LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Sections 5.4.1 to 5.4-14
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- 5.4 REACTOR VESSEL AND APPURTENANCES Section 5.4 is divided into four principal subsections: (1) Design Basis, (2) Description, (3) Evaluation, and (4) Tests and Inspections. 5.4.1 Design Basis The reactor vessel was designed and fabricated to Class A of the ASME Boiler and Pressure Vessel Code,Section III. Material specifications are discussed in Section 5. 2. 3. 1. Fracture toughness of the reactor vessel materials is discussed in Section 5.2.4. Design transients are discussed in Section 5.2.1. 5.4.2 Description The reactor vessel is cylindrical with a welded hemispherical bottom head and a removable, flanged and gasketed, hemispherical upper head. The vessel contains the core, core support structures, control rods, thermal shield, and other parts directly associated with the core. The reactor vessel closure head contains head adaptors. These head adaptors are tubular members, attached by partial penetration welds to the underside of the closure head. The upper end of these adaptors contain acme threads for the assembly of the control rod drive mechanisms and/or instrumentation adaptors. The seal arrangement at the upper end of these adaptors consists of a welded flexible canopy seal. The vessel has inlet and outlet nozzles located in a horizontal plane just below the vessel flange but above the top of the core. Coolant enters the inlet nozzles and flows down the core barrel-vessel wall annulus, turns at the bottom and flows up through the core to the outlet nozzles. The bottom head of the vessel contains penetration nozzles for connection and entry of the nuclear in-core detection instrumentation. Each tube is attached to the inside of the bottom head by a partial penetration weld. 5.4-1 SGS-UFSAR Revision 6 February 15, 1987 The reactor vessel is designed to provide the smallest and most economical volume required to contain the reactor core, control rods, and the necessary supporting and flow-directing internals. Inlet and outlet nozzles are spaced around the vessel. Outlet nozzles are located on opposite sides of the vessel to facilitate opti[!lum layout of the Reactor Coolant System (RCS) equipment. The inlet nozzles are tapered from the coolant loop-vessel interfaces to the vessel inside wall to reduce loop pressure drop. The reactor vessel flange and head are sealed by two hollow metallic 0-rings. Seal leakage is detected by means of two leakoff connections: one between the inner and outer ring, and one outside of the outer 0-ring. Piping and associated valving are provided to direct any leakage to the reactor coolant .drain tank. Leakage will be indicated by a high-temperature alarm from a detector in the leakoff line. Ring forgings have been used in the following areas of the reactor vessel: 1. (Not Used) 2. Vessel flange 3, Eight primary nozzles The cylindrical portion of the reactor vessel below the refueling seal ledge is permanently insulated with a metallic reflective-type insulation supported from the reactor coolant nozzles. This insulation consists of inner and outer sheets of stainless steel spaced 3 inches apart with multilayers of stainless steel as the insulating agent. Removable panels of the metallic reflective type insulation described above are provided for the reactor vessel head and closure region. These panels are supported on the refueling seal ledge and vent shroud support ring. The rest of the closure head is insulated with removable panels of at least 3 inches of the reflective insulation described 5,4-2 SGS-UFSAR Re:vision 22 May 5, 2006
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- or halide free insulating material. The bottom head is also insulated with reflective insulation, but it is not removable . A schematic of the reactor vessel is shown on Figure 5.1-1. The materials of construction are given in Table 5.2-27 and the design parameters are given in Table 5.2-3. The following summarizes those features which preclude wetting of the reactor vessel studs with boric acid. Refueling procedures include removal of studs before moving the head, and replacement only after the head is sealed. Hole plugs with 0-ring seals are placed in the stud holes whenever the head is off. The flange will be dried completely before replacing the head and removing the stud hole plugs. Spilling of boric acid onto the studs during venting will be precluded by the following precautions: 1. Detailed step-by-step venting procedures exist. 2. Personnel have accurate knowledge of RCS level during venting. 3. Only a small quantity of coolant is released at each venting step. 4. The small quantity of released coolant is piped into portable containers for collection. 5.4.3 Evaluation 5.4.3.1 Compliance With 10CFR50, Appendices G and H The Unit 2 reactor vessel was built to the 1965 Edition of the ASME Boiler and Pressure Vessel Code,Section III, and Addenda up to and including the Summer of 1966. Thus the ferritic materials 5.4-3 SGS-UFSAR Revision 6 February 15, 1987 in the reactor vessel were not tested by the vessel fabricator to meet later editions of Section III of the ASME Code as required by IOCFR50, Appendix G. Westinghouse performed tests, as part of the surveillance program, on the reactor vessel intermediate and lower shell course plates, which surround the effective height of the fuel assemblies. Full Charpy test curves were obtained on these plates from specimens oriented normal to the principal rolling direction. A summary of the results of these tests is shown in Table 5.4-1.
- Based on the test results shown in Table 5. 4-1, the core region shell plates have a minimum upper shelf energy greater than 75 ft-lb as required by Appendix G. The stress intensity factors for various reactor vessel locations were not calculated to determine if they are lower than the reference stress intensity factors specified in Appendix G of the Code. Westinghouse has performed these calculations for many older reactor vessels with similar properties and the results have always shown that the calculated stress intensity factors are lower than the reference stress intensity. Thus, based on past experience, Westinghouse is confident that if the calculation was performed, the results would be shown to be acceptable and be lower than the reference stress intensity factors as required by Appendix G of the ASME Code. Heatup and cooldown limit curves, including preoperational system leakage and hydrostatic pressure tests, were determined in accordance with the method described in Appendix G of the ASHE Code. Reactor vessel bolting material tests were not performed to demonstrate conformance with the minimum requirement of 25 mils lateral expansion, and 45 ft-lb at the preload temperature or at the lowest service temperature, whichever is lower. Tests were performed to meet 35 ft-lb at 10°F. The results of the tests are shown in Table 5.4-2. 5.4-4 SGS-UFSAR Revision 6 February 15, 1987 * * *
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- Table 5.4-2 shows that all the bolting material met the 45 ft-lb requirement at 10°F except for one end of one bar, which was used for closure head nuts and washers. It is expected that this bar would exhibit at least 45 ft-lb if tested at 50°F which is considered to be the lowest preload or service temperature. At the time the tests were conducted, lateral expansion measurements were not required. However, it is expected that these materials would exhibit at least 25 mils lateral expansion if tested at 50°F, based on test results from other bolting materials where both impact energy and lateral expansion data were obtained. The reactor vessel is designed to permit a thermal annealing treatment to recover material toughness properties of *ferritic materials in the reactor vessel beltline. Reactor vessel beltline region materials will be monitored by a surveillance program which includes eight surveillance capsules which will receive a neutron flux at least as high, but not more than three times as high as that received by the vessel inner surface. The surveillance program is in compliance with ASTM E-185-73 with the exception of the surveillance weld. The high flux region of the reactor vessel was fabricated from different combinations of weld wire and lots of welding flux for which sufficient tests and chemical analyses are not available to select surveillance weld metal as required by ASTM E-185-7 3. The surveillance weld, although fabricated using the same weld wire and flux lot number as used in core region vertical seams, may not be limiting material in the reactor vessel. Information relative to the changes in fracture toughness due to irradiation are discussed in Section 5.2.4 . 5. 4-5 SGS-UFSAR Revision 19 November 19, 2001 5.4.3.2 Radiation Analysis and Neutron Dosimetry Of Surveillance Capsules Knowledge of the neutron environment within the pressure vessel/surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced properties, changes observed in materials test specimens and the neutron environment (fluence, flux} to which the test specimens are exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship must be established between the environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is analytical techniques monitors contained in normally met by employing a combination of rigorous and measurements obtained with passive neutron flux each of the surveillance capsules. The latter information is derived solely from analysis. This section describes a discrete ordinates Sn transport analysis to determine the fast neutron (E >1. 0 MeV) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules. The analytical data is then used to develop lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. 5.4.3.3 Neutron Transport Methodology Fast neutron exposure calculations for the reactor geometry are carried out using appropriately benchmarked discrete ordinates transport techniques. Plant specific calculations are completed using the DORTtll two-dimensional discrete ordinates code and a benchmarked ENDF/B-VI based cross-section library. Both the BUGLE-93121 and BUGLE-96131 ENDF/B-VI multigroup cross-section libraries provide acceptable results for LWR applications. In the transport analyses, anisotropic scattering is treated with a P3 Legendre expansion at a minimum; and the angular discretization is modeled with at least an S8 order of angular quadrature. 5. 4-6 SGS-UFSAR Revision 19 November 19, 2001 * * *
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- In developing an analytical model of the reactor geometry, nominal design dimensions are employed for the various structural components. Likewise, water temperatures and, hence, coolant density in the reactor core and downcomer regions of the reactor are taken to be representative of nominal full power operating conditions. The reactor core itself is treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, etc. Sensitivities of the analytical results to tolerances in the internals dimensions as well as to fluctuations in water temperature are used to establish the uncertainties associated with the neutron exposure projections at the pressure vessel wall. For each operating fuel cycle, the spatial variation of the neutron source is obtained from a burnup weighted average of the power distributions occurring during the course of the fuel cycle. These spatial distributions include pinwise gradients for all fuel assemblies located at the periphery of the core and include a uniform or flat distribution for fuel assemblies interior to the core. The energy distribution of the source is likewise determined by selecting a fuel burnup representative of conditions averaged over the fuel cycle and an initial fuel assembly enrichment characteristic of the cycle specific core loading pattern. From the average burnup and initial enrichment, a fission split by isotope including 235U, 238U, 238Pu, 239Pu, 240Pu, and 241Pu is derived; and, from that fission split, composite values of energy release per determined. fission, neutron yield per fission, and fission spectrum are These composite values are then combined with the spatial distribution to produce the overall absolute neutron source for use in the transport calculations . 5.4-7 SGS-UFSAR Revision 19 November 19, 2001 5.4.3.4 Neutron Dosimetry The use of passive neutron sensors such as those included in the internal surveillance capsule dosimetry sets does not yield a direct measure of the energy dependent neutron flux level at the measurement location. Rather, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average flux level and, hence, time integrated exposure (fluence) experienced by the sensors may be developed from the measurements only if the sensor characteristics and the parameters of the irradiation are well known. In particular, the following variables are of interest: 1 -The measured specific activity of each sensor 2 -The physical characteristics of each sensor 3 -The operating history of the reactor 4 -The energy response of each sensor 5 -The neutron energy spectrum at the sensor location In this section the procedures used to determine sensor specific activities, to develop reaction rates for individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described. The specific activity of each of the radiometric sensors is determined using established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor is determined by means of a high purity germanium gamma spectrometer. In the case of the surveillance capsule multiple foil sensor sets, these analyses are performed by direct counting of each of the individual wires; or, as in the case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution and chemical separation of cesium from the sensor. The irradiation history of the reactor over its operating lifetime is determined from plant power generation records. In particular, operating data are extracted on a monthly basis from reactor startup to the end of the capsule irradiation period. For the sensor sets utilized in the surveillance capsule irradiations, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. 5.4-8 SGS-UFSAR Revision 19 November 19, 2001 * * *
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- Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, sensor reaction rates referenced to full power operation are then determined. This reaction rate calculation also includes the effects of varying fuel cycle dependent neutron flux at the locations of the sensor sets. Prior to using these measured reaction rates in the determination of fast neutron exposure parameters in terms of neutron fluence (E > 1.0 MeV) and Iron Atom Displacements {dpa), additional corrections are made to the U-238 measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. In addition to the corrections made for the presence of u-235 in the U-238 fission sensors, corrections are also made to both the U-239 and Np-237 sensor reaction rates to account for gamma ray induced fission reactions occurring over the course of the irradiation. These measured reaction rates, including all corrections, along with the results of the plant specific neutron transport calculations are then input to a least squares adjustment procedure to determine a best estimate neutron energy spectrum with associated uncertainties at the surveillance capsule location. Best estimates for key exposure parameters such as neutron fluence (E > 1.0 MeV) or iron atom displacements {dpa) along with their uncertainties are then easily obtained from the adjusted spectrum. The use of measurements in combination with the analytical results reduces the uncertainty in the calculated spectrum and acts to remove biases that may be present in the analytical technique. 5. 4-9 SGS-UFSAR Revision 19 November 19, 2001 5.4,3.5 Vessel Fluence Calculation For Fracture Toughness Determination for Salem Units Fracture toughness calculations Vessels (RPV) were performed development (section 5. 2. 4
- 2) , for Salem Units l and 2 Reactor Pressure for Pressure-Temperature Limit (P-T) Curve Pressurized Thermal Shock (PTS) evaluations (section 5.2.4.5), and Upper Shelf Energy (USE) projections (section 5.2.4.6). A key input in the calculation of RPV fracture toughness was the neutron fluence (n/cm2) at the*vessel clad-bare metal interface and at depths within the vessel wall corresponding to 25 and 75% of the wall thickness for each of the materials constituting the beltline region. The 25 and 75% wall thickness are commonly referred to as the and positions in the vessel wall. The fast neutron fluence at the vessel clad-bare metal*interface is used in the PTS evaluations, the and fluences are used in the P-T curve development, the USE projections use fluence. Vessel fluence calculations were performed using the methodology described in section 5.4.3.3 for end of life (EOL) conditions which defined as 32 EFPY. The results are presented in tables 5.4-7 and 5.4-8. 5. 4-10 SGS-UFSAR Revision 19 November 19, 2001 * * *
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- 5.4.3.6 Measurement of the Initial NOT Temperature of the Reactor Pressure Vessel Base Plate and Forging Material The unirradiated or initial NOT temperature of pressure vessel base plate and forging materials is presently measured by two methods. These methods are the drop weight test per ASTM E20B and the Charpy V-notch impact test (Type A} per ASTM E23. The NDT temperature is defined in ASTM-E208 as "the temperature at which a specimen is broken in a series of tests in which duplicate no break performance occurs at l0°F higher temperature." Using the Charpy V-notch test, the NDT temperature is defined as the temperature at which the energy required to break the specimen is a certain "fixed" value. For SA 5338 Class 1 and A508 Class 2 and Class 3 steel the ASME III Table N-421 specifies an energy value of 30 ft-lb. This value is based on a correlation with the drop weight test and is referred to as the 30 ft-lb "fix." A* curve of the temperature versus energy absorbed in breaking the specimen is plotted. To obtain this curve, 15 tests are performed which include three tests at five different temperatures. The intersection of the energy versus temperature curve with the 30 ft-lb ordinate is designated as the NOT temperature. As part of the Westinghouse surveillance program, Charpy V-impact tests, tensile tests, and fracture mechanics specimens are taken from the core region plates and forgings, and core region weldments including heat-affected zone material. The test locations are similar to those used in the tests by the fabricator at the plate mill. 5.4-11 SGS-UFSAR Revision 19 November 19, 2001 The uncertainties of measurement of the NDT of base plate are: 1. Differences in Charpy V-notch foot-pound values at a given temperature between specimens. 2. Variation of impact properties through plate thickness. The fracture toughness technology for pressure vessels and correlation with service failures based on Charpy V-notch impact data are based on the averaging of data. The Charpy V-notch 30 ft-lb "fix" temperature is based on multiple tests by the material supplier, the fabricator, and by Westinghouse as part of the surveillance program. In the review of available data, differences of 0 to approximately 40°F are observed in comparing curves plotted through the minimum and average values respectively. The value of NDT temperature derived from the average curve is judged to be representative of the material because of the averaging of at least 15 data points, consistent with the specified procedures of ASTM E23. In the case of the assessment of NDT temperature shift due to fast neutron flux, the displacement of transition curves is measured. The selection of maximum, minimum of average curves for this assessment is not significant since like curves are used. There are quantitative differences between the NDT temperature measurement at the surface, 1/4 thickness or the center of a plate. Differences in NDT temperature between 1/4 thickness and the center in heavy plates had been observed to vary from improvement in the NOT temperature to increases up to 85°F. The NDT temperature at the surface had been measured to be as much as 85°F lower than at 1/4 thickness. The 1/4 thickness location is considered conservative since the enhanced metallurgical properties of the surface are not used for the determination of NDT temperature. In addition, the limiting NDT temperature for the reactor vessel after operation is based on the NDT temperature shift due to irradiation. Since the fast 5.4-12 SGS-UFSAR Revision 6 February 15, 1987 * *
- neutron dose is highest at the inner surface, usaqe of the 1/4 thickness NDT temperature criterion is conservative. Data are beinq accumulated on the variation of NDT across heavy section steels at Westinghouse Nuclear Energy Systems. Similarly, the Pressure Vessel Research Committee sponsors an evaluation of properties of pressure vessel steels in plates and forginqs greater than 6 inches thick. Preliminary data show NOT temperature differences between 1/4 thickness and center of less than 20°F. The present criteria of usinq NDT temperature +60°F at the 1/4 thickness location without taking advantage of the enhanced properties at the surface of reactor vessel plates is conservative. To assess any possible uncertainties in the consideration of NDT temperature shift for welda, heat affected zone, and base metal, teat specimens of these three "material typesw are included in the reactor vessel surveillance program. 5.4.4 Tests and Inspections The inspections of the reactor vessel were governed by the ASME Code requirements. The reactor vessel inspections are summarized in Table 5.2-26. A preoperational volumetric examination utilizing ultrasonic techniques was performed on both reactor pressure vessels. This preoperational examination established a base line upon which the results of subsequent inservice inspections can be compared. 5.4-13 SGS-UFSAR Revision 16 January 31, 1998 All of the detailed examinations, as set forth in the Technical Specifications, were performed completely, as part of the preservice inspection program which included where practicable 100 percent of the pressure -retaining welds. Evaluations are made of any indications detected during any of the examinations which exceed the standards for materials and welds specified in the ASME Code,Section III Edition applicable to the construction of the component to determine disposition and/or the need to make repairs. The inservice inspection program is discussed in Section 5.2.8. 5.4.5 References for Section 5.4 1. RSICC Computer Code Collection CCC-650, "DOORS 3.1, One-Two-and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," August 1996. 2. RSIC Data Library Collection DLC-175, "BUGLE-93, Production and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data," April 1994. 3. RSIC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996. 4. Fracture Toughness Analysis for Salem Units 1 and 2 Reactor Pressure Vessels to Protect Against Pressurized Thermal Shock Events 10CFR50. 61, PSE&G Report NFU-060 Revision 0, dated January 10, 1986. 5.4-14 SGS-UFSAR Revision 19 November 19, 2001 * * *