LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 3.5, Missile Protection
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3.5 MISSILE PROTECTION structures, shields, and barriers are provided as protection against the effects of both internally and externally generated missiles. For turbine generated missiles, additional factors such as safe inspection and re-inspection schedules for low pressure turbine discs are included in the risk analysis. The discussion on turbine missiles is given in Section 3.5.4. The following section covers internally generated missiles. 3.5.1 Internally Generated Missiles A loss-of-coolant accident (LOCA) or other plant equipment failure might result in internally generated missiles. For such engineered safety features as are required to assure safety in the event of such an accident or equipment failure, protection from these missiles is considered in the layout of plant equipment and missile barriers. 3.5.1.1 Missile Types The types of missiles for which protection is provided include: 1. Valve stems 2. Valves 3. Instrument thimbles 4. Various types and sizes of nuts and bolts 5. Complete control rod drive mechanism, or parts thereof, except the MG set flywheels, which will not produce missiles under any anticipated accident condition. 3.5.1.2 Missile Protection Methods High pressure Reactor Coolant System equipment which could be the source of missiles is suitably shielded either by the concrete 3.5-1 SGS-UFSAR Revision 14 December 29, 1995 shield wall enclosing the reactor coolant and pressurizer loops or by the concrete operating floor to block any passage of missiles to the containment walls even though such postulated missiles are deemed most improbable. A structure is provided over the control rod drives to block any missiles generated from a fracture of equipment. Missile protection for the plant is provided to comply with the following criteria: 1. To protect the containment and lines from loss-of-function due to damage by such missiles as might be generated in a LOCA for break sizes up to and including the double-ended severance of a reactor coolant pipe. 2. To protect the engineered safety features 1 systems and components required to maintain containment integrity against loss-of-function due to damage by the missiles defined in Section 3.5.1.1. The following considerations are included in the design for missile protection to meet the above criteria: 1. The Reactor Coolant System is supported by steel structures designed to withstand the forces associated with a double-ended rupture of a reactor coolant pipe and shielded by concrete walls designed to stop the missiles. 2. The structural design of the missile shielding takes into account both static and impact loads. 3. Components of the Reactor Coolant System are examined to identify and to classify missiles according to size, shape, and kinetic energy for purposes of analyzing their effects. The Petry formula as described in Section 3. 5. 3 is used to check the missile penetration. The energy approach is used to determine the equivalent static load from the missile impact. The original missile shields located above the control rod drive mechanisms have been removed due to installation of the integrated head assembly (IHAl . The IHA utilizes an integrated steel missile shield that provides the necessary blockage of any missile that could be generated by the control rod drive mechanisms. To ensure the IHA missile shield is capable of withstanding a missile impact and adequately performing its design function, a penetration evaluation, as well as a finite element strain evaluation, were performed. 3.5-2 SGS-UFSAR Revision 22 May 5, 2006 The penetration evaluation uses the USNRC Standard Review Plan Section 3.5.3, Barrier Design Procedures, for guidance in determining the minimum required thickness of the IHA missile shield to ensure it can absorb the impact energy without perforation. A non-linear transient finite-element analysis evaluated the missile shield for the overall effects expected from the impact of a CRDM
missile. 3.5.2 Tornado Missiles Category I structures including Reactor Containments, Auxiliary Building, and Fuel Handling Buildings are designed to withstand tornado missiles. These structures are also designed such that under the impact of the most damaging tornado missile, they will not create a secondary missile of enough mass or velocity to penetrate any adjacent Category I structure.
3.5.2.1 Critical Missiles Selected for Evaluation For tornado generated missiles, a wooden utility pole has been used as the critical object for penetration analysis. The pole is 40 feet long, 12 inches in diameter, weighing 50 pounds per cubic foot, traveling in a vertical or
horizontal direction at 150 mph.
In addition, the exterior walls of the safety-related plant structures, which have a minimum thickness of 18 inches, have been evaluated against the following two missiles:
- 1. Steel rod, one inch diameter, three feet long, weight eight pounds, traveling horizontally at 316 feet per second and vertically at 252
feet per second, at all elevations, and
- 2. Utility pole, 13-1/2 inches diameter 35 feet long, weight 1490 pounds, traveling horizontally at 211 feet per second and vertically at 169 feet per second, at all elevations less than 30
feet above grade within 1/2 mile of the facility structures.
A passenger car missile is also used in the evaluation.
3.5.2.2 Missile Protection Methods
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The cumulative probability of missile strikes on all the unprotected components and openings for each of the two Salem units was determined to be less than 10 -G, which is acceptable per the guidance provided in Regulatory Guide 1.117 and SRP section 2.2.3. Penetrations into steel barriers are calculated using equations developed based on tests conducted at the Stanford Research Institute, summarized in Reference 7. The equation for penetration can be written in the form where e d B D = = = perforation thickness (in) effective projectile diameter (in) width of plate between rigid supports (in) wjd3, where w =missile weight in pounds missile striking velocity {ftjsec) uo s s = ultimate tensile strength of steel target (psi} The value of e calculated using the above equation is multiplied by 1.25 when designing steel barriers 1.25 e Steel missile barr-rers are designed using the following allowable ductility ratios: Flexural members Columns with slenderness ratio to 20 Columns Members where with slenderness ratio subjected to tension e = ultimate strain u e y = yield strain (1/r) (1/r) 1Jd 10.0 less than or equal 1Jd 1.3 greater than 20 j.Jd 1.0 1Jd 0.5 eu. eY 3.5.2.3 Safety Assurance Against Tornado Missile Induced Damages Category I systems outside the Category I buildings that may be damaged by tornado or secondary missiles are the refueling water storage and auxiliary feedwater storage tanks and systems. Sufficient precautions have been taken to assure a safe shutdown of the reactor and to maintain it in a safe condition. 3.5-S SGS-UFSAR Revision 17 October 16, 1998 3.5.2.3.1 Backup Water Sources for Category I Water Storage Tanks In the event that the integrity of Category I storage tanks are violated due to tornado induced missiles, sufficient backup water sources are available to assure a safe shutdown of the reactor. Primary system makeup requirements during cooldown are normally accomplished by the Chemical and Volume Control System (CVCS}, providing makeup water from the primary water storage tank. In the event that the primary water storage tank is the affected tank, primary system makeup can be provided by transferring water to the Reactor Coolant System from the three 8500 cubic foot capacity holdup tanks or the spent fuel pool via the eves. Approximately 24 I 000 gallons of water are required to maintain the level in the Reactor Coolant System in the transition from full power operation to cold shutdown. Removable borated water (31,000 gallons) is available in the spent fuel pool above the spent fuel pool pump suction line. In addition, another 13,000 gallons of borated water is available above the spent fuel pool pump suction line in the fuel transfer pool. Feedwater system makeup requirements during cooldown are normally supplied from the Main Feedwater System, providing makeup water from the condenser. In the event the Main Feedwater System becomes inoperable, the Auxiliary Feedwater system is placed in operation and feedwater makeup is supplied from the auxiliary feed storage tank. This Category I tank is adequately protected from the effects of earthquakes, tornado, wind loads, and floods. Backup water sources for the auxiliary feedwater pumps are the two demineralized water storage tanks (500,000 gallon capacity each), the two fire protection and domestic water storage tanks (350,000 gallon capacity each), and the station Service Water System. Because of the chloride content of the water (approximately 6000 ppm), the Service Water System would be used for this purpose only as a last resort in the event that other water sources become unavailable. The Service Water System will be used for decay heat removal as necessary, only until the normal or backup water sources are again made available. Backup water supply from the demineralized water storage tanks is available, upon filling and venting the associated suction piping. readily The fill process is accomplished using local/manual valves. provided to ensure the line is properly filled. Automatic vent valves are Once filled the alternate suction stops, AFS2's are opened from the Main Control Room to place this backup source of water in service. The fire protection and domestic water storage tanks can be used as a backup water source in the event that the demineralized water storage tanks become unavailable. Fire protection and domestic water can be provided to the auxiliary feed pumps only when a spool piece has been connected by station personnel. The Service Water System can be used as a backup water 3.5-6 I SGS-UFSAR Revision 17 October 16, 1998 -
source in the event that the other water sources become unavailable. Service Water can be provided to the Auxiliary Feed Pumps with a spectacle flange installed between the auxiliary feedwater and service water piping. Service water can be provided to the auxiliary feed pumps by rotating the spectacle flange into the full bore position. The time required to connect the two systems is less than 30 minutes. Operations involving alignment of normal and alternate sources of water to the auxiliary feedwater pumps are covered by station procedure. The backup water sources from the CVCS holdup tanks, the spent fuel pool, and the Service Water System are located in buildings/structures designed to withstand tornado induced missiles. The water storage tanks are located in different areas of the station site (refer to the arrangement drawings in Section 5) Although not specifically designed to withstand tornado induced missiles, the separated locations of the various tanks preclude the possibility that all tanks would be rendered unavailable due to tornado induced missiles. The systems which are required to bring the unit to a safe shutdown are enclosed in tornado protected buildings/structures and capable of being powered by the standby AC power systems. Plant Shutdown Operating Procedures In the unlikely event that a breach of any of the primary water sources were to occur (refueling water storage tank, auxiliary feedwater storage tank, or primary water storage tank), station operating personnel would start to shut down and cool the reactor below 350°F, using normal plant shutdown operating procedures. These procedures initiate immediate decay heat removal by steam dump to the condenser. As feedwater in the secondary plant cycle absorbs heat from the reactor coolant through the steam generators, it is converted to steam, giving up heat to the circulating water in the condenser. Hot shutdown can be maintained with Auxiliary Feedwater System operation and main steam atmospheric relief valves. Boration can be accomplished via the charging pumps, boric acid tanks, and boric acid trans fer pumps. In the event the Auxiliary Feed Storage Tank is unavailable, transfer to one of the backup water sources (previously discussed in this section) is accomplished by station procedure. No automatic water supply switchover is provided. The backup water sources for the Auxiliary Feedwater System water sources are addressed in the Technical Specifications (except for the service water backup source) . The unit would be maintained in a hot shutdown condition until station operating personnel have assessed operating conditions, damage, and availability of water sources to attain cold shutdown. Necessary repairs would be made and alternate water supplies provided as necessary for the CVCS to maintain Reactor Coolant System inventory and proceed to bring the unit to a cold shutdown condition. 3.5-7 SGS-UFSAR Revision 28 May 22, 2015 Shutdown procedures would be accomplished in as conditions permit utilizing alternate water methodology would be dependent upon the extent normal a fashion as operating sources as necessary. The of tornado missile damage as determined by operating personnel in their assessment of plant conditions. The arrangement of the service intake structure, service water pumps and piping is discussed in Section 9. 2. The service water pumps are located within the service intake structure, which is seismic Category I and designed to be tornado missile proof. The piping is buried below grade, thereby protected from tornado missiles. Indication is provided in the control room for all outdoor water storage tanks. Upon indication of loss of water from all outdoor water storage tanks such that they become unavailable for auxiliary feedwater supply (tornado induced failure), personnel would be dispatched to install the connection between the Service Water and Auxiliary Feedwater Systems. It has been determined by actual demonstration that two people can rotate the spectacle flange in 15 minutes. The installation requires no special tools. The spectacle flange bolts will be removed, the spectacle flange will be rotated, and then the bolts will be reinstalled to provide the required flow. An analysis was performed to determine the time period following a loss of ac power and main and auxiliary feedwater flow before the core becomes uncovered. The pertinent assumptions used in the analysis are as follows: 1. All ac power lost at time of incident 2. Rods assumed to begin dropping into core 2 seconds following incident 3. ANS standard decay heat curve assumed 4. Pressurizer relief and safety valves operative 5. Initial power is 1.006 times rated power 6. Loss of all main and auxiliary feedwater following incident. The calculation was subdivided into the following areas: 1. Heat required to raise primary to saturation temperature 3.5-8 SGS-UFSAR Revision 28 May 22, 2015
- 2. Heat required to uncover core after primary saturation temperature is reached 3. Secondary heat sink available Primary Saturation The amount of heat to *raise the Unit 1 primary inventory from the initial average temperature to the saturation to the pressurizer power operated relief valve setpoint was calculated to be approximately 16 full power seconds. The primary side inventory increases by 3.4% as a result of steam generator replacement thus; this value increases to approximately 16.5 full power seconds for Unit 2. The quantity of heat required to boil sufficient Unit 1 primary inventory in order to begin uncovering the core was found to be approximately 5.8 full power seconds. The primary side inventory increases by 3. 4% as a result of steam generator. replacement thus; this value increases to approximately 6. 0 full power seconds for Unit 2. Based upon nominal initial mass, the heat to the Unit 1 secondary inventory via the steam generator safety valves was calculated to be equivalent to 7 4. 8 full power seconds. The secondary side inventory decreases by 6.3% as a result of steam generator replacement thus; this value is reduced to approximately 70.1 full power seconds for Unit 2. The total heat generation, which results in uncovering the core, is then the sum of the three heat sinks itemized above. These components yield a total of 96.6 full power seconds for Unit 1 and 92.6 full power seconds for Unit 2. Assuming a standard ANS decay heat curve, the period of time following initiation of the incident was calculated to be approximately 4200 seconds or 70 minutes for Unit 1 and approximately 65 minutes for Unit 2. With this analysis taken into consideration, a conservative estimate of time history to initiate feedwater following a loss of outdoor water storage tanks follows: 3.5-9 SGS-UFSAR Revision 24 May 11, 2009 Time (Min.) 0 30 40 55 Event Tornado induces failure in all outdoor water storage tanks. (Low water indication in control room) Acknowledgement by control room, dispatch personnel to rotate spectacle flange. Spectacle flange rotation commences Service water connection complete; auxiliary feedwater initiated. This discussion demonstrates that sufficient water sources are available to bring the unit to a safe shutdown condition in the event of a breach of a primary water source due to a tornado induced missile. 3.5.2.3.2 Safety Evaluation of Loss of Suction to the Auxiliary Feedwater Pumps The design of the Auxiliary Feedwater (AFW) System was evaluated with regard to tornado-induced loss of suction to the AFW pumps. The Auxiliary Feedwater Storage Tank (AFST) and AFW pump suction piping have been designed to withstand the design basis tornado wind loadings as described in Section 3.3. The following discussion with respect to the evaluation therefore considers only tornado missiles. In the evaluation it was conservatively assumed that loss of pump suction pressure would result in instantaneous damage to the AFW pumps. The design basis tornado generated missile is a wooden utility pole described in Section 3. 5. 2. 1. The probability of a missile impact on the AFST outlet piping has been determined to be less SGS-UFSAR 3.5-10 Revision 28 May 22, 2015
-7 than 10 per year and thus no further has been performed regard. regard to the APST, it has been determined that the design basis utility pole striking the tank at its base would conservatively result in a 490 square inch opening. Such an opening would result in the AFST draining in approximately 5 minutes. Continuous AFST level indication is provided in the control room, as well as low level (100,000 gallon) and low low-level (30,000 gallon) audible and visual alarms. Additional audible and visual alarms are being provided to detect deviation from the technical specification minimum volume (200,000 gallons). The above indications and alarms prompt of damage to the AFST. The potential effects of such a loss of water to the damaged AFST on various modes of plant operation were also evaluated. In the event of a tornado forecast, the shift crew would operate the unit with an increased sensitivity toward the potential effects of a tornado, and thus, it was concluded that sufficient time is available to recognize and assess the damage, trip the AFW pumps from the control room, and change over to an alternate suction path without damage to the AFW pumps. In order to further enhance the overall design of the AFW System, a safety grade automatic low suction pressure trip for each AFW pump has been incorporated. To preclude inadvertent actuation, this modification is designed such that it can be made operable only during periods when a has been put in effect by the National Weather Service. Also, in accordance with I&E Bulletin 80-11, a wire mesh net protects the Unit 1 RWST and AFW tank in the event of adjacent wall collapse due to seismic loading. 3. 5-11 SGS-UFSAR Revision 7 July 22, 1987 3.5.3 Modified Petry Formula The penetration of the missile into concrete has been calculated by the Modified Petry formula: D 1 + A 215,000 Where: D penetration in feet -3 3 K coefficient equal to 4.76 x 10 ft /lb for reinforced concrete with a crushing strength of 3200 psi, and -3 3 2.82 x 10 ft /lb for reinforced concrete with a crushing strength of 5700 psi. W missile weight in pounds V striking velocity in feet per second A missile frontal area in square feet 3.5.4 Turbine Missile 3.5.4.1 Turbine Placement and Characteristics Plan and elevation views of the turbine building are provided on Plant Drawings 204811, 232445 and 232444. As described in Section 10, each unit consists of one tandem compound, six flow, four casing, condensing, 1800 RPM turbine. element is designated as a Siemens Westinghouse 13.9 The Unit 1 low pressure 2 m retrofit turbine. The Unit 2 low pressure element is designated as a Westinghouse building block 81R. 3.5-12 SGS-UFSAR Revision 27 November 25, 2013 3.5.4.1.1 *
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- High Pressure (HP) Turbine 'i'he HP turbine element, as shown on Figure 3. 5-4, is of a double flow design and is inherently thrust balanced. Steam from the four control valves enters at the denter. of the turbine element through four inlet pipes, two in the base and two in the cover. These feed an admission ring in the inner casing connected to the turbine casing. Steam leaves the admission ring and flows through the reaction blading. The reaction blading is mounted in the inner casing and the guide blade carriers shown on 3.5-5 which are mounted in the turbine casing. The HP rotor is made of NiCrMoV alloy steel. The mechanical properties are as follows: fled minimum Tensile Strength, N/nw2 (psi) Yield Strength, N/rnm2 (psi) 1 (0.2 percent offset) Elongation (10=5d), percent Reduction of Area, percent Impact Strength, Charpy V-Notch, J (min at room temperature) 50 Fracture Appearance Transition Temperature, °F, max :S:820 (.S:l1B931) 580-680(84122-98626) 100 :S:-22 The main body of the rotor weighs approximately 110,000 lb. The approximate values of the transverse centerline diameter, the maximum diameter, and the main body length are 4 5 inches, 64 inches, and 138 inches, respectively. The casing cover and base are made of carbon steel castings. The specified minimum mechanical properties are as follows: Tensile Strength, psi, min 3.5-13 70,000 Revision 21 December 6, 2004 Yield Strength. psi, min Elongation in 2 inches, percent, min Reduction of Area, percent, min 36,000 22 35 The ibend test specimen is capable of being bent cold through an angle of 90 degrees and around a pin 1 inch in diameter without cracking on. the outside of the bent portion. The blade and the inner casings are made of alloyed steel castings. The specified minimum mechanical properties are as follows: Tensile Strength, N/mm2 (psi} Yield Strength, N/mm2 (psi), (0.2 percent offset) Elongation (10=5d), percent Reduction of Area, 540-690 Impact Strength, Charpy V-Notch, J The approximate weights of the inner casing, two blade the casing cover, and the casing base are 46,000 lb, 26,500 lb, 115,000 lb, and 115,000 lb, respectively. The casing cover and base are tied together by means of more than 100 studs . The stud material is an alloy steel having the following mechanical properties: Tensile Strength, psi, min. Yield Strength, psi, min. (0.2 percent offset) Elongation in 2 inches, percent, min.
- Reduction of Area, percent min. SGS-UFSAR 2-1/2 Inches and Less 125,000 105,000 16 50 3.5-14 Over /2 To 4 Inches 115,000 95,000 16 so Over 4 110,000 85,000 16 45 Revision 21 December 6, 2004 * * *
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- The studs have from 17 to 66 inches and diameters ranging from 2.75 inches to 4.5 inches. About 90 percent of them have diameters ranging between 2.5 and 4 inches. The total stud cross-sectional area is approximately 900 in2 and the total stud free-length volume is approximately 36,000 in3 3.5.4.1.2 Low Pressure (LP) Turbine 1. Unit 1 The double flow LP turbine, shown on Figure 3. 5-6, high efficiency blading, diffuser type exhaust, and incorporates liberal exhaust hood design. The last stage rotating blade length is 4 6". The LP turbine cylinder is fabricated from steel plate to provide uniform wall thickness thus reducing thermal distortion to a minimum. The entire outer casing is subjected to low temperature exhaust steam. The LP turbines include an inner casing with the installed stationary blading and external heat shields. The heat shield mitigates heat losses by radiation from the inner to the condenser and also reduces thermal stressing of the inner casing components. Figure 3.5-7 shows the welded inner casing with its blade carriers and blade rings. The upgrade also features compartments for the inlet steam and steam extractions, which are designed to safely handle maximum pressure differences at steady state and transient operating conditions. The upgraded inner casing features optimal moisture removal at the circumferential outer diameter of the flow path. In addition, the last stage hollow stationary blades feature moisture removal, slots to take moisture from the blade surface and discharge it through the slots into the hollow blade and from there into the condenser. The fabricated inner is supported by the outer casing at the horizontal centerline and is fixed transversely at the top and bottom and axially at the centerline of the steam inlets, thus allowing freedom of expansion independent of the outer 3.5-15 SGS-UFSAR Revision 21 December 6, 2004 The outer cylinder and the inner cylinder are fabricated mainly of ASTM 515-GR65 material. The minimum specified properties are as follows . Tensile Strength, ' min 65,000 Yield Strength. psi, min 35,000 Elongation in 8 inch percent min 19 Elongation in 2 inch percent min 23 The LP rotors and discs are made of NiCrMoV alloy steel. The specified minimum mechanical properties are as follows: Tensile Strength, N/mm2 (psi) Yield Strength, N/mm2 (psi), (0.2 percent offset) Elongation (10=5d), percent Reduction of Area, percent Charpy V-Notch, J (min at room temperature) ::::;820 ($118931) 580-680{84122-98626) ;;;:16 250 100 50 Percent Fracture Appearance Transition S-22 Temperature, °F1 max There are discs shrunk per shaft with three per flow. 3.5-16 SGS-PFSAR Revision 21 December 6, 2004 * * *
.--2. Unit 2 SGS-UFSAR The Unit 2 LP turbine is a double flow design consisting of three elements. The original LP turbine rotors were of the built-up type consisting of individual discs shrunk on and keyed to a shaft. To reliability, optimize thermal performance and provide long term the turbine has been retrofitted with a monoblock rotor forging with fully integral discs and couplings. This design incorporates blading with integral shrouds. The last stage rotating blade length is 47", resulting in a large annuluf area. Figure 3.5-6A depicts the LP turbine. Steam will enter the LP turbine inlet at the cross-over pipe tee connection. Inlet flow guides will direct the steam in both directions of this double flow element. There are nine stages of controlled reaction blading. The first four stages on each end are housed in the first blade ring. The next two stages on each end are housed in the second blade ring. Both blade rings on each end are separately supported and made from stainless steel to minimize erosion/corrosion concerns. The last three stages on each end are of segmental assembly construction that are caulked into the inner cylinder. Advanced techniques were used to design the one-piece inner cylinder, resulting in minimized steam leakage and structural streamlining which minimizes blade path differential thermal expansion. The entire outer casing is subjected to low temperature exhaust steam. The outer and inner cylinders are fabricated mainly of ASTM 515-GR65 material. The minimum specified properties are as follows: Tensile Strength, psi, min. Yield Strength, psi, min. Elongation in 8 in., percent, Elongation in 2 in., percent, 3.5-17 min. min. 65,000 35,000 19 23 Revision 16 January 31, 1998 The LP rotors are made of 3. 5 NiCrMoV alloy steel. minimum mechanical properties are as follows: Tensile Strength, psi, min. Yield Strength, psi, min. Elongation in 2 in., percent, min.* Reduction of Area, percent, min. Impact Strength, Charpy V-Notch, ft-lb, min. at room temp. 50% fracture Appearance Transition Temp. °F, max. 115,000 100,000 17 50 120 35 The specified The first five stages of the rotors use a one-piece integral shroud blade. The first three rows are made of 12% Cr material. The next two are made of 17-4 PH stainless steel material. The last three rows of the rotors use shot peened profiled free-standing blades also made of 17-4 PH stainless steel. 3.5.4.1.3 Overspeed Protection System Testing The turbine overspeed protection system is tested by accelerating the turbine from. 1800 rpm until it trips. Should the turbine fail to trip automatically, it is manually tripped prior to exceeding 12 percent overspeed. The turbine stop and gdvernor valves are tested by actually cycling the valves. For Salem
- Unit-13 1 & 2, the Turbine Overspeed Protection requirements are discussed in 10.2.2.6, "Turbine Overspeed Protection". 3.5.4.2 of Missile Present manufacturing and inspection techniques for turbine rotor and disc forgings make the of an undetected flaw and subsequent catastrophic structural failure extremely remote. Forgings are subject to inspection and testing both at the forging suppliers and at Westinghouse. Current design procedures are well established and conservative, and analytical tools .such as finite element and fracture mechanics techniques allow in depth analysis of any potential trouble spots such as areas of stress concentration or inclusions which could give rise to crack propagation. SWPC has missile probabilities methodology that has been approved by the NRC, in NRC SER titled, "Siemens Westinghouse Topical Report, le Analysis Methodology General Electric (GE) Nuclear Steam Turbine Rotors by Siemens Westinghouse Power Corporation (SWPC), Project No. 721" 3.5-18 Revision 21 December 6, 2004 * * *
(reference ) . Although the title of this document references GE turbine rotors, the conclusions provide the following statement: "The approval of the Siemens methodology includes the use of specified values for the use of the PPDBURST and PDMISSILE computer programs, and the use of specified values for some key input and built-in parameters for those two programs for future plant-specific turbine missile probability analyses for GE and Siemens rotors". The results of SWPC missile analysis is documented in SWPC Technical Report, CT-27336, "Missile Probability Analysis PSEG Nuclear LLC Salem Unit 1" (reference 8) . This report is in full compliance with the requirements of the SER and meets the specific analysis requirements established by the SER conclusion section. The conclusion of the report states that the missile probabilities are within the limits required by Regulatory Guide 1. 115, Protection Against Low-Trajectory Turbine Missiles, Rev. 1, (1E-5 per year) for an unfavorably oriented low pressure turbine based on a six month turbine valve testing frequency with a 100,000 equivalent operating hours inspection frequency. Siemens further clarifies that the inspection intervals will be limited to 87,600 equivalent operating hours inspection frequency based on current NRC approved inspection frequency. 3.5.4.2.1 High Pressure Turbine Risk Analysis The design and fabrication of the HP rotor is such that the vendor does not require periodic inspection of the rotor to address the risk of missile generation. Ductile burst is unlikely, since it would require a rotational speed beyond the terminal speed of typical units and failure from this mechanism need not be considered. Failures due to high cycle fatigue fracture have not occurred in the past and the retrofit rotors have improved design safety factors making this mechanism unlikely. Failures due to low cycle fatigue are unlikely since LCF life is significantly greater than 10,000 start cycles for the original and retrofit rotors. Based on these factors, there are no requirements for periodic in-service inspections to address the risk of missile generation. For more details, refer to Reference 9. 3.5.4.2.2 Low Pressure Turbine Risk Analysis 1. Unit 1 The Salem Unit 1 low pressure turbines will be inspected on at least a 10 year (87, 600 equivalent operating hours) frequency to ensure that the probability of turbine generated missile remains within the requirements of Regulatory Guide 1.115. This inspection frequency requires that turbine valves (stop & control) be tested at six month intervals. 3.5-19 SGS-UFSAR Revision 27 November 25, 2013 Basically, the approach used is to make a conservative prediction of how a presumed or actual crack will grow and then schedule an inspection to the time the crack grows large enough to be of concern. Analytical components of this approach are Load, Crack Branching Fracture Toughness, Yield Strength, SCC Growth Rate, and Initial Crack Size. SWPC Technical Report, CT-27336, "Missile Probability Analysis PSEG Nuclear LLC Salem Unit 1" {reference 8) provides a discussion of these parameters and values utilized in the development of the Salem Unit 1 Low Pressure Turbine Generated Missile Probabilities. a. Load SWPC determines load for each disc through the use of finite element analysis. They assume that the analysis provides accurate results within 5% of tolerance due to the uncertainties in geometry as well as thermal and mechanical loads. A normal distribution is assumed. The mean values for the disc are: Disc #l -498 MPa Disc #2 -521 MPa Disc #3 -535 MPa b. Crack Branching Factor The branching factor k is assumed to be normally distributed with a mean of 0.65 and a standard deviation of 0.175, whereby j 0.65 {/' C'rack Depth s J in k =l ... 1 other1rise c. Fracture Toughness The normal distribution has been used in describing scatter in fracture toughness data with a mean of 219.8 and standard deviation of 10% of the mean value. , d. Yield Strength The yield strength values are assumed to be distributed normally with mean and standard deviation values on internal investigation data: Disc #1 815 MPa and std. deviation 30 MPa Discs #2 & #3 855 MPa and std. deviation = 30 MPa ; e. SCC Growth Rate The stress corrosion (SCC) rate is assumed to be independent of the stress intensity level. The main parameters influencing the SCC rate are temperature, material yield and water chemistry. The empirical 3.5-20 SGS-!p'FSAR Revision 21 December 6, 2004 * * *
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- equations for sec rates were developed based on SWPC field measurements and laboratory test data. The sec rate is given in inches/hour, temperature in °F1 and the material strength in ksi. da . 7302 dt = axp(-4.968-T + 460 + 0.0278
- O'y) The log normal distribution sec rate with a standard deviation of 0.578 is assumed. f. Initial Crack Size The initial crack size is assumed to be a non-varying variable with a value equal to 3 mm. 2. Unit 2 The Salem Unit 2 LP turbines are equipped with fully integral rotors which will be inspected at suitable intervals to ensure the probability of rotor burst is acceptably low . The LP turbine inspection schedule is based on the probability of rotor failure and subsequent missile generation. Inspections are scheduled before the probability of a failure increases beyond the threshold level. There are several mechanisms1 which may result in rotor failure (Reference 5): 1. Ductile burst resulting from extreme rotor over speed. 2. Fracture resulting from high cycle fatigue cracking. 3. Fracture resulting from low cycle fatigue cracking. 4. Fracture resulting from stress corrosion cracking (SCC). The probability of a turbine missile, assuming that destructive overspeed occurs from simultaneous failure of two valves on the same main steam line to close, has been analyzed in WCAP-11525 and WCAP-16054-P. The frequency of turbine valve testing has been used to maintain the probability of a turbine missile at or below 1 x 10-5 per year. The missile probability is a combination of the probability that the turbine valves remain open when the control system trips them and the probability of the failure of the control system to trip the valves. The failure of the turbine valves themselves is the most important component of that failure. The simultaneous failure of two valves on the same main steam line to close is the most likely event leading to a turbine missile (6). Scenarios where additional valves fail to close have lower probabilities than the failure of the two valves, since more simultaneous failures are required
- 3.5-21 SGS-UFSAR Revision 23 October 17, 2007 For the Salem fully integrated rotor, the assumptions used in WCAP-11525 and WCJ\P-16054-P, that one steam path remaining open through the HP turbine will cause a missile, become invalid. The probabilities calculated for a missile in WCAP-11525 and WCAP-16054-P include a small probability of a turbine missile at running speed and at design overspeed. overspeed probability is calculated. However, in the updates, a destructive This destructive overspeed probability now becomes the probability of reaching equilibrium overspeed. Reference 6 describes the calculation of the destructive and equilibrium overspeeds and the relationship between them. Of the four mechanisms listed above, the potential for stress corrosion cracking has the greatest influence on rotor integrity and therefore forms the basis for setting inspection schedules. The approach is to make a conservative prediction of how a presumed or actual crack will grow and then schedule an inspection prior to the time that the crack grows large enough to be of concern. Analytic components of this approach are: a. Probability of Crack Initiation b. Crack Growth Rate c. Critical Crack Size 'rhe Westinghouse criterion as given in Reference 5 for establishing each of these factors is as follows: a. Probability of Crack Initiation Westinghouse has used inspection data from bu:i.l t-up type rotors to calculate the probability of crack initiation for each disc number in the fully integral rotors. This approach results in conservative estimates since the built-up rotors have stresses and yield strengths which are significantly higher than the fully integral rotors. b. Crack Growth Rate Westinghouse has performed statistical studies using the field data on crack sizes and shapes as related to temperature of operation, crack location, material strength, and environment. They have used parameters for the crack growth rate model which are the same as those used for keyway stress corrosion crack growth rate in built-up rotors. This approach results in conservative estimates since the built-up rotors have stresses and yield strengths which are significantly higher than the fully integral rotors. 3.5-21a SGS-OFSAR Revision 23 October 17, 2007 * * *
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- c. Critical Crack Size Westinghouse used the between fracture toughness and stress intensity to calculate critical crack size. Finite element analysis was used to determine crack tip stress intensity factors for a single thru thickness radial crack emanating from the rim of a disc. The most severe thermal stress encountered alao considered in these calculations. a transient condition was The analysis determined that, at running speed, the stress intensity for all crack depths less than the total depth of the disc was well below the fracture toughness, where the total depth of the disc is defined as the distance from the disc rim to the point where the disc blends into the main body of the rotor. Westinghouse i::onservati vely concluded that the critical crack size is equivalent to the total depth of the disc. A limit load analysis confirmed that ductile fracture of the rotor would not occur under these conditions. Considering the factors above, combined with the number of discs and the proQability of the unit reaching a overspeed condition, analyses were made to determine the probability of rotor bursting. Results indicate that an inspection interval of 30 years will result in a probability of missile generation which is below the threshold value, However, as recommended by Westinghouse, an inspection schedule with shorter inspection intervals will be maintained. With the increased LP rotor component reliability (elimination of shrouds, rivets, and lashing wires), improved LP blade path seating, improved to withstand abnormal operating conditions and the one piece inner cylinder design (minimizes the effects of erosion/corrosion and less susceptible to , Westinghouse recommends a 10 year inspection interval for the LP This recommendation is based on the expected need for maintenance dtie to normal "wear and tear," and is not driven by nuclear safety concerns. 3.5-2lb SGS-UFSAR Revision 21 December 6, 2004 I 3.5.5 References for Section 3.5 1. DELETED 2. DELETED 3. DELETED 4. CT-25243, "Turbine Missile Report (Heavy Disc -Keyplate Design LP' s)" Revision 0, November 1985. 5. Westinghouse Report WSTG-4-P, "Analysis of the Probability of the Generation of Missiles from Fully Integral Nuclear Low Pressure Rotors," October, 1984. 6. Westinghouse Technical Report TM-95185, "Overspeed Analysis for Public Service Electric and Gas -Salem 2 Fully Integral Rotors Dated August 15, 1995" 7. w. B. Cottrell and A. w. Technology," ORNL-NSIC-5, Vol. Laboratory. Savolainen, "U.S. Reactor Containment 1, Chapter 6, Oak Ridge National 8. Siemens Technical Report CT-27336, "Missile Probability Analysis PSEG Nuclear LLC Salem Unit 1", Revision 1, dated November 5, 2003. 9. EC-022 62, "Missile Generation Risk Assessment for Original and Retrofit Nuclear HP Rotors", Siemens Westinghouse Power corporation, December 17, 2002. 10. "Siemens Westinghouse Topical Report, Missile Analysis Methodology for General Electric (GE) Nuclear Steam rotors by Siemens Westinghouse Power Corporation (SWPC), Project No. 721", dated May 16, 2002. 11. WCAP-16054-P, "Probabilistic Analysis Of Reduction in Turbine Valve Test Frequency For Nuclear Plants With Turbines," dated April 2003. 3.5-22 SGS-UFSAR Siemens-Westinghouse BB-95/96 Revision 27 November 25, 2013