LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Tables 5.5-1 Through 5.5-3
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TABLE 5.5-1 RESIDUAL HEAT REMOVAL SYSTEM DESIGN PARAMETERS Code Requirements Residual Heat Exchangers (Tube Side) (Shell Side) Residual Heat Removal Piping and Valves General Plant design life, years Component cooling water supply temperature design, °F Reactor coolant temperature at startup of decay heat removal °F Time to cool Reactor Coolant System from 350°F to 140°F, starting at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after shutdown, hr (1) Used for design. ASME III, Class C ASME VIII ANSI 831. 1. 0 (ll ANSI 831.7(2) 40 95 350 (2} For piping not supplied by the NSSS supplier, material inspection fabrication and quality control conform to ANSI 831.7. not possible to comply with ANSI 831.7, the requirements of ASME III-1971, which incorporated ANSI 831.7, were adhered to. (3) 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> was the original design value. With the 1.4% power uprate, reduction in temperature can be accomplished in 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. To cool down in 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> is not a design requirement. The design requirement is to cool down in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with a single train. 1 of 4 SGS-UFSAR Revision 19 November 19, 2001 I
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- TABLE 5.5-1 (Cont.) Refueling water storage temperature, °F Decay heat generation at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after shutdown, Btu/hr a3so3 concentration in refueling water storage tank, ppm boron COMPONENTS Residual Heat Exchangers Number Design heat transfer, Btu/hr Shell Design pressure, psig 150 Design temperature, OF 200 Design flow rate, lb/hr 2.475 Design outlet temperature, OF 108.8 Design inlet temperature, OF 95 X 106 Fluid Component cooling water Ambient 6
- 72.1 X 10 .. 2000 2 (per unit) 6 34,15 X 10 600 400 1.48 X 106 114 137 Reactor coolant (borated demineralized water)
- Original decay heat value used in the initial design 2 of 4 SGS-UFSAR Revision 19 November 19, 2001 I I TABLE 5.5-1 (Cont)
- Material of construction Carbon steel Residual Heat Removal Pumps Number Type Design pressure, psig Design temperature, °F Shutoff head, psi Design flow rate, gpm
- Design head, ft Available NPSH at design flow rate, ft Temperature of pump fluid, °F Normal fluid Fluid during LOCA recirculation phase Material of construction
- 3 of 4 SGS-UFSAR Austenitic stainless steel 2 (per unit) Vertical centrifugal 600 400 170 3,000 350 25 40 -350 Reactor coolant Radioactive borated water with H2 and NaOH in solution Austenitic stainless steel Revision 6 February 15, 1987
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- Piping and yalyes Residual heat removal loop (piping and valves in isolated loop): Design pressure, psig Design temperature, °F TABLE 5.5-1 (Cont.) lYme Sugtion Discharge 450* 600 400 400 Residual loop isolation valves and Design pressure, psig 2,485 Design temperature, °F 650
- Unit 2 piping downstream of 2RH75 & 76 are designed to 600 psig
- 4 of 4 SGS-UFSAR Revision 17 October 16, 1998 TABLE 5.5-2 RESIDUAL HEAT REMOVAL SYSTEM FAILURE ANALYSIS component 1. Residual heat removal pumps 2. Residual heat removal pump 3. Residual heat removal pump 4. Residual heat removal pump s. Remote operated valves inside containment in pump suction line SGS-UFSAR Malfunction Rupture of a pump casing Pump fails to start Motor operated valve on pump suction is closed Stop valve on discharge line closed or check valve sticks closed Valve fails to open Comments ana Consequences The casing ana shell are designed for 600 psi and 400°F. The pump is protected from overpressurization by two normally closed valves in the pump suction line ana by an open relief line, containing a relief valve, back to the containment sump. The pump is inspectable ana is located in the Auxiliary Building protected against credible missiles. Rupture is considered unlikely but in any event the pump can be isolated. One operating pump furnishes half of the flow required to meet design coolaown rate. This increases the time necessary for plant cooldown. This is prevented by prestartup ana startup ana operational checks. stop valves are locked open. Prestartup ana operational checks confirm position of valves. In the improbable event that one of the remote operated valves on the suction line to the residual heat removal pumps is inoperable, an attempt will be made to open it manually. If this is impossible, the plant will be cooled to about 280°F with steam dump from the steam generators, while additional recovery actions could be implemented basea on plant's abnormal ana emergency operating procedures, equipment availability and resources. 1 of 2 Revision 17 October 16, 1998 component 6. Remote operated valves inside containment on pump discharge line 7. Residual heat exchanger 8. Residual heat exchanger vent or drain valve SCS-UFSAR TABLE 5.5-2 (Cont.) Malfunction Valve fails to open Tube or shell rupture Left open 2 of 2 Comments and Consequences Pump discharge pressure gauge shows pump shut-off head indicating no flow. The low head safety injection lines may be opened and utilized to direct flow to the RCS hot legs. A reactor coolant pump must be operated. Rupture is considered unlikely, but in any event the faulty heat exchanger may be isolated. This is prevented by prestartup operational checks. Revision 16 January 31, 1998 TABLE 5.5-3 SALEM NUCLEAR GENERATING STATIONS UNIT NOS. 1 AND 2 STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS LOADING COMBINATION AND ALLOWABLE STRESS LIMITS LOADING COMBINATIONS SUPPORTS -ALLOWABLE STRESS LIMIT 1. 2. 3. Normal loads Normal loads + operating base earthquake (upset condition) Normal loads + design base earthquake + pipe rupture loads (faulted condition) Working stresses per AISC code 1-1/3 working stresses AISC code Yield stress of material, or AISC Code
- or ASME III, Subsection NF and Appendix F
- with increase factors consistent with the guidance of R. G. 1.124 1 of 1 SGS-UFSAR Revision 20 May 6, 2003