LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 6.2, Containment Systems

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Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 6.2, Containment Systems
ML17046A401
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6.2 CONTAINMENT SYSTEMS 6.2.1 Containment Functional Design 6.2.1.1 Design Basis The reactor containment completely encloses the entire Reactor Coolant System (RCS) and ensures that post-accident leakage is limited to a safe rate of 0.1 percent of the containment free volume per day at the design pressure of 47 psig. A steel liner and leak-tight penetrations are provided to ensure that the leakage limits are not exceeded. The structure provides biological shielding for both normal and accident situations. The reactor containment is designed to safely withstand the loading combinations described in Section 3.8. Containment and associated systems are designed, fabricated, and erected to quality and performance standards with appropriate testing and inspection requirements. Records of design, fabrication, construction, and testing of the containment are maintained throughout the life of the plant. The RCS is designed to maintain its capability in case of fire to safely shut down and isolate the reactor. The design pressure and temperature of the containment is equal to or greater than the peak pressure and temperature occurring as the result of the complete blowdown of the reactor coolant through any rupture of the RCS up to and including the complete severance of a reactor coolant pipe. Energy contribution from the steam generators is included in the calculation of the containment pressure transient due to the reverse heat transfer through the steam generator tubes. The RCS supports are designed to withstand the blowdown forces associated with the sudden severance of the reactor coolant piping but not steam piping since the coincidental rupture of the steam system is not credible. In addition, 6.2-1 SGS-UFSAR Revision 6 February 15, 1987 containment design pressure is not exceeded during any subsequent long-term pressure transient determined by the combined effects of heat sources such as residual heat and limited metal-water reactions, structural heat sinks and the operation of the engineered only the emergency onsite electric power supply. In a design basis accident (DBA), reactor coolant is released through a double-ended break of the largest reactor coolant pipe, causing a rapid pressure rise in the containment. The reactor coolant pipe used in the accident is the 29-inch inside diameter section because rupture of the 31-inch inside diameter section requires that the blowdown go through both the 29-inch and the 27 1/2 inch inside diameter pipes and would, result in a less severe transient. Additional energy release was considered from the following sources: 1. Stored heat in the reactor core. 2. Stored heat in the reactor vessel piping and other RCS components. 3. Residual heat production. 4. Limited metal-water reaction energy and reaction energy. hydrogen-oxygen Details of mass and energy releases are provided in Section 15.4.8. The containment is also designed to withstand credible external pressures. In the event of inadvertent spray actuation, the containment would depressurize until the temperature of the atmosphere was approximately the of the spray. A bounding calculation was to determine the maximum 6.2-2 SGS-OFSAR Revision 24 May 11, 2009 outside to inside pressure differential. The following initial conditions were assumed: 1. The containment is initially at 120°F which maximizes the temperature differential between the containment atmosphere and the spray, which is at a temperature of 40°F. 2. The containment pressure is 14.7 psia. 3. The relative humidity is at a maximum value of 100 percent. As the air temperature is reduced from 120 to 40°F, the partial pressure of the air decreases from 12.91 to 11.13 psi. The steam partial pressure decreases from 1.6927 to 0.12163 psi. Thus, a containment equilibrium pressure of 11.25 psia is produced. This causes a differential pressure (d/p) of 3.45 psi across the containment shell, with no credit taken for the operation of the containment Pressure-Vacuum Relief System. In the long-term, the Pressure-vacuum Relief System will be operated to return the containment pressure to normal. The d/p between the design and maximum calculated negative pressure is 0.05 psi. This margin is adequate due to the conservatism used in the external pressure analysis. The containment design provides limited access through personnel hatches with the reactor at power. This type of access is intended primarily for inspection and maintenance of the air recirculation equipment, incore ion chamber drives, seal table, operating deck, and reactor coolant drain tank. Opening of the containment equipment hatch or both doors in the personnel locks is limited by the Technical Specifications. After shutdown, the containment is purged to reduce the concentration of radioactive gases and airborne particulates. A 6.2-3 SGS-UFSAR Revision 6 February 15, 1987 purge system is provided to reduce the radioactivity level to an exposure of lees than 40 Derived Air Concentration-Hours (DAC-Hours), as defined by 10 CFR 20, in a 40-hour occupational work week, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after plant shutdown, based on !-percent fuel defects. To assure removal of particulate matter, the purge air is passed through a high-efficiency filter before being released to the atmosphere through the plant vent. The primary reactor shield is designed so that access to the primary equipment is limited by the activity of the primary system equipment and not the reactor. 6.2.1.2 Containment structural Acceptance Test 6.2.1.2.1 General Description The completed containment structure was tested for structural integrity by subjecting the structure to an air pressure test of 54 psig, which is equivalent to 115 percent of the design pressure. The basic requirements of Regulatory Guide 1.18, "Structural Acceptance Test for concrete Primary Reactor Containments," were satisfied in the performance of the test. Containment pressurization was accomplished in incremental steps to 12 psig, 24 psig, 36 psig, 47 psig, and a final test pressure of 54 psig. Except for the final pressure level, the containment pressure was increased to 1 paig above the level at which measurement readings were to be taken. The pressure was then reduced to the specified value and, after a minimum time delay of 10 minutes to permit equalization of strains in the structure, the observations and measurements were made. The final test pressure of 54 psig on the building was maintained for a period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. During this time, measurements and observations were made to verify the adequacy of the structural design. 6.2-4 SGS-UFSAR Revision 15 June 12, 1996 After the structural integrity test at 54 psig (held for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> minimum) the pressure was reduced in the same incremental steps to 0 psig prior to performance of the containment liner leakage test. Temperature, barometric pressure, and weather conditions were recorded hourly during the test period. Prior to the strength test, predicted stress and strain at various locations were developed for an internal pressure of 54 psig. Although strain gages were installed on designated areas of the liner and concrete reinforcement, the analytically derived strains were not used as acceptance figures for the actual value. Values obtained, however, were analyzed and evaluated to determine the magnitude and direction of principal strains. Test data in excess of the predicted extremes required resolution through review of the design, evaluation of measurement errors and material variability and, if necessary, exploration of the structure. Excessive crack widths, if any, observed during the test were required to be satisfactorily resolved in a manner similar to that discussed above for displacements. 6.2.1.2.2 Test Measurements and Instrumentation An instrumentation program to determine the degree of agreement between predicted and observed deflection values at various points on the pressurized structure was employed to verify the design. Radial and vertical growth of the cylinder was measured using linear motion transducers wired to electrical indicators along four approximately equally spaced meridians. Due to the equipment layout, it was not possible to run transducer wires across at six points at each circumference as recommended in Regulatory Guide 1.18. However, numerous additional strain gages were used on the liner plate and rebar to supplement the measurements. The radial deflections of the containment were measured at the spring line, mid-height of the cylinder and at 13.5 feet above the structural SGS-UFSAR Revision 6 February IS, 1987 mat. Vertical deflections were measured at the apex and spring line of the dome. Longitudinal and circumferential growth of the liner was measured by means of electrical strain gages attached to the exposed face of the liner in an area which is subjected solely to membrane forces (see Figure 6.2-1). Strain gages were attached to selected hoop and meridional bars in the cylindrical wall and dome, as well as selected radial and circumferential top and bottom bars in the base slab. Also, strain gages were attached to representative circumferential bars around the equipment access opening and around both of the personnel access openings. Approximately 200 sets of strain gages had been attached to reinforcing bars at various locations in the containment structure. Strain gages were attached to the steel liner to record strains at the junction with the mat liner, at mid-height, at the spring line, and in the dome. Additional strain gages were attached to the liner around the equipment access and personnel hatches. Redundancy of instrumentation was attained through multiplicity of points and gages at which measurements were made, such that loss or damage to any one position would not be critical. Two basic types of gages were used: (1) BLH, or equivalent, foil gages bonded to the members with epoxy cement, and (2) Microdot, Inc., weldable gages spot-welded to the members. Where possible, gages were installed on reinforcing bars in the laboratory and the bars cadwelded in place. Measurements around the personnel and equipment hatches were made using linear motion transducers between the hatches and the polar crane wall or other fixed supports as shown on Figure 6. 2-2. Twelve linear motion transducers at each equipment and personnel 6.2-6 SGS-UFSAR Revision 6 February 15, 1987 hatch were used to measure the deflections, in accordance with Regulatory Guide 1.18. During the structural acceptance test, all gages were read and recorded with a multichannel data acquisition system. Readings were obtained just prior to pressurization, at the various selected incremental pressures during pressurization and depressurization, and after depressurization. The Unit 2 containment is a nonprototype structure, not requiring strain measurement. However, a small number of rebar and liner strain gauges were read for comparison and study at locations that had exhibited high strain when the test was performed on Unit 1. Limited variable differential transmitter measurements were not taken on the Unit 2 personnel hatches, since the test performed on Unit 1 demonstrated that the personnel hatches were structurally loaded in a manner similar to the equipment hatch. Crack patterns in the concrete were measured and recorded at the quarter points of circumference at the maximum test pressure. A strain sensitive coating was used to make the crack pattern more discernible (see Figure 6. 2-3). Crack patterns in the areas of the large penetrations were visually checked to ascertain agreement with predicted stress patterns. The range of strains and deformations expected at the 54 psig test pressure were as follows: 1. Increase in containment diameter: not more than 1.75 inches. 2. Maximum vertical elongation of the structure: not more than 2 inches. 3. Maximum width of new cracks or increase in existing cracks: not more than 0.03 inch. 6.2-7 SGS-UFSAR Revision 6 February 15, 1987

4. Residual width of new cracks or increased width of existing cracks (after containment pressure is reduced to atmospheric): not more than 0.02 inch. Since the containment structure was expected to remain in the elastic range during the pressure test, there was not expected to be any permanent distortion in the liner or in the concrete once the pressure was reduced to atmospheric or below. However, it was fully expected that small residual cracks in the concrete would appear as a result of concrete creep during pressurization. 6.2.1.2.3 Acceptance Criteria The structural acceptance test determined whether the containment structure is capable of withstanding the magnitudes of loading used in the design. The acceptance criteria is that under the test load. The behavior of the structure under the test load must be such as to indicate its ability to withstand the loadings used for design. Were the test acceptance criterion to equal or exceed the stresses computed under the factored loadings, then destruction of some elements would result. It was not necessary to test up to design stresses to verify the structural integrity of the containment. Prediction and verification of deformation patterns, using the same design and analysis procedures for both design and test conditions, serves to verify the design. Tensile stresses in the liner plate during the structural acceptance test were expected to be greater than those which would occur under the accident condition. The reason for this was that there was no temperature rise associated with the test condition. Compressive stresses would be created by the high temperatures associated with an accident condition, which overcome the tension in the liner. Stresses in the reinforcing bars were expected to 6.2-8 SGS-UFSAR Revision 6 February 15, 1987 be lower during the test condition than the values calculated for the accident condition. With regard to the liner, the largest number and length of seams occurs in the cylinder and dome and, therefore, the greatest potential for leakage. The test condition was expected to yield tensile stresses in the dome and moat of the cylinder that are higher than the design condition. The exception was the lower cylinder wall, where design tensile stresses are expected to be higher. With the exception of this area, the test placed a greater stress condition on the potential leakage paths than any of the design conditions. The acceptance criterion requires demonstration that the overall structure exhibited elastic behavior throughout the test range. Inelastic behavior at localized stress concentrations was considered acceptable. Greatest agreement between the computed strains and those actually observed was anticipated to have been in the shell of the containment. Greater disparity between observed and calculated strains was contemplated around openings and at other discontinuities,. where theoretical analysis becomes more complex. The acceptance criterion for cracking was based on the width and spacing of cracks, as determined through review of predicted crack size and crack spacing. Data obtained during the test were evaluated and a comparison with the values predicted by design was made to assess the structural behavior of the containment with regard to local and overall response. 6.2.1.3 Containment Overall Integrated Leakage Rate Teets 6.2.1.3.1 Preoperational Test The preoperational containment overall integrated leakage rate test was performed following successful completion of the structural acceptance test. The test was performed to satisfy the requirements of lOCFRSO, Appendix J, "Primary Reactor Containment 6.2-9 SCS-UFSAR Revision 6 February 15, 1987 Leakage Testing for Water COOled Power Reactors," for Type A teats. The teat was performed according to the peak pressure teat program, uainq the "absolute" method, to ascertain that the lea)Cage rate did not exceed 0.1 percent of the containment free volume per day at the design pressure of 47 paig. The teat was performed at 47 paig. 6.2.1.3.2 Periodic Testa The overall integrated leakage rate teats shall be in accordance with 10CFR50.54(o) in conformance with Appendix J of lOCFRSO, Option a, using the methods and provisions of Regulatory Gui<le 1.163, September, 1995 as modifie<l by approved exemptions. If the Type A teat frequency is performed at 10 year intervals, two additional containment aurface inapectiona shall be performed at approximately equal intervals during shutdowns between Type A tests. The performance of these testa will be limited to periods when the plant is nonoperational and secured in the shutdown condition. The periodic testa will be performed at a peak pressure of 47 paiq. Detailed teat requirements are contained in the Technical Specifications. Should deviations become necessary, they will be the subject of License Change Requests {LCR) accompanied by appropriate justification. LCR 83-04, Public Service Electric & Gas (PSB&G) memo Liden to Varqa, dated July 22, 1983, documents such a request for Unit 1. 6.2.1.4 Penetration Leakage Rate Teats 6.2.1.4.1 Preoperational Teats Penetration leakage rate teats (Type B teats) were performed in accordance with lOCFRSO, Appendix J, "Primary Reactor containment Leakaqe Testing for Water cooled Power Reactors." Only the free volu=e of the double penetrations was included in the teat. 6.2-10 SGS-UFSAR Revision 16 January 31, 1998 Because this volume is very small when compared to the containment free volume, the sensitivity and accuracy attainable in this leakage rate test was increased correspondingly over that attainable through integrated leakage rate testing. All containment piping penetrations fitted with bellows are tested at Pa. Each bellow in penetrations utilizing more than one bellow is subjected to Type B testing. The penetration leakage rate tests were performed with the penetrations pressurized to 47 psig, and the Containment Building at atmospheric pressure. The combined leakage rate for the double penetrations and isolation valves was limited to less than 0.06 percent of the containment free volume per day. 6.2.1.4.2 Periodic Tests Periodic leakage rate testing for penetrations will be conducted in a manner similar to the preoperational tests. The periodic tests will be performed according to the required frequencies set forth in 10CFR50, Appendix J, "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors, n Option A, for Type B tests. 6.2.2 Containment Heat Removal Systems Adequate post-accident heat removal capability for the containment is provided by two separate, Engineered Safety Features (ESF) Systems. These are the Containment Spray System described in Section 6.2.2.1, and the Containment Fan Cooling System, described in Section 6.2,2.2. These systems are of different engineering principles and serve as independent sources of containment cooling to assure that post-accident containment atmospheric temperature and pressure do not rise beyond their design basis values. In addition to its ability to remove elemental iodine from the containment atmosphere, the heat removal function of the containment spray system is similar to that of the containment fan coil units. As described in section 15.4. 81 "Containment Pressure Analysis", a minimum of three containment fan coil units in operation with a single containment spray train is capable of maintaining post-accident containment temperature and pressure below their design basis values, assuming a worst-case single active failure. Thus, design margin exists for the containment heat removal system. 6.2-11 SGS-UFSAR Revision 20 May 6, 2003 6.2.2.1 Containment Spray System 6.2.2.1.1 Design Bases The primary purpose of the Containment Spray System is to spray cool water into the containment atmosphere in the event of a loss-of-coolant accident (LOCAl and thereby ensure that containment pressure does not exceed the design value of 47 psig at 271°F (100 percent relative humidity). This protection is afforded for all pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant pipe. Pressure and temperature transients for LOCA are presented in Section 15. Although the water in the core after a LOCA is quickly subcooled by the Safety Injection System (SIS), the Containment Spray System design is based on the conservative assumption that the core residual heat is released to the containment as steam. The Containment Spray System is designed to spray at least 2600 gpm of borated water into the Containment Building whenever two out of four (hi-hi) containment pressure signals occur or a manual signal is given. 6.2-12 SGS-UFSAR Revision 20 May 6, 2003
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  • Either of two subsystems containing a pump and associated valving and spray headers are independently capable of delivering 2600 gpm. The design basis is to provide sufficient heat removal capability to maintain the post-accident containment pressure below the design pressure assuming that the core residual heat is released to the containment as steam. A second purpose served by the Containment Spray System, including the recirculation phase, is to remove fission products (primarily iodine) from the containment atmosphere should it be released in the event of a LOCA. The analysis of offsite dose after a hypothetical LOCA is presented in Section 15. Iodine removal effectiveness is described in Section 6.2.3. The Containment Spray System is designed to operate over an extended time period, following a primary coolant maintain containment conditions at failure, as required to restore and near atmospheric pressure. It has the capability of reducing the containment post-accident pressure taking into account any reduction in capacity due to a single failure as defined in Section 6.2.2 . Portions of other systems which share functions and become part of the Containment Spray System, when required, are designed to meet the criteria of this section. Any failure of an active component in either spray subsystem does not degrade the minimum containment cooling as defined in Section 6. 2. 2 or fission product removal capability of the Containment Spray System, as the containment pressure-temperature analysis in Section 15 assumes the most restrictive single failure. Those portions of the spray systems located outside of the containment which are designed to circulate, under post-accident conditions, radioactively contaminated water collected in the containment meet the following requirements: 6.2-13 SGS-UFSAR Revision 23 October 17, 2007
1. Adequate shielding to maintain radiation levels within the limits of 10CFR50.67 (Section 11.2). 2. Collection of discharges from pressure relieving devices into closed systems. 3. Means to limit radioactivity leakage to the environs, consistent with limits set forth in 10CFR50.67. System active components are redundant. System piping located within the containment is redundant and separable in arrangement. All portions of the system located within containment are designed to withstand, without loss of functional performance, the post-accident containment environment and operate without benefit of maintenance for the duration of time to restore and maintain containment conditions at near atmospheric pressure. Table 6.2-1 tabulates the codes and standards to which the Containment Spray System components are designed. 6.2.2.1.2 System Design System Description Adequate containment cooling and iodine removal are provided by the Containment Spray System shown on Plant Drawings 205235 and 205335 whose components operate in sequential modes. These modes are: 1. Spray a portion of the contents of the refueling water storage tank (RWST) into the containment atmosphere using the containment spray pumps. During this mode, the contents of the spray additive tank (sodium hydroxide) are mixed into the spray stream to enhance the iodine removal capability of the Containment Spray System. 6.2-14 SGS-UFSAR Revision 27 November 25, 2013
2. Recirculation of water from the containment sump is provided by the diversion of a portion of the recirculation flow from the discharge of the residual heat removal (RHR) heat exchangers to the containment spray header after injection from the RWST has been The bases for the selection of the various conditions requiring system actuation are presented in Section 15. The principal components of the Containment Spray System are: two pumps, one spray additive tank, two eductors, spray ring headers and nozzles, and the necessary piping and valves. The containment spray pumps and the spray additive tank are located in the Auxiliary Building and the spray pump auctions are lined up to the RWST. Following an accident, the containment spray pumps are utilized until the water in the RWST is depleted. During the recirculation phase, the system utilizes the two RHR pumps, two residual heat exchangers and associated valves and piping of the SIS. The spray system is actuated by two out of four hi-hi containment pressure signals. The starting signal energizes the pumps and opens the discharge valves to the spray headers. The valves associated with the spray additive tank are opened on the same signal. If necessary, the operator can manually actuate the entire system from the control room. During the period of time that the spray pumps draw from the RWST, a small portion of the spray flow is diverted from the spray pump discharge line through the eductor and back to the pump suction. Valve CSl4 in the spray additive tank discharge line is provided with redundant position indication to assure effective chemical addition to the spray system. The liquid from the spray additive tank then mixes with the liquid entering the suction of the pumps. The result is a solution suitable for the removal of iodine from 6.2-15 SGS-UFSAR Revision 6 February 15, 1987 the containment atmosphere. The analysis of the iodine removal capability of the Containment Spray system, presented in Section 6.2.3, shows that most of the removable iodine in the containment atmosphere is washed out in the injection phase. After the injection operation, spray pump flow is discontinued when the water in the RWST is depleted. Containment pressure control can then be maintained with the RHR System functioning through the containment spray headers. If, for any reason, the containment pressure should be observed to increase, the operator can direct part of the discharge flow from the residual heat exchangers to the spray headers, thereby initiating recirculation spray flow. The procedure for the change-over from injection to recirculation and cooling water for the residual heat exchangers is described in Section 6.3. Comoonents All associated components, piping, structures, and power supplies of the Containment Spray System are designed to Class I (seismic) criteria. The Containment Spray System shares the RWST liquid capacity with the SIS. Refer to Section 6.3 for a detailed description of this tank. Pumps The two containment spray pumps are of the horizontal centrifugal type, driven by electric motors which can be supplied with power from the standby ac power supply. 6.2-16 SGS-UFSAR Revision 16 January 31, 1998 The design head of the pumps is sufficient to continue at rated capacity with a minimum level in the RWST against a head equivalent to the sum of the design pressure of the containment, the head to the uppermost nozzles, and the line and the nozzle pressure losses. Pump motors are direct-coupled and large enough for the maximum power requirements of the pumps. The materials of construction are stainless steel or equivalent corrosion-resistant material. Design parameters are presented in Table 6.2-2 and the pump head characteristic curve is presented on VTD 121398. The containment spray pumps are designed in accordance with the specifications discussed for the pumps in the SIS, Section 6.3. The pump motors are non-overloading to the end of the pump curve. Each containment spray pump motor is provided with a shroud to prevent water spray damage from MEL piping as described in Section 3.6.5.12.5. Details of the component cooling pumps and service water pumps, which serve the SIS, are presented in Section 9. Spray Headers and Nozzles The containment spray header piping arrangement is shown on Plant Drawings 207466 and 207467. These drawings illustrate the spray nozzle orientation, which has been designed to provide maximum spray coverage of the containment. The arrangement consists of four 360 degree ring headers at different elevations, with alternate headers connected. The header diameters are 101 feet at Elevation 244 feet-6 inches, 96 feet at Elevation 24 7 feet-0 inch, 53 feet at Elevation 266 feet-6 inches, and 48 feet at Elevation 269 feet-0 inch. The spray headers are stainless steel of a hollow-cone pressure nozzle design, with a 3/8-inch diameter orifice. The nozzles have no internal parts which would be subject to clogging. Sauter mean drop size of less The nozzles produce a drop size spectrum with a 6.2-17 SGS-UFSAR Revision 27 November 25, 2013 than 1000 microns with the spray pump operating at design conditions and the containment at full design pressure and temperature. The spray header supports are shown on Plant Drawings 223112, 223114 and 223123. These figures illustrate the relationship of the support steel to the headers and the containment building wall. The supports are designed such that interference with the spray pattern is kept to a minimum and their structural integrity under accident and seismic conditions is maintained. The design is such that the alternate connected ring headers and corresponding sections of riser (from the last anchor point on the containment wall) will act as a unit under design thermal and seismic conditions. The pipe hangers and restraints are designed to support and restrain the pipe under design thermal and seismic conditions. Spray Nozzles The spray nozzles are of a hollow-cone pressure nozzle design without any internal parts subject to clogging. The nozzles produce a drop size spectrum with a Sauter mean drop size less than 1000 microns with the spray pump operating at design conditions and the containment at design pressure and temperature. During spray recirculation operation, the water is screened through 1/12-inch (2.1 mm) diameter holes before leaving the containment sump. are stainless steel and have a 3/8-inch diameter orifice. The spray nozzles The nozzles are connected to four 360-degree ring headers of ring headers (alternating headers connected) of diameter 101 feet (Elevation 244 feet-6 inches), 53 feet (Elevation 266 feet-6 inches), 96 feet (Elevation 247 feet), 48 feet (Elevation 2 69 inches) . The nozzles and headers are so oriented as to maximize coverage of the containment volume. 6.2-18 SGS-UFSAR Revision 27 November 25, 2013 All stresses are within those allowed by ANSI B31.1 Piping Code. Heavier walled pipe is used at anchor points and points of restraint to eliminate high stress regions. Containment Dome Access System -Unit 2 Unit 2 utilizes a different design, the Containment Dome Access System. This system serves the dual purpose of supporting the Containment Spray System ring header piping and providing access for maintenance and inspection to the ring headers and the containment dome liner (see Vendor Technical Document 142864). This system consists of the following components: 1. An orbital inclined service bridge and trolley capable of carrying personnel and material, including an auxiliary hoist. It is designed to provide maximum coverage of both the containment dome liner and the spray header piping (see Vendor Technical Document 142864). 2. A structural steel girder, beam, and the support structure for the access bridge and spray piping (see Vendor Technical Document 142850) Both components are seismic Class I and have been statically dynamically designed to withstand the effects of the design basis earthquake. combined total weight of 394,000 lb. They have The support beams for this system penetrate the containment liner plate and are anchored into the concrete wall of the Reactor Containment Building. In order to maintain containment integrity, the penetrations through the liner plate are seal welded into place and vacuum box tested. A leak chase box is installed around each embedded beam to enable leak rate testing of the welds at any time (see Plant Drawing 224351) The orbital service bridge, the spray header support or basket, and the spray piping were mathematically modeled as a system of 6.2-19 SGS-UFSAR Revision 27 November 25, 2013 node points interconnected by various weightless springs. The springs were assigned and stiffness characteristics of the structural beam and functional pipe elements of the system. All weights and inertias were distributed among the nodes. The degrees of freedom of the nodes were chosen to closely simulate the response of the system to external loading; the materials were assumed to be linearly elastic. Static analysis was performed to obtain the maximum stresses under dead load and thermal variations. Using the above mathematical model, a dynamic modal analysis was also performed to determine the modal frequencies and mode shapes. Safe Shutdown Earthquake (SSE) response spectra with 0. 5 -percent damping factor at the proper structural elevations were used as the input for the response spectrum analysis. The element stresses of those modes with meaningful participation for a given excitation direction were summed as a square-root-of-the-sum-of-frequencies occurred within the-squares (SRSS). When mode 10 percent of each other, an absolute summation of stresses was made prior to root mean square (RMS) summation. The design stresses for the system are the summations of the maximum static and dynamic stresses for the respective members. The analysis assumed the orbital bridge was locked to the rail in its storage location, the personnel cage was locked in the down (stored) position on the bridge with no load on the hoist, and the containment spray piping empty of liquid. This analysis simulates actual conditions during reactor operation. The calculations performed on the Dome Access System indicate that none of the elements are subjected to loads beyond the allowable value of 32.40 ksi, which is 90 percent of the minimum yield strength of A36 steel. The loads obtained from the calculations for the Dome Access System were then used to design the Dome Access System containment SGS-UFSAR interface 6.2-20 tie-supports. Revision 9 July 22, 1989 These '"-"""
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  • which are made of A442 Grade 60 steel with an allowable stress of 19 ksi, will be ected to a stress of only 10.92 ksi. The allowable load on the access system will not be exceeded due to administrative control. The Dome Access System, consisting of the orbital service bridge and supporting basket and the spray header piping, was analyzed for an SSE using response spectrum curves at 0.5 percent damping with the bridge in the storage location. The bridge, basket, and piping were mathematically modeled as a multi-degree-of-freedom system with node points interconnected by various springs. ANSYS, a large analysis. general purpose computer program, was used to perform the modal Spray Additive Tank The Spray Additive Tank holds a solution sodium hydroxide. The concentration of this solution assures that the injection spray pH will be at least 8.5. The capacity of the tank is sufficient to contain enough sodium hydroxide solution which, upon mixing with the refueling water from the RWST, the boric acid from the boron ection tank {BIT) 1 the borated water contained within the accumulators, and coolant, will the containment sump to a pH greater than 7. 0. This assures adequate retention of the absorbed iodine in I the sump liquid, and minimizes chloride induced stress corrosion cracking of stainless steel. Although iodine removal capability is maintained under these conditions, no credit is taken for any iodine removed after decontamination factor limitations specified by the Standard Review Plan, Section 6.5.2 (Ref. 26) are reached during the injection and recirculation phases. A level indicating alarm is provided in the Control Room if, at any time, the solution tank contains less than the required amount of sodium hydroxide solution. Periodic sampling confirms that proper sodium hydroxide concentration exists in the tank. Also, a flow indication is provided in the Control Room to alert the if there is low flow from the tank when required. The tank design parameters are given in Table 6.2-3 . 6.2-21 SGS-UFSAR Revision 23 October 17, 2007 Beat Exchangers The two residual heat exchangers that are used during the recirculation phase are described in section 6.3. Valves The valves for the containment Spray System are designed in accordance with the specifications for the valves in the SIS. Valving descriptions and valve details are described in Section 6.3. Piping The piping for the containment Spray System is designed in accordance with the specifications for piping in the SIS (Section 6.3). The system piping is designed for 250 psig at l50°F. Motors for Pumps and Valves The motors for the Containment Spray system are designed in accordance with the specifications discussed for motors in the SIS (Section 6.3). 6.2.2.1.3 Design Evaluation Range of Containment Protection During the injection phase following the maximum LOCA (i.e., during the time that the containment spray pumps take their suction from the RWST) the Containment Spray System provides the design heat removal capacity for the containment. After the injection phase, each train of the Recirculation System provides sufficient cooled recirculated water to keep the core flooded as 6.2-22 SGS-UFSAR Revision 6 February 15, 1987
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  • well as providing, if sufficient flow to the containment spray headers to maintain the containment pressure below the design value. This applies for all reactor coolant pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant pipe. Only one spray header is required to operate for this capability at the earliest time recirculation is initiated. The Containment Spray and Fan Cooler Systems are capable of removing sufficient energy to maintain the pressure below the containment design pressure even in the event of a single failure. Each of these systems consists of independent equipment and components supplied from separate power sources. One containment spray train and three of five fan coolers, along with one train of the ECCS, is sufficient to ensure containment integrity. During the injection and recirculation phases, the spray water is raised to the temperature of the containment in falling through the steam-air mixture. The minimum fall path of the droplets is approximately 116 feet from the lowest spray ring headers to the operating deck. The actual fall path is due to the trajectory of the droplets sprayed out from the ring header. Heat transfer calculations show that thermal equilibrium is reached by all droplets in the first few feet of their fall. Thus, the spray water reaches essentially the containment saturation temperature. discussed in Section 15. The model for spray heat removal is In addition to heat removal, the Spray System is effective in scrubbing fission products from the containment atmosphere. Credit is taken for the removal of fission (primarily iodine) in the analysis of the hypothetical LOCA {Section 15). A discussion of the effectiveness of containment spray as fission product removal process is contained in Section 6.2.3. One containment spray pump and recirculation spray provide sufficient iodine scrubbing to ensure that post-accident fission product leakage 6.2-23 SGS-UFSAR Revision 23 October 17, 2007 I

{based on Reg. Guide 1.183 release fractions) would not result in doses the limits of 10CFR50.67. System Response *rhe starting sequence of the containment spray pumps and their related emergency power equipment is designed so that del:i very of the required spray into the containment is reached in 85 seconds following the appropriate ini t:Lating trip signal. 'J'his time delay for J.ni tiation of containment spray has included of signal delay, assumed loss of offsi te power, diesel start time, breaker closure, SEC sequencing and the time for the spray pumps to reach full speed and to fill the spray headers and piping. The above delay time is consistent with the safety analysis described in Chapter 15 and l:he Tech limit for containment spray pump response. Single Failure Analysis A failure has been made on alJ. active components of the system to show that the failure of any single component will not prevent fulfilling the design function, This analysis is summarized in 'l'able 6. 2-4. '!'he LOCA analysis presented in Section 15 n::lflects the single failure analysis. Reliance on Interconnected Systems The Containment Spray System initially operates independently of other ESF following a LOCA. It containment in combination with the Containment Fan Cooling System. For extended operation in the recircu1ation mode, water is supplied through the RHR pumps and heat exchangers. During the recirculation phase, some of the flow leaving the residual heat may be bled off and sent to either the discharge of the containment sp:ray pumps or to the suction of the ect:Lon pumps and centrifugal charging pumps. Minimum flow requirements will be set for the flow being senl: to the core and for the flow being sent to the containment spray pump 6.2-24 SGS-UFSAR Revision 23 October 17, 2007 * *

  • discharge. Sufficient flow instrumentation is provided so that the operator can perform appropriate flow adjustments with the remote throttle valves in the flow path. Shared Function Evaluation Table 6. 2-5 presents an evaluation of the main components which have been discussed previously and a brief description of how each component functions during normal operation and during the accident. Net Positive Suction Head (NPSH) and Spray Water Entrapment Spray recirculation has been evaluated considering loss of water through entrapment outside the containment sump. There are three areas within the containment where reactor coolant blowdown liquid and spray water may become trapped: the reactor cavity, the refueling canal, and the reactor instrumentation tunnel. The reactor cavity has ventilation openings around the reactor that would allow spray water to drain to the lower elevations of the containment. The refueling canal is normally isolated from the Fuel Handling Building and would trap no more than 9,500 gallons of liquid from Containment Spray System. The instrumentation tunnel has a water capacity of approximately 70,000 gallons, none of which would drain to the sump. The total quantity of water released to the containment at the beginning of the recirculation phase of the Containment Spray System operation, assuming a DBA with reactor coolant loop piping half full of water, is approximately 275,000 gallons. Discounting the water volume trapped in the refueling canal and the reactor instrumentation tunnel, the volume available at the suction of the RHR pump used for containment spray is approximately 190,000 gallons. The required NPSH for the RHR pump is a water level relative to the bottom (Elevation 70 feet) of the 8-foot deep containment sump. The indicated available water volume is a water level several feet above the containment sump top. There is 6.2-25 SGS-UFSAR Revision 6 February 15, 1987 therefore no significant effect on the required static head for the RHR pump. Available and required NPSH for the containment spray pumps and the RHR pumps are provided in Table 6.2-6. Compliance with Regulatory Guide 1.1 is discussed in Appendix 3A. During a moveable shield provides missile protection for the area immediately above the reactor vessel. The spray headers are therefore protected from missiles originating within the shield. Active components of the Containment Spray System are located outside the containment, and hence are not required to operate in the steam-air environment produced by the accident. Material Compatibility Parts of the system in contact with borated water, sodium hydroxide spray additive, or mixtures of the two are stainless steel or an equivalent corrosion-resistant material. 6.2.2.1.4 Tests and Inspections Inspection Capability Where practicable, all active components and passive components of the Containment Spray System are inspected periodically to assure system readiness. The pressure-containing systems are inspected for leaks from pump seals, valve packing, flanged joints, and safety valves. During operational testing of the containment spray pumps, the portions of the systems subjected to pump pressure are inspected for leaks. Design provisions for inspection of the SIS, which also functions as part of the Containment Spray System, are described in Section 6.3. 6.2-26 SGS-UFSAR Revision 6 February 15, 1987
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  • System and Component Testing Active components of the Containment Spray System were adequately tested both in pre-operational performance tests in the manufacturer's shop and in place after installation. Thereafter, periodic tests are also performed after component maintenance. Means are provided to test initially, under conditions as close to design as is practical, the full operational sequence that would bring the Containment Spray System into action. The containment spray pumps can be tested individually by opening the valves in the miniflow line. Each pump, in turn, can be started by operator action and checked for flow establishment. The spray injection valves can be tested with the pumps shut down. The spray additive tank valves can be opened periodically for testing. The contents of the tank will be periodically sampled to determine that the proper solution is present. During these tests, the equipment will be visually inspected for leaks. Leaking seals, packing, or flanges will be tightened to eliminate the leak. Valves and pumps will be operated and inspected after any maintenance to ensure proper operation. Permanent test lines for all spray loops are located so that the system, up to and including the isolation valves at the spray header, can be tested. These isolation valves can be checked separately. Flow bypass through the eductors was checked during the initial preoperational tests of the Spray System. Subsequent system tests will be made with the spray additive tank bypass valves closed. The air test lines for checking spray nozzles connect downstream of the isolation valves. Air flow through the nozzles is monitored as required by the Technical Specifications . 6.2-27 SGS-UFSAR Revision 18 April 26, 2000 The functional test of the ECCS described in Section 6.3 includes the operation of the Containment Spray System. A test signal simulating the containment spray initiating signal is used to demonstrate operation of the Spray System up to the isolation valve on the pump discharge. Spray Nozzles The spray nozzles are of a hollow-cone pressure nozzle design without any internal parts subject to clogging. The nozzles produce a drop size spectrum with a Sauter mean drop size less than 1000 microns with the spray pump operating at design conditions and the containment at design pressure and temperature. During spray recirculation operation, the water is screened through 1/12-inch {2.1 mm) diameter holes before leaving the containment sump. The spray nozzles are stainless steel and have a 3/8-inch diameter orifice. The nozzles are connected to four 360-degree ring headers of ring headers (alternating headers connected) of diameter 101 feet (Elevation 244 feet-6 inches), 53 feet (Elevation 266 feet-6 inches), 96 feet (Elevation 247 feet), 48 feet (Elevation 269 feet) . The nozzles and headers are so oriented as to maximize coverage of the containment volume. 6.2.2.2 Containment Fan Cooling System 6.2.2.2.1 Design Basis The Containment Fan Cooling System is designed to recirculate and cool the containment atmosphere in the event of a LOCA and thereby ensure that the containment pressure will not exceed its value of 47 psig at 271°F (100-percent relative humidity). Although the water in the core after a LOCA is quickly subcooled by the SIS, the Containment Fan Cooling System is designed on the 6.2-28 SGS-UFSAR Revision 24 May 11, 2009 conservative containment as steam. that the core residual heat is released to the The Containment Ventilation System (Section 9.4) which includes the Containment Fan Cooling System, is designed to remove the normal heat loss from equipment and piping in the reactor containment during plant operation and to remove sufficient heat from the reactor containment, the initial LOCA containment pressure to keep the containment pressure from the pressure. The fan cooler units continue to remove heat after the LOCA and reduce the containment pressure close to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. within the first In addition to the design bases specified above, the following objectives are met to provide ESF functions: 1. Each of the five fan-cooler units is normally of transferring heat at the rate of at least 44 x 106 Btu/hr from the containment atmosphere at post-accident peak conditions, i.e., a saturated air-stream mixture of 43.5 psig and 265.9°F. The accident analyses of Section 15 determined a minimum number of three fan-cooler units, along with other containment heat sinks, are needed to maintain containment integrity. This correlates to a cumulative heat transfer rate of at least 132 x 106 Btu/hr. This heat transfer rate exceeds the analyzed value assumed in the accident of Section 15. The establishment of basic heat transfer for the coils of the fan-cooler units, and the calculation by computer of the overall heat transfer capacity are discussed in Section 15.4. 2. In removing heat at the design basis rate, the cooler coils are capable of discharging the resulting condensate without impairing the air flow capacity of the fan coolers and without raising the exit temperature of the service water to the boiling point. Since condensation of water from the air-steam mixture is the principal mechanism for removal of heat from the post-;-accident containment SGS-UFSAR by the cooling coils, the coil fins will operate as wetted surfaces under these conditions. Entrained water added to the air-steam mixture, such as by operation of the Containment Spray System, will therefore have essentially no effect on the heat removal capability of the coils. 6.2-29 Revision 26 May 21, 2012 I In addition to the above bases, the is to at the post-accident conditions of 47 psig and 271°F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, followed by operation in an air-steam atmosphere at 20 psig, 219°F for an additional 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />. The design will permit subsequent of an air-steam at 5 an indefinite All components are capable of withstanding or are protected from d/ps which may occur during the rapid pressure rise to 47 psig in 10 seconds. Portions of other systems which share functions and become part of this Containment Cooling this section. Neither a the injection will nor the heat when are designed to meet the criteria of active failure in such systems failure the recirculation of cooling. Where portions of these other systems are located outside of containment, the following features are incorporated in the design for operation under post-accident conditions: 1. Means for isolation of any section. 2. Means to detect and control leakage into the environs, to the limits consistent with limits set forth in 10CFR50.67. 3. The RCFC (or CFCU) units are able to deliver their designed cooling capacity under all normal and abnormal conditions. Two-phase flow regions within the RCFC coo1ing coils following a LOCA/MSLB concurrent with a Loss of Offsite Power event are by the RCFC cooling coils water solid during all normal and abnormal conditions. The waterhammer issues and modifications to the SW system (see Section 9.2, Service Water System) addressed in Generic Letter 96-06, preclude the possibility of the detrimental heat transfer effects resulting from the development of two-phase flow regions within the RCFC cooling coils. 6.2-30 SGS-UFSAR Revision 23 October 17, 2007 6.2.2.2.2 System Description The Containment Fan Cooling System is illustrated on Plant Drawings 205238 and 205338. Individual system components and their supports meet the requirement for Class I (Seismic) structures and are isolated from fan vibration. The Containment Fan Cooling System consists of five air handling units, each including motor, fan, motor heat exchanger, cooling coils, roughing filters, dampers, duct distribution system, instrumentation, and controls. The units are located on the operating floor, between the containment wall and the polar crane wall. Each fan is designed to supply a nominal 110, 000 cfm during normal operation and 40,000 cfm during accident operation. The fans are direct driven, centrifugal type, and the coils are plate fintube type. Each fan-cooler unit is normally capable of removing at least 44 x 106 Btu/hr or a cumulative of 132 x 106 Btu/hr for three fan-cooler units from the containment atmosphere under accident conditions. A minimum of 1300 gpm of service (cooling) water is supplied to each unit during accident conditions. The design maximum river water inlet temperature is 90 oF, which results in an outlet temperature of 160°F under design basis conditions or 205°F for zero fouling case. Assuming a single active failure of the CFCU high speed breaker to open following an SEC MODE I or III accident signal, the outlet temperature could reach 209°F. The Section 15 accident analysis also assumes an additional degradation in the heat transfer rate of 10% for the first two minutes of diesel powered fan cooler operation. This assumption accounts for nitrogen gas that could be released from solution from the service water system accumulators that provide a part of the resolution of Generic Letter 96-06. 6.2-31 SGS-UFSAR Revision 27 November 25, 2013 Duct work distributes the cooled air to the various containment compartments and areas. During normal operation, the flow sequence through each air handling unit is as follows: inlet dampers, roughing filters, cooling coils, fan, discharge header. During post-accident operation, air is drawn through a moisture separator, a post-accident high-efficiency particulate air (HEPAJ filter section and cooling coils and is discharged to the duct header. Tight closing dampers isolate the post-accident filter section from the normally operating components. These dampers are tripped to the accident position upon either ma:t:lual or automatic actuation of the respective fan. Electrically operated four-way solenoid valves control instrument air to the damper control cylinders. On a loss of either control air or control power the dampers fail to the accident (open) position. The Fan Cooling System is actuated (in the post-accident mode) by a safety injection signal. The accident analysis assumes the CFCU initiating safety injection signal was containment high pressure because this is the limiting time delay case. Either all five fans or a minimum of three fans are started by the safeguards equipment controller, depending on the availability of emergency power. A flow switch at each fan indicates whether air is circulating in the intended normal or post-accident flowpath. Control Room. Indication and alarms are provided in the Flow Distribution and Flow Characteristics The location of the distribution ductwork outlets, together with the location of the fan cooler unit inlets, ensures that the air will be directed to all areas requiring ventilation before returning to the units. In addition to ventilating areas inside the periphery of the polar crane rail, the distribution system also includes branch ducts located at opposite extremes of the containment wall for ventilating th*e upper portion of the containment. These ducts extend upward along the containment wall as required to permit the throw of air from the ducts to reach the dome area and assure that the di.scharge air will mix with the atmosphere. I SGS-UFSAR 6.2-32 ----------------------------------Revision 18 April 26, 2000 * * *
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  • The air discharged inside the periphery of the polar crane rail circulates and rises above the operating floor:--t:hrough openings around the steam generators where it mixes with air displaced from the dome area. This mixture is returned to the fan coolers located on the operating floor. The temperature of this air will be essentially the design ambient for the containment vessel (120°F average maximum). The steam-air mixture from the containment entering the cooling coils initially during the accident will be at approximately 271°F and have a density of 0.172 .( pounds per cubic foot. Mast of the water vapor will condense on the cooling
  • coil, and the air leaving the fan cooler will be saturated at a temperature slightly below 271°F. With a flow rate of 39,000 cfm from each of 5 fans under accident conditions and a containment net free volume of 2,620,000 ft3, the recirculation rate with five fans operating is approximately 4.5 containment volumes per hour. Cooling Water for the Fan Cooler Units The cooling water requirements for all five fan cooler units during a LOCA and recovery are supplied by the Service Water System. The Service Water System is described in Section 9. The design basis river water temperature for service water to the containment fan coolers is 90°F, although river water temperatures throughout the year are normally less. The service water temperature rise through the containment fan coolers is approximately 8°F for normal operation. In the unlikely event of an accident, this*ternperature rise will be a maximum of approximately 70°F* for a period of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, after which it will decrease. It is not expected that any significant amount of calcium carbonate precipitation on the heat exchanger surfaces will occur at these temperatures, and, therefore, there will be no subsequent plugging of the fan coolers. As part of the issues addressed in Generic Letter 96-06, certain design constraints have been applied to the SW system for the Containment Cooling System CFCU (RCFC) *101°F for the zero fouling case and 131°F for the accident and single active failure of the flow controller to reset to accident flow . 6.2-33 SGS-OFSAR Revision 18 April 26, 2000 I units. These constraints and modifications are discussed in further detail in Section 9. 2 (Service Water System) . The following constraints apply to the cooling water for the fan cooler units: effects, or in elevation cannot be tolerated in the SW system. These locations must remain water solid during all operating conditions. 2. The pressure in the flowing portions of the SW system must remain above the fluid saturation pressure for all operating conditions. Flashing or boiling resulting from increased temperatures or decreased pressures cannot occur. This constraint precludes the possibility of waterhammer or other hydraulic events due to steam bubble collapse, two-phase flow, or steam propelled water slugs. 3. The SW containment and the containment closed must have pressure relief capability, or be shown not to be to large increases in internal pressure due to increased fluid temperatures. This constraint prevents failures of the containment boundary due to thermally induced overpressures. Service water discharge from the cooling coils is subsequently mixed with the water where radiation monitors R13A and R13B sample the effluent to to the river. An alarm is annunciated in the control room upon detection of high in an effluent line. Flow and temperature indication is provided outside containment for service water flow to and from each fan cooler unit. Abnormal flow alarms for inservice fan cooler units are provided in the control room. With the CFCU fixed resistance control scheme, the restricting orifices along with the flow control valve have been sized/set to establish a target flow of approximately 1900 gpm to 2100 gpm for each CFCU, on the service water header pressure at the CFCUs. This flow rate is than the nominal flow rate to maintain the containment ambient temperature less than or equal to the Tech Spec value of 120°F. This should help in maintaining lower containment ambient temperature during times of high river water temperature. During normal operation, the flow control valve will open to its open limit stop position (approx. 50% open) to provide the target flow rates described above. 6.2-34 SGS-UFSAR Revision 25 October 26, 2010 During a safety injection, the flow control valve will also open to its open limit stop position to provide minimum flow of 1300 gpm to each CFCU. The control valve closes when the fan cooler unit is not in use. The solenoid valve associated with the flow control valve is to apply control air header pressure to the valve operator (closing the valve) whenever the RCFC fans are not running, or DC power is lost. The solenoid valve would then be energized to apply the pneumatic control signal to the flow control valve operator when the RCFC fan is operating in high or low speed. Components Roughing and HEPA Filters The roughing filters in each fan cooler unit are designed to remove the larger particles of suspended dust and dirt from the containment atmosphere during normal power conditions. operation, normal reactor shutdown and loss of offsite power Removal of the particles also prevents buildup on the cooling coils, thus avoiding a reduction in heat transfer. The roughing filters are arranged in two banks, each consisting of structural steel frame and removable filter cells. Each filter cell contains a media which is capable of removing 90 percent of visible dust particles. The media efficiency is 70 percent on National Bureau of Standards type test ratings. The HEPA filters in each fan cooler are provided to remove any particulate matter from the containment atmosphere. The HEPA filters are arranged in a structural steel frame and are individually removable. The filter media is with asbestos separators and is capable of collecting 99 percent of particles 0.3 micron and larger in size from a saturated (100-percent relative humidity) 271°F atmosphere processed through the filter at 250-300 fpm. The HEPA filter media meets MIL-F-51079 and MIL-STD-282. Fan-Motor Units The five containment cooling fans are of the centrifugal, non-overloading direct drive type. Each fan provides a minimum flow rate of 39,000 cfm when operating against the system resistance existing during accident conditions (0.172 lb/ft3 density, a containment pressure of 47 psig, and temperature of 271 °F)
  • SGS-UFSAR 6.2-35 Revision 24 May 11, 2009 I The two-speed containment fan cooler motors are totally enclosed, fan cooled (TEFC), 300 hp (high speed), induction type, 3 phase, 60 cycle, 1200 RPM, 460 volt with ample insulation margin. At low speed the motor delivers 100 hp . Insulation is Class F (NEMA rated total temperature 1550C) Westinghouse Thermalastic. It is impregnated and coated to give a homogeneous insulation system which is highly impervious to moisture. Internal leads and the terminal box-motor interconnection are given special design consideration to assure that the level of insulation matches or exceeds that of the basic motor system. At incident ambient and/or accident load conditions (27l0F and 100 hp), the motor insulation hot spot temperature is not expected to exceed 113DC. Fan cooler motors are cooled by an air-to-water heat exchanger which is connected to the motor to form an entirely enclosed cooling system. Air movement is through the heat exchanger and is returned to the motor. Two vent valves permit containment ambient air to enter the cooling compartment (on increasing containment pressure) so the motor bearings will not be subjected to an excessive d/p. An open condensate drain line will enable the cooling compartment to equalize with the containment pressure as containment pressure is reduced by the motor heat exchanger. Cooling water is supplied by the Service Water System (SWS). The motors are equipped with high temperature grease lubricated ball bearings to withstand the design basis incident ambient temperatures. Continuous bearing temperature monitoring is provided which will alarm in the control room. Fan motor leads are brought out of the frame through a seal and into a motor junction box. The motor leads are spliced to the field cables using environmentally qualified splice kits. Overload protection for the fan motors is provided at the switchgear by overcurrent trip devices in the motor feeder breakers. The breakers can be operated from the Control Room and can be reclosed from the control room following a motor overload trip. In addition to the usual quality control tests which are performed to give assurance that the motors meet design specifications, special tests are performed to demonstrate that insulation margins are built in as expected. The completely wound stators are given a special high potential test to ground. The stators are immersed in water, meggered, and given a high potential test while immersed. After passing the water tests, the motor is baked and given a final coating dip. The stator and rotor are then baked again. 6.2-36 SGS-UFSAR Revision 19 November 19, 2001 * *
  • Coils are fabricated of AL-6X tubing. The heat removal capability of the cooling coils is at least 44 x 106 Btu/hr per fan cooler unit at saturation conditions (265.9°F, 43.5 psig). The design internal pressure of each coil is 200 psig and the coils can withstand postulated design basis. accident pressures and temperatures without damage. Each recirculating unit consists of 12 coil units mounted in two banks of 6 coils high. These banks are located one behind the other for horizontal series air flow, and the tubes of the coil are horizontal with vertical fans. A moisture separator in each fan cooler removes the larger droplets of suspended moisture from the containment atmosphere in the event of a LOCA. Removal of the droplets prevents any significant water deluge over the face of the HEPA filters and thus avoids a serious reduction in filter effectiveness. The separator consists of a structural steel frame with removal separator elements. Each element is of 95 of water 10 microns and in size. The coils are provided with drain pans and drain piping to prevent flooding during accident conditions. This condensate is drained to the containment sump. The ducts are designed to withstand the sudden release of RCS energy and energy from associated chemical reactions without failure due to shock or pressure waves by incorporation of damper along the ducts which open at slight overpressure ( 3. 0 psi) . The ducts are designed and supported to withstand thermal expansion during an accident. The seismic design and analysis methodologies used to qualify all ductwork and the contained equipment are described in Section 3.8.4.4.1. Ducts are of welded and flanged construction. All longitudinal seams are welded. Where joints are joints are with that are suitable for postulated design basis accident conditions. Ducts* are constructed of galvanized sheet metal. 6.2-37 SGS-UFSAR Revision 26 May 21, 2012 All air control dampers that are an integral part of the fan coolers are designed to Class I seismic criteria. The damper construction is designed to withstand the design basis earthquake (DBE) concurrent with the pressure transients, thermal energy, and chemical activity resulting from a LOCA. Each damper is constructed of painted steel, with multiple blades that operate in unison and seals to minimize air The backdraft damper at the of each fan cooler is a normally closed counter-weighted device that opens automatically when the fan operates. It is designed to remain intact and operable during any LOCA by withstanding an approximate 7-psi air pressure surge over a 10-second period. prevents the pressure surge from damaging the fan-motor assembly. This damper shut-off are provided at each fan cooler to divert air flow through the HEPA filters and moisture during any LOCA or through the roughing filters during normal operation. The roughing filter dampers are normally open and fail closed. The HEPA filter dampers are normally closed and fail open. Both sets of dampers revert to their fail positions after a safety injection signal. Each two-position shut-off damper is provided with redundant pneumatic that can provide 150 of the operating torque. Each damper assembly is to remain intact and operable during any LOCA a air pressure surge over a 10-second The fan coolers are equipped with pressure relief dampers in the filter enclosures. These dampers are normally closed counter-weighted devices that open progressively as d/p across them exceeds 0.25 psi. In the event of a LOCA, the pressure-relief dampers limit the d/p to 3 psi and thus maintain 6.2-38 SGS-UFSAR Revision 6 February 15, 1987 the structural integrity of the fan coolers during the pressure transient. 6.2.2.2.3 Design Evaluation Range of Containment Protection The Containment Fan Cooling System provides the design heat removal capacity for the containment following a LOCA assuming that the core residual heat is released to the containment as steam. The system accomplishes this by continuously recirculating the air-steam mixture through cooling coils to transfer heat from containment to service water. The heat removal function of the containment fan coil units is similar to that of the containment spray system. As described in section 15.4. 8, "Containment Pressure Analysis", a minimum of three containment fan coil units in operation with a single containment spray train is capable of maintaining post-accident containment temperature and pressure below their design basis values, assuming a worst-case single active failure. Thus, design margin exists for the containment heat removal system. The performance of the Containment Fan Cooler System in pressure reduction is discussed in Section 15. System Response Automatic starting of the standby fan cooler units (under design conditions, up to four of the fans, and two service water pumps operate during normal power operations for containment ventilation) and the related emergency power equipment is designed so that the required air flow and cooling water flow for an accident condition is reached within the time delay for starting fan cooler units assumed in the containment pressure analyses. 6.2-39 SGS-UFSAR Revision 20 May 6, 2003 (This text has been The water valves and air dampers are actuated to the accident position by closure of the fan cooler low speed breaker. Single Failure Analysis A failure analysis for a:j.l active components of the system shows that the failure of any single active component will not prevent fulfilling the design function. This analysis is summarized in Table 6.2-7. The analysis of the LOCA presented in Section 15 is consistent with the single failure analysis. Reliance on Interconnected Systems The Containment Fan Cooling System is dependent on the operation of the SWS. Cooling water to the coils is supplied from the SWS. Six service water pumps are provided, only two of which are required to operate during the post-accident period. Shared Function Evaluation Table 6.2-8 is an evaluation of the main components which have been discussed previously and a brief description on how each 6.2-40 SGS-OFSAR Revision 18 April 26, 2000 * *
  • component functions during normal operation and during the accident. Reliability Evaluation of the Fan Cooler Motor The design of the motor and motor heat exchanger is such that the accident environment is prevented, in a significant sense, from entering the motor winding. When entering in the very limited amount required to equalize motor interior pressure, the incoming atmosphere is directed to the heat exchanger coils where moisture is condensed. If some quantity of moisture should pass through the coil, the motor interior environment would 11clean up" since interior air continually recirculates through the heat exchanger. The motor insulation hot spot temperature is not expected to exceed Il3°C even under accident conditions; normal life would be expected with a continuous hot spot of 155°C. The insulation has resistance to moisture, and tests indicate that the insulation system would survive the accident ambient moisture condition without failure. The heat exchanger system of preventing moisture from reaching the winding keeps the winding in much more favorable conditions. In addition, the motors are furnished with an insulation voltage margin beyond the operating voltage of 480 V. To prove the effectiveness of the heat exchanger in inhibiting large quanti ties of the steam-air mixture from impinging on the winding and bearings, a full-scale motor of the same type was subjected to prolonged exposure to accident conditions. The test exposed the motor to a steam-air mixture as well as boric acid and alkaline spray at 80 psig and saturated temperature conditions. Insulation resistance, winding and bearing temperature, relative humidity, voltage and current, as well as heat exchanger water temperature and flow were recorded periodically during the test. Following the test, the motor was disassembled and inspected to further assure that the unit performed as designed. The post-testing inspection showed no degradation of the motor components (1). The fan motor bearings are designed to perform in 6.2-41 SGS-UFSAR Revision 6 February 15, 1987 the accident ambient temperature conditions. However, the interior bearing housings are cooled by the heat exchanger. It is expected that bearing temperatures would be 125°C to 140°C, under accident conditions. The heat exchanger is designed using a conservative 0.002 fouling factor. Throughout the lifetime of the plant, these motors perform the normal heat removal service and are loaded to approximately 275 hp. Environmental Protection All of the fan cooler units are located on the operating floor adjacent to the containment wall. The distribution header is located below the operating floor, between the polar crane wall and the containment wall. This arrangement provides missile protection for all components. System control and instrumentation devices required for post-accident operation are also installed in locations such as to minimize the danger of control loss due to missile damage. The fan motor enclosures, electrical insulation, and bearings are designed for operation during accident conditions. Surfaces in contact with the containment atmosphere are protected against corrosion. 6.2.2.2.4 Tests and Inspection Component and System Testing Each fan cooling unit was tested after installation for proper flow through the Duct Distribution System. The Containment Fan Cooling System is designed such that the components can be tested periodically, and after any component maintenance, for operability and functional performance. 6.2-42 SGS-UFSAR Revision 6 February 15, 1987 Four of the fan cooling units are in use during normal operation. The fan not in use can be started from the control room to verify readiness. The dampers directing flow through the post-accident filter section can be tested when the fan is running on low speed. The functional teat of the ECCS described in section 6.3 will demonstrate proper transfer of the fan units in the event of a loss-of-power. A test signal is used to initiate damper motion and fan starting. This test will verify proper functioning of the air flow switch provided for each fan. Inspection Access is available for visual inspection of the containment fan cooler components including fans, cooling coils, dampers, and ductwork. 6.2.3 Containment Atmosphere Iodine Removal 6.2.3.1 Introduction The containment Spray System is an Engineered Safety System employed to reduce pressure and temperature in the containment following a postulated LOCA. For this purpose, subcooled water is sprayed into the containment atmosphere through a large number of nozzles from spray headers located in the containment dome. Because of the large surface area between the spray solution and the containment atmosphere, the containment Spray System also serves as a removal mechanism for fission products postulated to be dispersed in the containment atmosphere. Radioiodine in its various forms is the fission product of primary concern in the evaluation of a LOCA. The major benefit of the containment spray is its capacity to absorb molecular iodine from the containment atmosphere. To enhance this iodine absorption capacity of the spray, the spray solution is adjusted to an alkaline pH which promotes iodine hydrolysis to nonvolatile forms. 6.2-43 SGS-UFSAR Revision 6 February 15, 1987 According to the known behavior of elemental iodine in highly dilute solutions, the hydrolysis reaction: t2 + OH HIO + I proceed& nearly to completion (2) at pH > 8. The iodide form is highly soluble, and BIO readily undergoes additional reactions to form iodate. The overall reaction is: Values for the spray removal half-life of the solecular iodine in a typical containment are on the order of minutes, or less. This makes the COntainment spray system a very efficient fission product removal system, in comparison to such alternatives as charcoal filtration systems. For the small break loss-of-coolant accidents for which containment spray is not automatically initiated, offsite dose analysis was performed by Westinghouse using the methodology suggested in Reference 21. The results of this analysis are presented in Reference 22. It has bean demonstrated that the consequences of a small break LOCA without containment spray actuation are bounded by those of the large break LOCA. 6.2.3.2 Iodine Removal Model Containment spray performance has been determined using the spray model developed by Westinghouse. This model includes the effects of spray drop size distribution, droplet coalescence, and liquid phase mass transfer resistance. Its use results in conservative values of spray iodine removal constants when compared with test results. Method of Calculation In order to eliminate the need of scale-up factors from experimental results to full-sized reactor containments, the size-dependent calculations in this model were prograamed for discrete size parameters, i.e., the calculations are repeated for incremental height steps, and for 40 different drop-size groups to represent the effects of the drop-size distribution. No 6.2-44 SGS-UFSAR Revision 16 January 31, 1998 significant number of effect on results was observed by increasing the with discrete groups. The resulting model size-dependent parameters has been programmed for a digital computer. In the computer code, the sprayed volume of the containment is divided into layers of incremental height and area equal to the total sprayed area at any height z. The height-dependent calculations, such as drop trajectories and the change in the drop size distribution due to coalescence, are performed for each height step, using the parameters calculated in the previous step as input for the next step. Drop-Size Distribution The drop-size distribution used in the model is based on data obtained from measurements of the actual size distribution from the Spraco 1713 nozzle for the range of pressure drops encountered during operation of the Spray System. The results obtained for 20, 30, 40, and 50 psi pressure drops across the nozzle have been used in this evaluation. Analysis of these drop-size measurements shows that the drop-size distribution from this nozzle may be represented by a continuous distribution function, which is used as the input to the computer code. Condensation As the spray solution enters the high temperature containment atmosphere, steam will condense on the spray drops. The amount of condensation is easily calculated by a mass balance on the drop: mh + m h = m'h c g f SGS-UFSAR 6.2-45 Revision 6 February 15, 1987 where: m and m' m c h the mass of the drop before and after lb the mass of condensate, lb the initial enthalpy of the drop, Btu/lb saturation enthalpy of water vapor and liquid, Btu/lb The increase in each drop diameter in the therefore, is by: where: v hg d d' the the the the the the specific volume of liquid at saturation, ft3 /lb specific volume of the drop before condensation, ft3/lb latent heat of evaporation, Btu/lb enthalpy of steam at Btu/lb drop em before condensation drop em after condensation This increase in drop size due to condensation is expected to be complete in a few feet of fall for the majority of drop sizes in the distribution. More detailed calculations by Parsley (3) show that even for the largest drops in the distribution thermal equilibrium is reached in less than half of the available drop fall height. The change in the drop-size distribution due to condensation was conservatively modeled by a 6.2-46 SGS-UFSAR increase to the Revision 25 October 26, 2010 equilibrium size immediately after the drops emerge from the nozzle. Drop Trajectories A description of the actual drop trajectories is required to obtain accurate drop residence times, and to obtain the trajectory angle required for the coalescence calculations described below. These trajectories are obtained by integrating the equations of motion for each drop size. The equations of motion were integrated numerically, with the drag coefficient being determined iteratively from Reynolds number and terminal velocity. These calculations yield the following results: 1. Spread and Nozzle Interference Trajectory results for a range of drop sizes show that the horizontal velocities of the drops are quickly attenuated. For the smaller drop sizes ( (4001-f), the trajectory essentially is a straight fall. Even for 10001-f drops, the horizontal velocity component diminished to less than 10 percent of the total velocity in less than 10 feet. The effect of temperature and pressure on drop trajectories has also been calculated. The resulting spray envelope is a smaller diameter at higher temperatures and pressure. 2. Drop Residence Time SGS-UFSAR For downward-directed spray nozzles, the initial vertical velocity is higher than the terminal velocity, resulting in a slightly shorter residence time than that calculated with the assumption of terminal velocity. An accurate account of the residence time is obtained from 6.2-47 Revision 6 February 15, 1987 consideration of the actual trajectories followed by the drop. Correction factors are calculated for each drop size in the spectrum, so that the drop fall-times used for the iodine removal calculations are the actual drop residence times. A measure of conservatism is added to the drop residence calculations by the use of the drop diameters after condensation. Actually, the drop velocities would have been attenuated to a fraction of the initial nozzle velocity by the time condensation is complete. Drop Coalescence This effect will tend to decrease the overall surface-to-volume ratio of the spray, thereby affecting the fission product removal capability of the system. Concern has been centered particularly on the effect of coalescence on scale-up factors applied to data obtained from small-scale experiments. The effects of this phenomenon are accounted for by a mathematical model which is dependent of the containment size. The mathematical model used to account for drop coalescence due to the effects of overlapping spray patterns and due to larger drops overtaking smaller ones shows the number of coalescences to be functions of the collision and coalescence efficiencies, as well as the trajectory angle, drop velocities, drop size, and drop density. The coalescence efficiency is the probability that a collision will result in the formation of a single larger drop. The collision efficiency describes the probability that two drops on a geometric collision course, (i.e., their centers of motion are separated by a distance less than the sum of the radii of the two drops), will actually collide. The results calculated with the drop coalescence model show that the smaller drops with diameters near the mode of the distribution 6.2-48 SGS-UFSAR Revision 6 February 15, 1987 are affected most. This is expected, since these sizes have the highest density of drop population. Due to the considerably larger volumes of the larger diameter drops, however, the increase in the larger drop population is not very pronounced. Mass Transfer The basic equation for the iodine concentration in the containment atmosphere is derived from a material balance of the elemental iodine in the containment. The iodine removal by the spray system may be expressed by: where: v c c g H F = containment free volume in cc = the iodine concentration in the containment atmosphere, gmfcc = the iodine partition coefficient, (gm/liter of liquid)/ (gm/liter of gas)
  • the spray flow rate, ccfsec The resulting change in the drop size distribution is taken into consideration in the mass transfer calculations described below. The variable E is the absorption efficiency, which may also be described as the fractional approach to saturation: E= SGS-UFSAR 6.2-49 Revision 6 February 15, 1987 where: eLl = the iodine concentration in the liquid entering the dispersed phase, gm/cc cL2 = the iodine concentration in the liquid leaving the dispersed phase, qmfcc C* = the equilibrium concentration in the liquid, gmfcc L This absorption efficiency may be calculated from the time-dependent diffusion equation for a rigid sphere, with the gas film mass transfer resistance as a boundary condition. This mass transfer model was suggested by L. F. Parsley {4), who gave the solution to the diffusion equation with the above-mentioned boundary condition as: where: Sh a k 9 DL at SGS-UFSAR 00 E=l E n""' 1 + Sh (Sh-1)] is the dimensionless group = the drop radius, em = the gas film mass transfer coefficient, em{ sec == the liquid diffusivity, 2 em jsec = the dimensionless drop residence time 6.2-50 Revision 15 June 12, 1996

... n = the eigenvalues of the solution It ia noted that this solution, which applies to the rigid drop modal ia baaed on the assumption that molecular diffusion is the only mechanism for the transport of iodine from the surface to the interior of the drop. Since a high degree of mixing is expected in the drops, particularly in the presence of sizable temperature and concentration gradients, it ia apparent that thia stagnant drop model presents a conservative approach to the calculation of iodine absorption by the drops. The absorption efficiency calculated with the model described above is a function of drop size. The removal constant, 1s in reciprocal hours, for the entire spray, therefore, is obtained by an appropriate summation over all drop size groups: n E i = 1 6.2.3.3 Experimental verification of the Iodine Removal Model To demonstrate that the ability of the model described above conservatively estimates actual spray performance, the Westinghouse model was applied to the teat runs made at Oak Ridge National Laboratories (ORNL) and Battelle Northwest Laboratories (BNWL). comparison of the results of these tests with the above described spray removal model show the spray removal model to be conservative in all cases. 6.2-51 SGS-UFSAR Revision 6 February 15, 1987 6.2.3.4 6.2.3.4.1 Injection Phase Operation 'l.'he analysis of iodine removal by containment spray water is based on the assumption that: l. One of two spray pumps :i.s operating. 2. One train of ECCS is operating at its maximum 'I'he duration of injection spray is 48 miriutes followed by recirculation spray. An eductor .system, described in Section 6. 2. 2 .1, is used to maintain the injection spray solution at a pH in the range of 8.5 to 10.0 to ensure efficient and rapid removal of the iodine from the containment 'l'he of the Spray System was conrH1rvati vely evaluated at the peak temperature and pressure resulting from a double-(*mded rupture of the HCS, with no credit taken for the subcooJing of the ECCS. These pressure and temperature conditions, lis1:ed in 'fable 6. 2-9, were assumed throughout the i.n:\ection and recirculation of operation of the Containment Spray System. The injection and recirculation phase spray flow rates per pump, used in the calculation of A., corresponding to this back-pressure in the containment are given in Table 6.2-9. Since this peak pressure condition is expected to exist at most for a few minutes, and since both mass transfer parameters and 6.2-52 SGS-UFSAR Revision 23 October 17, 2007

  • I* *
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  • spray flow rate improve with decreasing pressure, an appreciable conservatism is added to this evaluation by this assumption . The removal constants for the spray system, in the injection and recirculation phases, calculated with the model described and with the above mentioned assumptions, is shown in Table 6.2-9. 6.2.3.4.2 Recirculation Phase Under the assumptions stated in Section 6. 2. 3. 4 .1, the spray recirculation phase is analyzed to be initiated 58 minutes after the start of safety injection. Safety injection is assumed to last 48 minutes followed by an assumed conservative gap of 10 minutes without spray before recirculation spray is started. At this time, sump water would have reached its minimum equilibrium pH of at least 7. 0. The iodine removal capability remains high under these conditions and credit is taken for iodine removal by sprays during the recirculation phase as shown in Table 6.2-9. During the spray recirculation phase, the sump pH will remain at equilibrium pH since no additional water is added to the system. For those small-size primary breaks for which containment spray is not automatically actuated, sump solution pH will be adjusted to a minimum value of 7.0 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of switchover to cold leg recirculation mode. However, no credit it taken for retention of iodine in solution in the offsite dose analysis summarized in Reference 22. 6.2.3.4.3 Re-Evolution of Iodine Any re-evolution of dissolved iodine from the sump to the containment atmosphere is dependent on the concentration gradient between the liquid and vapor phases. The equilibrium between these concentrations is given by the partition coefficient, H, and, therefore, is a function of iodine concentration, pH, and temperature. A plot of the sump alkalinity, as a function of the time after the start of injection, is shown on Figure 6.2-14. The resulting partition coefficient, based on a constant iodine concentration equal to the concentration corresponding to a DF of 100 in the containment atmosphere, is shown on Figure 6.2-15 for sump temperatures of 150°F and 212°F. The equations given by Eggleton (5) were used to calculate the partition coefficient . Although the iodate reaction, i.e.: SGS-UFSAR 6.2-53 Revision 23 October 17, 2007 I is expected to contribute siqnificantly (5) to the iodine partition at the hiqh sump pH values, this reaction is conservatively neglected in these calculations. 3 From Figure 6.2-15 it is apparent that the partition coefficient of 4.3 x 10 , which is required to maintain a DF of 100 in the vapor phase, is exceeded at all times during the recirculation phase. 6.2.4 Containment Isolation System The Containment Isolation System provides the means of isolating the containment atmosphere and RCS as required to prevent the release of radioactivity to the outside environment in the event of a LOCA. 6.2.4.1 Design Bases The following conditions and definitions are used in the design of the Containment Isolation System to assure that subsequent to an accident, there will be two barriers between the atmosphere outside the containment and the containment atmosphere. 1. The design parameters of all piping and connected equipment within the isolated boundaries are equal to or greater than the DBA environment of the containment, 47 psiq, 271°F. 2. All valves and equipment which are isolation barriers are protected against missiles and water jets, both inside and outside the containment. 3. Lines which, due to safety considerations, must remain in service subsequent to certain accidents have, as a 6.2-54 SGS-OFSAR Revision 6 February 15, 1987 minimum, one manual isolation valve outside the containment. 4. All isolation valves and equipment are designed to Class I seismic criteria. 5. Per acceptance methods of General Design Criteria 55 and 56 and ANS l H27l-1976/ANS 56.2 the two barriers may consist of: {a) two closed piping systems or vessels, one inside and one outside the containment, (b) two automatic isolation valves, one inside and one outside the containment, (c) an automatic isolation valva inside the containment and a closed system outside the containment, (d) an automatic isolation valve outside the containment and a closed system inside the containment, or (a) an automatic isolation valve outside containment and a closed system outside the containment. 6. A check valve on an incoming line or a normally closed valve is considered an automatic valve. 6.2.4.2 system Description The following four classes of piping arrangement are provided in the containment Isolation System. These classes are illustrated on Fiqure 6.2-16. Class A Class A piping is connected to a normally closed system outside the containment, and is separated from the RCS and the containment atmosphere by a closed system inside the containment. For Class A piping, no additional valves are required for isolation. 6.2-55 SGS-UFSAR Revision 16 January 31, 1998 Class B Class B piping is connected to open systems outside the containment, and is connected to the RCS or is open to the containment atmosphere. Por Class B piping, the following is provided, as a minimum, for isolation: 1. Incominq Lines: Two auto-trip valves (one inside, one outside), or a check valve inside and an auto-trip valve outside. 2. OUtgoing Lines: Two auto-trip valves (one inside, one outside). Class c Class c piping is connected to open systems outside the containment, and is separated from the RCS and the containment atmosphere by a closed system. For Class C piping, the following is provided, as a minimum, for isolation: 1. Incoming Lines: One check valve or auto-trip valve outside. No valve inside. 2. OUtgoing Lines: one auto-trip valve outside. No valve inside. Claaa D Class D piping is connected to a closed system outside the containment, and is connected to the Res or is open to the containment atmosphere. 6.2-56 SGS-UFSAR Revision 6 February 15, 1987
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  • For D piping, the following is provided, as .a minimum, for isolation: 1. '2. Incoming Lines: One auto-trip valve or check valve inside. No valve outside. Outgoing Lines: One auto-trip valve inside and no valve outside. Alternately, one auto-trip valve outside and no valve inside. In addition to Classes B and C, for lines l-inch nominal pipe size and larger which penetrate the containment and which are connected to the RCS, at least two valves are provided inside the containment. The valves are normally closed or have automatic closure. For incoming lines, check valves are permitted and are considered as automatic. Piping which penetrates the containment, but which represents normally closed lines, also under this criterion. In this case, manual isolation valves are acceptable. In order to be considered a "closed" system inside containment, a. system must meet the following requirementsi 1. Does not communicate with either the RCS or the containment atmosphere . 2. Safety classification same as for engineered safety systems. 3. Must withstand external pressure and temperature equal to containment design pressure and temperature. 4. Must withstand accident transient and environment. 5. Must be missile protected. In order to be considered a "closed" system outside containment, a system must meet the following requirements: 6.2-57 SGS-cUFSAR Revision 21 December 6, 2004 I
1. Does not communicate with the atmosphere outside the containment. 2. Safety classification same as for engineered safety systems. 3. Internal design pressure and temperature must be at least equal to containment design pressure and temperature. For incoming lines to the containment, check valves are used whenever an additional barrier is provided. Use of check valves in this service is confined to either liquid lines or lines that are closed outside the containment. These check valves shut under a d/p when the higher pressure is on the containment side of the check valve. These isolation valving arrangements were designed in accordance with Atomic Energy Commission (AEC) proposed General Design Criteria published in 1967, which were in effect at the construction Permit stage. The valving arrangements that deviate from AEC General Design Criteria ss, 56, and 57 dated July 7, 1971, are the following: 1. RHR connections between the RCS and the RHR pumps. Redundant isolation protection is provided by a normally closed motor operated valve inside the containment and the closed system (RHR) outside the containment. 2. Seal water supply line from the seal water injection filters to the reactor coolant pump seals. Redundant isolation protection is provided by a check valve inside the containment and the closed system (CVCS) outside the containment. 6.2-58 SGS-UFSAR Revision 16 January 31, 1998 -............
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  • 3. 4. Safety injection recirculating suction line from the containment sump to the suction of the RHR pumps. Redundant isolation protection is provided by normally closed motor operated valves inside protective chambers outside of containment and the closed system (RHRl outside the containment. Containment instrument lines {see below). 5. The main feedwater lines are provided with one stopcheck valve (BF22) outside containment. These valves include remote-manual motor operators. 6. RHR pump discharge to cold leg Safety Injection. Redundant isolation is provided by the remote manual (SJ49) valves located outside containment and the RHR closed system outside containment. This is considered an acceptable isolation barrier per the nether defined basis" in ANSI N271-1976. This standard is endorsed by Regulatory Guide 1.141. 7. ECCS relief line discharge to the containment sump. Redundant isolation is provided by a check valve inside containment (PR25) and the closed system outside the containment. 8. Service Water system to and from the Containment Fan Coil Units. Redundant isolation is provided by remote manual valves outside containment and th.e closed Nuclear Class III system inside containment. Original system design complied with AEC General Design Criteria #53 and the system meets the definition for a Safety Class 2 system. 9. Component Cooling to and from the Excess Letdown Heat Exchanger. Redundant isolation is provided by automatic isolation valves outside containment and the closed Nuclear Class III system inside containment. Original system design complied with AEC General Design Criteria #53 and the system meets the definition for a Safety Class 2 system. 10. Main Steam supply lines to the Auxiliary Feed Pump Turbine, Radiation Monitors and the Steam Safety-valves support struts. These essential system branch lines off the Main Steam penetrations only utilize a single isolation barrier being the closed system inside containment;.._ The calculated release through these paths is already bounded by the accident analysis for a primary to secondary leak and a complete blowdown of the Steam Generator. 6.2-59 SGS-UFSAR Revision 18 April 26, 2000 I Instrument Lines Instrument lines which penetrate the containment are the following: 1. The containment pressure instrument used to initiate safeguards consists of four instrument lines penetrating the containment. Each line consists of a sealed, filled measuring system whose isolation consists of a diaphragm-type sensor which separates the containment atmosphere from the seal fluid and another diap?ragm in the transmitter which separates the seal from the atmosphere outside the containment. 2. The containment air sample radiation monitor normal inlet and outlet sample lines are each equipped with two automatic trip valves, one inside and one outside the containment, which close upon receipt of a containment isolation phase A signal. The backup inlet and outlet sample lines are normally closed and under administrative control with two remote operated isolation valves, one inside and one outside the containment for each line. 3. The containment pressure instrument used for wide range monitoring consists of two instrument lines penetrating the containment. Each line consists of a sealed, filled measuring system whose isolation consists of a diaphragm-type sensor, which separates the containment atmosphere from the seal fluid and another diaphragm in the transmitter, which separates the seal from the atmosphere outside containment. 6.2-59a SGS-UFSAR Revision 18 April 26, 2000 * * *
3. The pressurizer dead-weight pressure calibrator has a single line penetrating the containment. Isolation is accomplished with two manual valves located just outside the containment. These manual-valves are normally closed and are opened only under administratively controlled conditions. 4. Three lines penetrate the containment for instrumentation required for leak rate testing. Each line is isolated with two manual valves, one inside and one outside containment. These valves are normally closed and under administrative control. These provisions meet the requirements of Regulatory Guide 1.11. Containment Isolation Valve Summary Table 6.2-10 lists the major piping penetrations through the reactor containment for each fluid system and summarizes the specific isolation provisions for each penetration. Valve positions during normal operation, shutdown, and accident conditions are also listed. Isolation valving arrangements are shown graphically on Figures 6.2-17 through 6.2-46. The main steam isolation valves (MSIVs) fulfill their containment isolation function as remote-manual containment isolation valves. The automatic closure of the MSIVs is not required for containment isolation due to having a closed system inside containment. The remote-manual containment isolation function of the MSIVs can be accomplished through either the use of the hydraulic operator or when the MSIV has been tested in accordance with Technical Specification 4.7.1.5, the steam assist closure function can be credited. 6.2-59b SGS-UFSAR Revision 20 May 6, 2003 Valve closing time using the hydraulic actuator is approximately six minutes. The closure time for establishing containment isolation is that:. necessary to significantly limit the release of radioactivity to the environment. MSIV fast closure is not required for containment isolation in any operating Mode because the steam generator shell and main steam piping serve as the primary barrier for a LOCA. For the LOCA, the design basis does not assume a concurrent feedwater or steam line break. The main steam system does not directly connect to the reactor coolant system or the containment atmosphere. However, a steam generator tube break or rupture makes a connection between the RCS and the secondary side systems via the main steam system. The Chapter 15 SGTR accident analysis assumes a coincidental loss of offsite power that causes the steam dump valves to close, protecting the condensers. For the Mode 1 or 2 SGTR Chapter 15 accident analysis, isolation of the. faulted steam generator is assumed to occur within 30 minutes as necessary to limit the release of radioactivity to the environment via the steam generator PORV or safety relief valves. Isolation of the faulted steam generator also limits the spread of radioactivity to the interconnected steam generators, at least one of which will be used to cooldown the RCS until RHR can be initiated at 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> post-accident. During low temperature (<375°F) Mode 3 and Mode 4 operations, there is insufficient enerqy transferred to the secondary side in a SGTR or steam generator tube leak to result in lifting the steam generator I?ORVs or safety relief valves and there will be no release of radioactivity to the environment. Use of the remote-manual, hydraulic actuator for containment isolation in low temperature (<375°F) Mode 3 and Mode 4 is satisfactory because even if isolation of the faulted steam generator fails, the failure will not increase the dose consequences beyond the existing Chapter 15 SGTR accident analysis that remains bounding. The 20-inch inside diameter fuel transfer tube between th.e refueling canal inside the containment and the fuel transfer pool is sealed with a blind flange inside the containment, redundant isolation is provided by a double o-ring seal on the flange. The terminus of the tube outside the containment is closed by a gate valve which is not a containment isolation valve. The equipment hatch (door) is under administrative control to assure that it is properly closed and sealed whenever containment integrity is required. No instrumentation is provided for the equipment hatch. 6.2-60 SGS-UFSAR Revision 19 November 19, 2001 -, ( ' -f t '

Actuation Provisions Containment isolation i* actuated under the following conditione: 1. A safety injection signal generates the containment isolation signal (Phase A), which actuates most containment isolation valves. The Phase A isolation signal closes all trip valves which are located in lines which are connected to the reactor coolant loops and penetrate the containment, thereby preventing loss of reactor coolant through the lines in which the automatic trip valves are located. Normally closed motor operated containment isolation valves in the SIS are opened by the safety injection signal to permit SIS operation. 2. A rise in containment pressure to the high containment pressure set point also generates the Phase A isolation signal. 3. A further rise in containment pressure, indicating a major LOCA, results in a containment high-high pressure signal which generates both the containment spray and containment isolation Phase B signal. All normally open lines which penetrate the containment which are not closed by the Phase A isolation signal are closed by the Phase B isolation signal. Normally closed motor-operated Containment spray System valves are opened by the high-high containment pressure signal to permit Containment Spray System operation. 4. The CV68 and CV69 vlaves do not receive containment isolation signals (Phase A or Phase B). These valves get a close signal on a safety Injection (SI) signal. Lines which penetrate the containment and are normally closed by means of valves under administrative control are assumed to be already closed and do not receive an isolation signal. Automatic containment isolation valves can be actuated from the control room if any of the valves fail to close in response to the Phase A or Phase B isolation signal. 6.2-61 SGS-UFSAR Revision 16 January 31, 1998 6.2.4.3 Design Evaluation The following provisions apply to all lines penetrating the containment to prevent inadvertent opening of these lines to the atmosphere outside the containment: 1. Automatic isolation valves can be opened only upon manual reset of the solid state logic without cessation of the actuating signal. 2. Automatic isolation valves are capable of manual actuation from the control room with the limitations for reopening of the valve noted in Item 1 (above). 3. Remote manual valves are operated only under administrative control. 4. Manual valves are operated under administrative control. 5. Check valves open only when the fluid pressure is higher on the side outside the containment. 6. The design pressure of all piping and connecting components within the isolation boundary is not less than the design pressure of the containment, 47 psig. 7. Automatic valves, once opened by a safety injection signal, can only be closed upon cessation and manual reset of the actuating signal. For Items 1, 2, 3, and 4 (above}, and for flanged closures, specific administrative procedures define the positioning of these closures in the Containment Isolation System during normal operation, shutdown, and accident conditions. Instrumentation and adjunct control circuits associated with air operated automatic isolation valve closures are fail safe upon loss of voltage and/or control air. Such valves fail closed on loss of voltage, except for the outside containment isolation valves for the control air system (11, 12, 21 and 22CA330). The CA330's fail closed on loss of air, but fail as-is on loss of vital DC power. The control air system isolation valves inside containment (11, 12, 21 and 22CA360 check valves) prevent any single active failure from resulting in loss of the containment isolation function. The air operated isolation 6.2-62 SGS-UFSAR Revision 20 May 6, 2003 valves are air to open, spring return, diaphragm operated; thus providing a fail safe design. The automatic isolation valves inside the containment will function properly under all accident conditions. The isolation valve closing force is provided by a spring; control air is applied to the diaphragm of the isolation valve to open it. To close the isolation valve, an electrically operated solenoid valve located in the air supply line to the isolation valve operator vents the control air applied to the isolation valve diaphragm through the solenoid to the containment atmosphere, causing the spring to close the automatic isolation valve. Since the spring side of the isolation valve diaphragm is also vented to the containment atmosphere, the spring will force the valve to close when the solenoid vents the air line. Circuits which control redundant automatic valves are redundant in the sense that no single failure \\Jill preclude isolation. Means are provided to periodically test the functioning of the automatic isolation equipment such as the set point of sensors, speed of response, and operability of fail safe features. The containment isolation instrumentation is discussed in Section 7. Valves used for containment isolation are capable of tight shutoff against gas leakage from containment design pressure down to zero psig. Isolation valves and equipment are protected from missiles and water jets originating from the RCS. Missile protection for isolation valves, actuators, and controls is provided by locating isolation valves between the polar crane wall and the containment wall or locating isolation valves outside the containment structure. The pressure sensing devices which detect high containment pressure are located outside the containment. Location of the pressure sensing devices outside the containment protects them from missiles developed by a LOCA. Isolation valves and piping or vessels which provide one of the isolation barriers outside the containment are similarly protected. 6.2-63 SGS-UFSAR Revision 6 February 15, 1987 The closure times for the containment isolation valves are such that, in the event of a LOCA, no release of to the environment through containment can occur. In evaluating possible radioactive releases during a LOCA, the only release pathways considered were through those normally open penetrations associated with open systems outside the containment which are connected to the RCS or are open to the containment atmosphere (see Table 6.2-10). A loss of offsite power was assumed coincident with a LOCA. The diesel-generators were assumed to be ready for loading in 13 seconds. The closure time for valves is 10 seconds. Therefore, the total closing time for these valves is 23 seconds. It is conservatively assumed that these valves remain completely open for the time required to activate and completely close them. The closure time for air operated valves is conservatively estimated to be 10 seconds. Operation of these valves is initiated when a containment pressure of 4. 0 is reached. A conservative estimate of the time ::o attain this pressure, assuming a double ended cold break, is 3 seconds. Therefore, the total time for these air valves is 13 seconds. One is the isolation valve for containment pressure vacuum relief. This valve has a closure time of 2 seconds, resulting in a total closing time of 5 seconds for this analysis. The activity released to the containment during the time required to close all isolation valves is limited to that contained in the RCS prior to the accident. This is based on the time to close the isolation valves SGS-UFSAR small that no clad closed. 6.2-64 would occur before the valves were Revision 25 October 26, 2010

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  • Total LOCA doses calculated in Section 15 which include the contribution of release through isolation valves and are within the limit values of 10CFR50.67. Hence, it is concluded that the containment isolation valve closure times are sufficiently short and that there is no undue risk to the health and safety of the public. 6.2.4.4 Preoperational Tests Preoperational tests were performed on all valves in lines which penetrate the reactor containment and perform a containment isolation function to verify operability and leaktightness. Valve operability testing was conducted prior to leakage testing. Each isolation valve was tested to demonstrate proper closure of normally open valves (or opening and closing of normally closed valves) upon receipt of an isolation signal. Closure of 6.2-65 SGS-UFSAR Revision 23 October 17, 2007 containment isolation valves was accomplished by normal operation and without any preliminary exercising or adjustment. Valve leakage testing was performed by local pressurization in accordance with the applicable requirements of lOCFRSO, Appendix J, "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors, .. for Type C testa. Val vee were pressurized with air or nitrogen to a pressure of 47 psig. Where practical, pressure was applied in the same direction as the valve would experience during performance of its aafety function. Valve leakage was determined by measurement of the rate of pressure loss or by the flowrate of makeup air or nitrogen required to maintain teat pressure. The containment integrated leakage rate test procedure identified vent and drain valves which were opened in order to ensure exposure of the syatem piping penetrating containment to the full containment test pressure differentiaL Certain linea in the Service Water (SWS), Component Cooling Water, and RHR systema are required for containment environmental control or decay heat removal and were not included in the integrated leakage rate test. The isolation valves in these lines were tested separately using a Type c test and any detected leakage was added to the Type A containment integrated leakage rate test results. The combined leakage rate for the isolation valves and the double penetrations was limited to less than 0.06 percent of the containment free volume per day. Periodic Teats Periodic operability and leakage testa on isolation valves will be conducted throughout the lifetime of the plant according to the schedule specified in lOCP'RSO Appendix J, "Primary Reactor containment Leakage Testing for Water Cooled Power Reactors." The 6.2-66 SGS-UPSAR Revision 16 January 31, 1998 -

isolation valve tests will be requirements for preoperational testing. to test the in accordance with the of each isolation The piping arrangement valve consists of a on the main line downstream of the valve. To test for valve tightness, the main piping section of the valve is pressurized and evidence of leakage is checked at the downstream tap. When not in use, the monitoring lines are valved closed at the open end. Test pressures will be applied from the same direction as the pressure existing when the valve is required to perform its safety function. The of the majority of containment isolation valves is fully testable at power except for those valves listed below. The valves are checked for circuit up to and the valve actuator power operation by use of the Solid-State Protection System (SSPS) output test cabinet. Containment isolation valves: CA-330, SJ-12, SJ-13, CV-7 CV-68, CV-69, CV-116, CV-284, CC-118, CC-131, CC-136, CC-187, and CC-190 Main steam isolation valves: MS-167 All other valves can be operationally tested at power from the SSPS output test cabinet to simulate accident operating conditions and verify the valve closing logic. All valves can be tested from the main control console as operating conditions permit. SGS-UFSAR 6.2-67 Revision 26 May 21, 2012 6.2.5 Combustible Gas Control 6.2.5.1 Hydrogen Production Hydrogen accumulation in the containment atmosphere following the DBA can be the result of production from several sources. The potential sources of hydrogen are the zirconium-water reaction, corrosion of construction materials, and of the emergency core solution. The latter source, solution both core solution sump solution radiolysis. 6.2.5.1.1 Methods of Analysis The quantity of zirconium which reacts with the core cooling solution depends on the performance of the ECCS. The criteria for evaluation of the ECCS that the reaction be limited to 1 percent by weight of the total quantity of zirconium in the core. Emergency Core Cooling calculations have shown the reaction to be less than 0.1 percent, much less than required by the criteria. The use of aluminum inside the containment is limited, and is not used in safety-related components which are in contact with the recirculating core cooling fluid. Aluminum is much more reactive with the containment spray alkaline borate solution than other materials such as galvanized steel, copper and copper nickel By limiting the use of aluminum the source of hydrogen over the long term is rest'ricted to that from radiolytic decomposition of core and sump water. The upper limit. rate of such decomposition can be predicted with ample certainty t9 permit the design of effective countermeasures. It should be noted that the zirconium-water reaction and aluminum corrosion with containment spray are chemical reactions and thus essentially independent of the radiation field inside the containment a LOCA. decomposition of water 6.2-68 SGS-UFSAR Revision 6 15, 1987 is dependent on the radiation field intensity. The radiation field inside the containment is calculated for the maximum credible accident in which the fission product activities given in TID-14844 (6) are used. Two hydrogen generation calculations are performed: one using the Westinghouse model (7), the other using the AEC model discussed in Safety Guide 7 (8). 6.2.5.1.2 Zirconium-Water Reaction The zirconium-water reaction is described by the chemical equation: The hydrogen generation due to this reaction will be completed during the first day following the LOCA. The Westinghouse model assumes a 2-percent zirconium-water reaction and the AEC model assumes a 5-percent zirconium-water reaction. The hydrogen generated is assumed to be released immediately to the containment atmosphere. 6.2.5.1.3 Corrosion of Plant Materials Oxidation of metals in aqueous solution results in the generation of hydrogen gas as one of the corrosion products. Extensive corrosion testing has been conducted to determine the behavior of various metals used in the containment in the emergency core cooling solution at DBA conditions. Metals tested include Zircaloy, Inconel, aluminum alloys, coppernickel alloys, carbon steel, galvanized carbon steel, and copper. Tests conducted at ORNL (9, 10) have also verified the compatibility of the various metals (exclusive of aluminum) with alkaline borate solution. As applied to the quantitative definition of hydrogen production rates, the results of the corrosion tests have shown that only 6.2-69 SGS-UFSAR Revision 6 February 15, 1987 aluminum will corrode at a rate that will significantly add to the hydrogen accumulation in the containment atmosphere. The corrosion of aluminum may be described by the overall reaction: Therefore, three moles of hydrogen are produced for every two moles of aluminum that is oxidized. (Approximately 20 standard cubic feet of hydrogen for each pound of aluminum corroded.) The time-temperature cycle (Table 6.2-14) considered in the calculation of aluminum corrosion is based on a conservative step-wise representation of the postulated post-accident containment transient. The corrosion rate design curve is shown on Figure 6.2-47. Aluminum corrosion data points include the effects of temperature, alloy, and spray solution conditions. Based on these corrosion rates and the aluminum inventory given in Table 6.2-15, the contribution of aluminum following the DBA has been calculated. For conservative estimation, no credit was taken for protective shielding effects of insulation or enclosures from the spray, and complete and continuous immersion was assumed. Calculations based on Safety Guide 7 are performed by allowing an increased corrosion rate during the final step of the post-accident containment temperature transient (Table 6.2-14) corresponding to 200 mils/yr (15.7 mg/dm2/hr). The corrosion rates earlier in the accident sequence are the higher rates determined from Figure 6.2-47. Hydrogen is also produced through the corrosion of zinc inside containment. Sources of zinc within containment are the following: 1. Cable trays and hangers 6.2.-70 SGS-UFSAR Revision 6 February 15, 1987

2. Conduit 3. Junction Boxes 4. Ductwork These components are galvanized with approximately 2 oz/ft (2 of zinc, and the surface area and weight of zinc associated with each is as follows: Item Sq. Ft. Weight of Zn or Zinc 1. Cable trays and hangers 35,000 4, 375 lb 2. Conduit 15,000 1,875 lb 3. Junction Boxes 1,500 188 lb 4. Ductwork 10,813 lb The corrosion rate of zinc as a function of temperature is shown on 48. 6.2-The experimental data used as the basis for hydrogen production due to zinc was obtained from Reference 11. 6.2.5.1.4 Radiolysis Water radiolysis is a complex process involving reactions of numerous intermediates. However, the overall radiolytic process may be described by the reaction: Of interest here is the quantitative definition of the rates and extent of radiolytic hydrogen production following the DBA. 6.2-71 SGS-UFSAR Revision 24 May 11, 2009 I An extensive program has investigate the radiolytic solution following the DBA. been conducted by decomposition of Westinghouse to the core cooling In the course of this investigation, it became apparent that two separate radiolytic environments exist in the containment at DBA conditions. as a In one case, radiolysis of result of the decay energy the core cooling solution occurs of fission products in the fuel. In the other case, the decay of dissolved fission products, which have escaped from the core, results in the radiolysis of the sump solution. The results of these investigations are discussed in Reference 12. Core Solution Radiolysis As the emergency core cooling solution flows through the core, it is subjected to gamma radiation by decay of fission products in the fuel. This energy deposition results in solution radiolysis and the production of molecular hydrogen and oxygen. The initial production rate of these species will depend on the rate of energy absorption and the specific radiolytic yields. The energy absorption rate in solution can be assessed from knowledge of the fission products contained in the core, and a detailed analysis of the dissipation of the decay energy between core materials and the solution. The results of Westinghouse studies show essentially all of the beta energy will be absorbed within the fuel and cladding and that this represents approximately 50 percent of the total beta-gamma decay energy. This study shows further that of the gamma energy, a maximum of 7. 4 percent will be absorbed by the solution in core. Thus, an overall absorption factor of 3.7 percent of the total core decay energy (13 + y) is used to compute solution radiation dose rates and the time-integrated dose. Table 6. 2-16 presents the total decay energy (13 + y) of a reactor core, which assumes a full power operation time of 830 days prior to the accident. For the maximum credible accident case, the contained decay energy in the core accounts for the assumed TID-14844 release of SO percent halogens and 1 percent other fission products. To be conservative, the 6.2-72 SGS-UFSAR Revision 6 February 15, 1987 noble gases have been assumed to remain in the core, whereas in reality, the noble gases are assumed by the TID-14844 model to escape to the containment vapor space where little or no water radiolysis would result from decay of these nuclides. The radiolysis yield of hydrogen in solution has been studied extensively by Westinghouse and ORNL. The results of static capsule tests conducted by Westinghouse indicate that hydrogen yields much lower than the maximum of 0. 44 molecule per 100 eV would be the case in-core. With little gas space to which the hydrogen formed in solution can escape, the rapid back reactions of molecular radiolytic products in solution to reform water is sufficient to result in very low net hydrogen yields. However, it is recognized that there are differences between the static capsule tests and the dynamic condition in-core, where the core cooling fluid is continuously flowing. Such flow is reasoned to disturb the steady-state conditions which are observed in static capsule tests, and while the occurrence of back reactions would still be significant, the overall net yield of hydrogen would be somewhat higher in the flowing system. The study of radiolysis in dynamic systems was initiated by Westinghouse, which formed the basis for experimental work performed at ORNL. Both studies clearly illustrate the reduced yields in hydrogen from core radio lysis, i.e., reduced from the maximum yield of 0.44 molecule per 100 eV. These results were recently published (12, 13). For the purpose of this analysis, the calculations of hydrogen yield from core radiolysis are performed with the very conservative value of 0.44 molecule per 100 eV. That this value is conservative and a maximum for this type of aqueous solution and gamma radiation is confirmed by many published works. The Westinghouse results from the dynamic studies show 0. 44 to be a 6.2-73 SGS-UFSAR Revision 6 February 15, 1987 maximum at very high solution flow rates through the gamma radiation field. 'l'he referenced OHNI, ( 13) work also confirms this value as a maximum at high flow rates. A. 0. Allen {14) a very comprehensive review of work performed to confirm the primary hydrogen yield to be a maximum of 0.44 to 0.45 molecule per 100 eV. On the foregoing basis, the production rate and total hydrogen produced from core radiolysis, as a function of time, has been the maximum credible accident case. estimated for Calculations based on Safety Guide 7 assume a hydrogen yield value of 0. 5 molecule per 100 eV and that 10 perccmt of the gamma cmergy produced from fission products in the fuel rods is absorbed by the solution in the region of the core. Sump Solution Radiolysis Another. potential source o:E hydrogen assumed for. the post-accident period arises from water contained in the reactor containment sump being subjectE)d to radiolyt:Lc decomposition by fission products. In this consideration, an assessment must be made as to the decay energy in the solution and the radiolytic hydrogen yield, much in the same manner as given above for core radiolys:Ls. 'l'he energy deposited in solution J.s computed using the following basis: l. For the maximum credible accident, a TlD-14844 release model (which is more conservative than the guidance of Regulatory Guide 1.183 for fission product release assumptions) is assumed where 50 percent of the total core halogens and 1 of all other fission excluding noble gases, are released from the core to the sump solution. 2. 'l'he quantity of fission product: release is equal to that from a reactor operating at full power for 830 days prior to the accjdent . 6.2-74 SGS-UFSAR Revision 23 October 17, 2007 * * *
3. The total decay energy from the released fission products, both beta and gamma, is assumed to be fully absorbed in the Within the assessment of energy release by fission products in water, account is made of the decay of halogens, and a separate accounting for the slower decay of the 1 percent other fission products. To arrive at the energy deposition rate and egrated energy deposited, the contribution from each individual fission product class was computed. The overall contributions from each of the two classes of fission products is shown in Table 6.2-17. The yield of hydrogen from sump solution radiolysis is more nearly represented by the static capsule tests performed by Westinghouse and ORNL with the alkaline sodium borate solution. The differences between these tests and the actual conditions for the solution, however are important and render the capsule tests conservative in t1 ir predictions of radiolytic hydrogen yields. In this assessment, the sump solution will have considerable depth, which inhibits the ready diffusion of hydrogen from solutions, as compared to the case with shallow-depth capsule tests. This retention of hydrogen in solution will have a significant effect in reducing the hydrogen yields to the containment atmosphere. The build-up of hydrogen concentration in solution will enhance the back reaction to formation of water and lower the net hydrogen yield, in the same manner as a reduction in gas to liquid volume ratio will reduce the yield. This is illustrated by the data presented on Figure 6.2-49 for capsule tests with various gas to liquid volume ratios. The data show a significant reduction in the apparent or net hydrogen yield from the published primary maximum yield of 0.44 molecule per 100 eV. Even at the very highest ratios, where capsule solution depths are very low, the yield is less than 0.30, with the highest scatter data point at 0.39 molecule per 100 eV. 6.2-75 SGS-UFSAR Revision 6 February 15, 1987 With these considerations taken into account, a reduced hydrogen yield is a reasonable assumption to make for the case of sump radiolysis. Yhile it can be expected that the yield will be on the order of 0.1 or less, a conservative value of 0. 30 molecule per 100 eV has been used in the maximum credible accident case. Calculations based on Safety Guide 7 do not take credit for a reduced hydrogen yield in the case of sump radiolysis and a hydrogen yield value of 0.5 molecule per 100 eV has been used. 6.2.5.1.5 Coatings Keeler & Long's group of Nuclear Level One Qualified Coatings, which consist exclusively of epoxy products, have been used on carbon steel components in the nuclear reactor containment. These coating products afford resistance of the steel substrate to corrosion caused by accidental spillage, environmental agents and DBA conditions of temperature, moisture and chemistry. Tests conducted by the Franklin Institute Research Laboratories (15) cover the examination of the coating syste*s used in the reactor containment. These examinations were for chalking, flaking, peeling, craking, checking and rusting. Tests conducted by Keller & Long and ORNL (16), in compliance with ANSI NlOl. 2-1972 and ANSI NS12-1974 also demonstrate that the coating systeu are virtually unaffected by exposure to DBA test conditions. The question of hydrogen gas generation from coating systems exposed to DBA conditions is of concern only in cases where zinc-based coatings materials have been used on carbon steel components inside the reactor containment. Epoxy coatings, such as those used on carbon steel componenta inside the Salem containments. are not considered to be prone to hydrogen gas generation. Revision 13 June 12, 1994 -

Non-ferrous surfaces have been primed with a thin coat (0.5 mil dry thickness) of Con-Lux vinyl wash primer and catalyst 286.3 followed by a coat of Phenoline 305 finish. This primer con.tains 3. 08% zinc by weight. When spread at a rate less than one mil wet thickness, the amount of zinc contained in this material is negligible inside the containment where the area of non-ferrous surfaces is approximately 86,500 sq. ft. 6.2-76a SGS-UFSAR Revision 13 June 12, 1994 SGS-UFSAR THIS PAGE INTENTIONALLY LEFT BLANK Revision 13 June 12, 1994

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  • 6.2.5.1.6 Chemical and Volume Control System The source of hydrogen from the Chemical and Volume Control System is automatically cut-off upon receipt of a safety injection signal. 6.2.5.1.7 Results The results of the calculations for hydrogen production and accumulation from zirconium-water reactions, aluminum corrosion and radiolytic decomposition of core and sump solution are shown on Figures 6.2-50, 6.2-51, 6.2-52, and 6.2-53. Figures 6. 2-50 and 6. 2-51 show the hydrogen production rate as a function of time following a LOCA up to 100 days for the maximum credible accident. Similar information for the first 10 days is shown on Figure 6.2-54. Figures 6. 2-52 and 6. 2-53 show the total quantity of hydrogen accumulated in the containment as a function of time for the maximum credible accident case up to 100 days. The contribution of the individual source is also shown (note that zinc corrosion is not included) . Figure 6.2-55 shows the hydrogen production rate from aluminum and zinc corrosion for the first 10 days following a LOCA. Total hydrogen accumulated from all sources inside containment was reanalyzed following the Three Mile Island (TMI) accident to show compliance with 10CFR50. 44. The requirements for a hydrogen control system to mitigate a hydrogen release were eliminated when 10CFR50.44 was revised and it no longer defined a design-basis LOCA hydrogen release. 6.2.5.2 Hydrogen Control This section has been deleted based on Technical Specification Amendment numbers 281 and 264 to Facility Operating License numbers DPR-70 and DPR-75 . 6.2-77 SGS-UFSAR Revision 23 October 17, 2007 I 6.2.5.2.1 Hydrogen Recombiner Description This section has been numbers 281 and 264 to deleted based 6.2.5.2.2 Recombiner Test Program on Technical Amendment License numbers DPR-70 and DPR-75. This section has been deleted based on Technical Specification Amendment numbers 281 and 264 to Facility Operating License numbers DPR-70 and DPR-75. 6.2.5.2.3 Recombiner Inservice This section has been deleted based on Technical Specification Amendment numbers 281 and 264 to Facility Operating License numbers DPR-70 and DPR-75. 6.2.5.2.4 Hydrogen Purge There is no controlled purge in the Salem other than the three different and purge modes described in Section 9.4. There is, however, an inherent "backup" in the multiple exhaust fans and filters that are available in the Purge System. Purge System valve actuation periodic surveillance requirements are included in the Technical Specifications. 6.2.5.3 A Monitoring is provided for continuous measurement of hydrogen concentration at two locations within containment. Data from locations allows for diagnosing basis accidents. The system is designed in accordance with NUREG-0737 and Regulatory Guide 1.97. The analyzing unit is mounted inside containment such that only electrical penetrations are required. Equipmen:: located inside containment is operable under post-accident conditions of pressure, temperature, and radiation. All system components are seismically designed. 6.2-78 SGS-UFSAR Revision 25 October 26, 2010 Hydrogen concentration is measured by a hydrogen partial pressure sensor in conjunction with a total pressure sensor. The partial pressure sensor is galvanic in nature, consisting of a platinum black electrode and a platinum oxide counter electrode within a polysulfone housing. The range of measurement is 0 to 10 volume percent with an accuracy of 2 percent of full scale. Output is displayed in one Control Room. Alarms are provided for high hydrogen concentration, power failure, system error, and calibration mode. Power is supplied from vital sources. In addition to the Hydrogen Monitoring System, hydrogen concentration may be determined by taking a grab sample using the containment air particulate detector (APD) skid. In amendments 281 and 264 to Salem Units 1 and 2 Operating Licenses, the following commitments were made: In accordance with the commitments made pursuant to amendments 281 and 264 to Salem Units 1 and 2 Operating Licenses, the following actions are required: a) In Modes 1 and 2, two independent containment hydrogen analyzers shall be Functional; a grab sample from the containment air particulate detector (APD) skid is utilized as an alternate means during periods of non-functionality. b) With one hydrogen analyzer non-functional, restore the non-functional analyzer to Functional status within 30 days and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of its being determined to be non-functional, verify the capability to determine hydrogen concentration by taking a grab sample using the containment air particulate detector (APD) skid. If the analyzer is not restored to functionality within 30 days, then within the subsequent one (1) week develop an Action Plan, approved by the Operations Director, to restore the analyzer to Functionality. c) With both hydrogen analyzers non-functional, restore at least one analyzer to Functional status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of their being determined to be non-functional, determine hydrogen concentration by taking containment air particulate detector (APD) verify the capability to a grab sample using the skid. If at least one analyzer is not restored to functionality within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then within the subsequent 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> develop an Action Plan, approved by the Operations Director, to restore the analyzers to Functionality. 6.2-79 SGS-UFSAR Revision 29 January 30, 2017 Surveillance Requirements FUNCTIONALITY of each analyzer shall be determined by the performance of: a)a Channel Check at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; b)a CHANNEL FUNCTIONAL TEST at least once per 92 days; c)a gas calibration* at least once per 92 days using sample gases containing: 1. Two volume percent hydrogen (low span), balance Nitrogen, and 2. Six volume percent hydrogen (high span), balance Nitrogen. d) A CHANNEL CALIBRATION at least once per refueling using sample gases containing: 1. Two volume percent hydrogen (low span), balance Nitrogen, and 2. Six volume percent hydrogen (high span), balance Nitrogen.
  • The hydrogen sensor gas calibration shall consist of all CHANNEL CALIBRATION, with the exception that only a single elements of the point comparison installed plant check for reasonableness instrumentation) is required pressure sensors. (by comparison to to check the hydrogen 6.2.6 References for Section 6.2 other analyzer temperature and 1. Field, C. V., "Fan Cooler Motor Unit Test," WCAP-7829, April 1972. 2. Styrikovich, M. A. et al., "Attomnoyl Energiya," Volume 17, No. 1, pp. 45-49, (Translation in UDE-621.039.562.5), July 1964. 3. Parsley, Jr., L. F., "Design Considerations of Reactor Containment Spray Systems-Part VI." ORNL-TM-2412. Part 6, 1969. 4. Parsley, Jr., L. F., "Design Considerations of Reactor Containment Spray Systems-Part VII." ORNL-TM-2412, Part 7, 1970. 5. Eggleton, A. E. J., "A Theoretical Examination of Iodine-Water Partition Coefficient," AERE (R) -4887, 1967. 6.2-80 SGS-UFSAR Revision 29 January 30, 2017
  • 6. DiNunno, J. J.; Anderson, F. D.; Baker, R. E.; and Waterfield, R. L., "Calculations of Distance Factors for Power and Test Reactors," TID-14844 . 7. Bell, M. J.; Sulkowski, J. E.; and Picone, L. F., "Investigation of Chemical Additives for Reactor Containment Sprays," WCAP-7153-A, April 1975. 8. United States Atomic Energy Commission, "Safety Guide 7, Control and Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident," March 10, 1971. 9, ORNL Nuclear Safety Research and Development Program Bi-Monthly Report for July -August 1968, ORNL-TM-2368, p. 78. 10. ORNL Nuclear Safety Research and Development Program Bi-Monthly Report for September -October 1968, ORNL-TM-2425, p. 53, 11. Whyte, D. D. and Burchell, R. C., "Corrosion Study for Determining Hydrogen Generation from Aluminum and Zinc During Post-Accident Conditions," WCAP-8776, October 1976.
  • 12. Sejvar, J., "Distribution of Fission Product Decay Energy in PWR Cores," WCAP-7319-L (Proprietary), April 1969 and WCAP-7816 (Nonproprietary), December 1971.
  • 13. ORNL Nuclear Safety Research and Development Program Bi-Monthly Report for May -June 1969. 14. Allen, A. 0., "The Radiation Chemistry of Water and Aqueous Solutions." 15. Report F-C3217-11 "Test of Coatings in a Simulated Reactor Containment Environment, "Franklin Institute Research Laboratories, March 1972. 16. Report 78-0810-1, "Radiation Tolerance, Decontamination, Design Basis Accident, Physical Properties & Chemical Properties Tests for Carbon Steel & Concrete Coating Systems (1977 ORNL Test Series}," Keller & Long, Inc., August 1978. 6.2-81 SGS-UFSAR Revision 23 October 17, 2007 I 17. Deleted. 18. Deleted. 19. Cranston, G. v., "Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants, 11 Bechtel Topical Report BN-TOP-1 Revision 1, November 1, 1972 20. United States Atomic Energy Commission, Letter from R. C. DeYoung to R. D. Allen, February 1, 1973. 21. Nuclear Safety Advisory I1etter NSAL-93-016, Revision 1, "Containment Spray System Issues," Westinghouse, October 4, 1993. 22. Westinghouse Letter PSE-94 -500, J. Huckabee to E. S. Rosenfeld, PSE&G, "Small Break LOCA Offsite Dose Analysis," January 5, 1994. 23. VTD 326657, Cooling Fan Motor Drawing, August 20, 1994. 24. USNRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000 . 25. Section 50.67 of title 10 of the Code of Federal Regulations (10CFR 50,67), "Accident Source Term". 26. USNRC Standard Review Plan Section 6.5.2, "Containment Spray as a Fission Product Cleanup System", Revision 2, 1988. 27. Technical Specification Amendment Numbers 281 and 264 to Facility Operating License numbers DPR-70 and DPR-75. 6.2-82 SGS-UFSAR Revision 23 October 17, 2007 * * *
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