186.6 .601 44898.0 26473.8 21180.8 11822.1 .901 41387.1 25100.8 19367.0 10821.4 1.20 38232.0 23606.1 18928.0 10580.4 1.70 34320.6 21965.3 18905.3 10569.0 2.30 28812.4 19267.6 18555.8 10385.2 2.70 25338.9 17319.1 18824.2 10551.4 3.00 21136.7 14640.7 18262.6 10247.8 3.80 16809.7 11759.4 16575.7 9325.8 4.40 14536.5 10187.2 15689.3 8838.8 4.60 14136.0 9904.1 16858.6 9505.1 5.60 12420.5 8712.2 16234.2 9146.6 6.20 11982.8 8392.7 15921.2 8969.6 9.00 9684.1 6920.5 13988.9 7873.7 9.40 9877.1 7295.9 13944.1 7847.8 9.80 7860.5 6752.3 13413.0 7545.2 11.2 7195.3 5955.9 12055.2 6770.1 12.6 6392.5 5330.5 10878.9 6096.6 14.2 5270.8 4457.8 9458.3 5291.0 15.4 4624.5 4016.1 8372.2 4662.4 16.0 4276.8 3896.1 8226.6 4346.2 16.6 3798.5 3785.0 7662.0 3768.1 17.2 2917.2 3362.5 6194.4 2812.4 18.0 2096.5 2573.5 4799.2 1976.5 18.6 1633.2 2019.1 5400.4 2075.0 19.0 1355.9 1682.7 5037.7 1845.3 20.0 752.7 940.2 3059.0 969.9 21.0 338.0 423.7 67.7 19.5 22.4 257.7 323.5 0.0 0.0 22.6 224.5 282.0 23247.9 6138.9  Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-37 Revision:
Sheet: 10 2 of 2  Note: BREAK PATH NO. 1 FLOW BREAK PATH NO. 2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 22.8 218.1 274.0 0.0 0.0 24.2 0.0 0.0 0.0 0.0  Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-38 Revision:
Sheet: 10 1 of 2 1TABLE 6.2-38 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MINIMUM SAFETY INJECTION) BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 0.000 0.0 0.0 0.0 0.0 .100 41585.6 23222.0 22323.9 12416.9 .201 42362.9 23834.2 24356.8 13556.8
.400 44284.6 25474.7 23645.2 13186.6
.601 44898.0 26473.8 21180.8 11822.1
.901 41387.1 25100.8 19367.0 10821.4 1.20 38232.0 23606.1 18928.0 10580.4 1.70 34320.6 21965.3 18905.3 10569.0 2.30 28812.4 19267.6 18555.8 10385.2 2.70 25338.9 17319.1 18824.2 10551.4 3.00 21136.7 14640.7 18262.6 10247.8 3.80 16809.7 11759.4 16575.7 9325.8 4.40 14536.5 10187.2 15689.3 8838.8 4.60 14136.0 9904.1 16858.6 9505.1 5.60 12420.5 8712.2 16234.2 9146.6 6.20 11982.8 8392.7 15921.2 8969.6 9.00 9684.1 6920.5 13988.9 7873.7 9.40 9877.1 7295.9 13944.1 7847.8 9.80 7860.5 6752.3 13413.0 7545.2 11.2 7195.3 5955.9 12055.2 6770.1 12.6 6392.5 5330.5 10678.9 6096.6 14.2 5270.8 4457.8 9458.3 5291.0 15.4 4624.5 4016.1 8372.2 4662.4 16.0 4276.8 3896.1 8226.6 4346.2 16.6 3798.5 3785.0 7662.0 3768.1 17.2 2917.2 3362.5 6194.4 2812.4 18.0 2096.5 2573.5 4799.2 1976.0 18.6 1633.2 2019.1 5400.4 2075.0 19.0 1355.9 1682.7 5037.7 1845.0 20.0 752.7 940.2 3059.0 969.9 21.0 338.0 423.7 67.7 19.5 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-38 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 22.4 257.7 323.5 0.0 0.0 22.6 224.5 282.0 23247.9 6138.9 22.8 218.1 274.0 0.0 0.0 24.2 0.0 0.0 0.0 0.0
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-39 Revision:
Sheet: 10 1 of 2 TABLE 6.2-39 0.6 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BLOWDOWN MASS AND ENERGY RELEASE BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 0.000 0.0 0.0 0.0 0.0 .100 37543.2 20967.3 20214.7 11237.1 .201 37944.1 21308.4 21975.7 12225.4
.401 36638.0 20904.7 21828.7 12163.4
.700 33054.7 19396.9 19709.0 11000.9 1.00 29205.3 17472.1 18881.6 10546.6 1.30 28590.9 17281.0 18523.2 10348.9 1.80 27872.1 17065.2 18556.9 10369.9 2.40 25425.5 15927.1 18594.2 10403.9 2.60 24340.3 15406.0 19417.0 10873.5 3.00 22073.9 14311.3 19071.6 10693.9 4.20 15784.4 10768.9 16796.7 9450.3 4.80 13599.1 9382.0 15784.7 8891.4 5.20 12779.3 8831.1 15259.8 8598.8 5.40 12360.3 8537.3 16045.8 9047.2 5.80 11786.2 8140.7 16003.2 9020.3 6.60 11030.8 7652.3 15476.1 8722.3 7.20 10707.8 7427.5 15174.0 8555.3 8.80 9628.9 6747.4 14262.7 8045.8 10.0 9104.3 6657.1 13591.9 7667.0 10.4 7141.4 5960.4 13369.8 7542.4 11.0 6719.3 5667.2 12860.2 7254.0 12.4 6245.8 5141.1 11696.9 6590.9 13.4 5760.0 4782.4 10939.6 6158.4 15.4 4607.1 3914.9 9283.6 5223.1 17.2 3763.9 3443.8 7877.2 4395.5 17.8 3350.2 3288.6 7820.0 4113.4 18.6 2336.4 2719.4 6629.6 3083.0 19.6 1659.1 2042.2 5097.7 2117.1 20.2 1337.2 1656.2 5887.9 2305.3 20.8 1054.2 1312.7 4080.9 1518.5
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-39 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 21.4 793.8 992.3 3612.4 1187.6 22.8 250.8 315.8 501.5 143.9 23.4 227.4 350.0 8.6 2.4 25.4 0.0 0.0 0.0 0.0 1   
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-40 Revision:
Sheet: 10 1 of 2 TABLE 6.2-40 THREE-FOOT SQUARED PUMP SUCTION SPLIT BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO. 1 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC 0.000 0.0 0.0 .101 43200.0 24145.5 .301 42783.0 24131.5
.801 37700.1 21931.5 1.00 35905.1 21070.4 1.60 32918.7 19583.1 2.30 27472.4 16545.4 2.70 26963.0 16334.6 5.60 21735.5 13305.5 6.60 20831.6 12649.3 8.00 19175.3 11651.3 10.4 18279.7 11095.6 11.0 18588.4 11214.5 11.8 16835.9 10477.7 15.6 14537.2 9132.5 19.2 12060.2 7791.2 19.6 11802.9 7649.8 22.0 10229.8 6800.5
 
24.4 8417.9 5836.8 26.2 6696.8 5035.3 28.2 4227.0 3614.1 29.2 3443.9 2879.8 30.2 3043.1 2425.1 30.8 2476.7 1986.1 32.0 3360.7 1801.3 32.8 2831.4 1496.9 33.0 3522.0 1619.7 33.6 3227.6 1471.0 34.4 4423.5 1675.5 34.6 4552.6 1634.7 36.0 2084.9 700.7
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-40 Revision:
Sheet: 10 2 of 2  BREAK PATH NO. 1 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC 36.4 1762.9 602.4 36.6 1002.9 336.6 36.8 1110.8 396.7
 
37.2 0.0 0.0 
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-41 Revision:
Sheet: 10 1 of 2 TABLE 6.2-41 DOUBLE-ENDED HOT LEG GU ILLOTINE BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 0.000 0.0 0.0 0.0 0.0 .100 42412.0 27585.7 26857.8 17356.3 .201 36857.8 24038.5 23284.6 14946.6
.301 36716.0 23961.5 20962.8 13271.1
.400 35842.5 23421.4 19773.5 12306.1
.600 34749.8 22805.0 18504.8 11159.1
.800 33776.8 22305.6 17743.6 10457.2 1.00 32541.3 21657.0 17433.2 10108.4 1.50 30093.3 20391.3 17866.4 10093.6 1.90 28232.4 19272.8 18090.5 10122.4 2.50 25753.2 17648.4 17757.9 9898.8 2.90 24401.8 16701.8 17349.0 9673.7 3.70 22280.4 15138.3 16388.6 9161.6 4.40 20976.4 14165.7 15338.1 8612.7 5.00 20574.8 13790.2 14392.2 8125.0 5.20 20586.1 13722.5 14353.9 8116.9 6.20 20286.9 13372.7 13168.3 7469.0 7.80 19363.8 12704.9 11142.1 6359.3 8.00 14861.3 10628.5 10887.7 6219.4 8.80 14379.7 10294.9 9891.5 5672.8 10.2 13371.4 9390.4 8200.0 4756.8 10.8 12710.4 8915.7 7477.7 4375.0 11.6 12053.0 8399.5 6581.5 3910.8 12.6 10953.0 7664.2 5594.4 3416.6 13.4 9880.5 7046.4 4931.1 3094.3 14.2 8707.2 6450.8 4369.5 2826.8 14.8 7604.6 5945.2 3991.6 2648.9 15.2 6516.0 5488.7 3717.0 2521.5 15.8 4549.6 4601.4 3218.8 2305.3 16.2 3592.6 3958.6 2828.8 2134.9 17.0 2419.4 2853.6 1901.9 1791.0 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-41 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 17.8 1602.7 1940.9 1067.1 1304.1 19.0 1036.5 1268.7 542.1 673.1 20.4 303.3 353.9 242.4 303.3 21.0 0.0 0.0 0.0 0.0
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-42 Revision:
Sheet: 10 1 of 3 TABLE 6.2-42 DOUBLE-ENDED COLD LEG GUILLO TINE BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW SI SPILL PATH FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 0.000 0.0 0.0 0.0 0.0 0.0 0.0 .100 25295.8 14098.9 54132.4 30247.2 4429.2 389.8 .201 23118.0 12892.3 60165.0 33618.5 4384.0 385.8
.401 23004.3 12839.9 59305.1 33129.2 4293.9 377.9
.901 22213.9 12526.9 54065.6 30190.6 4068.5 358.0 1.20 21352.5 12166.1 52444.1 29323.8 3933.5 346.1 1.50 20915.4 12052.1 49796.5 27885.8 3798.3 334.3 2.00 19832.2 11605.5 42660.9 23983.9 3572.7 314.4 2.20 17682.0 10389.3 40816.5 22980.5 3482.9 306.5 2.30 16563.3 9748.7 43663.7 24595.2 3463.0 304.7 2.40 15811.8 9320.4 37164.6 20951.3 3445.8 303.2 2.60 14854.2 8778.0 39095.1 22064.1 3411.2 300.2 2.80 14140.2 8372.4 35745.6 20192.2 3376.5 297.1 3.30 13004.1 7737.5 33471.6 18931.5 3290.1 289.5 3.70 12331.6 7389.3 28681.3 16221.2 3221.0 283.4 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-42 Revision:
Sheet: 10 2 of 3  BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW SI SPILL PATH FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 3.90 12028.5 7246.4 27163.6 15356.4 3186.4 280.4 4.60 11000.3 6818.5 25391.6 14334.3 3055.6 268.9 6.00 8967.0 6022.2 23959.4 13546.2 2743.9 241.5 7.60 6796.1 5006.5 21012.0 12080.6 2547.5 224.2 8.80 6037.4 4350.0 18836.2 11072.6 2473.5 217.7 9.80 6077.5 4254.8 16674.0 10338.3 2316.1 203.8 10.2 6005.8 4468.8 14404.1 9361.3 2252.9 198.3 10.6 4750.8 4257.8 12026.7 8293.7 2189.8 192.7 11.2 4073.1 3795.5 9437.7 7068.3 2121.9 186.7 11.6 3887.8 3597.3 8317.5 6517.9 2114.0 186.0 13.8 2879.2 2879.0 5400.9 4716.3 2005.4 176.5 14.2 2658.9 2781.0 4022.0 4332.5 1953.6 186.0 15.0 1302.8 1577.4 3449.5 3296.8 1850.5 16.2 919.6 1139.2 5502.0 2353.2 1799.6 18.0 368.7 463.5 3956.5 1286.5 1756.5 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-42 Revision:
Sheet: 10 3 of 3  BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW SI SPILL PATH FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 18.8 221.2 280.3 3791.2 1120.4 1666.0 176.5 19.4 112.6 143.7 1522.3 438.2 1598.2 154.6 20.2 90.7 116.3 1031.1 293.5 1558.3 146.6 20.4 101.2 130.0 0.0 0.0 1561.0 137.4 22.0 0.0 0.0 0.0 0.0 1583.0 139.3 20.4 101.2 130.0 0.0 0.0 1561.0 137.4 22.0 0.0 0.0 0.0 0.0 1583.0 139.3
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-43 Revision:
Sheet: 10 1 of 2 TABLE 6.2-43 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MINIMUM SAFETY INJECTION) REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 24.2 0.0 0.0 0.0 0.0 24.7 .0 .0 146.7 12.9 25.0 .0 .0 146.7 12.9 25.3 44.3 52.3 146.7 12.9 27.3 135.9 160.7 146.7 12.9 28.3 164.2 194.1 146.7 12.9 29.3 610.0 725.5 5265.9 680.5 30.3 772.7 921.3 6546.7 920.9 31.3 755.9 901.0 6421.0 908.4 33.3 717.3 854.6 6123.1 874.5 35.3 682.1 812.2 5853.2 842.4 36.3 665.8 792.6 5724.2 827.4 38.3 635.8 756.5 5482.3 799.3 40.3 608.6 724.0 5259.5 773.5 42.3 583.9 694.3 5053.3 749.7 44.3 561.3 667.3 4861.9 727.5 46.3 540.6 642.3 4683.5 706.8 48.3 521.3 619.3 4516.5 687.4 50.3 503.4 597.8 4359.5 669.1 52.3 486.6 577.8 4211.5 651.7 56.3 456.0 541.2 3938.5 619.4
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-43 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 57.3 493.6 585.8 339.1 254.5 58.3 548.0 651.3 360.8 286.0 62.3 501.8 595.9 341.1 259.4 63.3 490.3 582.2 336.2 252.9 67.3 448.1 531.7 318.5 229.2 77.3 361.5 428.5 282.6 182.3 85.3 309.4 366.4 261.5 155.0 93.3 268.5 317.8 245.3 134.3 101.3 237.1 280.6 233.1 118.8 109.3 213.7 252.8 224.2 107.6 121.3 190.6 225.4 215.6 96.8 133.3 177.8 210.3 210.9 90.8 147.3 170.9 202.2 208.2 87.5 153.2 169.7 200.6 207.7 86.8
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-44 Revision:
Sheet: 10 1 of 2 TABLE 6.2-44 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MAXIMUM SAFETY INJECTION) REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 24.2 0.0 0.0 0.0 0.0 25.0 .0 .0 .0 .0 25.1 57.6 68.1 .0 .0 25.2 49.6 58.6 .0 .0 26.2 94.5 111.7 .0 .0 27.3 134.8 159.4 .0 .0 28.3 174.4 206.2 .0 .0 29.3 822.1 983.7 6923.0 941.1 30.3 830.2 990.6 6969.3 968.2 32.3 790.6 942.8 6675.9 937.2 33.3 772.0 920.4 6536.0 920.9 34.3 754.2 899.0 6401.2 905.2 36.3 721.5 859.6 6148.0 875.6 38.3 692.1 824.3 5916.0 848.7 40.3 665.6 792.4 5702.6 823.9 42.3 641.5 763.4 5505.6 801.1 43.3 630.2 749.9 5412.6 790.4 45.3 699.0 724.4 5236.1 769.9 47.3 589.3 700.8 5071.0 750.6 47.4 588.4 699.7 5063.0 749.7 49.3 571.0 678.9 4916.2 732.4
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-44 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 50.3 562.3 668.4 4842.3 723.7 52.3 545.7 648.6 4701.0 706.9 54.3 530.1 629.8 4567.4 690.8 56.3 515.3 612.1 4440.8 675.4 58.3 501.3 595.3 4320.5 664.7 59.4 201.1 237.9 1111.6 292.9 71.4 196.0 231.8 1119.7 289.7 80.4 192.6 227.8 1125.1 287.0 95.4 187.2 221.4 1134.3 282.4 99.4 185.8 219.8 1136.7 281.2 115.4 180.2 213.2 1146.6 276.0 119.4 178.9 211.5 1149.1 274.7 151.4 168.3 199.0 1168.7 263.5 155.1 167.1 197.6 1170.9 262.2
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-45 Revision:
Sheet: 10 1 of 2 TABLE 6.2-45 0.6 DOUBLE-ENDED PUMP S UCTION GUILLOTINE REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 25.4 0.0 0.0 0.0 0.0 25.9 .0 .0 146.7 12.9 26.3 42.0 49.6 146.7 12.9 28.5 120.2 142.1 146.7 12.9 29.5 144.9 171.3 146.7 12.9 30.5 455.8 541.1 4148.9 519.8 31.5 559.1 667.6 6207.2 897.6 32.5 317.2 378.3 6526.8 1138.9 32.7 538.4 642.9 6066.6 885.7 33.5 639.4 762.2 6244.6 825.6 34.5 502.8 600.1 5819.6 856.8 35.5 247.2 294.6 6136.8 1108.9 36.5 462.5 551.1 6212.9 716.4 36.8 414.1 493.5 6151.2 656.6 37.5 408.5 486.8 6102.7 631.8 38.6 586.9 698.8 5470.2 805.0 40.6 560.4 666.9 5245.5 779.0 42.6 537.1 638.9 5042.0 754.6 44.6 515.7 613.3 4852.6 731.9 46.6 496.2 589.9 4676.0 710.7 48.6 478.1 568.2 4510.6 690.9
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-45 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 49.6 469.6 558.0 4431.7 681.4 53.6 438.3 520.6 4138.2 646.1 57.6 410.8 487.6 3874.7 614.2 58.7 373.2 442.4 300.8 203.1 59.7 515.1 612.5 368.0 297.2 61.7 495.9 589.6 359.0 285.1 69.7 415.1 492.8 321.5 234.4 76.7 358.1 424.7 295.5 199.8 80.7 331.2 392.6 283.2 183.8 88.7 286.6 339.6 263.3 157.9 96.7 251.7 298.1 248.0 138.2 104.7 224.7 266.0 236.3 123.3 132.7 175.5 207.6 215.8 97.2 157.5 164.4 194.4 211.1 91.3
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-46 Revision:
Sheet: 10 1 of 2 TABLE 6.2-46 THREE-FOOT SQUARED PUMP SUCTION SPLIT REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC 37.2 0.0 0.0 37.7 146.7 12.9 39.3 248.4 133.8 41.2 308.7 206.1 42.2 3136.3 744.6 43.3 7095.2 1801.1 44.3 6988.0 1780.4 46.3 6679.0 1704.6 48.3 6386.4 1633.1 49.3 6248.8 1599.7 51.3 5990.4 1537.4 53.3 5752.2 1480.5 55.3 5531.5 1428.2 57.3 5326.5 1380.0 61.3 4956.3 1293.5 62.3 4870.9 1273.6 64.3 4707.5 1235.7 68.3 4406.5 1165.9 69.4 1018.4 1110.5 70.4 1351.3 1582.8 71.4 1320.5 1546.2 73.4 1106.1 1275.7 74.4 940.1 1022.8 75.4 813.5 845.6 80.4 673.9 662.2 84.4 613.4 585.1 88.4 563.8 522.3 98.4 472.0 407.4 106.4 425.0 349.2 118.4 384.2 299.0
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-46 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC 132.4 363.2 273.3 152.4 355.2 263.5 161.0 355.0 263.3 161.1 528.2 179.3 100000.0 528.2 70.2
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-47 Revision:
Sheet: 10 1 of 2 TABLE 6.2-47 DOUBLE-ENDED COLD LEG GUILLO TINE BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW SI SPILL PATH FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 22.0 0.0 0.0 0.0 0.0 0.0 0.0 22.5 .0 .0 .0 .0 1649.6 145167.4 22.9 .0 .0 .0 .0 1632.2 143630.8 23.0 35.6 42.0 .0 .0 1618.7 142449.4 23.3 17.0 20.1 .0 .0 1608.7 141562.7 24.0 33.5 39.6 .0 .0 1577.5 138820.2 26.0 58.7 69.4 .0 .0 1498.9 131900.0 28.0 75.6 89.4 .0 .0 1425.9 125481.1 29.0 83.5 98.8 264.7 36.6 1390.9 122398.8 30.0 113.4 134.2 5873.9 836.3 1362.4 119889.6 31.0 115.1 136.3 6051.5 872.8 1329.8 117020.7 32.9 112.1 132.7 5757.4 844.1 1270.5 111804.2 34.0 110.5 130.8 5594.0 827.7 1237.6 108912.0 35.0 109.1 129.1 5452.4 813.4 1208.6 106357.2 36.0 107.8 127.5 5316.8 799.7 1180.3 103868.7 37.0 106.5 126.1 5187.0 786.5 1152.8 101442.3 39.0 104.2 123.3 4943.2 761.6 1099.6 96761.5
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-47 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW SI SPILL PATH FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 40.0 103.1 122.0 4828.4 749.8 146.7 12908.2 42.0 101.1 119.7 4611.4 727.4 146.7 12908.2 44.0 99.3 117.5 4409.4 706.4 146.7 12908.2 46.0 97.6 115.5 4220.4 686.7 146.7 12908.2 47.1 99.4 117.6 236.4 279.5 146.7 12908.2 51.1 98.0 116.0 232.5 274.8 146.7 12908.2 68.1 93.5 110.6 219.3 259.3 146.7 12908.2 99.1 89.2 105.5 229.0 245.8 146.7 12908.2 103.1 88.7 104.9 230.7 244.3 146.7 12908.2 135.1 84.5 100.0 247.9 232.2 146.7 12908.2 137.1 84.3 99.7 249.1 231.4 146.7 12908.2 173.1 78.8 93.2 268.9 216.2 146.7 12908.2 203.1 73.9 87.4 283.2 202.7 146.7 12908.2 212.1 72.4 85.6 287.3 198.7 146.7 12908.2 212.2 114.1 134.8 322.6 28.4 146.7 12908.2 1000.0 79.0 93.3 357.8 31.5 146.7 12908.2 10000.0 41.7 49.3 395.0 34.8 146.7 12908.2 100000.0 21.7 25.7 415.0 36.5 146.7 12908.2
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-48 Revision:
Sheet: 10 1 of 2 TABLE 6.2-48 DOUBLE-ENDED PUMP S UCTION GUILLOTINE (MINIMUM SAFETY INJECTION) POST-REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 165.4 221.0 273.5 365.3 127.7 180.4 219.1 271.1 367.3 127.7 185.4 206.3 255.3 380.0 125.1 190.4 206.5 255.6 379.8 124.8 205.4 205.0 253.7 381.3 124.7 250.4 200.3 247.9 386.1 124.5 255.4 200.4 248.0 386.0 124.3 280.4 197.6 244.5 388.8 124.3 290.4 197.1 243.9 389.2 124.0 320.4 193.7 239.7 392.7 124.0 330.4 192.6 238.4 393.7 123.9 335.4 177.2 219.3 409.1 122.3 355.4 175.1 216.7 411.2 122.2 370.4 173.1 214.2 413.2 122.3 375.4 172.8 213.8 413.6 122.2 395.4 170.1 210.5 416.3 122.3 405.4 168.9 209.0 417.5 122.3 670.4 168.9 209.0 417.5 122.3 670.5 98.8 121.4 487.5 129.0 825.4 94.6 116.2 491.7 125.3 895.4 93.1 114.3 493.3 123.4 910.4 92.7 113.9 493.6 123.0 995.4 90.8 111.5 495.6 120.7
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-48 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 1065.4 89.6 110.1 496.7 118.6 1210.4 87.3 107.1 499.1 117.9 1305.4 85.7 105.2 500.6 114.2 1350.4 85.0 104.3 501.4 112.4 1565.4 81.8 100.4 504.6 108.9 1761.0 81.7 100.3 504.6 108.5 1761.1 87.4 112.2 493.9 55.4 3600.0 73.3 95.9 513.1 59.4 3600.1 57.8 68.2 528.6 46.8 10000.0 42.9 50.7 543.4 47.8 100000.0 22.3 26.4 564.0 49.6 1000000.0 10.7 12.7 575.7 50.7
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-49 Revision:
Sheet: 10 1 of 2 TABLE 6.2-49 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MAXIMUM SAFETY INJECTION) POST-REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 165.0 184.7 230.2 1276.8 216.8 173.0 183.8 229.1 1277.7 216.9 203.0 183.7 229.0 1277.8 215.7 208.0 183.0 228.1 1278.5 215.7 218.0 183.4 228.6 1278.1 215.2 233.0 183.0 228.1 1278.5 214.7 238.0 182.3 227.2 1279.2 214.7 258.0 182.7 227.7 1278.8 208.2 263.0 181.9 226.7 1279.6 208.2 278.0 182.0 226.8 1279.5 207.7 328.0 180.9 225.5 1280.6 206.2 333.0 181.5 226.2 1280.0 205.8 348.0 180.5 225.0 1281.0 205.5 378.0 180.6 225.1 1280.9 204.4 408.0 179.6 223.8 1281.9 198.1 413.0 180.0 224.3 1281.5 197.8 428.0 179.4 223.6 1282.1 197.5 438.0 179.7 224.0 1261.8 197.1 463.0 178.9 223.0 1282.6 196.5 678.0 178.8 222.8 1282.7 196.0 678.1 98.6 122.0 1362.9 212.1 688.0 98.4 121.6 1363.1 211.8 773.0 96.0 118.7 1365.5 209.3 808.0 95.0 117.5 1366.5 208.3 938.0 92.1 113.9 1369.4 204.0 1208.0 87.3 107.9 1374.2 197.3 1213.0 87.3 107.8 1374.2 197.1 1458.0 83.3 102.8 1378.2 190.3 1636.9 83.3 102.8 1378.2 190.3 1637.0 88.9 113.9 1372.6 130.3
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-49 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 3600.0 73.3 95.9 1388.2 134.6 3600.1 57.8 68.2 1403.7 123.5 10000.0 42.9 50.7 1418.6 124.8 100000.0 22.3 26.4 1439.2 126.6 1000000.0 10.7 12.7 1450.8 127.7
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-50 Revision:
Sheet: 10 1 of 2 TABLE 6.2-50 0.6 DOUBLE-ENDED PUMP SUCTION GUILLOTINE POST-REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 164.8 224.7 280.2 361.6 127.5 169.8 224.3 279.7 362.0 127.4 174.8 211.7 264.0 374.6 124.7 204.8 208.2 259.6 378.2 124.6 214.8 207.1 258.2 379.3 124.6 224.8 206.8 257.6 379.6 124.3 244.8 205.0 255.7 381.3 124.1 274.8 202.2 252.1 384.2 123.9 279.8 201.7 251.5 384.6 123.9 314.8 197.6 246.4 388.7 123.8 319.8 181.4 226.2 404.9 122.3 334.8 180.0 224.5 406.3 122.2 354.8 177.6 221.5 408.8 122.3 359.8 177.6 221.4 408.8 122.1 404.8 171.7 214.1 414.6 122.3 414.8 170.8 212.9 415.6 122.2 439.8 167.6 209.0 418.7 122.3 704.8 167.4 208.7 418.9 122.2 704.9 98.0 121.2 488.4 128.9 709.8 97.8 121.0 488.5 128.8 789.8 95.6 118.2 490.8 126.8 829.8 94.6 117.0 491.7 125.8 1074.8 89.6 110.8 496.8 118.9 1214.8 87.3 107.9 499.0 118.4 1314.8 85.7 105.9 500.7 114.5 1449.8 83.5 103.1 502.9 112.5 1454.8 83.4 103.0 502.9 112.2 1574.8 81.8 101.0 504.5 109.2 1769.8 81.8 101.0 504.6 108.9 1769.9 87.5 112.3 498.9 55.6
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION U F S A R ENGINEERED S AFETY FEATURES TABLE 6.2-50 Revision:
Sheet: 10 2 of 2 BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME SECONDS LBM/SEC THOUSAND BTU/SEC LBM/SEC THOUSAND BTU/SEC 3600.0 73.4 96.2 512.9 59.5 3600.1 57.8 68.2 528.6 46.5 10000.0 42.9 50.7 543.4 47.8 100000.0 22.3 26.4 564.0 49.6 1000000.0 10.7 12.7 575.7 50.7
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-51 Revision:
Sheet: 10 1 of 1 TABLE 6.2-51 DOUBLE-ENDED PUMP SUCTION GUILLOTI NE (MINIMUM SAFETY INJECTION) MASS BALANCE TIME (SECONDS) .00 24.20 24.20 153.24 675.36 1761.02 3600.00 MASS (THOUSAND LBM)
INITIAL IN RCS AND ACC 802.38 802.38 802.38 802.38 802.38 802.38 802.38 ADDED MASS PUMPED INJECTION 0.00 0.00 0.00 71.34 370.38 1006.97 2085.27  TOTAL ADDED 0.00 0.00 0.00 71.34 370.38 1006.97 2085.27
***  TOTAL AVAILABLE  *** 802.38 802.38 802.38 873.72 1172.76 1809.35 2887.65 DISTRIBUTION REACTOR COOLANT 505.43 96.42 96.47 167.92 167.92 167.92 167.92 ACCUMULATOR 296.95 208.05 208.01 -.00 0.00 0.00 0.00 TOTAL CONTENTS 802.38 304.47 304.47 167.92 167.92 167.92 167.92 EFFLUENT BREAK FLOW 0.00 497.90 497.90 705.79 1004.84 1641.43 2719.73 ECCS SPILL 0.00 0.00 0.00 0.00 0.00 0.00 0.00 TOTAL EFFLUENT 0.00 497.90 497.90 705.79 1004.84 1641.43 2719.73 ***  TOTAL ACCOUNTABLE  *** 802.38 802.37 802.37 873.71 1172.75 1809.34 2887.65
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-52 Revision:
Sheet: 10 1 of 1 TABLE 6.2-52 DOUBLE-ENDED PUMP SUCTION GUILLO TINE (MAXIMUM SAFETY INJECTION) MASS BALANCE  TIME (SECONDS) .00 24.20 24.20 155.09 682.99 1636.62 3600.00 MASS (THOUSAND LBM)
INITIAL IN RCS AND ACC 802.38 802.38 802.38 802.38 802.38 802.38 802.38 ADDED MASS PUMPED INJECTION 0.00 0.00 0.00 184.61 937.28 2331.36 5200.49  TOTAL ADDED 0.00 0.00 0.00 184.61 937.28 2331.36 5200.49 ***  TOTAL AVAILABLE  *** 802.38 802.38 802.38 986.98 1739.66 3133.74 6002.87 DISTRIBUTI
 
ON REACTOR COOLANT 505.43 96.42 96.47 177.04 177.04 177.04 177.04  ACCUMULATOR 296.95 208.05 208.01 -.00 0.00 0.00 0.00 TOTAL CONTENTS 802.38 304.47 304.47 177.04 177.04 177.04 177.04 EFFLUENT BREAK FLOW 0.00 497.90 497.90 809.94 1562.61 2956.69 5825.82 ECCS SPILL 0.00 0.00 0.00 0.00 0.00 0.00 0.00 TOTAL EFFLUENT 0.00 497.90 497.90 809.94 1562.61 2956.69 5825.82 ***  TOTAL ACCOUNTABLE  *** 802.38 802.37 802.37 986.98 1739.65 3133.73 6002.86 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of recor
: d. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-53 Revision:
Sheet: 10 1 of 1 TABLE 6.2-53 0.6 DOUBLE-ENDED PUMP SUCTION GUILLOTINE MASS BALANCE TIME (SECONDS) .00 25.40 25.40 157.46 709.83 1769.76 3600.00 MASS (THOUSAND LBM)
INITIAL IN RCS AND ACC 802.38 802.38 802.38 802.38 802.38 802.38 802.38 ADDED MASS PUMPED INJECTION 0.00 0.00 0.00 72.58 392.14 1013.65 2086.82  TOTAL ADDED 0.00 0.00 0.00 72.58 392.14 1013.65 2086.82
***  TOTAL AVAILABLE  *** 802.38 802.38 802.38 874.95 1194.52 1816.02 2889.20 DISTRIBUTION REACTOR COOLANT 505.43 89.41 100.05 171.60 171.60 171.60 171.60 ACCUMULATOR 296.95 214.19 203.55 -.00 0.00 0.00 0.00 TOTAL CONTENTS 802.38 303.61 303.61 171.60 171.60 171.60 171.60 EFFLUENT BREAK FLOW 0.00 498.77 498.77 703.35 1022.92 1644.42 2717.60 ECCS SPILL 0.00 0.00 0.00 0.00 0.00 0.00 0.00 TOTAL EFFLUENT 0.00 498.77 498.77 703.35 1022.92 1644.42 2717.60 ***  TOTAL ACCOUNTABLE  *** 802.38 802.37 802.37 874.95 1194.51 1816.02 2889.19 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-54 Revision:
Sheet: 10 1 of 1 TABLE 6.2-54 THREE-FOOT SQUARED PUMP SUCTION SPLIT MASS BALANCE TIME (SECONDS) .00 37.20 37.20 160.96  MASS (THOUSAND LBM)
INITIAL IN RCS AND ACC 802.38 802.38 802.38 802.38 ADDED MASS PUMPED INJECTION 0.00 0.00 0.00 65.37 TOTAL ADDED 0.00 0.00 0.00 65.37 ***  TOTAL AVAILABLE  *** 802.38 802.38 802.38 867.75 DISTRIBUTION REACTOR COOLANT 505.43 109.11 109.15 180.01  ACCUMULATOR 296.95 199.15 199.10 -.00 TOTAL CONTENTS 802.38 308.25 308.25 180.01 EFFLUENT BREAK FLOW 0.00 494.12 494.12 687.73  ECCS SPILL 0.00 0.00 0.00 0.00 TOTAL EFFLUENT 0.00 494.12 494.12 687.73
***  TOTAL ACCOUNTABLE  *** 802.38 802.37 802.37 867.74
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-55 Revision:
Sheet: 10 1 of 1 TABLE 6.2-55 DOUBLE-ENDED HOT LEG GUILLOTINE MASS BALANCE TIME (SECONDS)
.00 21.00 21.00  MASS (THOUSAND LBM)
INITIAL IN RCS AND ACC 802.38 802.38 802.38 ADDED MASS PUMPED INJECTION 0.00 0.00 0.00  TOTAL ADDED 0.00 0.00 0.00
***  TOTAL AVAILABLE  ***
802.38 802.38 802.38 DISTRIBUTION REACTOR COOLANT 505.43 96.42 96.50  ACCUMULATOR 296.95 232.00 231.92  TOTAL CONTENTS 802.38 328.42 328.42 EFFLUENT BREAK FLOW 0.00 473.96 473.96  ECCS SPILL 0.00 0.00 0.00 TOTAL EFFLUENT 0.00 473.96 473.96 ***  TOTAL ACCOUNTABLE  ***
802.38 802.38 802.38 
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-56 Revision:
Sheet: 10 1 of 1  TABLE 6.2-56 DOUBLE-ENDED COLD LEG GUILLOTINE MASS BALANCE TIME (SECONDS) .00 22.00 22.00 212.09  MASS (THOUSAND LBM)
INITIAL IN RCS AND ACC 802.38 802.38 802.38 802.38 ADDED MASS PUMPED INJECTION 0.00 0.00 0.00 110.90 TOTAL ADDED 0.00 0.00 0.00 110.90 ***  TOTAL AVAILABLE  *** 802.38 802.38 802.38 913.28 DISTRIBUTION REACTOR COOLANT 505.43 68.43 84.88 162.79  ACCUMULATOR 296.95 176.40 159.95 -.00 TOTAL CONTENTS 802.38 244.33 244.83 162.79 EFFLUENT BREAK FLOW 0.00 504.84 504.84 548.39  ECCS SPILL 0.00 52.70 52.70 101.89  TOTAL EFFLUENT 0.00 557.54 557.54 750.27
***  TOTAL ACCOUNTABLE  *** 802.38 802.37 802.37 913.06
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-57 Revision:
Sheet: 10 1 of 2 TABLE 6.2-57 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MINIMUM SAFETY INJECTION) ENERGY BALANCE TIME (SECONDS) .00 24.20 24.20 153.24 675.36 1761.02 3600.00  ENERGY (MILLION BTU)      INITIAL ENERGY IN RCS, ACC,S GEN 894.34 894.34 894.34 894.34 894.34 894.34 894.34 ADDED ENERGY PUMPED INJECTION 0.00 0.00 0.00 6.28 32.59 88.61 183.50  DECAY HEAT 0.00 8.24 8.24 24.94 73.37 155.29 278.68 HEAT FROM SECONDARY 0.00 -17.91 -17.91 -17.91 -9.20 -.33 -.33 TOTAL ADDED 0.00 -9.66 -9.66 13.31 96.77 243.58 461.85
*** TOTAL AVAILABLE *** 894.34 884.68 884.68 907.66 991.11 1137.92 1356.19 DISTRIBUTION REACTOR COOLANT 303.11 19.36 19.37 42.95 42.95 42.95 42.95  ACCUMULATOR 26.13 18.31 18.30 -.00 0.00 0.00 0.00  CORE STORED 26.85 13.84 13.84 5.09 4.91 4.32 3.33  PRIMARY METAL 132.98 125.38 125.38 108.68 80.13 61.82 46.50 SECONDARY METAL 121.87 118.64 118.64 108.10 89.06 64.06 47.45 STEAM GENERATOR 283.40 278.28 278.18 246.96 204.34 153.83 117.22 TOTAL CONTENTS 894.34 573.71 573.71 511.77 421.39 326.98 257.45
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-57 Revision:
Sheet: 10 2 of 2  TIME (SECONDS) .00 24.20 24.20 153.24 675.36 1761.02 3600.00  ENERGY (MILLION BTU)      EFFLUENT BREAK FLOW 0.00 310.98 310.98 395.93 569.76 810.98 1098.79  ECCS SPILL 0.00 0.00 0.00 0.00 0.00 0.00 0.00  TOTAL EFFLUENT 0.00 310.98 310.98 395.93 569.76 810.98 1098.79
*** TOTAL ACCOUNTABILITY ***
894.34 884.69 884.69 907.70 991.15 1137.96 1356.24
*** TOTAL ACCOUNTABILITY ***
894.34 884.69 884.69 907.70 991.15 1137.96 1356.24
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-58 Revision:
Sheet: 10 1 of 1 TABLE 6.2-58 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MAXIMUM SAFETY INJECTION) ENERGY BALANCE  TIME (SECONDS) .00 24.20 24.20 155.19 682.99 1636.86 3600.00 ENERGY (MILLION BTU)
INITIAL ENERGY IN RCS, ACC,S GEN 894.34 894.34 894.34 894.34 894.34 894.34 894.34 ADDED ENERGY PUMPED INJECTION 0.00 0.00 0.00 16.25 82.48 205.16 457.64  DECAY HEAT 0.00 8.24 8.24 25.15 73.93 146.81 279.47 HEAT FROM SECONDARY 0.00 -17.91 -17.91 -17.91 -9.12 -.33 -.33  TOTAL ADDED 0.00 -9.66 -9.66 23.49 147.30 351.64 736.79
***  TOTAL AVAILABLE  *** 894.34 884.68 884.68 917.83 1041.64 1245.99 1631.13 DISTRIBUTION REACTOR COOLANT 303.11 19.36 19.37 44.71 44.71 44.71 44.71  ACCUMULATOR 26.13 18.31 18.30 -.00 0.00 0.00 0.00 CORE STORED 26.85 13.84 13.84 5.09 4.91 4.39 3.33 PRIMARY METAL 132.98 125.38 125.38 106.67 79.26 62.36 46.55 SECONDARY METAL 121.87 118.64 118.64 108.18 87.65 64.36 47.54 STEAM GENERATOR 283.40 278.18 278.18 247.16 200.80 154.57 117.46 TOTAL CONTENTS 894.34 573.71 573.71 511.80 417.33 330.38 259.59 EFFLUENT BREAK FLOW 0.00 310.98 310.98 406.09 624.37 915.67 1371.61  ECCS SPILL 0.00 0.00 0.00 0.00 0.00 0.00 0.00  TOTAL EFFLUENT 0.00 310.98 310.98 406.09 624.37 915.67 1371.61
***  TOTAL ACCOUNTABILITY  *** 894.34 884.69 884.69 917.89 1041.70 1246.05 1631.19 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-59 Revision:
Sheet: 10 1 of 1 TABLE 6.2-59 0.6 DOUBLE-ENDED PUMP S UCTION GUILLOTINE ENERGY BALANCE TIME (SECONDS) .00 25.40 25.40 157.46 709.83 1769.76 3600.00  ENERGY (MILLION BTU)      INITIAL ENERGY IN RCS, ACC,S GEN 894.34 894.34 894.34 894.34 894.34 894.34 894.34 ADDED ENERGY PUMPED INJECTION 0.00 0.00 0.00 6.39 34.51 89.20 183.64  DECAY HEAT 0.00 8.62 8.62 25.60 76.99 155.58 279.32 HEAT FROM SECONDARY 0.00 -17.09 -17.09 -17.09 -7.79 .48 .48  TOTAL ADDED 0.00 -8.48 -8.48 14.89 103.70 246.26 463.44
***  TOTAL AVAILABLE  *** 894.34 885.87 885.87 909.23 998.05 1140.61 1357.78 DISTRIBUTION REACTOR COOLANT 303.11 18.73 19.66 44.54 44.54 44.54 44.54  ACCUMULATOR 26.13 18.85 17.91 -.00 0.00 0.00 0.00  CORE STORED 26.85 13.79 13.79 5.09 4.91 4.33 3.33  PRIMARY METAL 132.98 125.80 125.80 109.62 79.92 62.03 46.66 SECONDARY METAL 121.87 119.72 119.72 109.46 88.73 64.37 47.73 STEAM GENERATOR 283.40 281.41 281.41 250.76 204.08 154.60 117.94 TOTAL CONTENTS 894.34 578.30 578.30 519.47 422.17 329.86 260.20 EFFLUENT BREAK FLOW 0.00 307.58 307.58 389.94 576.05 810.92 1097.77  ECCS SPILL 0.00 0.00 0.00 0.00 0.00 0.00 0.00  TOTAL EFFLUENT 0.00 307.58 307.58 389.94 576.05 810.92 1097.77
***  TOTAL ACCOUNTABLE  *** 894.34 885.88 885.88 909.41 998.22 1140.79 1357.96 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-60 Revision:
Sheet: 10 1 of 1 TABLE 6.2-60 THREE-FOOT SQUARED PUMP SUCTION SPLIT ENERGY BALANCE TIME (SECONDS) .00 37.20 37.20 160.96 ENERGY (MILLION BTU)
INITIAL ENERGY IN RCS, ACC,S GEN 894.34 894.34 894.34 894.34 ADDED ENERGY PUMPED INJECTION 0.00 0.00 0.00 5.75  DECAY HEAT 0.00 14.31 14.31 29.85 HEAT FROM SECONDARY 0.00 -15.58 -15.58 -15.58 TOTAL ADDED 0.00 -1.27 -1.27 20.02
***  TOTAL AVAILABLE  *** 894.34 893.08 893.08 914.37 DISTRIBUTION REACTOR COOLANT 303.11 22.71 22.72 44.95  ACCUMULATOR 26.13 17.52 17.52 -.00  CORE STORED 26.85 12.29 12.29 5.09  PRIMARY METAL 132.98 125.51 125.51 110.80 SECONDARY METAL 121.87 121.47 121.47 111.52  STEAM GENERATOR 283.40 286.72 286.72 256.62 TOTAL CONTENTS 894.34 586.22 586.22 528.97 EFFLUENT BREAK FLOW 0.00 306.86 306.86 385.43  ECCS SPILL 0.00 0.00 0.00 0.00 TOTAL EFFLUENT 0.00 306.86 306.86 385.43
***  TOTAL ACCOUNTABLE  *** 894.34 893.08 893.08 914.40
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-61 Revision:
Sheet: 10 1 of 1 TABLE 6.2-61 DOUBLE-ENDED HOT LEG GUILLOTINE ENERGY BALANCE TIME (SECONDS) .00 21.00 21.00  ENERGY (MILLION BTU)
INITIAL ENERGY IN RCS, ACC,S GEN 894.34 894.34 894.34 ADDED ENERGY PUMPED INJECTION 0.00 0.00 0.00  DECAY HEAT 0.00 8.41 8.41 HEAT FROM SECONDARY 0.00 -17.95 -17.95 TOTAL ADDED 0.00 -9.54 -9.54
***  TOTAL AVAILABLE  *** 894.34 884.80 884.80 DISTRIBUTION REACTOR COOLANT 303.11 23.95 23.95  ACCUMULATOR 26.13 20.42 20.41  CORE STORED 26.85 11.74 11.74  PRIMARY METAL 132.98 124.74 124.74  SECONDARY METAL 121.87 117.45 117.45  STEAM GENERATOR 283.40 274.67 274.67  TOTAL CONTENTS 894.34 572.97 572.97 EFFLUENT BREAK FLOW 0.00 311.84 311.84  ECCS SPILL 0.00 0.00 0.00  TOTAL EFFLUENT 0.00 311.84 311.84
***  TOTAL ACCOUNTABLE  *** 894.34 884.81 884.81
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-62 Revision:
Sheet: 10 1 of 1 TABLE 6.2-62 DOUBLE-ENDED COLD LEG GUILLOTINE ENERGY BALANCE TIME (SECONDS) .00 22.00 22.00 212.09  ENERGY (MILLION BTU)
INITIAL ENERGY IN RCS, ACC,S GEN 894.34 894.34 894.34 894.34 ADDED ENERGY PUMPED INJECTION 0.00 0.00 0.00 9.76  DECAY HEAT 0.00 6.35 6.35 29.85 HEAT FROM SECONDARY 0.00 -17.03 -17.03 -17.03 TOTAL ADDED 0.00 -10.68 -10.68 22.58
***  TOTAL AVAILABLE  *** 894.34 883.67 883.67 916.92 DISTRIBUTION REACTOR COOLANT 303.11 14.54 15.99 43.95  ACCUMULATOR 26.13 15.52 14.08 -.00  CORE STORED 26.85 16.34 16.34 5.09  PRIMARY METAL 132.98 126.38 126.38 110.65 SECONDARY METAL 121.87 119.53 119.53 112.12 STEAM GENERATOR 283.40 280.82 280.82 259.05 TOTAL CONTENTS 894.34 573.14 573.14 530.85 EFFLUENT BREAK FLOW 0.00 305.90 305.90 377.14  ECCS SPILL 0.00 4.64 4.64 8.97  TOTAL EFFLUENT 0.00 310.54 310.54 386.11
***  TOTAL ACCOUNTABLE  *** 894.34 883.68 883.68 916.96
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-63 Revision:
Sheet: 10 1 of 2 TABLE 6.2-63 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MINIMUM SAFETY INJECTION) PRINCIPAL PARAMETER TRANSIENTS TIME SECONDS FLOODING TEMP  &deg; F FLOODING RATE IN/SEC CARRYOVER FRACTION CORE HEIGHT FT DOWNCOMER HEIGHT  FT FLOW FRACTION INJECTION TOTAL ACCUMULATOR SPILL (CUBIC FEET PER SECOND) ENTHALPY BTU/LBM 24.2 204.9 0.000 0.000 0.00 0.00 .250 0.0 0.0 0.0 0.00 24.8 202.5 27.110 0.000 .52 1.76 .000 12030.7 12353.9 0.0 88.00 25.0 200.4 35.797 .000 1.05 1.97 .000 12305.6 12223.9 0.0 88.00 26.0 199.1 3.133 .314 1.51 5.32 .339 12103.3 11516.6 0.0 88.00 27.0 198.9 3.013 .455 1.66 8.73 .360 11570.7 10234.0 0.0 88.00 30.3 197.7 6.744 .674 2.15 15.00 .659 7041.8 7503.7 0.0 88.00 31.2 197.1 6.429 .693 2.31 16.00 .658 7712.7 7257.8 0.0 88.00 33.3 196.1 5.899 .725 2.62 16.00 .653 7279.1 5310.9 0.0 88.00 37.3 195.2 5.313 .743 3.11 16.00 .642 6615.5 5130.0 0.0 88.00 41.3 195.1 4.923 .749 3.54 16.00 .630 6073.4 5575.8 0.0 88.00 46.3 195.7 4.547 .753 4.03 16.00 .616 5521.8 5003.7 0.0 88.00 52.3 197.1 4.193 .754 4.57 16.00 .601 4968.7 4437.4 0.0 88.00 56.3 198.3 3.922 .754 4.91 16.00 .591 4651.0 4112.2 0.0 88.00 57.3 196.5 4.251 .756 4.98 15.95 .622 536.7 0.0 0.0 88.00 58.3 199.0 4.594 .756 5.07 15.74 .623 515.7 0.0 0.0 88.00 68.3 201.3 4.069 .754 5.51 14.81 .623 522.6 0.0 0.0 88.00 70.3 205.3 3.543 .752 6.06 13.02 .614 545.2 0.0 0.0 88.00 77.3 210.3 3.120 .748 6.54 13.11 .604 555.2 0.0 0.0 88.00 85.3 215.2 2.740 .745 7.04 12.59 .501 363.2 0.0 0.0 88.00 95.3 223.7 2.331 .741 7.58 12.25 .575 560.0 0.0 0.0 88.00
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-63 Revision:
Sheet: 10 2 of 2 TIME SECONDS FLOODING TEMP  &deg; F FLOODING RATE IN/SEC CARRYOVER FRACTION CORE HEIGHT FT DOWNCOMER HEIGHT  FT FLOW FRACTION INJECTION TOTAL ACCUMULATOR SPILL (CUBIC FEET PER SECOND) ENTHALPY BTU/LBM 105.3 231.1 2.125 .733 8.07 12.17 .558 574.2 0.0 0.0 88.00 115.3 233.2 1.951 .737 8.52 12.27 .545 576.8 0.0 0.0 88.00 127.3 245.5 1.323 .737 9.01 12.54 .533 578.5 0.0 0.0 88.00 141.3 252.7 1.744 .733 9.55 12.99 .525 570.5 0.0 0.0 88.00 153.2 256.2 1.713 .741 10.00 13.42 .522 570.9 0.0 0.0 88.00
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-64 Revision:
Sheet: 10 1 of 1 TABLE 6.2-64 DOUBLE-ENDED PUMP SUCTION GUILLOTINE (MAXIMUM SAFETY INJECTION) PRINCIPAL PARAMETER TRANSIENTS TIME SECONDS FLOODING TEMP  &deg; F FLOODING RATE IN/SEC CARRYOVER FRACTION CORE HEIGHT  FT DOWNCOMER HEIGHT  FT FLOW FRACTION INJECTION  TOTAL ACCUMULATOR SPILL (CUBIC FEET PER SECOND) ENTHALPY BTU/LBM 24.2 204.9 0.00 0.000 0.00 0.00 .250 0.0 0.0 0.0 0.00 24.8 202.5 27.621 0.000 .51 1.89 .000 13816.3 12353.7 0.0 88.00 25.0 200.4 37.282 .000 1.86 2.21 .000 13522.5 12219.6 0.0 88.00 25.9 199.2 3.314 .311 1.50 5.58 .339 13522.5 11559.9 0.0 88.00 25.9 198.9 3.113 .457 1.66 9.37 .360 12485.1 11022.5 0.0 88.00 29.3 198.0 7.453 .642 2.02 15.98 .672 8865.8 7721.2 0.0 88.00 30.7 196.8 6.853 .696 2.30 16.00 .670 8290.7 7149.1 0.0 88.00 32.3 195.9 6.406 .721 2.56 16.00 .667 7946.7 6780.6 0.0 88.00 36.3 194.3 5.764 .743 3.59 16.00 .657 7267.1 6049.9 0.0 88.00 40.3 194.0 5.349 .750 3.56 16.00 .647 6726.0 5470.8 0.0 88.00 45.3 194.3 4.961 .754 4.09 16.00 .636 6171.2 4879.5 0.0 88.00 50.3 195.2 4.650 .755 4.58 16.00 .625 5707.6 4390.2 0.0 88.00 55.3 196.5 4.385 .755 5.04 16.00 .616 5310.1 3972.2 0.0 88.00 59.4 197.8 2.527 .738 5.38 16.00 .420 1454.8 0.0 0.0 88.00 62.4 198.9 2.511 .739 5.55 16.00 .420 1454.7 0.0 0.0 88.00 71.4 202.8 2.456 .739 6.03 16.00 .422 1454.6 0.0 0.0 88.00 80.4 207.9 2.401 .739 6.51 16.00 .424 1454.6 0.0 0.0 88.00 91.4 213.9 2.334 .740 7.07 16.00 .426 1454.5 0.0 0.0 88.00 101.4 224.9 2.274 .741 7.57 16.00 .429 1454.4 0.0 0.0 88.00 111.4 226.2 2.214 .742 8.06 16.00 .432 1454.3 0.0 0.0 88.00 121.4 232.2 2.154 .742 8.53 16.00 .435 1454.3 0.0 0.0 88.00 133.4 239.2 2.082 .744 9.07 16.00 .439 1454.2 0.0 0.0 88.00 143.4 244.8 2.022 .745 9.51 16.00 .443 1454.1 0.0 0.0 88.00 155.1 250.8 1.954 .746 10.00 16.00 .448 1454.0 0.0 0.0 88.00 Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy re lease data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-65 Revision:
Sheet: 10 1 of 1 TABLE 6.2-65 SYSTEM PARAMETERS Plant model  4 loop, 12 ft core Core power (Mwt) 3652.8 Core inlet temperature ( F) 566.5 Steam pressure (psi) 1000 Assumed containment backpressure (psia) 74.7 
 
Note:  This table presents mass and/or energy release data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-66 Revision:
Sheet: 10 1 of 2  TABLE 6.2-66 FULL DOUBLE-ENDED MSLB AT HOT SHUTDOWN (WITH SAF IN CONTAINMENT SPRAY SYSTEM) MASS AND ENERGY RELEASES (FORWARD FLOW)
Tim e (sec) Mass Flow (lbm/sec) Enthalpy (Btu/lbm) 0.0 3229 1188.8 0.253 3145 1190.2 0.254 3145 1190.2 1.0 2896 1194.1 2.0 2549 1197.7 3.0 2272 1200.2 4.0 2044 1201.9 5.0 1855 1203.1 6.0 1698 1203.9 6.5 1631 1204.2 7.0 1565 1204.4 7.95 1457 1204.7 9.0 1354 1204.8 10.0 1268 1204.8 11.0 1208 1204.7 12.0 1199 1204.7 13.0 1187 1204.7 14.0 1171 1204.7 15.0 1154 1204.6 16.0 1134 1204.6 17.0 1113 1204.5 18.0 1092 1204.4 19.0 1071 1204.3 20.0 1050 1204.2 21.0 1045 1204.1 22.0 1025 1204.0 23.0 1006 1203.9 24.0 988 1203.8  Note: This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-66 Revision:
Sheet: 10 2 of 2 Time (sec) Mass Flow (lbm/sec) Enthalpy (Btu/lbm) 25.0 971 1203.7 26.0 954 1203.6 27.0 939 1203.5 28.0 924 1203.4 29.0 910 1203.2 30.0 897 1203.1 35.0 842 1202.6 40.0 800 1202.2 45.0 769 1201.8 50.0 746 1201.5 60.0 715 1201.0 70.0 698 1200.7 80.0 687 1200.6 90.0 681 1200.5 100.0 677 1200.4 120.0 671 1200.3 140.0 666 1200.2 160.0 661 1200.1 180.0 656 1200.0 200.0 650 1199.9 220.0 644 1199.8 240.0 635 1199.4 260.0 627 1199.3 280.0 615 1199.1 281.0 613 1199.1 281.2 32.46 1199.1 1800.0 32.46 1199.1 1800.1 0 0    0 0 
 
Note:  This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-67 Revision:
Sheet: 10 1 of 1 TABLE 6.2-67 FULL DOUBLE-ENDED MSLB AT HOT SHUTDOWN (WITH SAF IN CONTAINMENT SPRAY SYSTEM) MASS AND ENERGY RELEASES (REVERSE FLOW)  Tim e  (sec) Mass Flow  (lbm/sec) Enthalpy  (Btu/lbm) 0.0 9639 1188.8 0.253 9639 1188.8 0.254 4543 1189.4 1.0 4298 1191.3 2.0 4047 1193.5 3.0 3827 1195.3 4.0 3630 1196.8 5.0 3453 1198.1 6.0 3295 1199.2 6.5 3223 1199.4 7.0 2112 1199.4 7.95 0 1199.4  0 0.0 
 
Note:  This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-68 Revision:
Sheet: 10 1 of 2 TABLE 6.2-68 FULL DOUBLE-ENDED MSLB AT 102%
POWER (WITH SAF OF BROKEN LOOP MSIV) MASS AND ENERGY RELEASES (FORWARD FLOW) Time  (sec) Mass Flow  (lbm/sec) Enthalpy  (Btu/lbm) 0.0 2884 1193.2 0.250 2823 1194.0 0.251 2823 1194.0 1.0 2641 1196.2 2.0 2435 1198.3 3.0 2270 1200.0 4.0 2138 1200.9 5.0 2053 1201.5 6.0 2012 1201.8 6.5 1993 1202.0 7.0 1974 1202.1 8.0 1938 1202.4 9.0 1900 1202.6 10.0 1858 1202.9 11.0 1813 1203.2 12.0 1773 1203.4 13.0 1727 1203.6 14.0 1680 1203.9 15.0 1636 1204.0 16.0 1610 1204.2 17.0 1569 1204.4 18.0 1528 1204.5 18.6 1504 1204.5 19.0 1488 1204.6 20.0 1450 1204.7 Note:  This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-68 Revision:
Sheet: 10 2 of 2 Time  (sec) Mass Flow  (lbm/sec) Enthalpy  (Btu/lbm) 21.0 1414 1204.7 22.0 1379 1204.8 23.0 1346 1204.8 24.0 1315 1204.8 25.0 1285 1204.8 26.0 1257 1204.8 27.0 1231 1204.7 28.0 1206 1204.7 29.0 1183 1204.7 30.0 1161 1204.6 35.0 1071 1204.3 40.0 1007 1204.0 45.0 962 1203.7 50.0 931 1203.5 60.0 889 1203.2 70.0 862 1202.9 80.0 840 1202.7 90.0 823 1202.5 100.0 808 1202.3 120.0 787 1202.1 140.0 770 1201.8 160 755 1201.6 173.1 746 1201.5 173.1 32.46 1201.9 1800.0 32.46 1201.9
 
1800.1 0 0  0 0  Note:  This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-69 Revision:
Sheet: 10 1 of 1 TABLE 6.2-69 FULL DOUBLE-ENDED MSLB AT 102%
POWER (WITH SAF OF BROKEN LOOP MSIV) MASS AND ENERGY RELEASES (REVERSE FLOW)
Time  (sec) Mass Flow  (lbm/sec) Enthalpy  (Btu/lbm) 0.0 8611 1193.2 0.250 8611 1193.4 0.251 4058 1193.5 1.0 3944 1195.1 2.0 3813 1196.5 3.0 3711 1197.6 4.0 3630 1198.3 5.0 3562 1198.8 6.0 3505 1199.1 6.5 3479 1199.3 7.0 3335 1199.4 8.0 3098 1199.5 9.0 2760 1199.8 10.0 2473 1200.0 11.0 2185 1200.2 12.0 1898 1200.5 13.0 1610 1200.8 14.0 1323 1201.2 15.0 1035 1201.6 16.0 748 1202.2 17.0 460 1202.9 18.0 173 1203.8 18.6 0 1204.5  0 1204.5 Note:  This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-69 A Revision:
Sheet: 10 1 of 1 TABLE 6.2-69A MASS AND ENERGY BALANCE PEAK CALCULATED CONTAINMENT PRESSURE FOR MSLB ENERGY BALANCE (X 10 6 Btu)  Initial  0 sec Peak Pressure 281 sec End of Blowdown 1800 sec Containment atmosphere 3.2 200.5 122.2 Containment sump 1.1 20.9 161.1 Total Energy 4.3 221.4 283.3 Initial energy 4.3 4.3 4.3 Energy added by blowdown 0.0 260.3 345.1 Energy added by sprays 0.0 3.6 44.7 Energy removed by heat sinks 0.0 46.8 86.3 Total Energy 4.3 221.4 307.8 MASS BALANCE (X 10 3 lbm)  Initial  0 sec Peak Pressure 281 sec End of Blowdown 1800 sec Containment atmosphere 203.9 374.5 311.2 Containment sump 12.5 111.5 830.1 Total Mass 216.4 486.0 1141.3 Initial Mass 216.4 216.4 216.4 Mass added by blowdown 0.0 216.9 287.6 Mass added by sprays 0.0 52.3 657.5 Total Mass 216.4 485.6 1161.5  Note:  This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-69 B Revision:
Sheet: 10 1 of 1 TABLE 6.2-69B MASS AND ENERGY BALANCE PEAK CALCULATED CONTAINMENT TEMPERATURE FOR MSLB ENERGY BALANCE (X 10 6 Btu)  Initial  0 sec Peak Temperature 82 sec End of Blowdown 1800 sec Containment atmosphere 3.2 155.0 57.2 Containment sump 1.1 4.7 264.6 Total Energy 4.3 159.7 321.8 Initial energy 4.3 4.3 4.3 Energy added by blowdown 0.0 175.1 324.7 Energy added by sprays 0.0 0.0 91.1 Energy removed by heat sinks 0.0 19.7 63.8 Total Energy 4.3 159.7 356.3 MASS BALANCE (X 10 3 lbm)  Initial  0 sec Peak Temperature 82 sec End of Blowdown 1800 sec Containment atmosphere 203.9 330.4 252.8 Containment sump 12.5 31.8 1545.2 Total Mass 216.4 362.2 1797.0 Initial Mass 216.4 216.4 216.4 Mass added by blowdown 0.0 145.8 270.2 Mass added by sprays 0.0 0.0 1340.0 Total Mass 216.4 362.2 1826.6  Note:  This table presents mass and energy release data related to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the MSLB containment response at an analyzed core power level of 3659 MWt.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-75 Revision:
Sheet: 13 1 of 2 TABLE 6.2-75 CONTAINMENT SPRAY SYSTEM PARAMETERS Spray Additive Tank Quantity 1 Type Vertical cylinder Volume 10,700 gal Material Austenitic stainless steel Design code ASME Section III, Class 3 ANSI N18.2 safety class Class 3 Operating pressure Atmospheric Design temperature 100 F Maximum fluid temperature 98 F NaOH concentration 20% by weight Containment Spray Pumps Quantity 2 Type Centrifugal Horsepower 600 Hp Design flow 3010 gpm Operating flow (minimum) 2808 gpm NPSH required 21 ft Operating flow (recirculation) 3660 gpm NPSH required 23.6 ft. NPSH available 23.76 ft. Design temperature 300 F Design pressure 350 psig Design code ASME III, Class 2 ANSI N18.2 safety class Safety Class 2 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-75 Revision:
Sheet: 13 2 of 2 Refueling Water Storage Tank Quantity 1 Capacity 475,000 gal Material Austenitic stainless steel Type Vertical cylinder Design code ASME III, Class 2 ANSI N18.2 safety class Class 2 Concentration of boron 2400-2600 ppm boron Design temperature 100 F Maximum fluid temperature 98 F* Operating pressure Atmospheric Containment Spray Heat Exchanger Quantity 2 Type Shell and tube Design codes:
Shell side ASME III, Class 3 Tube side ASME III, Class 2 ANSI N18.2 safety class:
Shell side 3 Tube side 2
 
Material:
Shell side Carbon steel Tube side Austenitic stainless steel Design pressure:
Shell side 150 psig Tube side 346 psig Design temperature:
Shell side 200 F Tube side 300 F Containment Spray Headers and Nozzles Nozzle quantity 396 Material Austenitic Stainless Steel Design flow rate, per nozzle, gpm 15.2 Pressure drop, per nozzle, psi 40
* Maximum fluid temperature within the mixing chamber during the injection phase is 100F due to the exothermal reaction between sodium hydroxide and boric acid.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-76 Revision:
Sheet: 8 1 of 1 TABLE 6.2-76 HEAT EXCHANGER PERFORMANCE DATA Containment Spray Heat Exchanger
 
Type Vertical U-tube Number 2 Capacity (each) 96.7x10 6 Btu/hr Heat Transfer Area (each) 3468 ft 2 (Effective) Units Overall Heat Transfer Coefficient 431 (Service) 758 (Clean)
Tube (Hot) side fl ow rate (each) 1.5x10 6 lb/hr Shell (Cold) side flow rate (each) 2.35x10 6 lb/hr Tube side inlet temperature 244.8 F Tube side outlet temperature 180.8 F Shell side inlet temperature 119.9 F (a)Shell side outlet temperature 161.5 F  (a) For the design basis accident, this equipment will experience a 6F cooling water supply temperature transient (120F to 126F to 120F) over a 11/2 hour period, or 3F for a period of 5 hours (cooling tower operation), which will have an insignificant impact on analysis of piping stresses.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-77 Revision:
Sheet: 12 1 of 1 TABLE 6.2-77 CONTAINMENT COOLING - ACTIVE COMPONENT FAILURE ANALYSIS Component  Failure  Comments and Consequences Suction or discharge valve for RHR pump  Fails to open  Two parallel paths; only one required.
* Suction or discharge valve for spray pump  Fails to open  Two parallel paths; only one required.
* RHR pump  Fails to start; mechanical or loss of power  Two parallel paths; only one required.
* Spray pump  Fails to start; mechanical or loss of power  Two parallel paths; only one required.
* Containment penetration line  Cracked or clogged  Two parallel paths; only one required.
* Spray nozzles  Clogged  The strainer in sump will prevent large particles from entering spray system. 198 nozzles/train ensure no
 
significant decreases of flow. Automatic valve between spray additive tank and RWST  Fails to open  Two valves in parallel; each capable of permitting full flow.
* 1 containment spray heat exchanger and 1 residual heat removal heat exchanger in operation provide 100% of minimum required cooling S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-78 Revision:
Sheet: 12 1 of 1 TABLE 6.2-78 DELETED
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-79 Revision:
Sheet: 12 1 of 2 TABLE 6.2-79 CODES, STANDAR DS AND GUIDES APPLICABLE TO THE CONTAINMENT SPRAY SYSTEM A. USNRC Regulatory Guides Title Reg. Guide No. Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps 1.1  Assumptions Used for Evaluating the Potential Radiological Consequences of a LOCA for PWRs 1.4  Periodic Testing of Protection System Actuation Functions 1.22  Quality Group Classification and Standards 1.26  Seismic Design Classification 1.29  Control of Stainless Steel Welding 1.31  Protection Against Pipe Whip Inside Containment 1.46  Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems 1.47  Design Limits and Loading Combinations for Seismic Category I Fluid System Components 1.48  In-service Inspection of ASME Code Class 2 and 3 Nuclear Power Plant Components 1.51  Application of the Single Failure Criterion to Nuclear Power Protection Systems 1.53  Manual Initiation of Protection Actions 1.62  Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants LWR Edition 1.70  Sumps for Emergency Core Cooling and Containment Spray Systems 1.82  Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Station will be as Low as is Reasonably Achievable 8.8 B. 10 CFR Part 100, Reactor Site Criteria C. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code:
Material Specifications Section II  Nuclear Power Plant Components Section III  Nondestructive Examination Section V Pressure Vessels Division I Section VIII Welding and Brazing Qualifications Section IX  Rules for In-service Inspection of Nuclear Power Plant Section XI
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-79 Revision:
Sheet: 12 2 of 2 D. Institute of Electrical and Electronics Engineers (IEEE) Standards:
Title Std. No. Criteria for Protection Systems for Nuclear Power Generating Stations IEEE-279  Criteria for Class 1E Power Systems for Nuclear Power Generating Stations IEEE-308  Qualifying Class 1E Electric Equipment for Nuclear Power Generating Stations IEEE-323  Installation, Inspection and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations IEEE-336  Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE-338  Trial-Use Guide for Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Stations IEEE-344  Trial-Use Guide for General Principles for Reliability Analysis of Nuclear Power Generating Station Protection Systems IEEE-352  Trial-Use Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection Systems IEEE-379 E. American National Standard Institute (ANSI) Standards:
Pipe Threads B2.1  Steel Pipe Flanges and Flanged Fittings B16.5  Steel Butt Welding Fittings B16.9 Steel Weld Fittings B16.11  Butt Welding Ends B16.25  Power Piping B31.1  Nuclear Safety Criteria for Design of Stationary Pressurized Water Reactor Plants N18.2  Protective Coatings for Light Water Nuclear Reactor Containment Facilities N101.2 F. Tubular Exchanger Manufacturers Association (TEMA)
Mechanical Standards TEMA Class "R" Heat Exchangers G. Hydraulic Institute Standard for Rotary, Reciprocating, and Centrifugal Pumps H. American Institute of Steel Construction (AISC)
Steel Construction Manual I. American Society for Testing and Materials (ASTM)
Materials Specification J. Nuclear Energy Institute (NEI)
NEI No. Pressurized Water Reactor Sump Performance Evaluation Methodology 04-07 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-80 Revision:
Sheet: 8 1 of 1 TABLE 6.2-80 CONTAINMENT SPRAYED VOLUME Volume (10 6 ft 3) a) Bldg. volume above elevation 25'-0"  Cylinder (+) 1.447  Dome (+) 0.718 b) Annulus below elevation 25'-0" (+) 0.132 c) Refueling Canal
(+) 0.045 d) Pressurizer cubicle above elevation 25'-0" (-) 0.005 e) Steam generators above elevation 25'-0" (-) 0.015 f) Missile shield shadow
(-) 0.010 g) Equipment hatch
(-) 0.002 Total Sprayed Volume 2.310 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-81 Revision:
Sheet: 8 1 of 1  TABLE 6.2-81 EFFECT OF DROP SIZE ON SPRAY EFFICIENCY Drop  Size (microns) % Drops of Smaller  Size Terminal Velocity (fps) Residence Time (sec) Equil. Time  (sec) Distance Traveled  (ft) 500 68.9 6.7 20.0 0.5 < 3.3 1000 96 12.8 10.5 1.7 < 21.7 1500 99.99 17.9 7.5 3.8 < 68.0 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-82 Revision: 10 Sheet  1 of 1 TABLE 6.2-82 CONTAINMENT ENCLOSURE DESIGN AND PERFORMANCE DATA Free Volume, ft 3  Containment Enclosure Annulus 524,344 Electrical Penetration Areas 84,035 Mechanical Penetration Areas 70,320 RHR & SI Equipment Vaults 102,816 Containment Enclosure Equipment Area 92,568 Charging Pump Areas 12,000 Pressure, Inches H 2 O gauge  Normal Operation 0 Post Accident
-0.25* Temperature,  F  Normal Operation 50-104 Post Accident, Maximum 148 Exhaust Fans Number 2 Type Centrifugal Nominal Flow, scfm 2,000 Filters  Number of Trains 2 Moisture Separator, per train 1 HEPA, per train 2 Carbon Adsorber, per train 1 Thickness of Containment, in.
42-54 Thickness of Containment Enclosure, in.
15-36 Containment Wall Characteristics Coefficient of Linear Expansion, in./in.- F 6.5x10-6 Modulus of Elasticity, psi 3.12x10 6 (3000 psi concrete) 3.61x10 6 (4000 psi concrete) Thermal Conductivity, Btu/hr-ft- F 0.83 Thermal Capacitance, Btu/ft3- F 29.0
* A negative pressure of 0.685 iwg has to be established at the 21' -0" elevation of the Containment Enclosure Equipment Area to ensure that the required design negative pressure differential of 0.25 iwg is established at the top of the Containment Enclosure Annulus for the full range of design outside ambient temperatures.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 1 of 16  TABLE 6.2-83 CONTAINMENT ISOLATION SYSTEM DESIGN INFORMATION Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-1 II Main Steam (Loop I) Secondary Steam 30 Yes No 9 30"x24"x30"-MS-V86 ONo 68-6 Gate Elec-Hyd O CL CL FAI MSIS, CSW 5(4) A&B Yes 6    6"-MS-PV3001 ONo 68-9 Globe Pneumatic CL - CL FC CSW 70(4) A Yes 4    4"-MS-V393 ONo 65-6 Globe Pneumatic CL CL O FO CSW 20(4) A Yes 6    6"x10"-MS-V6,7,8,9,10 ONo 60-0 Safety Self CL CL CL - - - - No    4    4"x2-1/2"-MS-V204 ONo 72-0 Globe Motor LC LC LC FAI MSIS, CSW(13) 30(4) A Yes
 
1    1"-MSD-V44 ONo 24-7 Globe Motor O CL CL FAI MSIS, CSW 25(4) A -
X-2 II Main Steam (Loop II) Secondary Steam 30 Yes No 9 30"x24"x30"-MS-V88 ONo 71-9 Gate Elec-Hyd O CL CL FAI MSIS, CSW 5(4) A&B Yes 6    6"-MS-PV3002 ONo 66-9 Globe Pneumatic CL - CL FC CSW 70(4) B Yes 4    4"-MS-V394 ONo 68-8 Globe Pneumatic CL CL O *FO CSW 20(4) A&B Yes 6    6"x10"-MS-V22.23, 24,25,26 ONo 86-0 Safety Self CL CL CL - - - - No
* On Train A control power failure, valve position is FC; On Train B control power failure, valve position is FO; On instrument air failure, valve position is FO.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 2 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication 4    4"x2-1/2"-MS-V205 ONo 112-0 Globe Motor LC LC LC FAI MSIS, CSW(13) 30(4) A Yes
 
1    1"-MSD-V45 ONo 24-7 Globe Motor O CL CL FAI MSIS, CSW 25(4) A -
X-3 II Main Steam (Loop III) Secondary Steam 30 Yes No 9 30"x24"x30"-MS-V90 ONo 77-7 Gate Elec-Hyd O CL CL FAI MSIS, CSW 5(4) A&B Yes 6    6"-MS-PV3003 ONo 74-7 Globe Pneumatic CL - CL FC CSW 70(4) A Yes 6    6"x10"-MS-V36,37,38,39,40 ONo 72-0 Safety Self CL CL CL - - - - No 4    4"x2-1/2"-MS-V206 ONo 82-0 Globe Motor LC LC LC FAI MSIS, CSW(13) 30(4) A Yes 1    1"-MSD-V46 ONo 18-5 Globe Motor CL CL CL FAI MSIS, CSW 25(4) A -
X-4 II Main Steam (Loop IV) Secondary Steam 30 6  6 4  1 Yes No 9 30"x24"x30"-MS-V92 6"-MS-PV3004 6"X10"-MS-V50,51,52,53,54 4"x2-1/2"-MS-V207 1"-MSD-V47 O O O O O No No No No No 83-5 78-7 72-0  82-5  18-5 Gate Globe Safety
 
Globe
 
Globe Elec-Hyd Pneumatic Self  Motor  Motor O CL CL LC CL CL- CL LC CL CL CL CL LC CL FAI FC -  FAI  FAI MSIS, CSW CSW -
: MSIS, CSW(13)
MSIS, CSW 5(4) 70(4)-  30(4) 25(4) A&B B -  A  A Yes Yes No  Yes  -
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 3 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-5 II Main Feed Water (Loop I) Secondary Water 18 Yes No 9 18"-FW-V30 ONo 13-7 Gate Pneu/ Elec-Hyd O CL CL FAI S, RXT, SG, CSW 10(4) A&B Yes 4    4"-FW-V76 ONo 26-7 Stop/ Check Manual/ Self CL CL O - No X-6 II Main Feed Water (Loop II) Secondary Water 18 Yes No 9 18"-FW-V39 ONo 14-9 Gate Pneu/ Elec-Hyd O CL CL FAI S, RXT, SG, CSW 10(4) A&B Yes 4    4"-FW-V82 ONo 18-0 Stop/ Check Manual/ Self CL CL O - No X-7 II Main Feed Water (Loop III) Secondary Water 18 Yes No 9 18"-FW-V48 ONo 28-6 Gate Pneu/ Elec-Hyd O CL CL FAI S ,RXT, SG, CSW 10(4) A&B Yes 4    4"-FW-V88 ONo 20-6 Stop/ Check Manual/ Self CL CL O - No X-8 II Main Feed Water (Loop IV) Secondary Water 18 Yes No 9 18"-FW-V57 ONo 27-7 Gate Pneu/ Elec-Hyd O CL CL FAI S, RXT, SG, CSW 10(4) A&B Yes 4    4"-FW-V94 ONo 20-6 Stop/ Check Manual/ Self CL CL O - No X-9 E Residual Heat Removal Pump Suction (Loop I/ Hot Leg) Reactor Coolant 12 No No 2 12"-RC-V23 I Yes 21-7 Gate Motor LC O CL FAI CSW 120(4) A Yes 3    3"x4"-RC-V24 I Yes 8-4 Relief Self - - - - - - - No
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 4 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-10 E Residual Heat Removal Pump Suction (Loop IV/ Hot Leg) Reactor Coolant 12 No No 2 12"-RC-V88 I Yes 21-0 Gate Motor LC O CL FAI CSW 120(4) A Yes 3    3"x4"-RC-V89 I Yes 9-0 Relief Self - - - - - - - No X-11 I,E Residual Heat Removal Borated Water 8 Yes No 4 8"-RH-V14 ONo 0-8 Gate Motor O CL O FAI CSW 15(4) A Yes 6    6"-RH-V31 I No 106-2 Check Self - - - - - - - No 6    6"-RH-V15 I No 109-5 Check Self - - - - - - - No 3/4    3/4"-RH-V28 I No 35-0 Globe Pneumatic CL CL CL FC CSW, T 10 B Yes X-12 I,E Residual Heat Removal Borated Water 8 Yes No 4 8"-RH-V26 ONo 2-8 Gate Motor O CL O FAI CSW 15(4) B Yes 6    6"-RH-V30 I No 130-6 Check Self - - - - - - - No 6    6"-RH-V29 I No 128-1 Check Self - - - - - - - No 3/4    3/4"-RH-V27 I No 34-6 Globe Pneumatic CL CL CL FC CSW, T 10 A Yes
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 5 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-13 I,E Residual Heat Removal Borated Water 8 Yes No 3 8"-RH-V32 ONo 6-6 Gate Motor CL CL CL FAI CSW 40(4) B Yes 3/4    3/4"-RH-V49 I No 34-1 Globe Pneumatic CL CL CL FC CSW, T 10 A Yes 8    8"-RH-V70 ONo 6-5 Gate Motor CL CL CL FAI CSW 40(4) A Yes 8    8"-RH-V50 I No 53-5 Check Self - - - - - - - No 8    8"-RH-V51 I No 52-11 Check Self - - - - - - - No X-14 I,E Cont. Spray Borated Water 8 Yes No 7 8"-CBS-V11 OYes 2-8 Gate Motor CL CL O FAI CSW, CSAS 13(4) A Yes 8    8"-CBS-V12 I Yes 4-4 Check Self - - - - - - - No X-15 I,E Cont. Spray Borated Water 8 Yes No 7 8"-CBS-V17 OYes 2-8 Gate Motor CL CL O FAI CSW, CSAS 13(4) B Yes 8    8"-CBS-V18 I Yes 4-4 Check Self - - - - - - - No X-16 I Cont. On-Line Purge (Exhaust)
Cont. Atmosphere 8 No Yes 10 8"-COP-V4 OYes 2-9 Butterfly Pneumatic CL O CL FC CVIS, CSW 2 A Yes    8    8"-COP-V3 I Yes 1-4 Butterfly Pneumatic CL O CL FC CVIS, CSW 2 B Yes S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 6 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-17 I Equip. Vent (RCDT) Vent Gas 2 2 No Yes 7 2"-VG-FV-1661 2"-VG-FV-1712 O I Yes Yes 12-5 13-7 Globe Globe Solenoid Solenoid O O O O CL CL FC FC T,CSW T,CSW 2 2 A B Yes Yes X-18 I Cont. On-Line Purge (Supply)
Cont. Atmosphere 8
8 No Yes 10 8"-COP-V1 8"-COP-V2 O I Yes Yes 3-0 1-4 Butterfly Butterfly Pneumatic Pneumatic CL CL O O CL CL FC FC CVIS,CSW CVIS,CSW 2 2 A B Yes Yes X-19 I Post Accident Monitoring Sample Borated Water 1/2  1/2 No Yes 1 1/2"-SS-FV-2857 1/2"-SS-V273 O I Yes Yes 6-8  3-5 Globe  Check Solenoid  Self CL - CL - CL (11) - FC  - T,CSW  - 2  - B  - Yes  No X-20 I Primary Component Cooling Water Loop A (Supply) Demineralized Water (Corrosion Inhibitor) 12 12 1-1/2 No Yes 6 12"-CC-V168 12"-CC-V57 1-1/2"x2"-CC-V845 O I I Yes Yes Yes 5-6 36-5 31-4 Butterfly Butterfly Relief Pneumatic Pneumatic Self O O - O O - CL CL- FC**
FC** - P,CSW P,CSW - 10 10 - B A - Yes Yes No X-21 I Primary Component Cooling Water Loop A (Return) Demineralized Water (Corrosion Inhibitor) 12 12 1-1/2 No Yes 6 12"-CC-V122 12"-CC-V121 1-1/2"x2"-CC-V410 O I I Yes YesYes 4-11 28-10 23-4 Butterfly Butterfly Relief Pneumatic Pneumatic Self O O - O O - CL CL- FC**
FC** - P,CSW P,CSW - 10 10 - B A - Yes Yes No X-22 I Primary Component Cooling Water Loop B (Return) Demineralized Water (Corrosion Inhibitor) 12 12 1-1/2 No Yes 6 12"-CC-V257 12"-CC-V256 1-1/2"x2"-CC-V474 O I I Yes Yes Yes 7-9 32-11 29-4 Butterfly Butterfly Relief Pneumatic Pneumatic Self O O - O O - CL CL- FC**
FC** - P,CSW P,CSW - 10 10 - A B - Yes Yes No X-23 I Primary Component Cooling Water Loop B (Supply) Demineralized Water (Corrosion Inhibitor) 12 12 1-1/2 No Yes 6 12"-CC-V175 12"-CC-V176 1-1/2"x2"-CC-V840 O I I Yes Yes Yes 9-4 52-5 44-2 Butterfly Butterfly Relief Pneumatic Pneumatic Self O O - O O - CL CL- FC**
FC** - P,CSW P,CSW - 14 14 - A B - Yes Yes No X-24 I,E Safety Injection (Hi Head) Borated Water 3
4 4 3/4 Yes - No No - 3 3"-SI-V140 4"-SI-V138 4"-SI-V139 3/4"-SI-V158 I O O I No No NoNo 5-0 7-8 9-2 12-6 Check Gate Gate Globe Self Motor Motor Pneumatic CL CL CL CL CL CL CL CL O O O CL - FAI FAI FC - CSW,S CSW,S CSW,T - 10(4) 10(4) 10 - A B B No Yes Yes Yes X-25 I,E Safety Injection (Hi Head) Borated Water 4 2 2 3/4 Yes No 3 4"-SI-V102 2"-SI-V106 2"-SI-V110 3/4"-SI-V160 O I I
I No No No No 2-0 74-2 51-7 42-3 Gate Check Check Globe Motor Self Self Pneumatic CL-
- CL CL-
- CL O -
- CL FAI -
- FC CSW -
- T,CSW 12(4) -
- 10 A -
- B Yes No No Yes ** On electric power failure, valve position is FAI; On instrument air failure, valve position is FC.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 7 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Valve Loc Inside/ Outside (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-26 I,E Safety Injection (Hi Head) Borated Water 4 2 2 3/4 Yes No 3 4"-SI-V77 2"-SI-V81 2"-SI-V86 3/4"-SI-V134 O I I I No No NoNo 10-10 224-6 182-8 62-1 Gate Check Check Globe Motor Self Self Pneumatic CL-
- CL CL-
- CL O -
- CL FAI -
- FC CSW -
- CSW,T 12(4) -
- 10 B -
- A Yes No No Yes X-27 I,E Safety Injection (Hi Head) Borated Water 4
2 2 2 2 3/4 Yes No 4 4"-SI-V114 2"-SI-V118 2"-SI-V122 2"-SI-V126 2"-SI-V130 3/4"-SI-V131 O I I I I
I No No No No No No 3-3 75-0 144-4 72-7 108-2 33-11 Gate Check Check Check Check Globe Motor Self Self Self Self Pneumatic O - - -
- CL O - - -
- CL O - - -
- CL FAI - - -
- FC CSW - - -
- CSW,T 10(4) - - -
- 10 A - - -
- B Yes No No No No Yes X-28 I,E Chemical & Volume Control (RCP 1A Seal Water) Borated Water 2
2 (10) No 5 2"-CS-V166 2"-CS-V4 O I No No 3-9 2-0 Globe Check Motor Self O - O - O - FAI - CSW - 10(4) - A - Yes No X-29 I,E Chemical & Volume Control (RCP 1B Seal Water) Borated Water 2
2 (10) No 5 2"-CS-V162 2"-CS-V20 O I No No 2-10 2-0 Globe Check Motor Self O - O - O - FAI - CSW - 10(4) - A - Yes No X-30 I,E Chemical & Volume Control (RCP 1C Seal Water) Borated Water 2
2 (10) No 5 2"-CS-V158 2"-CS-V36 O I No No 3-9 6-3 Globe Check Motor Self O - O - O - FAI - CSW - 10(4) - A - Yes No X-31 I,E Chemical & Volume Control (RCP 1D Seal Water) Borated Water 2
2 (10) No 5 2"-CS-V154 2"-CS-V52 O I No No 3-3 3-6 Globe Check Motor Self O - O - O - FAI - CSW - 10(4) - A - Yes No X-32 I Equipment and Floor Drainage (RCDT) Mildly Contaminated Waste Water 3
3 1-1/2 No Yes 7 3"-WLD-V82 3"-WLD-V81 1-1/2"x2"-WLD-V213 O I I Yes Yes Yes 1-0 14-6 13-4 Globe Globe Relief Pneumatic Pneumatic Self O O - O O - CL CL- FC FC - T,CSW T,CSW - 10 10 - A B - Yes Yes No X-33 I,E Chemical & Volume Control (Charging) Borated Water 3
3 No No 5 3"-CS-V143 3"-CS-V144 O I No No 3-5 4-7 Gate Check Motor Self O - O - CL- FAI - S,CSW B - Yes No S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 8 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-34 I Equipment and Floor Drainage (RC Sump) Mildly Contaminated Waste Water 2 2 3/4 No Yes 6 2"-WLD-FV-8331 2"-WLD-FV-8330 3/4"x1"-WLD-V209 I O I Yes Yes Yes 17-3 3-0 17-1 Globe Globe Relief Solenoid Solenoid Self O O - O O - CL CL- FC FC - T,CSW T,CSW - 2 2 - B A - Yes Yes No X-35A I Safety Injection (Test Line) Borated Water 3/4 3/4 3/4 3/4 No Yes 3 3/4"-SI-V70 3/4"-SI-V62 3/4"x1"-SI-V247 3/4"-SI-V157 I O I O Yes Yes Yes Yes 8-10 3-11 10-0 59-8 Globe Globe Relief Globe Pneumatic Pneumatic Self Pneumatic CL CL- CL CL CL- CL CL CL- CL FC FC - FC T,CSW T,CSW - T,CSW 10 10 - 10 B A - A Yes Yes No Yes X-35B I Reactor Coolant (Pressurizer Steam/ Liquid Sample) Steam & Borated Water 1/2 1/2 1/2 3/4 No Yes 8 1/2"-RC-FV-2830 1/2"-RC-FV-2831 1/2"-RC-FV-2840 3/4"x1"-RC-V312 I
I O I Yes Yes Yes Yes 61-5 63-4 4-10 69-8 Globe Globe Globe Relief Solenoid Solenoid Solenoid Self CL CL CL- CL CL CL- CL CL CL- FC FC FC - T,CSW T,CSW T,CSW - 2 2
2 - B B A - Yes Yes Yes No X-35C I Reactor Coolant (RC Sample Loop I) Borated Water 1/2  1/2  1/2  3/4 No Yes 8 1/2"-RC-FV-2874 1/2"-RC-FV-2894 1/2"-RC-FV-2832 3/4"x1"-RC-V314 O O I I Yes Yes Yes Yes 5-1  5-0  84-9  75-11 Globe  Globe  Globe  Relief Solenoid  Solenoid  Solenoid  Self CL LC CL - CL LC CL - CL (11) LC(11) CL(11) - FC  FC  FC  - T,CSW  CSW  T,CSW  - 2  2(4)  2  - B  A  A  - Yes  Yes  Yes  No X-35D I Reactor Coolant (RC Sample Loop III) Borated Water 1/2  1/2  1/2  3/4 No Yes 8 1/2"-RC-FV-2876 1/2"-RC-FV-2896 1/2"-RC-FV-2833 3/4"x1"-RC-V337 O
O I I Yes Yes Yes Yes 7-5  11-10  67-6  75-4 Globe  Globe  Globe  Relief Solenoid  Solenoid  Solenoid  Self CL LC CL - CL LC CL - CL (11) LC(11) CL(11) - FC  FC  FC  - T,CSW  CSW  T,CSW  - 2  2(4)  2  - A  B  B  - Yes  Yes  Yes  No X-36A I Demineralized Water DM Water 1 1 1-1/2 No Yes 2 1"-DM-V4 1"-DM-V5 1-1/2"x2"-DM-V18 O I I Yes YesYes 2-3 16-5 15-10 Gate Gate Relief Manual Manual Self LC LC- -
- - - - - - - - - - - - - - - - - No No No X-36B I Nitrogen Gas (H1 Pressure)
N 2 Gas 1 1 No Yes 4 1"-NG-V13 1"-NG-V14 O I Yes Yes 3-9 8-11 Globe Globe Pneumatic Pneumatic CL CL CL CL CL CL FC FC T,CSW T,CSW 10 10 B A Yes Yes X-36C I Reactor Makeup Water DM Water 3 3 No Yes 1 3"-RMW-V30 3"-RMW-V29 O I Yes Yes 2-4 8-11 Globe Check Pneumatic Self O - CL- CL- FC - T,CSW B - Yes No S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 9 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-37A I Chemical & Volume Control (Letdown) Borated Water 3 3 No Yes 5 3"-CS-V150 3"-CS-V149 O I Yes Yes 3-5 113-0 Globe Gate Pneumatic Motor O O O O CL CL FC FAI T,CSW T,CSW 10 10 B A Yes Yes X-37B I Chemical & Volume Control (Excess Letdown) Borated Water 2
2 3/4 No Yes 5 2"-CS-V168 2"-CS-V167 3/4"x1"-CS-V794 I O I Yes YesYes 7-10 4-10 8-1 Globe Globe Relief Motor Motor Self O O - O O - CL CL- FAI FAI - T,CSW T,CSW - 10 10 - B A - Yes Yes No X-38/ 76A I Fire Protection Dry 4 4 No Yes 2 4"-FP-V592 4"-FP-V588 O I Yes Yes 7-11 3-0 Gate Check Manual Self LC- -
- -
- -
- -
- -
- -
- No No X-38/ 76B I Combustible Gas Control Containment Atmosphere 10 10 2 2 No Yes 8 10"-CGC-V46 10"-CGC-V45 2"-CGC-V43 2"-CGC-V44 I O O O Yes Yes YesYes 2-7 47-7 45-3 46-9 Check Gate Gate Gate Self Manual Manual Manual - LC LC LC -
- - - -
- - - -
- - - -
- - - -
- - - -
- - - No No No No X-39 I Spent Fuel Pool Cooling and Cleanup Dry 2 2 3/4 No Yes 2 2"-SF-V86 2"-SF-V87 3/4"x1"-SF-V101 I O I Yes YesYes 3-11 3-8 2-8 Gate Gate Relief Manual Manual Self LC LC- -
- - - - - - - - - - - - - - - - - No No No X-40A I Nitrogen Gas (Low Pressure)
N 2 Gas 1 1 No Yes 1 1"-NG-FV-4609 1"-NG-FV-4610 O I Yes Yes 6-1 8-3 Globe Globe Solenoid Solenoid CL CL CL CL CL CL FC FC T,CSW T,CSW 2 2 A B Yes Yes X-40B I PRT Sample Vent Gas 1/2 1/2 No Yes 8 1/2"-RC-FV-2836 1/2"-RC-FV-2837 I O Yes Yes 8-3 5-4 Globe Globe Solenoid Solenoid CL CL CL CL CL CL FC FC T,CSW T,CSW 2 2 B A Yes Yes X-41  Spare X-42  Spare X-43A  Spare                   
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 10 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-43B  Spare X-43C  Spare X-43D  Spare X-43E  Spare X-44  Spare X-45  Spare X-46  Spare X-47A  Spare X-47B  Spare X-47C  Spare X-47D  Spare               
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 11 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-48A II Primary Component Cooling Water Thermal Barrier Loop B (Supply) Demineralized Water (Corrosion Inhibitor) 6 No No 6 6"-CC-V1092 ONo 5-5 Butterfly Motor O O O FAI CSW - B Yes X-48B II Primary Component Cooling Water Thermal Barrier Loop B (Return) Demineralized Water (Corrosion Inhibitor) 6 No No 6 6"-CC-V1095 ONo 5-1 Butterfly Motor O O O FAI CSW - B Yes X-49A II Primary Component Cooling Water Thermal Barrier Loop A (Supply) Demineralized Water (Corrosion Inhibitor) 6  No No 6 6"-CC-V1101 ONo 4-6 Butterfly Motor O O O FAI CSW - A Yes X-49B II Primary Component Cooling Water Thermal Barrier Loop A (Return) Demineralized Water (Corrosion Inhibitor) 6 No No 6 6"-CC-V1109 ONo 4-4 Butterfly Motor O O O FAI CSW - A Yes X-50 III E Safety Injection PT-936 Demineralized Water  Yes No 9 N/A  No X-51  Spare X-52A I Air Sample Supply Containment Atmosphere 1/2 1/2 No No 11 1/2"-CAH-FV-6572 1/2"-CAH-FV-6573 O I Yes Yes 1-0 1-0 Gate Gate Solenoid Solenoid O O CL CL CL CL FC FC T,CSW T,CSW 2 2 A B Yes Yes X-52B I Air Sample Supply Containment Atmosphere 1/2 1/2 No No 11 1/2"-CAH-FV-6574 1/2"-CAH-V12 O I Yes Yes 1-0 1-0 Gate Check Solenoid Self O - CL- CL- FC - T,CSW A - Yes No S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 12 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X 56  (Numbers not used)
X-57A  Spare X-57B III E Safety Injection PT-937 Demineralized Water  Yes No 9 N/A  No X-57C  Spare X-58  Spare X-59  Spare X-60 E Containment Spray Borated Water 16 Yes No 2 16"-CBS-V14 ONo 2-0 Encap- sulated Gate Motor CL CL O FAI Recirc. CSW 34(4) B Yes X-61 E Containment Spray Borated Water 16 Yes No 2 16"-CBS-V8 ONo 2-0 Encap- sulated Gate Motor CL CL O FAI Recirc., CSW 34(4) A Yes X-62***
X-63 II Steam Generator Blowdown Secondary Water 3 No No 9 3"-SB-V9 ONo 7-2 Gate Pneumatic O O CL FC T,CSW 10 A&B Yes X-64 II Steam Generator Blowdown Secondary Water 3 No No 9 3"-SB-V10 ONo 10-0 Gate Pneumatic O O CL FC T,CSW 10 A&B Yes *** Penetration for fuel transfer tube, Type B test applicable.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 13 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-67 I Service Air Air 2 2 No Yes 7 2"-SA-V229 2"-SAV-V1042 O I Yes Yes 2 Gate Globe Manual Manual LC LC LC LC LC LC -
- -
- -
- -
- No No X-65 II Steam Generator Blowdown Secondary Water 3 No No 9 3"-SB-V11 ONo 8-0 Gate Pneumatic O O CL FC T,CSW 10 A&B Yes X-66 II Steam Generator Blowdown Secondary Water 3 No No 9 3"-SB-V12 ONo 6-4 Gate Pneumatic O O CL FC T,CSW 10 A&B Yes X-68  Instr. Air Air 2 No Yes 12 2"-IA-V530 2"-IA-V531 O I Yes Yes 14-11 Globe Check Pneumatic Self CL CL CL CL CL CL CL - T A - Yes No X-69  Spare X-70  Spare X-71A/ 74A E Combustible Gas Train B H 2 Analyzer Inlet Containment Atmosphere 1 No No 8 1"-CGC-V32 ONo 10-7 Globe Manual LC LC O - - - - No X-71B/ 74B I Combustible Gas Train B H 2 Analyzer Return Containment Atmosphere 1
1 No No 8 1"-CGC-V24 1"-CGC-V25 O I No No 10-7 10-2 Globe Check Manual Self LC- LC- O - -
- -
- -
- -
- No No X-71C/ 74C  X-76/ 73 I Combustible Gas Containment Exhaust Containment Atmosphere 2
2 No No 8 2"-CGC-V36
 
2"-CGC-V28 O I Yes Yes 27-6 13-1 Globe Globe Manual  Motor LC CL LC CL O (12)  O (12) -
FAI -
T,CSW -
12 -
B No Yes X-71D/ 74D I Leak Detection Containment Detection 1/2 1/2 No Yes 11 1/2"-LD-V1 1/2"-LD-V2 I O Yes Yes 2-1 1-6 Globe Globe Manual Manual LC LC LC LC LC LC -
- -
- -
- -
- No No S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 14 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-71E/ 74E  Spare X-72A/ 75A E Combustible Gas Train A H 2 Analyzer Inlet Containment Atmosphere 1 No No 8 1"-CGC-V10 ONo 10-7 Globe Manual LC LC O - - - - No X-72B/ 75B I Combustible Gas Train A H 2 Analyzer Return Containment Atmosphere 1 1 No No 8 1"-CGC-V3 1"-CGC-V4 O I No No 10-7 13-3 Globe Check Manual Self LC- LC- O - -
- -
- -
- -
- No No X-72C/ 75C  X-76/ 73 I Combustible Gas Containment Exhaust Containment Atmosphere 2  2 No No 8 2"-CGC-V15
 
2"-CGC-V14 O I Yes Yes 9-0 18-2 Globe Globe Manual Motor LC CL LC CL O (12)  O (12)  -
FAI -
T,CSW -
12 -
A No Yes X-72D/ 75D  Spare X-72E/ 75E  Spare X-72F/ 75F  Spare HVAC-1 (9) E Containment Air-Purge Containment Atmosphere 36 36 No  10    0-6 0-6            HVAC-2 (9) E Containment Air-Purge Containment Atmosphere 36 36 No  10 10    0-6 0-6 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 15 of 16  Valve Valve Position (5) Cnt. Penetration (8) Appl. Gen. Design Criteria or Reg. Guide (1) System  Name Fluid Contained Line Size (in.) Essential Sys. Pot. Bypass or Through-Line Leakage Figure 6.2-91 Sheet Isolation Valve Number Va lve Loc Ins id e/ Outs id e (2) Type C Tests Length of Pipe (ft-in) (3) Type Operator Normal Shutdown Post Accident Power Failure Containment Isolation Signal (6) Valve Close Time (Sec)
(7) Power Source Bus A or B Position Indication X-77A III E Reactor Coolant System LT-1321 LT-1322 LIS-1321 Demineralized Water  Yes No 9 N/A  No X-77B III E Reactor Coolant System LT-1321 LT-1322 LIS-1322 PT-403 Demineralized Water  Yes No 9 N/A  No X-77C III E Safety Injection PT-934 Demineralized Water  Yes No 9 N/A  No X-77D III E Safety Injection PT-2576 Demineralized Water  Yes No 9 N/A  No X-78A III E Reactor Coolant System LT-1311 LT-1312 LIS-1311 Demineralized Water  Yes No 9 N/A  No X-78B III E Reactor Coolant System LT-1311 LT-1312 LIS-1312 PT-405 Demineralized Water  Yes No 9 N/A  No X-78C III E Safety Injection PT-935 Demineralized Water  Yes No 9 N/A  No X-78D III E Safety Injection PT-2577 Demineralized Water  Yes No 9 N/A  No       
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-83 Revision:
Sheet: 14 16 of 16  NOTES FOR TABLE 6.2-83
: 1. I. - General Design Criteria 55 and 56  II - General Design Criterion 57  III - Sealed fluid instrument lines (Section 6.2.4.1.d.3)  E - Exception, as discussed in Subsection 6.2.4.2.m 2. I - Location inside the containment  O - Location outside the containment 3. Length of pipe from containment to outermost isolation valve 4. The listed valve close time is not essential for containment isolation. This valve does not receive a containment isolation signal. 5. O - Open FAI - Fail as is LO - Locked open CL - Closed FC - Fail closed LC - Locked closed 6. Containment Isolation Signals T - Phase A containment isolation signal P - Phase B containment isolation signal  CVIS - Containment ventilation isolation signal
 
Other  MSIS - Main steam line isolation signal S - Safety injection signal  RXT - Reactor trip signal coincident with a low reactor coolant Tavg signal  SG - Steam generator hi-hi signal  CSW - Control Switch  CSAS - Containment Spray actuation signal  RECIRC - Refueling water storage tank lo-lo level signal coincident with an S signal.
: 7. The closing times given are those that are specified as maximum
: 8. Alpha designations on penetration numbers are for clarification only and are not reflected in other documents
: 9. Testable blind flanges are installed on the outboard side of penetrations HVAC-1 and HVAC-2 during modes 1, 2, 3, and 4. These blind flanges form part of the containment boundary and are subject to Type B testing.
: 10. The seal injection function is non-essential, but this seal injection path does provide an alternate passive path for boration and RCS inventory control, which is credited for safe shutdown.
: 11. These sampling valves may be open intermittently to obtain a post accident sample (not a safety-related function). 12. These CGC valves would only be opened post accident for a containment building purge, a backup function. This would only be required if both safety-related hydrogen recombiners failed or if the post LOCA hydrogen generation rate was significantly greater than the design basis generation rate. 13. The MSIV bypass valves are normally locked closed with the breakers locked open. During startup and surveillance testing, the valves may be open (breakers closed), which allows MSIS and CSW operation.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-84 Revision:
Sheet: 8 1 of 1                                                            TABLE 6.2-84 HYDROGEN GAS ANALYZER DESIGN PARAMETERS A. General  Safety Classification 1E  Sensor Type Thermal catalytic Scale, % 0 to 10  Indication Local (4% H 2 concentration to be cl early indicated in red)
Accuracy, %
21/2 full scale Controls Off-Standby-Analyze Response Time, seconds 60, 90% step change Min. Sample Flow, scfm 0.08 @ 60 psig containment pressure Max. Sample Transit Time, minutes 10.0 B. Normal Sampling Temperature,  F 50 to 120 Pressure, psig
-3.5 to 1.5 Gamma Radiation, Rad/hr 50  Relative Humidity, % 5 to 95 C. LOCA and Post-LOCA Sampling Temperature,  F 380 maximum Pressure, psig -5 to 60  Gamma Radiation, Rads - water    Rads - air 1.5x10 8 6.3x10 7  Water 1900 ppm boron, pH of 10.5 Relative Humidity, %
100 D. Normal Operating Environment Temperature,  F 50 to 150 Pressure, psig 0 to 3  Humidity, % 10 to 95  Radiation Dose, Rads in40 year period 5x10 6    Time from sample entering the sample line in the containment until it reaches the hydrogen analyzer.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-85 Revision:
Sheet: 8 1 of 1 TABLE 6.2-85 APPLICABLE CODES, STANDARDS, AND REGULATORY GUIDES USED IN THE DESIGN OF THE ELECTRIC HYDROGEN RECOMBINER
: 1. NRC Regulatory Guides 1.7 (March 10, 1971) 1.28 (Safety Guide 28, 6/7/72) 1.29 (Rev. 2, 2/76) 1.38 (Rev. 1, 10/76)
: 2. 10 CFR 50, Appendix A, GDC 2, 41, 42, 43
: 3. Industry Codes ASME IX (Welding and Brazing Requirements)
National Electric Code National Electric Manufacturing Association National Fire Protection Association
: 4. Underwriters' Laboratories, Inc. 5. Institute of Electrical and Electronics Engineers IEEE 308-1971 IEEE 323-1974
 
IEEE 334-1974
 
IEEE 383-1974
 
IEEE 344-1975
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-86 Revision:
Sheet: 8 1 of 1                                                            TABLE 6.2-86 ELECTRIC HYDROGEN RE COMBINER TYPICAL PARAMETERS Power (maximum), kW 75 Capacity (minimum), scfm 100 Heaters  Number 4 Heater surface area/heater, ft 2 35 Maximum heat flux, Btu/hr-ft 2 2850 Maximum sheath temperature,  F 1550 Gas Temperature Inlet,  F 80 to 155 In heater section,  F 1150 to 1400 Materials Outer structure 300-Series SS Inner structure Incoloy-800 Heater element sheath Incoloy-800 Dimensions, ft Height 8 Width 3.9 Depth 4.6 Weight, lb 4500    Power can be controlled by SCR.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-87 Revision:
Sheet: 12 1 of 1 TABLE 6.2-87 CONTAINMENT BUIL DING ALUMINUM INVE NTORY Item Exposed Surface (ft
: 2) (Note 1 & 2)
Jib Crane 3 Nuclear Instrumentation 83 Flux Map Drive System 88 Rod Position Indicator Connectors 81 CRDM Connectors 69 Miscellaneous Valves 86 Containment Elevator Fans 127.9 Excore Detector Supports 28.6 CRD Fans, Blades, and Hubs 19.1 Polar Gantry Crane 40.8 I&C Transmitters & Misc.
34.9 Electrical Fixtures 27.6 Test Pump 3 Refueling Machine and Transfer System 26.4 Signs (HP, etc.)
8 Gaitronics System 140 Work Control Allocation (Note 2) 500 Contingency (Note 2) 433.7 Total 1800 (Note 1 & 2) Note 1: The 1800 ft 2 total value is the design basis value for hydrogen generation. Allocation among the specified categories may vary due to design changes in progress. Refer to Calculation 4.3.16.13F for actual allocation. Note 2: The aluminum contingency values in Table 6.2-87 are not the most restrictive values following OR12. Calculation C-S-1-83814, Seabrook Post Accident Chemical Product Formation, has more limiting values. Calculation C-S-1-83814 removed the Work Control Contingency. The limiting values are based on the effect of aluminum corrosion product on the containment sump strainers. Allocation of the design contingency may vary due to design changes in progress.
S EABROOK S TATION ENGINEERED S AFETY FEATURES TABLE 6.2-88 Revision:
Sheet: 11 1 of 1 TABLE 6.2-88 CONTAINMENT BUIL DING ZINC INVENTORY Item Exposed Surface (ft
: 2) Ductwork, Angles and Supports 48,357 Decking 7,850 Grating 24,395 Cable Trays 19,295 Tray Supports 86,190
 
Conduit 23,387 Conduit Supports 98,256 Instrument Tube Trays and Supports 21,040 Electrical Box Supports 8,416 Wire Mesh Doors 1,182 Scaffolding Components 3,400 Wire Mesh Tool Crib 648 Misc. Sheet Metal and Structural Members 700 Refueling Machine and Fuel Transfer System 52
 
Debris Interceptor Hardware 25 Work Control Allocation 6,000 Contingency 5,598 Total 355,000 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-89 Revision:
Sheet: 10 1 of 1 TABLE 6.2-89 PARAMETERS USED TO EVALUATE CGCS PERFORMANCE Reactor Power Level, MW t 3650.6 Initial Volume of Containment Atmosphere, scf 2.55x10 6 Mass of Zircaloy Cladding, lb 43,336 Hydrogen in Primary Coolant, scf 1127 Hydrogen Production Rates (scfh) from Corrosion:  Per the methodology of Reference 27, Attachment II.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-90 Revision:
Sheet: 8 1 of 1 TABLE 6.2-90 SYSTEMS OPEN TO CONTAINM ENT ATMOSPHERE AND/OR VENTED PRIOR TO AND DURING TYPE "A" TESTING
: 1. Reactor Coolant 2. Containment Online Purge 3. Equipment Vent - Hydrogenated 4. Equipment and Floor Drainage 5. Combustible Gas Control (including H 2 Analyzers) 6. Fire Protection
: 7. Containment Air Purge 8. Sample (Including Post-Accident) 9. SI Accumulators (Including Test/Fill Line) 10. Demineralized Water
: 11. Nitrogen Gas 12. Reactor Makeup Water 13. CS Purification/Letdown 14. Letdown Return
: 15. Reactor Cavity Cleanup 16. Containment Air Handling 17. Service Air
: 18. Instrument Air
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-91 Revision:
Sheet: 8 1 of 1 TABLE 6.2-91 SYSTEMS NOT VENTED DURING TYPE "A" TESTING System Justification High Pressure Safety Injecti on System is normally filled w ith water and operating under post-accident conditions Low Pressure Safety Injection System is normally filled with water and operating under post-accident conditions  Containment Spray System is normally filled with water and operating under post-accident conditions. Portion of the system inboard of the isolation valves is automatically vented inside containment through the spray nozzles. Chemical and Volume System maintains plant in safe Control (including RCP Seal Water) condition during the test - normally filled with water Residual Heat Removal System maintains plant in safe condition during the test - normally filled with water Cooling Water to Containment Fan C oolers System is normally filled with water under post-accident conditions. Fan coolers are used to maintain temperature during the test. Steam Generator Blowdown Closed system inside containment Main Steam Closed system inside containment Feedwater Closed system inside containment PCCW to Thermal Barrier Cooling Closed system inside containment
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-92 Revision:
Sheet: 8 1 of 6 TABLE 6.2-92 CONTAINMENT LINER PENETRATIONS
* Piping Penetration No. Essential (E) or Nonessential (NE)
System Selection Basis
 
X-1 E Main Steam from SG E 11A Decay Heat Removal X-2 E Main Steam from SG E 11B Decay Heat Removal X-3 E Main Steam from SG E 11C Decay Heat Removal X-4 E Main Steam from SG E 11D Decay Heat Removal X-5 E Feedwater to SG E 11A Decay Heat Removal X-6 E Feedwater to SG E 11B Decay Heat Removal X-7 E Feedwater to SG E 11C Decay Heat Removal X-8 E Feedwater to SG E 11D Decay Heat Removal X-9 NE RHR Pump Suction from HL #1 X-10 NE RHR Pump Suction from HL #4 X-11 E RHR to Safety Injection Low Pressure Injection X-12 E RHR to Safety Injection Low Pressure Injection X-13 E RHR to Safety Injection Hot Leg Injection X-14 E Containment Bldg. Spray Containment Spray X-15 E Containment Bldg. Spray Containment Spray X-16 NE Containment Online Purge X-17 NE Hydrogenated Vent Hdr.
X-18 NE Containment Online Purge X-19 NE Post-Accident Sample Req'd for Post-Accident Sampling X-20 NE* Primary Component Cooling Water, Loop A Desirable for Some Accidents; Isolate on Hi-2 Containment Pressure X-21 NE* Primary Component Cooling Water, Loop A Desirable for Some Accidents; Isolate on Hi-2 Containment
 
Pressure X-22 NE* Primary Component Cooling Water, Loop B Desirable for Some Accidents; Isolate  on Hi-2 Containment
 
Pressure X-23 NE* Primary Component Cooling Water, Loop B Desirable for Some Accidents; Isolate on Hi-2 Containment
 
Pressure
* Although these systems are nonessential, they are valuable in accident m onitoring and control.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-92 Revision:
Sheet: 8 2 of 6 Piping Penetration No. Essential (E) or Nonessential (NE)
System Selection Basis X-24 E SI from Charging Pumps High Pressure SI X-25 E SI, High Head Medium Pressure SI X-26 E SI, High Head Medium Pressure SI X-27 E SI, High Head Medium Pressure SI X-28 NE Seal Water to RC Pump 1A X-29 NE Seal Water to RC Pump 1B X-30 NE Seal Water to RC Pump 1C X-31 NE Seal Water to RC Pump 1D X-32 NE RC DRN TK to Primary DRN TK X-33 NE Chemical and Volume Control X-34 NE Floor Equipment Drain X-35 NE* RCS Sampling Required for Post-Accident Sampling Manual Bypass Operation X-35 NE SI Test  X-36 NE Demineralized Water  X-36 NE Nitrogen Gas  X-36 NE Reactor Makeup Water X-37 NE Letdown HX  X-37 NE RCP Seal Water Return X-38 NE* Combustion Gas Control May be Required for Purging Following Some Accident-Manual Operation X-38 NE Fire Protection X-39 NE Refueling Cavity Purification X-40 NE Nitrogen to PRT X-40 NE PRT Gas Sample X-41 NE Spare  X-42 NE Spare  X-43 E Press. Protection Containment Containment Monitoring X-44 NE Spare  X-45 NE Spare S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-92 Revision:
Sheet: 8 3 of 6 Piping Penetration No. Essential (E) or Nonessential (NE)
System Selection Basis X-46 NE Spare  X-47 E Press. Protection Monitoring Containment X-48 NE PCCW-Thermal Barrier Cooling X-49 NE PCCW-Thermal Barrier Cooling X-50 E Press. Protection Containment Monitoring X-51 NE Spare  X-52 E Containment Air Sample Containment Monitoring X-53-56  (Numbers Not Used)
X-57 E Press. Protection Containment Monitoring X-58 NE Spare  X-59 NE Spare  X-60 E Containment Recirc. Sump CBSTK-10B Containment Spray and SI Recirculation X-61 E Containment Recirc. Sump CBSTK-10A Containment Spray and SI Recirculation X-62 NE Fuel Transfer Tube X-63 NE Steam Generator Blowdown E11A X-64 NE Steam Generator Blowdown E11B X-65 NE Steam Generator Blowdown E11C X-66 NE Steam Generator Blowdown E11D X-67 NE Service Air  X-68 NE Instrument Air  X-69 NE Spare  X-70 NE Spare  X-71 NE* Combustible Gas Control Required for H 2  Post-Accident Sampling X-71 NE Spare  X-72 NE* Combustible Gas Control Required for H 2 Post-Accident Sampling X-72 NE Spare  X-77 E Reactor Vessel Level Indication System Required for Post-Accident Monitoring X-78 E Reactor Vessel Level Indication System Required for Post-Accident S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-92 Revision:
Sheet: 8 4 of 6 Piping Penetration No. Essential (E) or Nonessential (NE)
System Selection Basis Monitoring HVAC-1 NE Containment Air Purge Supply HVAC-2 NE Containment Air Purge E-1 NA Electrical NA E-2 NA Electrical NA E-3 NA Electrical NA E-4 NA Electrical NA E-5 NA Electrical NA E-6 NA Electrical NA E-7 NA Electrical NA E-8 NA Electrical NA E-9 NA Spare Header NA E-10 NA Spare Header NA E-11 NA Electrical NA E-12 NA Spare Header NA E-13 NA Spare Header NA E-14 NA Electrical NA E-15 NA Spare Header NA E-16 NA Electrical NA E-17 NA Electrical (Conax) NA E-18 NA Electrical NA E-19 NA Electrical NA E-20 NA Electrical NA E-21 NA Electrical NA E-22 NA Electrical NA E-23 NA Electrical NA E-24 NA Electrical NA E-25 NA Electrical NA E-26 NA Electrical NA E-27 NA Electrical NA E-28 NA Electrical NA S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-92 Revision:
Sheet: 8 5 of 6 Piping Penetration No. Essential (E) or Nonessential (NE)
System Selection Basis
 
E-29 NA Electrical NA E-30 NA Electrical NA E-31 NA Electrical NA E-32 NA Electrical NA E-33 NA Spare Header NA E-34 NA Electrical NA E-35 NA Electrical NA E-36 NA Spare Header NA E-37 NA Spare Header NA E-38 NA Electrical NA E-39 NA Electrical NA E-40 NA Electrical NA E-41 NA Electrical NA E-42 NA Electrical NA E-43 NA Electrical NA E-44 NA Spare Header NA E-45 NA Electrical NA E-46 NA Spare Header NA E-47 NA Electrical NA E-48 NA Electrical NA E-49 NA Electrical NA E-50 NA Electrical NA E-51 NA Electrical (Conax) NA E-52 NA Electrical NA E-53 NA Electrical NA E-54 NA Electrical NA E-55 NA Electrical NA E-56 NA Electrical NA E-57 NA Instrument X-77 NA E-58 NA Mechanical Spare ***NA  *** Flange welded to nozzle inside and outside containment. Blank test flange bolted to flange inside containment.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.2-92 Revision:
Sheet: 8 6 of 6 Piping Penetration No. Essential (E) or Nonessential (NE)
System Selection Basis
 
E-59 NA Mechanical Spare ***
NA E-60 NA Unused **NA E-61 NA Unused ** NA E-62 NA Unused ** NA E-63 NA Unused ** NA E-64 NA Instrument X-78 NA  ** Plate welded to nozzle inside containment.
NOTE 1 Viewed from outside containment looking south.
NOTE 2 Electrical penetrations are Westinghouse except for E-17 and E-51 which are Conax.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-1 Revision:
Sheet: 12 1 of 2 TABLE 6.3-1 EMERGENCY CORE COOLING SYSTEM COMP ONENT DESIGN PARAMETERS Accumulators Number 4 Design pressure (psig) 700 Design temperature (F) 300 Operating temperature ( F) 100 to 150 Normal operating pressure (psig) 650 Minimum operating pressure (psig) 585 Total volume (ft
: 3) 1350 each Nominal operating water volume (ft
: 3) 850 each Volume N 2 gas (ft 3) 500 Boric acid concentration (ppm) 2600-2900 Relief valve setpoint (psig) 700 Centrifugal Charging Pumps Number 2 Design pressure (psig) 2800 Design temperature (F) 300 Design flow (a) (gpm) 150 Design head (ft) 5800 Maximum flow (gpm) 550 Head at maximum flow (ft) 1400 Discharge head at shutoff (ft) 6200 Motor rating (hp) 600 Required NPSH at maximum flow (ft) 28 Available NPSH (ft) 40 Safety Injection Pumps Number 2 Design pressure (psig) 1750 Design temperature (F) 300 Design flow (gpm) 425 (a) Includes miniflow
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-1 Revision:
Sheet: 12 2 of 2 Safety Injection Pumps Design head (ft) 2700 Maximum flow (gpm) 669 Minimum head at maximum flow (ft) 1700 Minimum head at shutoff (ft) 3545 Motor rating (b) (hp) 450 Required NPSH at maximum flow (ft) 18 Available NPSH (ft) 43.4 Residual Heat Removal Pumps Number 2 Design pressure (psig) 600 Design temperature (F) 400 Design flow (gpm) 3000 Design head (ft) 375 Maximum flow (gpm) 5150 (c) Minimum head at maximum flow (ft) 275 Minimum head at shutoff (ft) 460 NPSH required at 4388 gpm (ft) (d) 18 NPSH available at 4388 gpm (ft) (d) 25.61 Power (hp) 400 Residual Heat Exchangers (See Subsection 5.4.7 for design parameters)  Motor-Operated Valves Maximum Opening or Closing Time Up to and including 8 inches, time (sec) Over 8 inches (e) 15 (f)(g)    (b) 1.15 service factor not included. (c) 5150 GPM corresponds to maximum pump flow on manufacturer's certified pump curve. Calculated maximum system flow is 4500 GPM during injection mode and 4388 GPM during recirculation. (d) These conditions reflect the most limiting suction conditions for the residual heat removal pumps during post-LOCA recirculation from the containment sump. (e) Closing time varies dependent upon size, type valve and type of actuator. (f) Does not include valves RH-V32 and RH-V70. (g) Active valves that do not receive an automatic signal for operation and that do not have a required stroke time in any analysis may have a longer maximum opening or closing time.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-2 Revision:
Sheet: 8 1 of 1 TABLE 6.3-2 EMERGENCY CORE COOLING SYSTEM RELIEF VALVE DATA Description Fluid Discharge Fluid Inlet Temperature Normal (&deg;F)
Set Pressure (psig)
Backpressure Constant (psig)
Maximum Total Backpressure (psig)
Capacity Outside Containment N 2 supply to accumulators N 2 gas 120 700 0 0 1500 scfm Safety injection pump discharge Water 120 1750 0 to 15 50 20 gpm Residual heat removal pump safety injection line Water 120 600 0 to 15 50 20 gpm Safety injection pumps suction header Water 100 220 0 to 15 50 25 gpm Accumulator to containment N 2 gas 120 700 0 0 1500 scfm Inside Containment N 2 supply to accumulators (NNS-B31.1 Portion)
N 2 gas 120 800 0 50 1184 scfm S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-3 Revision:
Sheet: 15 1 of 3 TABLE 6.3-3 MOTOR-OPERATED ISOLATION VALVES IN THE EMERGENCY CORE COOLING SYSTEM Location Valve Identification Interlocks Automatic Features Position Indication Alarms Accumulator isolation valves  (See Dwg. NHY-503907) SI-V3, -V17, -V32, -V47 "S" signal, RCS pressure > SI unblock pressure Opens on "S" signal closed, MCC power
 
available, and
 
RCS MCB Yes-out of position Safety injection pump suction from RWST CBS-V47, -V51 CBS-V49, -V53 None None MCB Yes-out of position RHR suction from RWST (See Dwg. NHY-503255) CBS-V2, -V5 Cannot be opened manually by control switch unless sump valve closed and
 
RHR recir. valve closed Opens on "S" signal MCB Yes-out of position RHR discharge to safety injection/charging RH-V35, -V36 Cannot be opened unless None safety injection pump miniflow isolated and RHR suction valve from RCS closed None MCB Yes-out of position* Safety injection  hot leg injection (See Dwg. NHY-503909) SI-V77, -V102 None None MCB Yes-out of position* RHR hot leg injection (See Dwg. NHY-503769) RH-V32, -V70 None None MCB Yes-out of position* Containment sump isolation valve (See Dwg. NHY-503252) CBS-V8, -V14 Cannot be opened in normal operation unless RHR suction valves from RCS
 
closed Opens on RWST low-low-1 with "S" signal MCB Yes-out of position*
* Out-of-position alarm is provided as part of the Bypass and Inoperable Status Alarm System MCB - Main Control Board S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-3 Revision:
Sheet: 15 2 of 3 Location Valve Identification Interlocks Automatic Features Position Indication Alarms CVCS (See Dwg. NHY-503335) suction from RWST LCV-112 D&E "S" signal and CVC tank low level Opposite Train VCT Isolation Valve not full open.
Opens on "S" signal and CVS tank low level Opens on opposite train VCT valve not full open (not credited for the ECCS function). MCB Yes-out of position Safety injection pump to cold leg (See Dwg. NHY-503909) SI-V114 None None MCB Yes-out of position CVCS normal discharge (See Dwg. NHY-50337) CS-V142, -V143 "S" signal Closes on "S" signal MCB Yes-out of position Cold leg isolation SI-V138, -V139 "S" signal Opens on "S" signal MCB Yes-out of position Charging pump/safety injection pump crossover (See Dwg. NHY-503338) CS-V460, -V461 CS-
 
V475 None None MCB Yes-out of position RHR to RCS cold legs (See Dwg. NHY-503769) RH-V14, -V26 None None MCB Yes-out of position* Safety injection pump miniflow (See Dwg. NHY-503911 and
 
NHY-503901) SI-V89, -V90, -V93 Cannot be opened unless None RHR discharge to safety injection and to charging pumps closed  MCB Yes-out of position* RHR cross connect (See Dwg. NHY-503765) RH-V21, -V22 None None MCB Yes-out of position*
* Out-of-position alarm is provided as part of the Bypass and Inoperable Status Alarm System MCB - Main Control Board S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-3 Revision:
Sheet: 15 3 of 3 Location Valve Identification Interlocks Automatic Features Position Indication Alarms Safety injection pump cross connect  (See Dwg. NHY-503912) SI-V111, -V112 None None MCB Yes-out of position* Charging pump miniflow (See Dwg. NHY-503398 &
 
503380) CS-V196, -V197 "S" signal in conjunction with charging pump Hi flow Closes on "S" signal in conjunction with charging pump
 
Hi flow MCB Yes-out of position S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-4 Revision:
Sheet: 8 1 of 2 TABLE 6.3-4 MATERIALS EMPLOYED FO R EMERGENCY CORE COOLING SYSTEM COMPONENTS Component Material Accumulators Carbon steel clad with austenitic stainless steel Pumps  Centrifugal charging Austenitic stainless steel Safety injection Austenitic stainless steel Residual heat removal Austenitic stainless steel Residual heat exchangers Shell Carbon steel Shell end cap Carbon steel Tubes Austenitic stainless steel Channel Austenitic stainless steel Channel cover Austenitic stainless steel Tube sheet Austenitic stainless steel Valves Motor-operated valves containing radioactive fluids Pressure containing parts Austenitic stainless steel or equivalent Body-to-bonnet bolting and nuts Low alloy steel Seating surfaces Stems Stellite No. 6 or equivalent Austenitic stainless steel or 17-4 pH stainless Motor-operated valves containing nonradioactive, boron-free fluids  Body, bonnet and flange Carbon steel Stems Corrosion resistant steel Diaphragm valves Austenitic stainless steel Accumulator check valves Parts contacting borated water Austenitic stainless steel Clapper arm shaft 17-4 pH stainless
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-4 Revision:
Sheet: 8 2 of 2 Relief valves Stainless steel bodies Au stenitic stainless steel Carbon steel bodies Carbon steel All nozzles, discs, spindles and guides Austenitic stainless steel Bonnets for stainless steel valves without balancing bellows Stainless steel or plated carbon steel All other bonnets Carbon steel Piping  All piping in contact with borated water Austenitic stainless steel
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-5 Revision:
Sheet: 15 1 of 6  TABLE 6.3-5 FAILURE MODE AND EFFECT S ANALYSIS - EMERGENCY CORE COOLING SYSTEM - ACTIVE COMPONENTS Component Failure Mode ECCS Operation Phase *Effect on System Operation
**Failure Detection Method Remarks 1. Motor-operated gate valve LCV-112 B (LCV-112C analogous)
Fails to close on demand.
Injection - cold legs of RC loops. Failure reduces redundancy of providing VCT discharge isolation. No effect on safety for system operation; isolation valves LCV-112C and CS-V192 provide backup tank discharge isolation. Valve position indication (open to closed position change) at MCB. Valve closed position monitor light for group monitoring of components at MCB.
Valve is electrically interlocked with isolation valve LCV-112D. Valve closes on actuation by a SI "S" signal provided isolation valve LCV-112D is at a full open position. 2. Motor-operated gate valve LCV-112 D (LCV-112E analogous)
Fails to open on demand.
Injection - cold legs of RC loops. Failure reduces redundancy of providing fluid flow from RWST to suction of HHSI/CH pumps. No safety effect on system operation. Alternate isolation valve (LCV-112E) opens to provide backup flow path to suction of HHSI/CH pumps. Same methods of detection as that stated for item #1 except open position monitor light for group monitoring of components at MCB and
 
closed to open position change indication at
 
MCB. Valve is electrically interlocked with the instrumentation that monitors fluid level of the VCT. Valve opens upon actuation by a SI "S" signal or upon
 
actuation by a low-low-level VCT signal. Valve is electrically interlocked with isolation valve LCV-112C. Valve opens when LCV-112C (LCV-112B analogous) leaves the full open position (not credited for the ECCS function). 3. Centrifugal charging pump CS-P-2A (CS-P-2B analogous)
Fails to deliver working fluid. Injection and
 
recirculation cold legs of RC loops. Failure reduces redundancy of providing emergency coolant to the RCS at prevailing incident RCS pressure.
Fluid flow from HHSI/CH pump CS-P-2A will be lost.
Minimum flow requirements at prevailing high RCS pressures will be met by HHSI/CH pump CS-P-2B
 
delivery. HHSI/CH pump discharge header flow (FI-917) at MCB. Trip/closed pump switchgear circuit breaker indication on MCB.
Circuit breaker close position monitor light for group monitoring of components at MCB.
Breaker trip alarm at MCB. One HHSI/CH pump is used for normal charging of RCS during plant operation. Circuit breaker aligned to close on actuation by a SI "S" signal. 4. Motor-operated globe valve CS-V196 (CS-V197 analogous)
Fails to close on demand.
Injection-cold legs of RC loops. Failure reduces redundancy of providing isolation of HHSI/CH pump miniflow line. No effect on safety for system operation. Valve position indication (open to closed position change) at MCB. Valve closed position monitor light and alarm for group monitoring of components at MCB.
Valve aligned close upon actuation to by a SI "S" signal.
* See list at end of table for definition of acronyms and abbreviations used.
** As part of plant operation, periodic tests, surveillance inspections and instrument calibrations are made to monitor equipment and performance. Failures may be detected during such monitoring of equipment in addition to detection.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-5 Revision:
Sheet: 15 2 of 6  Component Failure Mode ECCS Operation Phase *Effect on System Operation
**Failure Detection Method Remarks 5. Motor-operated gate valve CS-V143 (CS-V142 analogous)
Fails to close on demand.
Injection-cold legs of RC loops. Failure reduces redundancy of providing isolation of HHSI/CH pump discharge to normal charging line of CVCS. No effect on safety for system operation. Alternate isolation valve (CS-V142) provides backup normal CVCS charging line isolation. Same method of detection as that stated for item #1. Valve aligned to close upon actuation by a SI "S" signal. 6. Motor-operated gate valve SI-V138 (SI-V139 analogous)
Fails to open on demand.
Injection-cold legs of RC loops. Failure reduces redundancy of fluid flow paths from HHSI/CH pumps to the RCS. No effect on safety for system operation. Alternate (SI-V139) open to provide backup flow path from HHSI/CH pumps to RCS. Same method of detection as that stated for item #4. Valve aligned to open upon actuation by a SI "S" signal. 7. Motor-operated globe valve RH-FCV-610 (RH-FCV-611 analogous) a. Fails to close on demand. b. Fails closed. Injection-cold legs of RC loops.
 
Injection-cold legs of RC loops. a. Failure reduces working fluid delivered to RCS from RHR pump RH-P-8A. Minimum flow requirements for LHSI will be met by LHSI/RHR pump RH-P-8B delivering working fluid to RCS. b. Failure results in an insufficient fluid flow through LHSI/RHR pump RH-P-8A for a small LOCA or steam line break resulting in possible pump damage. If pump becomes inoperative minimum flow requirements for LHSI will be met by LHSI/RHR pump HR-P-8B delivering working fluid to RCS. a. Valve position indication (open to closed position change) at MCB. RHR pump
 
return line to cold legs flow indication (FI-618) at MCB. b. Same as that stated above for failure mode "Fails to close on demand" except closed to open position change at MCB. Valve is regulated by signal from flow transmitter located in pump discharge
 
header. The contro l valve opens when the RHR pump discharge flow is less than ~700 gpm and closes when the flow exceeds ~1400 gpm. 8. Residual heat removal pump RH-P-8A(Pump RH-P-8A) analogous)
Fails to deliver working fluid.
Injection-cold legs of RC loops. Failure reduces redundancy of providing emergency coolant to the RCS from the RWST at low RCS pressure (195 psig). Fluid flow from LHSI/RHR pump RH-P-8A will be lost. Minimum flow requirements for LHSI will be met by LHSI/RHR pump RH-P-8B delivering working fluid. RHR pump return line to cold legs flow indication (FI-618) at MCB. RHR pump discharge pressure (PI-614) at MCB. Open pump switchgear circuit breaker indication at MCB. Circuit breaker close position monitoring light and alarm for group monitoring of components at MCB. Breaker trip alarm at MCB. The RHR pump is sized to deliver
 
reactor coolant through the RHR heat exchanger to meet plant cooldown requirements and is used during plant cooldown and startup operations. The pump circuit breaker is aligned to close on actuation by a SI "S" signal. 9. Safety injection pump SI-P-6A(Pump SI-P-6B analogous)
Fails to deliver working fluid.
Injection-cold legs of RC loops. Failure reduces redundancy of providing emergency coolant to the RCS from the RWST at high RCS pressure
 
(1520 psi). Fluid flow from HHSI/SI pump SI-P-6A will be lost. Minimum flow requirements for HHSI will be met by HHSI/SI pump SI-P-6B delivering working fluid. SI pumps discharge pressure (PI-919) at MCB.
SI pump discharge flow (FI-918) at MCB.
Open pump switchgear circuit breaker
 
indication at MCB. Circuit breaker close position monitor light and alarm for group monitoring of components at MCB. Breaker trip alarm at MCB. Pump aligned too close on circuit breaker actuation by a SI "S" signal.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-5 Revision:
Sheet: 15 3 of 6  Component Failure Mode ECCS Operation Phase *Effect on System Operation
**Failure Detection Method Remarks 10. Motor-operated globe valve CBS-V8 (CBS-V14 analogous)
Fails to open on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing fluid from the Containment Sump to the RCS during recirculation. LHSI/
RHR pump RH-P-8A will not provi de recirculation flow. Minimum LHSI flow requirements will be met through
 
opening of isolation valve CBS-V14 and recirculation of fluid by LHSI/RHR pump RH-P-8B. Same method of detection as that stated for item #6. In addition failure may be detected through monitoring of RHR pump return line to cold legs flow indication (FI-618) and RHR pump discharge pressure (PI-614) at MCB. Valve is actuated to open by SI "S" SIGNAL in coincidence with two-out-of-four "Low low Level" RWST signals. Valve is electrically interlocked from remotely being opened from MCB by isolation valves CBS-V2, RC-V23 and RC-V22. 11. Motor- operated gate valve CBS-V2 (CBS-V5 analogous)
Fails to close on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing flow isolation of Containment Sump from RWST. No effect on safety for system operation. Alternate check isolation valve CBS-V55 provides backup isolation. Same method of detection as that stated for item #4. Valve is electrically interlocked with isolation valve CBS-V8 and RH-V35 and may not be opened unless these valves are closed, for manual operation from main control board. Valve opens automatically on "S" signal. 12. Motor- operated gate valve RH-V14.
Fails to close on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing LHSI/RHR pump discharge flow path isolation of RCS. No effect on safety for system operation. Alternate isolation valve RH-V26
 
will be closed to isolate altern ate flow path to cold legs. Same method of detection as that stated for item #4. 13. Motor-operated globe valve SI-V93 Fails to close on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing isolation of HHSI/SI pump's miniflow line isolation from RWST. No effect on safety for system ope ration. Alternate isolation valves SI-V89 and SI-V90 in each pumps' miniflow line provide backup isolation. Same method of detection as that stated for item #4. Valve is electrically interlocked with isolation valves RH-V35 and RH-V36 and may not be opened unless these
 
valves are closed. 14. Motor-operated globe valve SI-V90 (SI-V89 analogous)
Fails to close on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing isolation of HHSI/SI pump SI-P-6A mini- flow isolation from RWST.
No effect on safety for system operation. Alternate isolation valve SI-V93 in main miniflow line provides backup isolation. Same method of detection as that stated for item #4. Same remark as stated for item #16. 15. Motor-operated gate valve RH-V35 Fails to open on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing NPSH to suction of HHSI/CH pumps from LHSI/RHR pumps. No safety effect on system operation. Minimum NPSH to HHSI/CH pump suction will be met by flow from LHSI/RHR pump RH-P-8B via cross-tie line and opening of isolation valve CS-V460 or CS-V461 and isolation valve RH-V36. Same method of detection as that stated for item #4. Valve is electrically interlocked with isolation valves SI-V90, SI-V89, SI-V93, RC-V23, RC-V22 and CBS-V8. Valve cannot be opened unless valve SI-V93 or SI-V90 and
 
SI-V89 valves are closed; valve RCS-V23 or RCS-V22 is closed, and CBS-V8 is open.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-5 Revision:
Sheet: 15 4 of 6  Component Failure Mode ECCS Operation Phase *Effect on System Operation
**Failure Detection Method Remarks 16. Motor-operated gate valve RH-V36.
Fails to open on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing NPSH to suction of HHSI/SI pumps from LHSI/RHR pumps. No effect safety for system operation. Minimum NPSH to HHSI/SI pump suction will be met by flow from LHSI/RHR pump RH-P-8A via cross-tie line and opening of isolation valve CS-V460 or CS-V461 and isolation valve RH-V35. Same method of detection as that stated for item #4. Valve is electrically interlocked with isolation valves, SI-V90, SI-V89 SI-V93, and CBS-V14, RC-V88 and RC-V87. Valve cannot be opened unless valve SI-V93 or SI-V90 and
 
SI-V89 valves are closed; valve RC-V88 or RC-V87 is closed and valve CBS-V14 is open. 17. Motor-operated gate valve CS-V460 (CS-V461 analogous)
Fails to open on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing fluid flow through cross-tie between suction of HHSI/CH pumps and HHSI/SI pumps. No effect on safety of system operation. Alternate isolation valve (CS-V461 opens to provide backup flow path through cross- tie line. Same method of detection as that stated for item #4. 18. Motor-operated gate valve CBS-V47 (CBS-V51 analogous)
Fails to close on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing flow isolation of HHSI/SI pump suction from RWST. No effect on safety for system operation. Alternate check isolation valve (CBS-V48) provides backup isolation. Same method of detection as that stated for item #4. 19. Motor-operated gate valve LCV-112D (LCV-112E analogous)
Fails to close on demand.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing flow isolation of suction of HHSI/CH pumps from RWST. No effect on safety for system operation. Alternate check isolation valve (CBS-V58) provides backup isolation. Same method of detection as that stated previously for failure of item during injection phase of ECCS operation. The MCB RWST-VCT Suction Interlock NORMAL/BLOCK Selector Switch must be operated to BLOCK before the RWST valves can be closed. 20. Residual heat pump RH-P-8A (pump RH-P-8B analogous)
Fails to deliver working fluid.
Recirculation-cold legs of RC loops. Failure reduces redundancy of providing recirculation of coolant to the RCS from the Containment Sump. Fluid flow from LHSI/RHR pump RH-P-8A will be lost.
Minimum recirculation flow requirements for LHSI flow will be met by LHSI/RHR pum p RH-P-8B delivering fluid. Same method of detection as that stated previously for failure of item during injection phase of ECCS operation.
: 21. Safety injection pump SI-P-6A (pump SI-P-6B analogous)
Fails to deliver working fluid.
Recirculation-cold or hot legs of RC loops. Failure reduces redundancy of providing recirculation of coolant to the RCS from the Containment Sump to cold legs of RC loops via RHR and SI pumps. Fluid flow from HHSI/SI pump SI-P-6A will be lost. Minimum recirculation flow requirements for HHSI flow will be met by HHSI/SI pump SI-P-6B delivering working fluid. Same method of detection as that stated previously for failure of item during injection phase to ECCS operation.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-5 Revision:
Sheet: 15 5 of 6  Component Failure Mode ECCS Operation Phase *Effect on System Operation
**Failure Detection Method Remarks 22. Motor-operated gate valve RH-V14 Fails to close on demand. Recirculation-hot legs of RC loops. Failure reduces redundancy of providing recirculation of coolant to the RCS from the Containment Sump to hot legs of RC loops. Fluid flow from LHSI/RHR pump RH-P-8A
 
will continue to flow to cold legs for RC loops. Closure of backup isolation valve RH-V22 permits minimum recirculation flow requirement to hot legs of RC loops to be met by LHSI/RHR pump RH-P-8B recirculating fluid to RC hot legs directly and via HHSI/SI pumps. Same method of detection as that stated for item #4. 23. Motor-operated gate valve RH-V32 (RH-V70 analogous)
Fails to open on demand. Recirculation-hot legs of RC loops. Failure reduces redundancy of providing recirculation of coolant to the RCS from the containment sump to the hot legs of RC loops. No effect on safety for system operation. 
 
Alternate isolation valve (RH-V70) opens to provide flow path to RCS hot legs via LHSI/RHR pumps. Same method of detection as that stated for item #4. 24. Motor-operated gate valve RH-V26 Fails to close on demand. Recirculation-hot legs of RC loops. Failure reduces redundancy of providing recirculation of coolant to the RCS from the Containment Sump to hot legs of RC loops. Fluid flow from LHSI/RHR pump RH-P-8B
 
will continue to flow to cold legs of RC loops. Closure of backup isolation valve RH-V21 permits minimum recirculation flow requirements to hot legs of RC loops to be met by LHSI/RHR pump RH-P-8A recirculating fluid to RC hot legs directly and via HHSI/SI pumps. Same method of detection as that stated for item #4. 25. Motor-operated gate valve SI-V112 (SI-V111 analogous)
Fails to close on demand. Recirculation-hot legs of RC loops. Failure reduces redundancy of providing flow isolation of HHSI/SI pump flow to cold legs of RC loops. No effect on safety for system operation valve SI-V114 provides backup isolation against flow to cold legs of RC loops. Same method of detection as that stated for item #4. 26. Motor-operated gate valve SI-V102 (SI-V77 analogous)
Fails to open on demand. Recirculation-hot legs of RC loops. Failure reduces redundancy of providing recirculation of coolant to the hot legs of RCS from the Containment Sump via HHSI/SI pumps. Minimum recirculation flow requirements to hot legs of RC loops will be met by LHSI/RHR pump RH-P-8A and RH-P-8B recirculating fluid from Containment Sump to hot legs of RC loops and HHSI/SI pump SI-P-6B recircula ting fluid to hot legs 2 and 3 of RC loops through the opening of isolation valve SI-V77. Same method of detection as that stated for item #6. In addition, SI pump discharge pressure (PI-919) and flow (FI-918) at MCB.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-5 Revision:
Sheet: 15 6 of 6  Component Failure Mode ECCS Operation Phase *Effect on System Operation
**Failure Detection Method Remarks 27. Motor-operated gate valve SI-V114 Fails to close on demand. Recirculation-hot legs of RC loops. Failure reduces redundancy of providing flow isolation of HHSI/SI pump flow to cold legs of RC loops. No effect on safety for system operation. Alternate isolation valves SI-V112 and SI-V111 in cross-tie line between HHSI/SI pumps provides backup isolation against flow to cold legs of RC loops. Same method of detection as that stated for item #4. 28. Residual heat removal pump RH-P-8A  (Pump RH-P-8B analogous)
Fails to deliver working fluid. Recirculation-hot legs of RC loops. Failure reduces redundancy of providing recirculation of coolant to the RCS from the Containment sump to the hot legs of RC loops. Fluid flow from LHSI/RHR pump RH-P-8A will be lost. Minimum flow requirements to hot legs directly and via HHSI/SI pumps. Same method of detection as that stated previously for failure of item during injection phase of ECCS operation except flow indication is not available.
List of abbreviations and acronyms CBS Containment Spray CH, CS Charging HHSI High Head Safety Injection LHSI Low Head Safety Injection LOCA Loss of Coolant Accident MCB Main Control Board NPSH Net Positive Suction Head RC Reactor Coolant RCS Reactor Coolant System RHR, RH Residual Heat Removal RWST Refueling Water Storage Tank SI Safety Injection VCT Volume Control Tank
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3 -6 Revision:
Sheet: 8 1 of 1 TABLE 6.3-6 EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS LONG-TERM PHASE Flow Path Indication of Loss of Flow Path Alternate Flow Path Low Head Recirculation From containment sump to low head injection header via the residual heat removal pumps and
 
the residual heat exchangers  Accumulation of water in a residual heat removal pump compartment or auxiliary building sump Via the independent, identical low head flow path
 
utilizing the second residual heat exchanger and residual heat removal pump High Head Recirculation From containment sump to the high head injection header via residual heat removal pump, residual heat exchanger and the high head injection pumps Accumulation of water in a residual heat removal pump and safety injection pump compartment or the auxiliary building sump or charging pump compartments From containment sump to the high head injection headers via alternate residual heat removal pump, residual heat exchanger, safety injection or charging pump S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-7 Revision:
Sheet: 15 1 of 3  TABLE 6.3-7 MANUAL ACTIONS REQUIRED FOR ECCS OPERATION (BASED ON NO SINGLE FAILURE) During the injection mode of ECCS operation, the operators are not required to take any manual actions. All equipment operates automatically as designed. The minimum time calculated for the injection mode is approximately 26 minutes. During this time, the operators verify that all ECCS pumps are operating, and monitor the RWST and Containment Building recirculation sump levels in anticipation of the transfer to the recirculation mode of ECCS operation. Component cooling water flow to the residual heat removal heat exchangers is automatically initiated on a 'T' signal. The operator verifies that this has occurred before the transfer to recirculation begins. This table summarizes the manual operator actions required to complete the transfer from the injection mode of ECCS operation to the recirculation mode. This table assumes that no single failures have occurred before the transfer begins and also that none occur during the transfer.
The actions listed are also based on the design basis large break LOCA, all ECCS pumps operating, and any loss of off-site power occurring at the initiation of the safety injection signal.
This table is not intended to be a summary of the plant procedures that are used during this event. It is simply a summary of the manual actions required to support ECCS operation without any single failures being assumed to occur. This table assumes that all equipment operates as designed.
TRANSFER TO COLD LEG RECIRCULATION The RWST "Lo-Lo" level signal in conjunction with an 'S' single initiates the automatic opening of the containment sump isolation valves. Once operators are alerted to this, they perform the following manual actions:
: 1. Reset the 'S' signal.
: 2. Verify the containment sump isolation va lves (CBS-V8/V14) are open and close both RWST suction valves (CBS-V2/V5) to the RHR and CBS pumps.
: 3. Close the three safety injection pump miniflow valves (SI-V89 /V90 /V93)
: 4. Restore power to one of the two RHR to co ld leg injection isolation valves (CS-9787-1 for RH-V14 or CS-9787 for RH-V26).
: 5. Close the selected RHR to cold leg inj ection isolation valve (RH-V14 or RH-V26).
: 6. Open the two parallel safety injection pump to charging pump suction cross-connect valves (CS-V460 / V461).
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-7 Revision:
Sheet: 15 2 of 3  7. Open both RHR pump discharge valves to the safety injection pump and charging pump suctions (RH-V35 / V36). NOTE:  After the completion of the manual actions outlined above, all ECCS pumps are operating with their suctions being provided from the containment sump. The ECCS is now aligned for cold leg recirculation with both RHR pumps taking suction from the containment sump and delivering flow directly to two RCS cold legs and to the suctions for the safety injection and charging pumps. The safety injection and charging pumps are delivering this flow to the RCS cold legs.
The following actions provide additional isolation of the RWST from the recirculation fluid.
: 8. Close the two RWST suction valves to the safety injection pumps (CBS-V47 / V51).
: 9. Place the two RWST-VCT Suction Interlock Selector Switches to BLOCK.
: 10. Close the two RWST suction valves to the charging pumps (CS-LCV-112D and E) and de-energize.
: 11. Remove power from the RHR to cold leg isolation valves. At approximately 5 to 6 hours after the initiation of the accident, ECCS operation is shifted from the cold leg recirculation mode to the hot leg recirculation mode via the following manual actions:
: 1. Restore power to the two motor control centers required (CS-9787 or CS-9787-1).
: 2. Close the second RHR to cold leg injection isolation valve (RH-V14 or V26).
: 3. Open both RHR to hot leg injection isolation valves (RH-V32 / V70).
: 4. Stop No. 1 Safety injection pump.
: 5. Close No. 1 safety injection pump discharge cross connect valve (SI-V112).
: 6. Open No. 1 safety injection pump discharge hot leg recirculation isolation valve (SI-V102).
: 7. Start No. 1 safety injection pump.
: 8. Stop No. 2 safety injection pump.
: 9. Close No. 2 safety injection pump discharge cross connect valve (SI-V111).
: 10. Close the safety injection pumps common discharge isolation valve (SI-V114).
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-7 Revision:
Sheet: 15 3 of 3  11. Open No. 2 safety injection pump discharge hot leg recirculation isolation valve (SI-V77).
: 12. Start No. 2 safety injection pump.
: 13. De-engergize the two motor control centers The ECCS is now aligned for hot leg recirculation with both RHR pumps taking suction from the containment sump and delivering flow directly to two RCS hot legs and to the suctions of the safety injection and charging pumps. The safety injection pumps are now delivering flow to two RCS hot legs and the charging pumps are delivering flow to two RCS cold legs.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-8 Revision:
Sheet: 8 1 of 1 TABLE 6.3-8 EMERGENCY CORE COOLING SY STEM SHARED FUNCTIONS EVALUATION Component Normal Operating Arrangement Accident Arrangement Refueling water storage tank Lined up to suction of safety injection and residual heat  removal pumps Lined up to suction of centrifugal charging, safety injection and residual heat removal pumps Centrifugal charging pumps Lined up for charging service; suction from volume control tank, discharge via normal charging. Suction from refueling water storage tank, discharge lined up to cold leg injection. Valves for realignment meet single failure criteria. Residual heat removal pumps Lined up to cold legs of reactor coolant piping Lined up to cold legs or hot legs or
 
reactor coolant piping Residual heat exchangers Lined up to cold legs of reactor coolant piping Lined up to cold legs or hot legs of reactor coolant piping
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.3-9 Revision:
Sheet: 10 1 of 1 TABLE 6.3-9 NORMAL OPERATING STATUS OF EMERGENCY CORE COOLING SYSTEM COMPONENTS FOR CORE COOLING Number of safety injection pumps operable 2 Number of charging pumps operable 2 Number of residual heat removal pumps operable 2 Number of residual heat exchangers operable 2 Refueling water storage tank volume (gal) 477,000 (min.) Boron concentration in refueling water storage tank (ppm) 2,400-2,600 Boron concentration in accumulator (ppm) 2,300-2,600 Number of accumulators 4 Minimum accumulator pressure (psig) 585 Nominal accumulator water volume (ft
: 3) 850 System valves, interlocks, and piping required for the above components which are operable All SEABROOK STATION UFSAR ENGINEERED SAFETY FEATURES TABLE 6.4-1 Revision: Sheet:    1 8 1 of 3  TABLE 6.4-1 CONTROL ROOM COMPLEX SAFETY-RELATED VENTILATION SYSTEM MALFUNCTION ANALYSIS Component Malfunction Comments Normal Makeup Air Fan Fan trips Two 100 percent capacity fans are provided for normal makeup air and control room pressurization requirements. The redundant fan may be manually actuated to supply the makeup air. Each fan is powered from a separate emergency bus. Loss of air flow is alarmed. Fan status lights are provided on the main control board.
Normal Makeup Air Discharge Damper Damper fails to open, or fails to close If the damper fails to open, the redundant damper may be opened and its associated fan actuated to provide the makeup air. If both normal, makeup air trains are unavailable because of a vital bus outage, the emergency makeup air and filtration subsystem may be utilized.
If the damper fails to close, the crosstrain control scheme design ensures that the associated fan trips to ensure isolation.
Both normal makeup air dampers are provided with manual handwheel override actuators. Indicating lights on the main control board monitor all damper positions via limit switches on the damper linkage.
Emergency Makeup Air Fan Fan fails to actuate on high intake radiation,
'S' signal, or manual Actuation Two 100 percent capacity fans are provided with a fully redundant filter unit associated with each fan. Both fans are automatically actuated on high radiation or an 'S' signal. If one fan fails on manual actuation, the redundant fan may be manually actuated. Flow indication and alarms are provided for each filter/fan train. Each fan is powered from separate emergency buses. Fan status lights are provided on the main control board.
 
SEABROOK STATION UFSAR ENGINEERED SAFETY FEATURES TABLE 6.4-1 Revision: Sheet:    1 8 2 of 3  Component Malfunction Comments Emergency Makeup Air Discharge Damper Damper fails to open on high intake radiation, 'S' signal or manual Actuation Both dampers are automatically opened on high radiation or an 'S' signal. If one damper fails to open on manual actuation, the redundant damper and associated fan may be actuated.
Indicating lights on the main control board monitor all damper positions via limit switches on the damper linkage.
Water (35-40% ethylene glycol by volume) Chiller Water Chiller trips Two 100% capacity chillers, each supplied from a separate emergency bus, are provided. Secure the tripped train and place the redundant train in service.
Chilled Water Pump Chilled Water Pump trips Four 100% capacity pumps, two per train with each train supplied from a separate emergency bus, are provided. The diverse pump can be manually aligned and started to provide chilled water flow or the redundant train can be placed into service.
Air Handling Unit (including cooling coils)
Air Handling Unit trips Two 100% capacity air handling units, each supplied from a separate emergency bus, are provided. The redundant train may be started to supply conditioned air to the control room.
Air Handling Unit Discharge Damper Fails to open Each of the two 100% capacity units is provided with its own discharge damper. When the damper fails to open, the air handling unit and the corresponding chiller and the chilled water pump will all trip. The redundant train may be started.
Indicating lights on the air handling unit control panel monitor all damper positions via limit switches on the damper linkage.
Fails to close The damper may be manually positioned through the use of the handwheel on the actuator.
Indicating lights on the air handling unit control panel monitor all damper positions via limit switches on the damper linkage.
 
SEABROOK STATION UFSAR ENGINEERED SAFETY FEATURES TABLE 6.4-1 Revision: Sheet:    1 8 3 of 3  Component Malfunction Comments 3-Way Temperature Control Valve (TCV)
TCV fails Each train has its own TCV, and it is designed to fail
-safe to "full flow to the air handling unit cooling coils". The failed train may be secured and the redundant train may be started.
Chiller Condenser Exhaust Fan Exhaust fan fails Two 100% capacity exhaust fans, each supplied from a separate emergency bus, are provided. The redundant train may be started.
Exhaust Fan Control Damper Damper fails to close on actuation of emergency makeup air subsystem and/or a high radiation signal Two redundant isolation dampers are configured in series. Both dampers close upon actuation of the emergency makeup air fans or generation of a remote intake high radiation signal ensuring isolation of the exhaust subsystem.
Indicating lights on the main control board monitor all damper positions via limit switches on the damper actuator.
Cable Spreading Room Exhaust Fan Exhaust Fan Belt failure and continued motor operation Since the cable spreading room ventilation system is only placed in service using administrative controls, operator action to secure the associated cable spreading room supply fan can be performed promptly. Securing the supply fan will ensure that the Control Room Complex remains at the required differential pressure with respect to the cable spreading room.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-1 Revision:
Sheet: 8 1 of 3 TABLE 6.5-1 COMPLIANCE OF CONTAINM ENT ENCLOSURE AIR CLEANING UNITS TO REGULATORY GUIDE 1.52, REV. 2, MARCH 1978 Regulatory Guide Section Applicability to this System Comment Index Regulatory Guide Section Applicability to this System Comment Index C.1.a Yes Note 1 C.3.f Yes --- C.1.b Yes --- C.3.g Yes Note 11 C.1.c Yes --- C.3.h Yes --- C.1.d Yes --- C.3.i Yes Note 17 C.1.e Yes --- C.3.j Yes Note 12 C.2.a Yes Note 2 C.3.k Yes --- C.2.b Yes --- C.3.l Yes Note 13 C.2.c Yes --- C.3.m Yes --- C.2.d Yes Note 3 C.3.n No Note 5 C.2.e Yes --- C.3.o Yes --- C.2.f Yes --- C.3.p Yes --- C.2.g Yes Note 15 C.4.a Yes Note 14 C.2.h Yes Note 7 C.4.b Yes --- C.2.i Yes --- C.4.c Yes --- C.2.j Yes --- C.4.d Yes --- C.2.k No Note 4 C.4.e Yes --- C.2.l Yes --- C.5.a Yes Note 16 C.3.a Yes Note 8 C.5.b Yes Note 16 C.3.b No Note 9 C.5.c Yes Note 16 C.3.c No Note 2 C.5.d Yes Note 16 C.3.d Yes Note 6 C.6.a Yes Note 16 C.3.e Yes Note 10 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-1 Revision:
Sheet: 8 2 of 3 NOTES:  1. The design basis LOCA and ruptur e of a CRDM housing are the postulated design basis accidents. 2. Demisters will also serve as prefilters. No other prefilters are provided. Regulatory Guide 1.52, Revision 2, Section C.2, permits the use of demisters as prefilters for HEPA filters in an ESF System. The demister element consists of a 51/2" thick, 0.006" diameter, 340 stainless steel mesh, and fibrous glass fill. The "dry" efficiency of the demister is approximately 45 percent when tested in
 
accordance with NBS Dust Spot Test. This efficiency is similar to that of a prefilter. The demister will therefore serve a dual purpose, that of a demister and a prefilter; and there is no need to consider HEPA filter particulate loading
 
without a prefilter for this application. Flow instrumentation is provided in common ductwork downstream of containment enclosure emergency exhaust filter fans. One channel of volumetric flow indication is indicated and alarmed at the MCB as well as indicated locally.
Secondary flow indication is available using filter train differential pressure and
 
fan status indication. 3. No significant pressure surges to this system are foreseen; thus, no special protective devices are needed. 4. There are no outdoor air intakes that could affect the operation of the system. 5. The system is located in the containmen t enclosure, the area served. Therefore, any leakage will eventually be re-routed through the cleanup system before being expelled to the atmosphere. 6. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2.
HEPA filters are designed and qualif ied to MIL-F-51068, MIL-F-51069 and UL-586. There is no need to withstand iodine removal sprays. 7. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. Regulatory Guide 1.89 - IEEE Std. 323, Regulatory Guide 1.30 - IEEE Std. 336, Regulatory Guide 1.100 - IEEE Std. 344, Regulatory Guide 1.118 - IEEE Std.
338, Regulatory Guide 1.32 - IEEE 308. 8. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. Demisters are qualified to MSAR 71-45.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-1 Revision:
Sheet: 8 3 of 3 NOTES:  9. Heaters are not required for this unit. The ambient design conditions for the site are 88F dry bulb with a maximum relative humidity of 74.4 percent. Section 5.5 of ANSI 509-1976 states that "approximately 70%  RH" is required upstream of the moisture separator. Only 33 percent of the total air supplied to the containment enclosure area is outside air. The remainder is recirculated by the containment enclosure cooling units as explained in Section 9.4.6. The cooling units will maintain the space temperature at or below 153 F at the outside design conditions. Therefore, since no moisture is added to the Supply Air System, the relative humidity will not exceed 50 percent; which is less than the 70 percent RH required by ANSI N509. 10. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. Filter and adsorber mounting frames are constructed and designed in accordance
 
with Section 4.3 of ERDA 76-21. 11. Equipment was purchased prior to issuance or Regulatory Guide 1.52, Rev. 2.
Filter housings, floors and doors are constructed in accordance with ERDA 76-21. 12. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2.
Tray type adsorber cells are designed, cons tructed and tested in accordance with AACC CS-8T. 13. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. The system fan, its mounting, and the ductwork connections are designed, constructed and tested in accordance with the requirements of Regulatory Guide
 
1.52, Rev. 1 and ANSI N509-1980 and ANSI N510-1980. 14. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. Accessibility of components and maintenance are compatible with the intent of Regulatory Guide 1.52, Rev. 1. 15. Pertinent pressure drops (high delta-P across filter unit) are indicated, alarmed, and recorded at MCB. 16. In-place inspection and testing is performed in accordance with ANSI N510-11980. 17. Original charcoal was tested in acco rdance with ANSI N509-1976. All testing of replacement charcoal and future required periodic testing of charcoal will be in
 
accordance with ASTM D3803-1989.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5 -2 Revision:
Sheet: 8 1 of 3 TABLE 6.5 -2 COMPLIANCE OF FUEL ST ORAGE BUILDING AIR CLEANING UNITS TO REGULATORY GUIDE 1.52, REV. 2, MARCH 1978 Regulatory  Guide Section Applicability to this System Comment Index Regulatory  Guide Section Applicability to this System Comment Index C.1.a Yes Note 1 C.3.f Yes  - - - C.1.b Yes  - - - C.3.g Yes Note 11 C.1.c Yes  - - - C.3.h Yes  - - - C.1.d Yes  - - - C.3.i Yes Note 16 C.1.e Yes  - - - C.3.j Yes Note 12 C.2.a Yes Note 2 C.3.k Yes  - - - C.2.b Yes  - - - C.3.l Yes Note 13 C.2.c Yes  - - - C.3.m Yes  - - - C.2.d Yes Note 3 C.3.n No Note 5 C.2.e Yes  - - - C.3.o Yes  - - - C.2.f Yes  - - - C.3.p Yes  - - - C.2.g Yes  - - - C.4.a Yes Note 14 C.2.h Yes Note 7 C.4.b Yes  - - - C.2.i Yes  - - - C.4.c Yes  - - - C.2.j Yes  - - - C.4.d Yes  - - - C.2.k No Note 4 C.4.e Yes Note 15 C.2.1 Yes  - - - C.5.a Yes Note 15 C.3.a Yes Note 8 C.5.b Yes Note 15 C.3.b Yes Note 9 C.5.c Yes Note 15 C.3.c No Note 2 C.5.d Yes Note 15 C.3.d Yes Note 6 C.6.a Yes Note 15 C.3.e Yes Note 10 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5 -2 Revision:
Sheet: 8 2 of 3 NOTES:  1. A fuel handling accident is the postulated design basis accident. 2. Demisters will also serve as prefilters. No other prefilters are provided. Regulatory Guide 1.52, Revision 2, Section C.2, permits the use of demisters as prefilters for HEPA filters in an ESF System. The demister element consists of a 51/2" thick, 0.006" diameter, 340 stainless steel mesh, and fibrous glass fill. The "dry" efficiency of the demister is approximately 45 percent when tested in accordance with NBS Dust Spot Test. This efficiency is similar to that of a prefilter. The demister will therefore serve a dual purpose, that of a demister and a prefilter; and there is no need to consider HEPA filter particulate loading without a prefilter for this application. 3. No significant pressure surges to this system are foreseen; thus, no special protective devices are needed. 4. There are no outdoor air intakes that could affect the operation of the system. 5. The system is located in the Fuel Storage Building, the area served. Therefore, any leakage will eventually be re -routed through the cleanup system before being expelled to the atmosphere. 6. Equipment was purchased prior to i ssuance of Regulatory Guide 1.52, Rev. 2. HEPA filters are designed and qualified to MIL -F -51068, MIL -F -51069 and UL -586. There is no need to withstand iodine removal sprays. 7. Equipment was purchased prior to issu ance of Regulatory Guide 1.52, Rev. 2. Regulatory Guide 1.89  -IEEE Std. 323, Regulatory Guide 1.30  - IEEE Std. 336, Regulatory Guide 1.100  - IEEE Std. 344, Regulatory Guide 1.118  - IEEE Std. 338, Regulatory Guide 1.32  - IEEE 308. 8. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. Demisters are qualified to MSAR 71 -45. 9. Equipment was purchased prior to issuan ce of Regulatory Guide 1.52, Rev. 2. Heaters are constructed to industry standards and are seismically qualified. 10. Equipment was purchased prior to issu ance of Regulatory Guide 1.52, Rev. 2. Filter and adsorber mounting frames are constructed a nd designed in accordance with Section 4.3 of ERDA 76 -21. 11. Equipment was purchased prior to i ssuance or Regulatory Guide 1.52, Rev. 2. Filter housings, floors and doors are constructed in accordance with ERDA 76 -21. 12. Equipment was purchased prior to issuan ce of Regulatory Guide 1.52, Rev. 2. Tray type adsorber cells are designed, constructed and tested in accordance with AACC CS -8T.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5 -2 Revision:
Sheet: 8 3 of 3  13. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. The system fan, its mounting, and the ductwork connections are designed, constructed and tested in accordance with the requirements of Regulatory Guide 1.52, Rev. 1 and ANSI N509 and N510 -1980. 14. Equipment was purchased prior to issuance of Regulatory Guide 1.52, Rev. 2. Accessibility of components and maintenance are compatible with the intent of Regulatory Guide 1.52, Rev. 1. 15. In -place inspection and testing is performed in accordance with ANSI N510 -1980. 16. Original charcoal was tested in acco rdance with ANSI N509 -1976. All testing of replacement charcoal and future required pe riodic testing of charcoal will be in accordance with ASTM D3803 -1989.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-3 Revision:
Sheet: 13 1 of 4 TABLE 6.5-3 COMPLIANCE OF CONTRO L ROOM EMERGENCY FILTRATION SUBSYSTEM TO REGULATORY GUIDE 1.52, REV. 2. MARCH 1978 REGULATORY GUIDE SECTION APPLICABILITY TO THIS SYSTEM COMMENT INDEX C.1.a Yes ------- C.1.b Yes Note 1 C.1.c Yes ------- C.1.d Yes ------- C.1.e Yes ------- C.2.a Yes Note 2 C.2.b Yes ------- C.2.c Yes ------- C.2.d No Note 3 C.2.e Yes ------- C.2.f Yes ------- C.2.g Yes Note 4 C.2.h Yes ------- C.2.i Yes ------- C.2.j Yes ------- C.2.k Yes ------- C.2.l Yes Note 8 C.3.a No Note 2 C.3.b Yes Note 5 C.3.c Yes Note 5 C.3.d Yes Note 5 C.3.e Yes Note 9, 19 C.3.f Yes ------- C.3.g Yes Note 10 C.3.h Yes ------- C.3.i Yes Note 11, 19 C.3.j Yes Note 12, 19 C.3.k Yes Note 6 C.3.l Yes Note 17 C.3.m Yes ------- C.3.n Yes  C.3.o Yes -------- C.3.p Yes Note 18 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-3 Revision:
Sheet: 13 2 of 4 REGULATORY GUIDE SECTION APPLICABILITY TO THIS SYSTEM COMMENT INDEX C.4.a Yes Note 16 C.4.b Yes Note 13 C.4.c Yes Note 14 C.4.d Yes Note 7 C.4.e Yes ------- C.5.a Yes Notes 7, 15 C.5.b Yes Notes 7, 15 C.5.c Yes Notes 7, 15 C.5.d Yes Notes 7, 15 C.6.a Yes Notes 7, 15 NOTE 1: The control room emergency exhaust unit is located in environmental zone CB-3 at Elevation 75' where the total of 40-year integrated radiation level is less than or equal to 1.0x10 3 rads. The other system components upstream and downstream of the unit are also located in "mild" radiation environmental zones. NOTE 2: Demisters are not included in the filter design. The filter components for each redundant train include prefilters, an electric air heater, upstream HEPAs, an activated carbon adsorber bank, and downstream HEPAs. The piping which processes outside makeup air to the filters includes long vertical runs which will remove droplets entrained in the air stream. NOTE 3: No accident which will cause pressure surges is postulated for the area in which the filters are located. NOTE 4: The flow rate for each redundant filter train is indicated in the control room. High and low flow are alarmed in the control room. The pressure drop across the combined internal components for each filter train is indicated in the control room with high differential pressure generating an alarm. Pressure drops across each individual filter component are indicated locally in the control room HVAC equipment room. Additional details on system instrumentation are provided in
 
Subsection 6.5.1.5 NOTE 5: The prefilters, air heaters, and HEPA filters were designed, constructed, and tested per ANSI N509-1980.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-3 Revision:
Sheet: 13 3 of 4 NOTE 6: The filter system design includes a low-flow air bleed system. A 4-inch diameter pipe with a manual isolation valve is provided to cross-connect the two redundant filter units at a point downstream of the carbon adsorber sections. During single train operation, this alignment will provide approximately 50 cfm of air flow through the carbon adsorber section of th e inactive train. The configuration ensures that this low-flow cooling air is low humidity recirculation air. This satisfies the 70 percent maximum relative humidity criteria given a single active failure. NOTE 7: System operability verification and surveillance testing will be performed in accordance with plant Technical Specifications. NOTE 8: The atmosphere cleanup system housi ngs and ductwork have been designed to exhibit on test a maximum, total leakage rate as defined in Section 4.12 of ANSI N509-1980. Leak tests are performed in accordance with Section 6 of ANSI N510-1980. NOTE 9: The filter and adsorber mounting fram es for Train A, except for the prefilter mounting frame, were designed and constructed prior to the issuance of this Regulatory Guide and meet the intent of ANSI N509-1980. The filter and adsorber mounting frames for Train B along with the prefilter mounting frame for Train A were designed and constructed per ANSI N509-1980. NOTE 10: The filtration unit for Train A including floor/drains, et C., were designed and constructed prior to issuance of this Regulatory Guide and meet the intent of ANSI N509-1980. The Train B filtration un it was designed and constructed to ANSI N-509-1980. NOTE 11: Carbon has been qualified to ANSI N509-1980
*. The Train B adsorber cell is a four-inch deep bed with a minimum residence time of 0.25 seconds. The design iodine removal efficiency is consistent with that of a two-inch deep bed (95%). NOTE 12: The adsorber cells for Train A ha ve been designed and constructed per ANSI N509-1976 and meet the intent of ANSI N509-1980. The adsorber cells for
 
Train A have been tested per ANSI N510-1980. The adsorber cell for Train B has been designed, constructed, and tested per ANSI N509-1980. NOTE 13: Train B replaceable components are designed for removal from outside the filter unit.
* Information is historical and applies to the original charcoal installed. All testing of replacement charcoal and future required periodi c testing of charcoal will be in accordance with ASTM D3803-1989.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-3 Revision:
Sheet: 13 4 of 4 NOTE 14: Meets requirements of Section 4.11 of ANSI N509-1980. NOTE 15: All in-place testing/inspection is per ANSI N510-1980 requirements, with acceptance criteria of ANSI N509-1980 as applicable. NOTE 16: The system layout for Train A was performed prior to the issuance of this Regulatory Guide, however, it meets the requirements of Section 4.7 of ANSI N509-1980 and the intent of Subs ection 2.3.8 of ERDA 76-21. Train B meets the requirements of ANSI N509-1980 and ERDA 76-21. NOTE 17: The system fan motor, mountings and ductwork connections for Train A were procured prior to issuance of this Regulat ory Guide. However, they are designed and constructed to meet the intent of Sections 5.7 and 5.8 of ANSI N509-1976.
They are field tested/inspected per Section 8 of ANSI N510-1980, with the acceptance criteria of ANSI N509-1980. Trai n B has been designed, constructed, and tested per ANSI N509-1980. NOTE 18: The system dampers were procured prior to the assurance of this Regulatory Guide. However, they are designed, cons tructed, and tested per the intent of Section 5.9 of ANSI N509-1976 and ANSI N510-1975. They are field tested/inspected as a part of ductwork per ANSI N510-1980 requirements, with
 
the acceptance criteria of ANSI N509-1980. NOTE 19: Carbon adsorber cells ordered after May 27, 2009 have been designed and constructed in accordance with Regulatory Guide 1.52, Revision 3, June 2001.
Revision 3 endorses ASME AG-1-1997, "Code on Nuclear Air and Gas Treatment" and ASME N509-1989, "Nuclear Power Plant Air Cleaning Units and Components."  Revision 3 also notes that ESF Atmosphere Cleanup Systems Designed to ASME N509-1989 (or earlier versions) and tested to ASME N510-1989 (or earlier versions) are considered adequate to protect public health and safety. This exception to Regulat ory Guide 1.52, Revision 2, applies only to carbon adsorbent units.
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-4 Revision:
Sheet: 8 1 of 2  TABLE 6.5-4 CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANING SYSTEM MATERIALS Component Parameter 1) Moisture Separator (2)
Filtration efficiency at design air flow (2000 cfm) 99% entrained moisture  Water capacity at design air flow (2000 cfm) 3 lb/min entrained water  Case Material Type 304 Stainless Steel  Media Material Type 304 Stainless Steel & Fiberglass 2) HEPA Filter (4)
Efficiency at rated flow(2000 cfm), 20% and 120% rated flow 99.97% at 0.3 microns of DOP  Dust holding capacity 4 lbs. Case Material 409 Stainless Steel  Media Material Fiberglass  Separator Material None 3) Carbon Adsorber
: a. Lot Requirements (See Note 2)
Efficiency  Low Temperature 99%  Ambient Pressure Methyl Iodide at 95%
RH and 25 C    Low Temperature 99.9%  Ambient Pressure Elemental Iodine at 95%
RH and 30 C    High Temperature 99%  Ambient Pressure Methyl Iodide at 95%
RH and 80 C    Methyl Iodine in 98%  Containment at 95% RH
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-4 Revision:
Sheet: 8 2 of 2  Component Parameter  130C, 3.7 Atm., 1 Hour Load and 4 Hour Post-Sweep
: b. Batch Requirements (See Note 2)
Low Temperature 97%  Ambient Pressure Methyl Iodide at 95% RH and 30 C    High Temperature 99%  Ambient Pressure Methyl Iodide at 95% RH and 80 C    Except Pre and Post Sweep at 25 C    Elemental 99.9% Loading  Iodide Retention at 180 C 99.5% Retentivity  Media Activated Coconut Shell  Carbon    Impregnating Material KI 3  Ignition Temperature (ASTM D3466) 330 C  Density (ASTM D2854) 0.38g/cc (min)  Hardness (ASTM D3802) 97%  Mesh Size (ASTM D2862) 5% Maximum Retention on 8 90-100% thru 8 on 16 (8x12 Mesh 40-60%)  (12x16 Mesh 40-60%) 5% maximum thru 16; 1% Maximum thru 18  Depth of carbon bed 4 inches  Total weight of carbon 804 lbs  Carbon Bed Envelope Material Type 304 Stainless Steel 4) Filter Mounting Frames Type 304 Stainless Steel 5) Filter System Housing Epoxy Coated Carbon Steel
: 6) Ductwork Galvanized Steel
: 7) Fan Carbon Steel Note 1: Refer to Chapter 15 Appendix B for filter efficiencies assumed for design basis accidents. Note 2: Testing information is histori cal and applies to the original charcoal installed. All testing of replacement charcoal and future required peri odic testing of charcoal will be in accordance with ASTM D3803-1989.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-5 Revision:
Sheet: 8 1 of 3 TABLE 6.5-5 FUEL STORAGE BUILDING EMER GENCY AIR CLEANING SYSTEM MATERIALS Component Parameter 1) Moisture Separator (12)
Filtration efficiency at design air flow (17,000 cfm) 99% entrained moisture  Water capacity at design air flow (17,000 cfm) 24 lb/min entrained water  Case Material Type 304 Stainless Steel  Media Material Type 304 Stainless Steel & Fiberglass 2) Medium Efficiency Filter (12)
Filtration efficiency at design  air flow (17,000 cfm) 80%  Dust Holding Capacity 340 grams  Case Material 409 Stainless Steel  Media Material Fiberglass  Separator Material None 3) HEPA Filters (24)
Efficiency at rated flow(17,000 cfm), 20% and 120% 99.97% at 0.3 microns of DOP rated flow  Dust Holding Capacity 4 lbs  Case Material 409 Stainless Steel  Media Material Fiberglass  Separator Material None 4) Carbon Adsorber
: a. Lot Requirements (See Note 2)
Efficiency  Low Temperature 99%  Ambient Pressure
 
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-5 Revision:
Sheet: 8 2 of 3 Component Parameter  Methyl Iodide at 95% RH and 25 C    Low Temperature 99.9%  Ambient Pressure Elemental Iodine at 95%
RH and 30 C    High Temperature 99%  Ambient Pressure Methyl Iodine at 95%
RH and 80 C  b. Batch Requirements (See Note 2)
Low Temperature 97%  Ambient Pressure Methyl Iodide at 95% RH and 30 C    High Temperature 99%  Ambient Pressure Methyl Iodide at 95% RH and 80 C    Except Pre and Post Sweep at 25 C    Elemental 99.9% Loading  Iodine Retention at 180 C 99.5% Retentivity  Media Activated Coconut Shell  Carbon    Impregnating Material KI 3  Ignition Temperature (ASTM D3466) 330 C S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-5 Revision:
Sheet: 8 3 of 3 Component Parameter  Density (ASTM D2854) 0.38 g/cc (min)  Hardness (ASTM D3802) 97%  Mesh Size (ASTM D2862) 5% Maximum Retention on 8 90-100% thru 8 on 16 (8x12 Mesh 40-60%) (12x16 Mesh 40-60%) 5% Maximum thru 16; 1% Maximum thru
 
18  Depth of carbon bed 4 inches  Total weight of carbon 6500 lbs  Carbon bed envelope material Type 304 Stainless Steel 5) Filter Mounting Frames Type 304 Stainless Steel 6) Filter System Housing Epoxy Coated Carbon Steel
: 7) Ductwork Galvanized Steel
: 8) Fan Carbon Steel
 
Note 1: Refer to Chapter 15 Appendix B for filter efficiencies assumed for design basis accidents.
Note 2: Testing information is histori cal and applies to the orig inal charcoal installed. All testing of replacement charcoal and future required peri odic testing of charcoal will be in accordance with ASTM D3803-1989.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-6 Revision:
Sheet: 8 1 of 2 TABLE 6.5-6 CONTROL ROOM EMERGENCY MAKEUP AIR AND FILTRATION SUBSYSTEM MATERIALS TRAIN A FILTER Component Parameter 1) Medium Efficiency Filter (1)
Filtration Efficiency at Design Air Flow (1100 cfm) 80%  Dust Holding Capacity 340 grams  Case Material 409 Stainless Steel  Media Material Fiberglass  Separator Material None 2) HEPA Filters (4)
Filtration Efficiency at Design Air Flow (1100 cfm), 20% and 120% Rated Flow 99.97% at 0.3 Microns of DOP  Dust Holding Capacity 4 lbs. Case Material 409 Stainless Steel  Media Material Fiberglass  Separator None 3) Carbon Adsorber (7 trays)
: a. Lot Requirement (See Note 1)    Methyl Iodide, 80C, 95% RH (ASTM D3803) 1% Penetration, Maximum  Molecular Iodine, 30C, 95% RH (ASTM D3803) 0.1 Penetration, Maximum  b. Batch Requirement (See Note 1)
Molecular Iodine, 180C (ASTM D3803) 99.5 Retentivity, Minimum  Methyl Iodine, 30C, 95% RH (ASTM D3803) 3% Penetration, Maximum  Media Carbon Activated Coconut Shell  Impregnating Material Iodine Salts & Tertiary Amines  Ignition Temperature (ASTM D3466) 330C Minimum  Density (ASTM D2854) 0.38 g/cc Minimum  Hardness (ASTM D3802) 92 Minimum Note 1 Testing information is histor ical and applies to the original charcoal installed. All testing of replacement charcoal and future required peri odic testing of charcoal will be in accordance with ASTM D3803-1989.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-6 Revision:
Sheet: 8 2 of 2 TRAIN A FILTER Component Parameter  b. Batch Requirement (See Note 1)    Mesh Size (ASTM D2862) Retained on #6 Sieve:  0.1% Maximum Retained on #8 Sieve:  5.0% Maximum Through #8, on #12 Sieve:  60% Maximum Through #12, on #16 Sieve:
40% Minimum Through #16 Sieve:
5.0% Maximum Through #18 Sieve:
1.0% Maximum  Depth of Carbon Bed 2 Inches (tray type-6 trays,  1 test tray)  Total Weight of Carbon 300 lbs. (not including test  canisters)  Carbon Tray Envelope Material Type 304 Stainless Steel
: 4) Filter Mounting Frames Type 304 Stainless Steel
: 5) Filter System Housing Epoxy Coated Carbon Steel
: 6) Ductwork/Piping Galvanized Steel/Carbon Steel
: 7) Fan Carbon Steel Housing, Aluminum Blades
 
and Hub TRAIN B FILTER Same as Train A filters with the following exceptions:
: 1) HEPA Filters (2)
: 2) Carbon Adsorber Depth of Carbon Bed 4 Inches  Total Weight of Carbon 390 lbs. (not including test  canisters
: 3) Filter System Housing Type 304 Stainless Steel
: 4) Fan                                                          Carbon Steel Note 1 Testing information is histor ical and applies to the original charcoal installed. All testing of replacement charcoal and future required peri odic testing of charcoal will be in accordance with ASTM D3803-1989.
S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-7 Revision:
Sheet: 8 1 of 1  TABLE 6.5-7 VOLUMES OF CONTAINMENT ENCLOSURE AREAS AREA VOLUME  (ft 3) Containment Enclosure Annulus 524,344 Electrical Penetration Areas 84,035 Mechanical Penetration Areas 70,320 RHR and SI Equipment Vaults 102,816 Containment Enclosure Equipment Area 92,568 Charging Pump Areas 12,000 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.5-8 Revision:
Sheet: 8 1 of 1 TABLE 6.5-8 CONTAINMENT OPERATION FO LLOWING A DESIGN BASIS ACCIDENT General  - Type of Structure Concrete cylinder with hemispherical  dome and welded steel liner - Fission Product Removal System Water/NaOH building spray - Free Volume 2.715x10 6 ft 3 - Hydrogen Purge Mode Containment internal recombiner s with backup charcoal filtered purge to environment Time Dependent Parameters Anticipated Conservative
- Leak Rate of Containment  (percent of building volume per day) 0-24 hours 0.10 0.20 24 hours-30 days 0.05 0.10 - Containment Enclosure Emergency Exhaust Filter Bypass (percent of primary containment leakage) 0-24 hours 7.5 15.0 24 hours-30 days 3.75 7.5 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.8-1 Revision:
Sheet: 8 1 of 1 TABLE 6.8-1 EMERGENCY FEEDWATER PUMP DATA Total Number Per Unit 2 Electric Motor-Driven 1 Turbine-Driven 1 Design Flow (each) 710 gpm Design Basis Flow (each) 650 gpm Design Head 3050 ft. (1320 psi) Feedwater Design Temperature 50-100 o F Required BHP 770 Motor Size, HP 900 Turbine Rating, HP 900 S EABROOK S TATION UFSAR ENGINEERED S AFETY FEATURES TABLE 6.8-2 Revision:
Sheet: 8 1 of 1 TABLE 6.8-2 EMERGENCY FEEDWATER SYSTEM FAILURE ANALYSIS Active Failure Concurrent with a Feedwater System Pipe Break (One SG
* Faulted) COMPONENT ACTIVE FAILURE SYSTEM RESPONSE Emergency Feed Pump Pump fails to start (loss of power source) Second pump starts and provides required flow to the intact SGs. Flow sensing elements identify and isolate faulted SG. Flow Control Valve One of two valves in line to faulted SG fails to close No effect on system response. The second (redundant) valve in line closes to isolate the faulted
 
SG. Flow Control Valve Valve in line to intact SG closes on spurious signal Two pumps start; flow sensing elements identify and isolate faulted SG. Pumps provide the required flow to the two intact SGs. Check Valve in Pump Discharge Piping Check valve fails to open; flow from one pump blocked Second pump provides required flow to the three intact SGs. Flow sensing elements identify and isolate faulted SG.
* SG - steam generator SEABROOK STATION UPDATED FINAL SAFETY ANALYSIS REPORT C HAPTER 6 ENGINEERED SAFETY FEATURES  FIGURES  6.2-1 Containment Pressure Response Following DE Pump Suction Guillotine (Minimum Safety Injection) 6.2-2 Containment Temperature Response Following DE Pump Suction Guillotine (Minimum Safety Injection) 6.2-3 Sump Water Temperature Response Following DE Pump Suction Guillotine (Minimum Safety Injection) 6.2-4 Containment Pressure Response Following DE Pump Suction Guillotine (Maximum Safety Injection) 6.2-5 Containment Temperature Response Following DE Pump Suction Guillotine (Maximum Safety Injection) 6.2-6 Sump Water Temperature Response Following DE Pump Suction Guillotine (Maximum Safety Injection) 6.2-7 Containment Pressure Response Following 0.6 DE Pump Suction Guillotine (Minimum Safety Injection) 6.2-8 Containment Temperature Response Following 0.6 DE Pump Suction Guillotine (Minimum Safety Injection) 6.2-9 Sump Water Temperature Response Following 0.6 DE Pump Suction Guillotine (Minimum Safety Injection) 6.2-10 Containment Pressure Response Following 3 ft2 Pump Suction Split (Minimum Safety Injection) 6.2-11 Containment Temperature Response Following 3 ft2 Pump Suction Split (Minimum Safety Injection) 6.2-12 Sump Water Temperature Response Following 3 ft2 Pump Suction Split (Minimum Safety Injection) 6.2-13 Containment Pressure Response Following DE Cold
-Leg Guillotine (Minimum Safety Injection)
 
6.2-14 Containment Temperature Response Following DE Cold
-Leg Guillotine (Minimum Safety Injection) 6.2-15 Sump Water Temperature Response Following DE Cold
-Leg Guillotine (Minimum Safety Injection) 6.2-16 Containment Pressure Response Following DE Hot
-Leg Guillotine (Minimum Safety Injection) 6.2-17 Containment Temperature Response Following DE Hot
-Leg Guillotine (Minimum Safety Injection) 6.2-18 Sump Water Temperature Response Following DE Hot
-Leg Guillotine (Minimum Safety Injection) 6.2-19 Condensing Heat Transfer Coefficient Following DE Pump Suction Guillotine with Maximum Injection and Minimum Spray (DBA) 6.2-20 Containment Response Following Full DE Main Steam Line Break at Hot Shutdown with One Spray Train Failure 6.2-21 Containment Temperature Response Following Full DE Main Steam Line Break at Hot Shutdown with One Spray Train Failure 6.2-22 Containment Pressure Response Following Full DE Main Steam Line Break at 102% Power with Broken
-Loop MSIV Failed 6.2-23 Containment Temperature Response Following Full DE Main Steam Line Break at 102% Power with Broken
-Loop MSIV Failed 6.2-24 Effective Heat Transfer Coefficients for Passive Heat Sinks
- Full DE Main Steam Line Break at Hot Shutdown with One Spray Train Failed 6.2-25 Effective Heat Transfer Coefficients for Passive Heat Sinks for Full DE
 
Main Steam Line Break at 102% Power with Broken
-Loop MSIV Failed 6.2-26 Vertical Section through Reactor Cavity 6.2-27 Plan View of Reactor Cavity at Elevation of Nozzles 6.2-29 Section Drawing of Steam Generator Compartment Parallel to Hot Leg 6.2-30 Section Drawing of Steam Generator Compartment Parallel to Cold Leg 6.2-31 Plan Drawing of Steam Generator Compartment at Elevation (-)1'-7" 6.2-32 Plan Drawing of Steam Generator Compartment at Elevation 16'
-6" 6.2-33 Plan Drawing of Steam Generator Compartment at Elevation 22'
-0" 6.2-34 Plan Drawing of Steam Generator Compartment at Elevation 25'
-6" 6.2-35 Section Drawing of Pressurizer Compartment 6.2-36 Plan Drawing of Pressurizer Compartment from Elevation 16'
-6" to Elevation 0'
-0" 6.2-37 Plan Drawing of Pressurizer Compartment from Elevation 22'
-0" to Elevation 16'
-0" 6.2-38 Plan Drawing of Pressurizer Compartment from Elevation 25'
-0" to Elevation 22'
-0" 6.2-39 Plan Drawing of Pressurizer Compartment from Elevation 45'
-0" to Elevation 25'
-0" 6.2-40 Plan Drawing of Pressurizer Compartment at Elevation 57'
-0" 6.2-41 Plan and Elevation Drawings of Pressurizer Skirt Cavity 6.2-42 Elevation Drawing of Pressurizer Skirt Cavity 6.2-54 Pressurizer Compartment Nodalization (15 Node, 23 Node, 29 Node) 6.2-55 Horizontal Pressure Profiles around Pressurizer at Break Location 6.2-56 Pressurizer Vertical Pressure Profile Passing through Break Node 6.2-57 Transient Forces on Pressurizer (Y
-Force, X-Force) 6.2-58 Transient Moments on Pressurizer (X
-Moment Y-Moment) 6.2-59 5 Node Pressurizer Skirt Cavity Nodalization for Surge Line Rupture 6.2-60 6 Node Pressurizer Skirt Cavity Nodalization for Surge Line Rupture 6.2-69 Pressure Transient in Pressurizer Compartment Following Spray Line Rupture at Elevation 34'
-0" 6.2-70 Differential Pressure Transients in Pressurizer Compartment Following Spray Line Rupture at Elevation 34'
-0" 6.2-71 Pressurizer Skirt Cavity Pressure Response (6
-Node Model) 6.2-74 Containment Spray System 6.2-75 Engineered Safety Features Flow Diagram 6.2-76 Containment Structure Piping Zone 57F
- Plan at Elevation 119'
-0" 6.2-77 Containment Structure Piping Zones 57E and F Sections 6.2-78 Spray Pump
- NPSH 6.2-79 Containment Steel Recirculation Sump Screen Details 6.2-80 Loop A Spray Pattern
- Plan at Elevation 25'
-0" 6.2-81 Loop B Spray Pattern
- Plan at Elevation 25'
-0" 6.2-82 Spray Loop A Coverage Pattern 6.2-83 Spray Loop B Coverage Pattern 6.2-84 Particle Size vs. Number Percentage for #1713A Nozzle 6.2-85 Particle Size vs. Volume Percentage for #1713A Nozzle 6.2-86 Particle Size vs. Terminal Velocity 6.2-87 Energy Content of the Containment Atmosphere after Design Basis LOCA 6.2-88 Energy Content of the Containment Sump after Design Basis LOCA 6.2-89 Integrated Energy Absorbed by Structural Heat Sinks after Design Basis LOCA 6.2-90 Integrated Energy Removed from the Containment by Heat Exchangers after Design Basis LOCA 6.2-91 Isolation Valve Diagrams 6.2-92 Combustible Gas Control System 6.2-94 Electric Hydrogen Recombiner 6.2-95 Schematic Electric Recombiner System 6.2-96 Containment Hydrogen Concentration with and without Recombiner 6.2-97 Containment Hydrogen Concentration with and without Purging 6.3-1 Safety Injection System Overview 6.3-2 Safety Injection System Low Head Injection System Detail 6.3-3 Safety Injection System Low Head Injection Accumulators Detail 6.3-4 Safety Injection System Intermediate Head Injection System Detail 6.3-5 Safety Injection System High Head Injection System Detail 6.3-6 Residual Heat Removal System Process Flow Diagram 6.3-7 Residual Heat Removal Pump Performance Curve 6.3-8 Charging Pump Performance Curve 6.3-9 Safety Injection Pump Performance Curve 6.3-10 Refueling Water Storage Tank 6.4-1 East Makeup Air Intake Protection 6.4-2 West Makeup Air Intake 6.8-1 Emergency Feedwater System Overview 6.8-2 Emergency Feedwater System Details 6.8-3 Head and Capacity Curves for Emergency Feedwater Pumps and Startup Feedwater Pump
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Pressure Response Following DE Pump Suction Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-1
 
Note: The figure presents results rela ted to the long-term LO CA containment response analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Temperature Response Following DE Pump Suction Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-2
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Sump Water Temperature Response Following DE Pump Suction Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-3
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Pressure Response Following DE Pump Suction Guillotine (Maximum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-4
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt. Containment Temperature Response Following DE Pump Suction Guillotine (Maximum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-5
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Sump Water Temperature Response Following DE Pump Suction Guillotine (Maximum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-6
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Pressure Response Following 0.6 DE Pump Suction Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-7
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Temperature Response Following 0.6 DE Pump Suction Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-8
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Sump Water Temperature Response Following 0.6 DE Pump Suction Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-9
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Pressure Response Following 3 ft 2 Pump Suction Split (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-10
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment response results at analyzed core power level of 3659 MWt.
Containment Temperatur e Response Following 3 ft 2 Pump Suction Split (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-11
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Sump Water Temperature Response Following 3 ft 2 Pump Suction Split (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-12
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Pressure Response Following DE Cold-Leg Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-13
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Temperature Response Following DE Cold-Leg Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-14
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analysis core power level of 3659 MWt.
Sump Water Temperature Response Following DE Cold-Leg Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-15
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Containment Pressure Response Following DE Hot-Leg Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-16
 
Note: This figure presents results related to the th e long-term LOCA containmen t response analysis of record. Section 6.2.1.8 contains a discu ssion of the results of this analys is of record, as compared with the long-term LOCA containment response results at an analyzed core power level of 3659 MWt.
Containment Temperature Re sponse Following DE Hot-Leg Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-17
 
Note: This figure presents results related to the long-term LOCA containment res ponse analysis of record.
Section 6.2.1.8 contains a disc ussion of the results of this analysis of record, as compared with the long-term LOCA containment resp onse results at an analyzed core power level of 3659 MWt.
Sump Water Temperature Response Following DE Hot-Leg Guillotine (Minimum Safety Injection)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-18
 
Note: This figure presents results related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the results of this analysis of record, as compared with the long-term LOCA containment response results at an analyzed core power level of 3659 MWt.
Condensing Heat Transfer Coefficient Following DE Pump Suction Guillotine with Maximum Injection and Minimum Spray (DBA)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-19
 
Note: This figure presents results related to th e MSLB containment response analysis of record.
Section 6.2.1.8 contains a discussi on of the results of this anal ysis of record, as compared with the MSLB containment res ponse results at an analyzed core power level of 3659 MWt.
Containment Response Following Full DE Main Steam Line Break at Hot Shutdown with One Spray Train Failure S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-20
 
Note: This figure presents results re lated to the MSLB containment response analysis of record. Section 6.2.1.8 contains a discussion of th e results of this analysis of record, as co mpared with the MSLB c ontainment response results at an analyzed core power level of 3659 MWt.
Containment Temperature Response Following Full DE Main Steam Line Break at Hot Shutdown with One Spray Train Failure S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-21
 
Note: This figure presents results related to the MS LB containment response anal ysis of record. Section 6.2.1.8 contains a discussion of the re sults of this analysis of reco rd, as compared with the MSLB containment response results at an an alyzed core power level of 3659 MWt. Containment Pressure Respons e Following Full DE Main Steam Line Break at 102% Power with Broken-Loop MSIV Failed S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-22
 
Note: This figure presents results related to th e MSLB containment response analysis of record.
Section 6.2.1.8 contains a discussion of the results of this analysis of record, as compared with the MSLB containment response results at an analyzed core power level of 3659 MWt. Containment Temperature Response Following Full DE Main Steam Line Break at 102% Power with Broken-Loop MSIV Failed S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-23
 
Effective Heat Transfer Coeffici ents for Passive Heat Sinks - Full DE Main Steam Line Break at Hot Shutdown with One Spray Train Failed S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-24
 
Effective Heat Transfer Coeffi cients for Passive Heat Sinks for Full DE Main Steam Line Break at 102% Power with Broken-Loop MSIV Failed S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-25
 
Vertical Section through Reactor Cavity S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-26
 
Plan View of Reactor Cavity at Elevation of Nozzles S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-27
 
Section Drawing of Steam Generator Compartment Parallel to Hot Leg S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-29
 
Section Drawing of Steam Generator Compartment Parallel to Cold Leg S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-30
 
Plan Drawing of Steam Generator Compartment at Elevation (-)1'-7" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-31
 
Plan Drawing of Steam Generator Compartment at Elevation 16'-6" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-32
 
Plan Drawing of Steam Generator Compartment at Elevation 22'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-33
 
Plan Drawing of Steam Generator Compartment at Elevation 25'-6" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-34
 
Section Drawing of Pressurizer Compartment S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-35
 
Plan Drawing of Pressurizer Compartment from Elevation 16'-6" to Elevation 0'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-36
 
Plan Drawing of Pressurizer Compartment from Elevation 22'-0" to Elevation 16'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-37
 
Plan Drawing of Pressurizer Compartment from Elevation 25'-0" to Elevation 22'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-38
 
Plan Drawing of Pressurizer Compartment from Elevation 45'-0" to Elevation 25'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-39
 
Plan Drawing of Pressurizer Compartment at Elevation 57'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-40
 
Plan and Elevation Drawings of Pressurizer Skirt Cavity S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-41
 
Elevation Drawing of Pressurizer Skirt Cavity S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-42
 
Pressurizer Compartment Noda lization (15 Node, 23 Node, 29 Node) [3 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-54 Sh. 1 of 3
 
Pressurizer Compartment Noda lization (15 Node, 23 Node, 29 Node) [3 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-54 Sh. 2 of 3
 
Pressurizer Compartment Noda lization (15 Node, 23 Node, 29 Node) [3 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-54 Sh. 3 of 3
 
Horizontal Pressure Profiles around Pressurizer at Break Location S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-55
 
Pressurizer Vertical Pressure Profile Passing through Break Node S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-56
 
Transient Forces on Pressurizer (Y-Force, X-Force)
[2 Sheets]
S EABROOK S TATION  UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure    6.2-57 Sh. 1 of 2
 
Transient Forces on Pressurizer (Y-Force, X-Force)
[2 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure    6.2-57 Sh. 2 of 2
 
Transient Moments on Pressurizer (X-Moment, Y-Moment)
[2 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-58 Sh. 1 of 2
 
Transient Moments on Pressurizer (X-Moment, Y-Moment)
[2 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-58 Sh. 2 of 2
 
5 Node Pressurizer Skirt Cavity Nodalization for Surge Line Rupture S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-59
 
6 Node Pressurizer Skirt Cavity Nodalization for Surge Line Rupture S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-60
 
Pressure Transient in Pressurizer Compartment Following Spray Line Rupture at Elevation 34'-0" [3 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-69 Sh. 1 of 3
 
Pressure Transient in Pressurizer Compartment Following Spray Line Rupture at Elevation 34'-0" [3 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-69 Sh. 2 of 3
 
Pressure Transient in Pressurizer Compartment Following Spray Line Rupture at Elevation 34'-0" [3 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-69 Sh. 3 of 3
 
Differential Pressure Transients in Pressurizer Compartment Following Spray Line Rupture at Elevation 34'-0" [2 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-70 Sh. 1 of 2
 
Differential Pressure Transients in Pressurizer Compartment Following Spray Line Rupture at Elevation 34'-0" [2 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-70 Sh. 2 of 2
 
Pressurizer Skirt Cavity Pressure Response (6-Node Model)
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-71
 
Containment Spray System S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.2-74 See PID-1-CBS-B20233
 
Engineered Safety Features Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.2-75 See 1-NHY-804979
 
Containment Structure Piping Zone 57 F - Plan at  Elevation 119'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.2-76 See 805146
 
Containment Structure Pipi ng Zones 57E and F Sections S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.2-77 See 805147 Spray Pump - NPSH S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 12 Figure          6.2-78 Deleted Containment Recirculation Sump and Strainer S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 12 Figure 6.2-79
 
Loop A Spray Pattern - Plan at Elevation 25'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-80
 
Loop B Spray Pattern - Plan at Elevation 25'-0" S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-81
 
Spray Loop A Coverage Pattern S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-82
 
Spray Loop B Coverage Pattern S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-83
 
Particle Size vs. Number Percentage for #1713A Nozzle S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-84
 
Particle Size vs. Volume Percentage for #1713A Nozzle S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-85
 
Particle Size vs. Terminal Velocity S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-86
 
Note: This figure presents energy data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contai ns a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containmen t response at an analyzed co re power level of 3659 MWt.
Energy Content of the Containment Atmosphere after Design Basis LOCA S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-87
 
Note: This figure presents energy data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a discussion of the mass and energy release data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment res ponse at an analyzed core power level of 3659 MWt.
Energy Content of the Containment Sump after Design Basis LOCA S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-88
 
Note: This figure presents energy data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a di scussion of the mass and energy rel ease data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
Integrated Energy Absorbed by Structural Heat Sinks after Design Basis LOCA S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-89
 
Note: This figure presents energy data related to the long-term LOCA containment response analysis of record. Section 6.2.1.8 contains a di scussion of the mass and energy rel ease data associated with this analysis of record, as compared with the mass and energy release data associated with the long-term LOCA containment response at an analyzed core power level of 3659 MWt.
Integrated Energy Removed from the Containment by Heat Exchangers after Design Basis LOCA S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-90
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 1 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 2 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 3 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 4 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 5 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 6 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 7 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 8 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 9 of 12 IRCORC(MS  &  FW PIPE CHASE)* MSD-V44  &  45 PENS X-1 & X-2 ARE  NORMALLY OPEN; MSD-V46  & V47    PENS X-3  & X-4 ARE NORMALLY CLOSED.MAIN STEAMMAIN FEED WATERSTEAM GENERATOR BLOWDOWNSTEAMGENERATORCONTAINMENT PRESSURE  &  COOLANT PRESSURE  /  LEVEL INSTRUMENTATION LINESEALED, FLUID FILLED SYSTEMHYDRAULICISOLATORORC  (CONTAINMENT ENCLOSURE)IRC PTLT/PTRCSX-50X-77CX-77DX78DX-57BX-78CX-77AX-77BX-78AX-78B 2 2 2 TC DR FC NG MX-5X-6X-7X-8X-63X-64X-65X-66 23 FC TC LOBLOWDOWNTANKEMERGENCY FEEDWATERMAIN FEEDWATERFAIS, RXT, SGEMERGENCY FEED PUMPTURBINE LOOPS 1  2 ONLY(5)(5)FT FT FC NG PI G TC DRMSISMAINSTEAMLOOPS I-I/
G 2 3 LO LO 2MSIS PTMSIS  LC**
2 2X-1X-2X-3X-4 2 M*M LC FOFA1 T**  During startup and surveillances, the MSIV bypass valves may be open with breakers closed, which allows MSIS and control switch operation.006_Figure 02-91 Sh9
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 10 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 11 of 12
 
Isolation Valve Diagrams [12 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 13 Figure          6.2-91 Sh. 12 of 12
 
Combustible Gas Control System S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.2-92 See PID-1-CGC-B20612
 
Electric Hydrogen Recombiner S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-94
 
Schematic Electric Recombiner System S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-95
 
Containment Hydrogen Concentration With and Without Recombiner 0 1 2 3 4 5 6 7050100150200250300350400450500550600650700750800Time (Hr)Volume Percent HydrogenRecombiner At 3.5V/%Without HCS Control Containment Hydrogen Concen tration with and without Recombiner S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-96
 
Containment Hydrogen Concentration With and Without Purge 0 1 2 3 4 5 6 7 8 90100200300400500600700800900100011001200130014001500Time (hr)Volume Percent HydrogenWithout Purge0.03 Vol/day Purge begin at 3.8%
Containment Hydrogen Concen tration with and without Purging S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.2-97
 
Safety Injection System Overview S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.3-1 See PID-1-SI-B-20445
 
Safety Injection System Low Head Injection System Detail
[2 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.3-2 Sh. 1 of 2 See PID-1-SI-B-20448 Safety Injection System Low Head Injection System Detail
[2 Sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.3-2 Sh. 2 of 2 See PID-1-SI-B-20449
 
Safety Injection System Low Head Injection Accumulators Detail S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.3-3 See PID-1-SI-B20450
 
Safety Injection System Intermediate Head Injection System Detail S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.3-4 See PID-1-SI-B20446
 
Safety Injection System High Head Injection System Detail S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.3-5 See PID-1-SI-B20447
 
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 1 of 14
 
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 2 of 14
 
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 3 of 14 MODES OF OPERATOR MODE A - INJECTION This mode presents the process conditions for the case of maximum safeguards, i.e., all pumps operating, following accumulator delivery. Two residual heat removal (RHR) pumps, two safety injection (SI) pumps, and two cen trifugal charging (CC) pumps operate, taking suction from the refueling water storage tank and de livering to the reactor through th e cold leg connections. Note that the flow from each pump is less than its maximum runout since the pump discharge piping is shared by the two pumps of each subsystem. Note also that the SI pump branch connections to the residual lines are close to their discharge into the accumulator lines, thereby minimizing any increase in the RHR branch line head loss due to the combined flows of the RHR and SI pumps. MODE B - COLD-LEG RECIRCULATION This mode presents the process conditions for the case of cold-leg recirculation assuming residual heat removal (RHR) pump No. 2 operati ng, safety injection pumps 1 and 2 operating, and centrifugal charging (CC) pumps 1 and 2 operating. It is assumed that the spray pumps have emptied the RWST at this time. In this mode the safeguards pumps operate in se ries, with only the RHR pump capable of taking suction from the containment sump. The recirculated coolant is then delivered by the RHR pump to both of the SI pumps which deliver to the reactor through their co ld-leg connections and to both of the CC pumps which deliver to the r eactor through thei r cold-leg connections. The RHR pump also delivers fl ow directly to the reac tor through two cold legs. MODE C - HOT-LEG RECIRCULATION This mode presents the process conditions for the case of hot-leg recirculation, assuming residual heat removal (RHR) pump No. 1 operating, centrifugal charging (CC) pumps 1 and 2 operating, and safety injection (SI) pumps 1 and 2 operating. In this mode, the safeguards pumps again operate in series with only the RHR pump taking suction from the containment sump. The recirculated coolant is then delivered by the RHR pump to both of the CC pumps which continue to deliver to the reactor through thei r cold-leg connections and to both of the SI pumps which deliver to the reactor through their hot-leg connections. The RHR pump also delivers directly to the reac tor through two hot-leg connections.
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 4 of 14 Valve Alignment Chart Operational Modes Valve Injection (A)
Cold Leg Recirculation (B)
Hot Leg Recirculation (C) 1 O C C 2 O C C 3 O C C 4 O C C 5 O C C 6 O O C 7 O O C 8 C C O 9 C C O 10 C C C 11 C C C 12 C O O 13 C O O 14 C C C 15 C C C 16 C C C 17 C C C 18 O O O 19 O O O 20 C O O 21 C O O 22 O O O 23 O O O 24 O O C 25 C C O 26 O C C 27 C C C O = Open C = Closed Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 5 of 14 Valve Alignment Chart Operational Modes Valve Injection (A)
Cold Leg Recirculation (B)
Hot Leg Recirculation (C) 28 O C C 29 C O O 30 C C C 31 C C C 35 O O O Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 6 of 14 Mode A - Injection Phase (Runout Conditions Following Accumulator Delivery)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 1 Refueling Water ATM Tank 100 - 450,000 2 Refueling Water
* 100 13,687 - 3 Refueling Water 13 Psia 100 12,839 -
4 Refueling Water - 100 6,839 - 5 Refueling Water - 100 6,000
** - 6 Refueling Water 11 Psia 100 848 - 7 Refueling Water - 100 6,000 - 8 Refueling Water >10 Psia 100 839 - 9 Refueling Water >10 Psia 100 424 - 10 Refueling Water 10 Psia 100 424 - 11 Refueling Water 1165 100 424 -
12 Refueling Water <25 100 26 - 13 Refueling Water 10 Psia 100 424 -
14 Refueling Water 1165 100 424 - 15 Refueling Water 25 100 26 -
16 Refueling Water - 100 52 - 17 Refueling Water 1050 100 796 - 18 Refueling Water 73 100 199 - 19 Refueling Water - 100 1,699 - 20 Refueling Water - 100 1,699 -
21 Nitrogen 0 100 0 - 22 Nitrogen 0 100 0 850 23 Nitrogen 0 100 0 500 24 Reactor - 120 0 - 25 Refueling Water 0 100 3,000 - 26 Refueling Water 138 100 3,000 -
27 Refueling Water - 100 3,000 -
* At reference conditions 100 &#xba;F and 0 psig.
**  Estimated spray pump flow. Actual runout flow is 3300 gpm per spray pump.
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 7 of 14 Mode A - Injection Phase (Runout Conditions Following Accumulator Delivery)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 28 Refueling Water 47 100 3,000 - 29 Refueling Water 86 100 0 - 30 Refueling Water - 100 0 -
31 Refueling Water - 100 3,000 - 32 Refueling Water 86 100 3,000 - 33 Refueling Water 86 100 0 -
34 Reactor - 120 0 - 35 Refueling Water 0 100 3,000 - 36 Refueling Water 138 100 3,000 -
37 Refueling Water - 100 3,000 - 38 Refueling Water 47 100 3,000 - 39 Refueling Water ~86 100 0 -
40 Refueling Water - 100 0 - 41 Refueling Water - 100 3,000 - 42 Refueling Water 86 100 3,000 -
43 Recirc Coolant Containment Pressure 120 0 - 44 Recirc Coolant Containment Pressure 0 -  45 Recirc Coolant Containment Pressure 120 0 - 46 Refueling Water Low 100 0 -
47 Refueling Water Low 100 0 -
48 Refueling Water Low 100 0 - 49 Refueling Water Low 100 0 - 50 Refueling Water Low 100 0 -
51 Refueling Water Low 100 0 - 52 Refueling Water Low 100 0 - 53 Refueling Water >10 Psia 100 839 -
54 Refueling Water - 100 0 -
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 8 of 14 Mode A - Injection Phase (Runout Conditions Following Accumulator Delivery)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 55 Refueling Water 1519 100 419 - 56 Refueling Water - 100 0 - 57 Refueling Water 10 Psia 100 419 -
58 Refueling Water 10 Psia 100 419 - 59 Refueling Water 1519 100 419 - 60 Refueling Water 1516 100 124 -
61 Refueling Water 1516 100 124 - 62 Refueling Water 1456 100 714 - 64 Refueling Water 1396 100 714 -
65 Refueling Water 1008 100 178.5 - 66 Refueling Water 388 100 178.5 -
 
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 9 of 14 Mode B - Cold-Leg Recirculation (RHR Pump No. 2 Operating)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 1 Refueling Water ATM Tank 100 - - 2 Refueling Water - 100 0 - 3 Refueling Water - 100 0 -
4 Refueling Water - 100 0 - 5 Refueling Water - 100 0 - 6 Recirc Coolant - 182 0 -
7 Refueling Water - 100 0 - 8 Refueling Water - 100 0 - 9 Recirc Coolant ~35 182 1,263 - 10 Recirc Coolant ~35 182 424 - 11 Recirc Coolant ~1165 182 424 - 12 Refueling Water - 100 0 -
13 Recirc Coolant ~35 182 424 - 14 Recirc Coolant ~1165 182 424 - 15 Refueling Water - 100 0 -
16 Refueling Water - 100 0 - 17 Recirc Coolant 1050 182 848 - 18 Recirc Coolant 73 182 212 - 19 Recirc Coolant - 182 1,618 - 20 Recirc Coolant - 182 212 -
21 Nitrogen 0 Ambient 0 - 22 Nitrogen 0 Ambient 0 850
** 23 Nitrogen 0 Ambient 0 500 24 Recirc Coolant - 212 0 - 25 Recirc Coolant ~12 212 4,500 - 26 Recirc Coolant 113 212 4,500 -
27 Recirc Coolant - 212 4,500 -
* At reference conditions 212 &#xba;F and 0 psig.
** Minimum Water/Volume At Operating Conditions.
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 10 of 14 Mode B - Cold-Leg Recirculation (RHR Pump No. 2 Operating)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 28 Recirc Coolant 29 182 2,813 - 29 Recirc Coolant 56 182 0 - 30 Recirc Coolant 60 182 1,687 -
31 Recirc Coolant 65 182 4,500 - 32 Recirc Coolant 55 182 2,813 - 33 Recirc Coolant 0 182 0 -
34 Recirc Coolant - 212 0 - 35 Refueling Water - 100 0 - 36 Refueling Water - 100 0 -
37 Refueling Water - 100 0 - 38 Refueling Water - 100 0 - 39 Refueling Water - 100 0 -
40 Refueling Water - 100 0 - 41 Refueling Water - 100 0 - 42 Refueling Water - 100 0 -
43 Recirc Coolant Containment Pressure 212 0 450,000 44 Recirc Coolant Containment Pressure 212 4,500 - 45 Recirc Coolant Containment Pressure 212 0 - 46 Refueling Water Low 100 0 -
47 Refueling Water Low 100 0 -
48 Refueling Water Low 100 0 - 49 Refueling Water Low 100 0 - 50 Refueling Water Low 100 0 -
51 Refueling Water Low 100 0 - 52 Recirc Coolant - 182 0 - 53 Recirc Coolant - 182 0 -
54 Recirc Coolant - 182 0 -
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 11 of 14 Mode B - Cold-Leg Recirculation (RHR Pump No. 2 Operating)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 55 Recirc Coolant ~1519 182 ~419 - 56 Recirc Coolant >30 182 838 - 57 Recirc Coolant ~30 182 419 -
58 Recirc Coolant ~30 182 419 - 59 Recirc Coolant ~1519 182 ~419 - 60 Recirc Coolant 1516 182 124 -
61 Recirc Coolant 1516 182 124 - 62 Recirc Coolant 1456 182 714 - 64 Recirc Coolant 1396 182 714 -
65 Recirc Coolant 1008 182 178.5 - 66 Recirc Coolant 388 182 178.5 -
 
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 12 of 14 Mode C - Hot-Leg Recirculation (RHR Pump No. 1 Operating)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 1 Refueling Water ATM Tank 100 - - 2 Refueling Water - 100 0 - 3 Refueling Water - 100 0 -
4 Refueling Water - 100 0 - 5 Refueling Water - 100 0 - 6 Recirc Coolant - 182 0 -
7 Refueling Water - 100 0 - 8 Refueling Water - 100 0 - 9 Recirc Coolant ~25 182 650 - 10 Recirc Coolant ~25 182 650 - 11 Recirc Coolant ~715 182 650 - 12 Refueling Water - 100 0 -
13 Recirc Coolant ~25 182 650 - 14 Recirc Coolant ~715 182 650 - 15 Refueling Water - 1000 -
16 Refueling Water - 100 0 - 17 Recirc Coolant 0 182 0 - 18 Recirc Coolant - 182 0 - 19 Recirc Coolant - 182 0 - 20 Recirc Coolant - 182 0 -
21 Nitrogen - Ambient 0 - 22 Nitrogen 0 Ambient 0 850
** 23 Nitrogen 0 Ambient 0 500 24 Recirc Coolant - 212 0 - 25 Recirc Coolant - <212 0 - 26 Recirc Coolant - <212 0 -
27 Recirc Coolant - <212 0 -
* At reference conditions 212 &#xba;F and 0 psig.
** Minimum water volume at operating conditions.
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 13 of 14 Mode C - Hot-Leg Recirculation (RHR Pump No. 1 Operating)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 28 Recirc Coolant - <182 0 - 29 Recirc Coolant - <182 0 - 30 Recirc Coolant - <182 0 -
31 Recirc Coolant - <182 0 - 32 Recirc Coolant - <182 0 - 33 Recirc Coolant 50 182 2,362 -
34 Recirc Coolant - 212 0 - 35 Recirc Coolant ~12 212 4,500 - 36 Recirc Coolant 113 212 4,500 -
37 Recirc Coolant - 212 4,500 - 38 Recirc Coolant - <182 0 - 39 Recirc Coolant 55 182 0 -
40 Recirc Coolant 60 182 2,138 - 41 Recirc Coolant 65 182 4,500 - 42 Recirc Coolant 55 182 2,362 -
43 Recirc Coolant Containment Pressure 212 - - 44 Recirc Coolant Containment Pressure 212 0 - 45 Recirc Coolant Containment Pressure 212 4,500 - 46 Recirc Coolant 7 182 2,362 -
47 Recirc Coolant 5 182 1,181 -
48 Recirc Coolant 645 182 650 - 49 Recirc Coolant - 182 1,506 - 50 Recirc Coolant 645 182 650 -
51 Recirc Coolant - 182 325 - 52 Recirc Coolant - 182 0 - 53 Recirc Coolant - 182 0 -
54 Recirc Coolant - 182 2,138 -
Residual Heat Removal System Process Flow Diagram S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT Rev. 14 Figure 6.3-6 Sh. 14 of 14 Mode C - Hot-Leg Recirculation (RHR Pump No. 1 Operating)
Location Fluid Pressure (psig) Temperature(&#xba;F) Flow (gpm)* Volume (gal) 55 Recirc Coolant ~1519 182 419 - 56 Recirc Coolant <35 182 1,300 - 57 Recirc Coolant ~35 182 419 -
58 Recirc Coolant 35 182 419 - 59 Recirc Coolant 1519 182 419 - 60 Recirc Coolant 1516 182 124 -
61 Recirc Coolant 1516 182 124 - 62 Recirc Coolant 1456 182 714 - 64 Recirc Coolant 1396 182 714 -
65 Recirc Coolant 1008 182 178.5 - 66 Recirc Coolant 338 182 178.5 -
 
Residual Heat Removal Pump Performance Curve S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.3-7
 
Charging Pump Performance Curve S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.3-8
 
Safety Injection Pump Performance Curve S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.3-9
 
Refueling Water Storage Tank S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.3-10
 
East Makeup Air Intake Protection S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.4-1
 
West Makeup Air Intake [2 sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.4-2 sh 1 of 2
 
West Makeup Air Intake [2 sheets]
S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.4-2 sh 2 of 2
 
Emergency Feedwater System Overview S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.8-1 See PID-FW-B20685
 
Emergency Feedwater System Details S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure 6.8-2 See PID-FW-B20688
 
Head and Capacity Curves for Emergency Feedwater Pumps and Startup Feedwater Pump S EABROOK S TATION UPDATED F INAL SAFETY ANALYSIS R EPORT  Figure          6.8-3}}

Latest revision as of 00:38, 8 January 2025

8 to Updated Final Safety Analysis Report, Chapter 6, Engineered Safety Features
ML17310A412
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 10/27/2017
From:
NextEra Energy Seabrook
To:
Office of Nuclear Reactor Regulation
Shared Package
ML17310A361 List:
References
SBK-L-17157
Download: ML17310A412 (504)
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