(a) Definitions. For the purposes of this section:
(1) ASME Code means the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division I, ``Rules for the Construction of Nuclear Power Plant Components,'' edition and addenda and any limitations and modifications thereof as specified in Sec. 50.55a.
(2) Pressurized Thermal Shock Event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel.
(3) Reactor Vessel Beltline means the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.
(4) RTNDT means the reference temperature for a reactor vessel material, under any conditions. For the reactor vessel beltline materials, RTNDT must account for the effects of neutron radiation.
(5) RTNDT(U) means the reference temperature for a reactor vessel material in the pre-service or unirradiated condition, evaluated according to the procedures in the ASME Code, Paragraph NB-2331 or other methods approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate.
(6) EOL Fluence means the best-estimate neutron fluence projected for a specific vessel beltline material at the clad-base-metal interface on the inside surface of the vessel at the location where the material receives the highest fluence on the expiration date of the operating license.
(7) RTPTS means the reference temperature, RTNDT, evaluated for the EOL Fluence for each of the vessel beltline materials, using the procedures of paragraph (c) of this section.
(8) PTS Screening Criterion means the value of RTPTS for the vessel beltline material above which the plant cannot continue to operate without justification.
(b) Requirements. (1) For each pressurized water nuclear power reactor for which an operating license has been issued under this part or a combined license issued under Part 52 of this chapter, other than a nuclear power reactor facility for which the certification required under Sec. 50.82(a)(1) has been submitted, the licensee shall have projected values of RTPTS or RTMAX-X, accepted by the NRC, for each reactor vessel beltline material. For pressurized water nuclear power reactors for which a construction permit was issued under this part before February 3, 2010 and whose reactor vessel was designed and fabricated to the 1998 Edition or earlier of the ASME Code, the projected values must be in accordance with this section or Sec. 50.61a. For pressurized water nuclear power reactors for which a construction permit is issued under this part after February 3, 2010 and whose reactor vessel is designed and fabricated to an ASME Code after the 1998 Edition, or for which a combined license is issued under Part 52, the projected values must be in accordance with this section. When determining compliance with this section, the assessment of RTPTS must use the calculation procedures described in paragraph (c)(1) and perform the evaluations described in paragraphs (c)(2) and (c)(3) of this section. The assessment must specify the bases for the projected value of RTPTS for each vessel beltline material, including the assumptions regarding core loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation for each beltline material. This assessment must be updated whenever there is a significant \2\ change in projected values of RTPTS, or upon request for a change in the expiration date for operation of the facility.---------------------------------------------------------------------------
\2\ Changes to RTPTS values are considered significant if either the previous value or the current value, or both values, exceed the screening criterion before the expiration of the operating license or the combined license under Part 52 of this chapter, including any renewed term, if applicable for the plant.---------------------------------------------------------------------------
(2) The pressurized thermal shock (PTS) screening criterion is 270 [deg]F for plates, forgings, and axial weld materials, and 300 [deg]F for circumferential weld materials. For the purpose of comparison with this criterion, the value of RTPTS for the reactor vessel must be evaluated according to the procedures of paragraph (c) of this section, for each weld and plate, or forging, in the reactor vessel beltline. RTPTS must be determined for each vessel beltline material using the EOL fluence for that material.
(3) For each pressurized water nuclear power reactor for which the value of RTPTS for any material in the beltline is projected to exceed the PTS screening criterion using the EOL fluence, the licensee shall implement those flux reduction programs that are reasonably practicable to avoid exceeding the PTS screening criterion set forth in paragraph (b)(2) of this section. The schedule for implementation of flux reduction measures may take into account the schedule for submittal and anticipated approval by the Director, Office of Nuclear Reactor Regulation, of detailed plant-specific analyses, submitted to demonstrate acceptable risk with RTPTS above the screening limit due to plant modifications, new information or new analysis techniques.
(4) For each pressurized water nuclear power reactor for which the analysis required by paragraph (b)(3) of this section indicates that no reasonably practicable flux reduction program will prevent RTPTS from exceeding the PTS screening criterion using the EOL fluence, the licensee shall submit a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent potential failure of the reactor vessel as a result of postulated PTS events if continued operation beyond the screening criterion is allowed. In the analysis, the licensee may determine the properties of the reactor vessel materials based on available information, research results, and plant surveillance data, and may use probabilistic fracture mechanics techniques. This analysis must be submitted at least three years before RTPTS is projected to exceed the PTS screening criterion.
(5) After consideration of the licensee's analyses, including effects of proposed corrective actions, if any, submitted in accordance with paragraphs (b)(3) and (b)(4) of this section, the Director, Office of Nuclear Reactor Regulation, may, on a case-by-case basis, approve operation of the facility with RTPTS in excess of the PTS screening criterion. The Director, Office of Nuclear Reactor Regulation, will consider factors significantly affecting the potential for failure of the reactor vessel in reaching a decision.
(6) If the Director, Office of Nuclear Reactor Regulation, concludes, pursuant to paragraph (b)(5) of this section, that operation of the facility with RTPTS in excess of the PTS screening criterion cannot be approved on the basis of the licensee's analyses submitted in accordance with paragraphs (b)(3) and (b)(4) of this section, the licensee shall request and receive approval by the Director, Office of Nuclear Reactor Regulation, prior to any operation beyond the criterion. The request must be based upon modifications to equipment, systems, and operation of the facility in addition to those previously proposed in the submitted analyses that would reduce the potential for failure of the reactor vessel due to PTS events, or upon further analyses based upon new information or improved methodology.
(7) If the limiting RTPTS value of the plant is projected to exceed the screening criteria in paragraph (b)(2), or the criteria in paragraphs (b)(3) through (b)(6) of this section cannot be satisfied, the reactor vessel beltline may be given a thermal annealing treatment to recover the fracture toughness of the material, subject to the requirements of Sec. 50.66. The reactor vessel may continue to be operated only for that service period within which the predicted fracture toughness of the vessel beltline materials satisfy the requirements of paragraphs (b)(2) through (b)(6) of this section, with RTPTS accounting for the effects of annealing and subsequent irradiation.
(c) Calculation of RTPTS. RTPTS must be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for the material. RTPTS must be evaluated using the same procedures used to calculate RTNDT, as indicated in paragraph (c)(1) of this section, and as provided in paragraphs (c)(2) and (c)(3) of this section.
(1) Equation 1 must be used to calculate values of RTNDT for each weld and plate, or forging, in the reactor vessel beltline. Equation 1: RTNDT = RTNDT(U) + M + [Delta]RTNDT
(i) If a measured value of RTNDT(U) is not available, a generic mean value for the class \3\ of material may be used if there are sufficient test results to establish a mean and a standard deviation for the class.---------------------------------------------------------------------------
\3\ The class of material for estimating RTNDT(U) is generally determined for welds by the type of welding flux (Linde 80, or other), and for base metal by the material specification.---------------------------------------------------------------------------
(ii) For generic values of weld metal, the following generic mean values must be used unless justification for different values is provided: 0 [deg]F for welds made with Linde 80 flux, and -56 [deg]F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.
(iii) M means the margin to be added to account for uncertainties in the values of RTNDT(U), copper and nickel contents, fluence and the calculational procedures. M is evaluated from Equation 2.[GRAPHIC] [TIFF OMITTED] TR19DE95.003
(A) In Equation 2, [sigma]U is the standard deviation for RTNDT(U). If a measured value of RTNDT(U) is used, then [sigma]U is determined from the precision of the test method. If a measured value of RTNDT(U) is not available and a generic mean value for that class of materials is used, then [sigma]U is the standard deviation obtained from the set of data used to establish the mean. If a generic mean value given in paragraph (c)(1)(i)(B) of this section for welds is used, then [sigma]U is 17 [deg]F.
(B) In Equation 2, [sigma][Delta] is the standard deviation for [Delta]RTNDT. The value of [sigma][Delta] to be used is 28 [deg]F for welds and 17 [deg]F for base metal; the value of [sigma][Delta] need not exceed one-half of [Delta]RTNDT.
(iv) [Delta]RTNDT is the mean value of the transition temperature shift, or change in RTNDT, due to irradiation, and must be calculated using Equation 3. Equation 3: [Delta]RTNDT = (CF)f(0.28-0.10 log f)
(A) CF ([deg]F) is the chemistry factor, which is a function of copper and nickel content. CF is given in table 1 for welds and in table 2 for base metal (plates and forgings). Linear interpolation is permitted. In tables 1 and 2, ``Wt - % copper'' and ``Wt - % nickel'' are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging. For a weld, the best estimate values will normally be the mean of the measured values for a weld deposit made using the same weld wire heat number as the critical vessel weld. If these values are not available, the upper limiting values given in the material specifications to which the vessel material was fabricated may be used. If not available, conservative estimates (mean plus one standard deviation) based on generic data \4\ may be used if justification is provided. If none of these alternatives are available, 0.35% copper and 1.0% nickel must be assumed.---------------------------------------------------------------------------
\4\ Data from reactor vessels fabricated to the same material specification in the same shop as the vessel in question and in the same time period is an example of ``generic data.''---------------------------------------------------------------------------
(B) f is the best estimate neutron fluence, in units of 10\19\ n/cm\2\ (E greater than 1 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question. As specified in this paragraph, the EOL fluence for the vessel beltline material is used in calculating KRTPTS.
(v) Equation 4 must be used for determining RTPTS using equation 3 with EOL fluence values for determining [Delta]RTPTS. Equation 4: RTPTS = RTNDT(U) + M + [Delta]RTPTS
(2) To verify that RTNDT for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program \5\ results.---------------------------------------------------------------------------
\5\ Surveillance program results means any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR part 50, appendix H.---------------------------------------------------------------------------
(i) Results from the plant-specific surveillance program must be integrated into the RTNDT estimate if the plant-specific surveillance data has been deemed credible as judged by the following criteria:
(A) The materials in the surveillance capsules must be those which are the controlling materials with regard to radiation embrittlement.
(B) Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions must be small enough to permit the determination of the 30-foot-pound temperature unambiguously.
(C) Where there are two or more sets of surveillance data from one reactor, the scatter of [Delta]RTNDT values must be less than 28 [deg]F for welds and 17 [deg]F for base metal. Even if the range in the capsule fluences is large (two or more orders of magnitude), the scatter may not exceed twice those values.
(D) The irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the cladding/base metal interface within 25 [deg]F.
(E) The surveillance data for the correlation monitor material in the capsule, if present, must fall within the scatter band of the data base for the material.
(ii)(A) Surveillance data deemed credible according to the criteria of paragraph (c)(2)(i) of this section must be used to determine a material-specific value of CF for use in Equation 3. A material-specific value of CF is determined from Equation 5.[GRAPHIC] [TIFF OMITTED] TR19DE95.004
(A) Surveillance data deemed credible according to the criteria of paragraph (c)(2)(i) of this section must be used to determine a material-specific value of CF for use in Equation 3. A material-specific value of CF is determined from Equation 5.[GRAPHIC] [TIFF OMITTED] TR19DE95.004
(B) In Equation 5, ``n'' is the number of surveillance data points, ``Ai'' is the measured value of [Delta]RTNDT and ``fi'' is the fluence for each surveillance data point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of [Delta]RTNDT must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld.
(iii) For cases in which the results from a credible plant-specific surveillance program are used, the value of [sigma][Delta] to be used is 14 [deg]F for welds and 8.5 [deg]F for base metal; the value of [sigma][Delta] need not exceed one-half of [Delta]RTNDT.
(iv) The use of results from the plant-specific surveillance program may result in an RTNDT that is higher or lower than those determined in paragraph (c)(1).
(3) Any information that is believed to improve the accuracy of the RTPTS value significantly must be reported to the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate. Any value of RTPTS that has been modified using the procedures of paragraph (c)(2) of this section is subject to the approval of the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, when used as provided in this section.
Table 1--Chemistry Factor for Weld Metals, [deg]F----------------------------------------------------------------------------------------------------------------
Nickel, wt-%
Copper, wt-% ------------------------------------------------
0 0.20 0.40 0.60 0.80 1.00 1.20----------------------------------------------------------------------------------------------------------------0.............................................................. 20 20 20 20 20 20 200.01........................................................... 20 20 20 20 20 20 200.02........................................................... 21 26 27 27 27 27 270.03........................................................... 22 35 41 41 41 41 410.04........................................................... 24 43 54 54 54 54 540.05........................................................... 26 49 67 68 68 68 680.06........................................................... 29 52 77 82 82 82 820.07........................................................... 32 55 85 95 95 95 950.08........................................................... 36 58 90 106 108 108 1080.09........................................................... 40 61 94 115 122 122 1220.10........................................................... 44 65 97 122 133 135 1350.11........................................................... 49 68 101 130 144 148 1480.12........................................................... 52 72 103 135 153 161 1610.13........................................................... 58 76 106 139 162 172 1760.14........................................................... 61 79 109 142 168 182 1880.15........................................................... 66 84 112 146 175 191 2000.16........................................................... 70 88 115 149 178 199 2110.17........................................................... 75 92 119 151 184 207 2210.18........................................................... 79 95 122 154 187 214 2300.19........................................................... 83 100 126 157 191 220 2380.20........................................................... 88 104 129 160 194 223 2450.21........................................................... 92 108 133 164 197 229 2520.22........................................................... 97 112 137 167 200 232 2570.23........................................................... 101 117 140 169 203 236 2630.24........................................................... 105 121 144 173 206 239 2680.25........................................................... 110 126 148 176 209 243 2720.26........................................................... 113 130 151 180 212 246 2760.27........................................................... 119 134 155 184 216 249 2800.28........................................................... 122 138 160 187 218 251 2840.29........................................................... 128 142 164 191 222 254 2870.30........................................................... 131 146 167 194 225 257 2900.31........................................................... 136 151 172 198 228 260 2930.32........................................................... 140 155 175 202 231 263 2960.33........................................................... 144 160 180 205 234 266 2990.34........................................................... 149 164 184 209 238 269 3020.35........................................................... 153 168 187 212 241 272 3050.36........................................................... 158 172 191 216 245 275 3080.37........................................................... 162 177 196 220 248 278 3110.38........................................................... 166 182 200 223 250 281 3140.39........................................................... 171 185 203 227 254 285 3170.40........................................................... 175 189 207 231 257 288 320----------------------------------------------------------------------------------------------------------------
Table 2--Chemistry Factor for Base Metals, [deg]F----------------------------------------------------------------------------------------------------------------
Nickel, wt-%
Copper, wt-% ------------------------------------------------
0 0.20 0.40 0.60 0.80 1.00 1.20----------------------------------------------------------------------------------------------------------------0.............................................................. 20 20 20 20 20 20 200.01........................................................... 20 20 20 20 20 20 200.02........................................................... 20 20 20 20 20 20 200.03........................................................... 20 20 20 20 20 20 200.04........................................................... 22 26 26 26 26 26 260.05........................................................... 25 31 31 31 31 31 310.06........................................................... 28 37 37 37 37 37 370.07........................................................... 31 43 44 44 44 44 440.08........................................................... 34 48 51 51 51 51 510.09........................................................... 37 53 58 58 58 58 580.10........................................................... 41 58 65 65 67 67 670.11........................................................... 45 62 72 74 77 77 770.12........................................................... 49 67 79 83 86 86 860.13........................................................... 53 71 85 91 96 96 960.14........................................................... 57 75 91 100 105 106 1060.15........................................................... 61 80 99 110 115 117 1170.16........................................................... 65 84 104 118 123 125 1250.17........................................................... 69 88 110 127 132 135 1350.18........................................................... 73 92 115 134 141 144 1440.19........................................................... 78 97 120 142 150 154 1540.20........................................................... 82 102 125 149 159 164 1650.21........................................................... 86 107 129 155 167 172 1740.22........................................................... 91 112 134 161 176 181 1840.23........................................................... 95 117 138 167 184 190 1940.24........................................................... 100 121 143 172 191 199 2040.25........................................................... 104 126 148 176 199 208 2140.26........................................................... 109 130 151 180 205 216 2210.27........................................................... 114 134 155 184 211 225 2300.28........................................................... 119 138 160 187 216 233 2390.29........................................................... 124 142 164 191 221 241 2480.30........................................................... 129 146 167 194 225 249 2570.31........................................................... 134 151 172 198 228 255 2660.32........................................................... 139 155 175 202 231 260 2740.33........................................................... 144 160 180 205 234 264 2820.34........................................................... 149 164 184 209 238 268 2900.35........................................................... 153 168 187 212 241 272 2980.36........................................................... 158 173 191 216 245 275 3030.37........................................................... 162 177 196 220 248 278 3080.38........................................................... 166 182 200 223 250 281 3130.39........................................................... 171 185 203 227 254 285 3170.40........................................................... 175 189 207 231 257 288 320---------------------------------------------------------------------------------------------------------------- [60 FR 65468, Dec. 19, 1995, as amended at 61 FR 39300, July 29, 1996; 72 FR 49500, Aug. 28, 2007; 73 FR 5722, Jan. 31, 2008; 75 FR 23, Jan. 4, 2010] Sec. 50.61a Alternate fracture toughness requirements for protectionagainst pressurized thermal shock events.
(a) Definitions. Terms in this section have the same meaning as those presented in 10 CFR 50.61(a), with the exception of the term ``ASME Code.''
(1) ASME Code means the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division I, ``Rules for the Construction of Nuclear Power Plant Components,'' and Section XI, Division I, ``Rules for Inservice Inspection of Nuclear Power Plant Components,'' edition and addenda and any limitations and modifications thereof as specified in Sec. 50.55a.
(2) RTMAX-AW means the material property which characterizes the reactor vessel's resistance to fracture initiating from flaws found along axial weld fusion lines. RTMAX-AW is determined under the provisions of paragraph (f) of this section and has units of [deg]F.
(3) RTMAX-PL means the material property which characterizes the reactor vessel's resistance to fracture initiating from flaws found in plates in regions that are not associated with welds found in plates. RTMAX-PL is determined under the provisions of paragraph (f) of this section and has units of [deg]F.
(4) RTMAX-FO means the material property which characterizes the reactor vessel's resistance to fracture initiating from flaws in forgings that are not associated with welds found in forgings. RTMAX-FO is determined under the provisions of paragraph (f) of this section and has units of [deg]F.
(5) RTMAX-CW means the material property which characterizes the reactor vessel's resistance to fracture initiating from flaws found along the circumferential weld fusion lines. RTMAX-CW is determined under the provisions of paragraph (f) of this section and has units of [deg]F.
(6) RTMAX-X means any or all of the material properties RTMAX-AW, RTMAX-PL, RTMAX-FO, RTMAX-CW, or sum of RTMAX-AW and RTMAX-PL, for a particular reactor vessel.
(7) [phi]t means fast neutron fluence for neutrons with energies greater than 1.0 MeV. [phi]t is utilized under the provisions of paragraph (g) of this section and has units of n/cm\2\.
(8) [phi] means average neutron flux for neutrons with energies greater than 1.0 MeV. [phi] is utilized under the provisions of paragraph (g) of this section and has units of n/cm\2\/sec.
(9) [Delta]T30 means the shift in the Charpy V-notch transition temperature at the 30 ft-lb energy level produced by irradiation. The [Delta]T30 value is utilized under the provisions of paragraph (g) of this section and has units of [deg]F.
(10) Surveillance data means any data that demonstrates the embrittlement trends for the beltline materials, including, but not limited to, surveillance programs at other plants with or without a surveillance program integrated under 10 CFR part 50, appendix H.
(11) TC means cold leg temperature under normal full power operating conditions, as a time-weighted average from the start of full power operation through the end of licensed operation. TC has units of [deg]F.
(12) CRP means the copper rich precipitate term in the embrittlement model from this section. The CRP term is defined in paragraph (g) of this section.
(13) MD means the matrix damage term in the embrittlement model for this section. The MD term is defined in paragraph (g) of this section.
(b) Applicability. The requirements of this section apply to each holder of an operating license for a pressurized water nuclear power reactor whose construction permit was issued before February 3, 2010 and whose reactor vessel was designed and fabricated to the ASME Boiler and Pressure Vessel Code, 1998 Edition or earlier. The requirements of this section may be implemented as an alternative to the requirements of 10 CFR 50.61.
(c) Request for approval. Before the implementation of this section, each licensee shall submit a request for approval in the form of an application for a license amendment in accordance with Sec. 50.90 together with the documentation required by paragraphs (c)(1), (c)(2), and (c)(3) of this section for review and approval by the Director of the Office of Nuclear Reactor Regulation (Director). The application must be submitted for review and approval by the Director at least three years before the limiting RTPTS value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61 for plants licensed under this part.
(1) Each licensee shall have projected values of RTMAX-X for each reactor vessel beltline material for the EOL fluence of the material. The assessment of RTMAX-X values must use the calculation procedures given in paragraphs (f) and (g) of this section. The assessment must specify the bases for the projected value of RTMAX-X for each reactor vessel beltline material, including the assumptions regarding future plant operation (e.g., core loading patterns, projected capacity factors); the copper (Cu), phosphorus (P), manganese (Mn), and nickel (Ni) contents; the reactor cold leg temperature (TC); and the neutron flux and fluence values used in the calculation for each beltline material. Assessments performed under paragraphs (f)(6) and (f)(7) of this section, shall be submitted by the licensee to the Director in its license amendment application to utilize Sec. 50.61a.
(2) Each licensee shall perform an examination and an assessment of flaws in the reactor vessel beltline as required by paragraph (e) of this section. The licensee shall verify that the requirements of paragraphs (e), (e)(1), (e)(2), and (e)(3) of this section have been met. The licensee must submit to the Director, in its application to use Sec. 50.61a, the adjustments made to the volumetric test data to account for NDE-related uncertainties as described in paragraph (e)(1) of this section, all information required by paragraph (e)(1)(iii) of this section, and, if applicable, analyses performed under paragraphs (e)(4), (e)(5) and (e)(6) of this section.
(3) Each licensee shall compare the projected RTMAX-X values for plates, forgings, axial welds, and circumferential welds to the PTS screening criteria in Table 1 of this section, for the purpose of evaluating a reactor vessel's susceptibility to fracture due to a PTS event. If any of the projected RTMAX-X values are greater than the PTS screening criteria in Table 1 of this section, then the licensee may propose the compensatory actions or plant-specific analyses as required in paragraphs (d)(3) through (d)(7) of this section, as applicable, to justify operation beyond the PTS screening criteria in Table 1 of this section.
(d) Subsequent requirements. Licensees who have been approved to use 10 CFR 50.61a under the requirements of paragraph (c) of this section shall comply with the requirements of this paragraph.
(1) Whenever there is a significant change in projected values of RTMAX-X, so that the previous value, the current value, or both values, exceed the screening criteria before the expiration of the plant operating license; or upon the licensee's request for a change in the expiration date for operation of the facility; a re-assessment of RTMAX-X values documented consistent with the requirements of paragraph (c)(1) and (c)(3) of this section must be submitted in the form of a license amendment for review and approval by the Director. If the surveillance data used to perform the re-assessment of RTMAX-X values meet the requirements of paragraph (f)(6)(v) of this section, the licensee shall submit the data and the results of the analysis of the data to the Director for review and approval within one year after the capsule is withdrawn from the vessel. If the surveillance data meet the requirements of paragraph (f)(6)(vi) of this section, the licensee shall submit the data, the results of the analysis of the data, and proposed [Delta]T30 and RTMAX-X values considering the surveillance data in the form of a license amendment to the Director for review and approval within two years after the capsule is withdrawn from the vessel. If the Director does not approve the assessment of RTMAX-X values, then the licensee shall perform the actions required in paragraphs (d)(3) through (d)(7) of this section, as necessary, before operation beyond the PTS screening criteria in Table 1 of this section.
(2) The licensee shall verify that the requirements of paragraphs (e), (e)(1), (e)(2), and (e)(3) of this section have been met. The licensee must submit, within 120 days after completing a volumetric examination of reactor vessel beltline materials as required by ASME Code, Section XI, the adjustments made to the volumetric test data to account for NDE-related uncertainties as described in paragraph (e)(1) of this section and all information required by paragraph (e)(1)(iii) of this section in the form of a license amendment for review and approval by the Director. If a licensee is required to implement paragraphs (e)(4), (e)(5), and (e)(6) of this section, the information required in these paragraphs must be submitted in the form of a license amendment for review and approval by the Director within one year after completing a volumetric examination of reactor vessel materials as required by ASME Code, Section XI.
(3) If the value of RTMAX-X is projected to exceed the PTS screening criteria, then the licensee shall implement those flux reduction programs that are reasonably practicable to avoid exceeding the PTS screening criteria. The schedule for implementation of flux reduction measures may take into account the schedule for review and anticipated approval by the Director of detailed plant-specific analyses which demonstrate acceptable risk with RTMAX-X values above the PTS screening criteria due to plant modifications, new information, or new analysis techniques.
(4) If the analysis required by paragraph (d)(3) of this section indicates that no reasonably practicable flux reduction program will prevent the RTMAX-X value for one or more reactor vessel beltline materials from exceeding the PTS screening criteria, then the licensee shall perform a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent the potential for an unacceptably high probability of failure of the reactor vessel as a result of postulated PTS events. In the analysis, the licensee may determine the properties of the reactor vessel materials based on available information, research results and plant surveillance data, and may use probabilistic fracture mechanics techniques. This analysis and the description of the modifications must be submitted to the Director in the form of a license amendment at least three years before RTMAX-X is projected to exceed the PTS screening criteria.
(5) After consideration of the licensee's analyses, including effects of proposed corrective actions, if any, submitted under paragraphs (d)(3) and (d)(4) of this section, the Director may, on a case-by-case basis, approve operation of the facility with RTMAX-X values in excess of the PTS screening criteria. The Director will consider factors significantly affecting the potential for failure of the reactor vessel in reaching a decision. The Director shall impose the modifications to equipment, systems and operations described to meet paragraph (d)(4) of this section.
(6) If the Director concludes, under paragraph (d)(5) of this section, that operation of the facility with RTMAX-X values in excess of the PTS screening criteria cannot be approved on the basis of the licensee's analyses submitted under paragraphs (d)(3) and (d)(4) of this section, then the licensee shall request a license amendment, and receive approval by the Director, before any operation beyond the PTS screening criteria. The request must be based on modifications to equipment, systems, and operation of the facility in addition to those previously proposed in the submitted analyses that would reduce the potential for failure of the reactor vessel due to PTS events, or on further analyses based on new information or improved methodology. The licensee must show that the proposed alternatives provide reasonable assurance of adequate protection of the public health and safety.
(7) If the limiting RTMAX-X value of the facility is projected to exceed the PTS screening criteria and the requirements of paragraphs (d)(3) through (d)(6) of this section cannot be satisfied, the reactor vessel beltline may be given a thermal annealing treatment under the requirements of Sec. 50.66 to recover the fracture toughness of the material. The reactor vessel may be used only for that service period within which the predicted fracture toughness of the reactor vessel beltline materials satisfy the requirements of paragraphs (d)(1) through (d)(6) of this section, with RTMAX-X values accounting for the effects of annealing and subsequent irradiation.
(e) Examination and flaw assessment requirements. The volumetric examination results evaluated under paragraphs (e)(1), (e)(2), and (e)(3) of this section must be acquired using procedures, equipment and personnel that have been qualified under the ASME Code, Section XI, Appendix VIII, Supplement 4 and Supplement 6, as specified in 10 CFR 50.55a(b)(2)(xv).
(1) The licensee shall verify that the flaw density and size distributions within the volume described in ASME Code, Section XI,\1\ Figures IWB-2500-1 and IWB-2500-2 and limited to a depth from the clad-to-base metal interface of 1-inch or 10 percent of the vessel thickness, whichever is greater, do not exceed the limits in Tables 2 and 3 of this section based on the test results from the volumetric examination. The values in Tables 2 and 3 represent actual flaw sizes. Test results from the volumetric examination may be adjusted to account for the effects of NDE-related uncertainties. The methodology to account for NDE-related uncertainties must be based on statistical data from the qualification tests and any other tests that measure the difference between the actual flaw size and the NDE detected flaw size. Licensees who adjust their test data to account for NDE-related uncertainties to verify conformance with the values in Tables 2 and 3 shall prepare and submit the methodology used to estimate the NDE uncertainty, the statistical data used to adjust the test data and an explanation of how the data was analyzed for review and approval by the Director in accordance with paragraphs (c)(2) and (d)(2) of this section. The verification of the flaw density and size distributions shall be performed line-by-line for Tables 2 and 3. If the flaw density and size distribution exceeds the limitations specified in Tables 2 and 3 of this section, the licensee shall perform the analyses required by paragraph (e)(4) of this section. If analyses are required in accordance with paragraph (e)(4) of this section, the licensee must address the effects on through-wall crack frequency (TWCF) in accordance with paragraph (e)(5) of this section and must prepare and submit a neutron fluence map in accordance with the requirements of paragraph (e)(6) of this section.---------------------------------------------------------------------------
\1\ For forgings susceptible to underclad cracking the determination of the flaw density for that forging from the licensee's inspection shall exclude those indications identified as underclad cracks.---------------------------------------------------------------------------
(i) The licensee shall determine the allowable number of weld flaws in the reactor vessel beltline by multiplying the values in Table 2 of this section by the total length of the reactor vessel beltline welds that were volumetrically inspected and dividing by 1000 inches of weld length.
(ii) The licensee shall determine the allowable number of plate or forging flaws in their reactor vessel beltline by multiplying the values in Table 3 of this section by the total surface area of the reactor vessel beltline plates or forgings that were volumetrically inspected and dividing by 1000 square inches.
(iii) For each flaw detected in the inspection volume described in paragraph (e)(1) with a through-wall extent equal to or greater than 0.075 inches, the licensee shall document the dimensions of the flaw, including through-wall extent and length, whether the flaw is axial or circumferential in orientation and its location within the reactor vessel, including its azimuthal and axial positions and its depth embedded from the clad-to-base metal interface.
(2) The licensee shall identify, as part of the examination required by paragraph (c)(2) of this section and any subsequent ASME Code, Section XI ultrasonic examination of the beltline welds, any flaws within the inspection volume described in paragraph (e)(1) of this section that are equal to or greater than 0.075 inches in through-wall depth, axially-oriented, and located at the clad-to-base metal interface. The licensee shall verify that these flaws do not open to the vessel inside surface using surface or visual examination technique capable of detecting and characterizing service induced cracking of the reactor vessel cladding.
(3) The licensee shall verify, as part of the examination required by paragraph (c)(2) of this section and any subsequent ASME Code, Section XI ultrasonic examination of the beltline welds, that all flaws between the clad-to-base metal interface and three-eights of the reactor vessel thickness from the interior surface are within the allowable values in ASME Code, Section XI, Table IWB-3510-1.
(4) The licensee shall perform analyses to demonstrate that the reactor vessel will have a TWCF of less than 1 x 10-6 per reactor year if the ASME Code, Section XI volumetric examination required by paragraph (c)(2) or (d)(2) of this section indicates any of the following:
(i) The flaw density and size in the inspection volume described in paragraph (e)(1) exceed the limits in Tables 2 or 3 of this section;
(ii) There are axial flaws that penetrate through the clad into the low alloy steel reactor vessel shell, at a depth equal to or greater than 0.075 inches in through-wall extent from the clad-to-base metal interface; or
(iii) Any flaws between the clad-to-base metal interface and three-eighths \2\ of the vessel thickness exceed the size allowable in ASME Code, Section XI, Table IWB-3510-1.---------------------------------------------------------------------------
\2\ Because flaws greater than three-eights of the vessel wall thickness from the inside surface do not contribute to TWCF, flaws greater than three-eights of the vessel wall thickness from the inside surface need not be analyzed for their contribution to PTS.---------------------------------------------------------------------------
(5) The analyses required by paragraph (e)(4) of this section must address the effects on TWCF of the known sizes and locations of all flaws detected by the ASME Code, Section XI, Appendix VIII, Supplement 4 and Supplement 6 ultrasonic examination out to three-eights of the vessel thickness from the inner surface, and may also take into account other reactor vessel-specific information, including fracture toughness information.
(6) For all flaw assessments performed in accordance with paragraph (e)(4) of this section, the licensee shall prepare and submit a neutron fluence map, projected to the date of license expiration, for the reactor vessel beltline clad-to-base metal interface and indexed in a manner that allows the determination of the neutron fluence at the location of the detected flaws.
(f) Calculation of RTMAX-X values. Each licensee shall calculate RTMAX-X values for each reactor vessel beltline material using [phi]t. The neutron flux ([phi][t]), must be calculated using a methodology that has been benchmarked to experimental measurements and with quantified uncertainties and possible biases.\3\---------------------------------------------------------------------------
\3\ Regulatory Guide 1.190 dated March 2001, establishes acceptable methods for determining neutron flux.---------------------------------------------------------------------------
(1) The values of RTMAX-AW, RTMAX-PL, RTMAX-FO, and RTMAX-CW must be determined using Equations 1 through 4 of this section. When calculating RTMAX-AW using Equation 1, RTMAX-AW is the maximum value of (RTNDT(U) + [Delta]T30) for the weld and for the adjoining plates. When calculating RTMAX-CW using Equation 4, RTMAX-CW is the maximum value of (RTNDT(U) + [Delta]T30) for the circumferential weld and for the adjoining plates or forgings.
(2) The values of [Delta]T30 must be determined using Equations 5, 6 and 7 of this section, unless the conditions specified in paragraph (f)(6)(v) of this section are not met, for each axial weld, plate, forging, and circumferential weld. The [Delta]T30 value for each axial weld calculated as specified by Equation 1 of this section must be calculated for the maximum fluence ([phi]tAXIAL-WELD) occurring along a particular axial weld at the clad-to-base metal interface. The [Delta]T30 value for each plate calculated as specified by Equation 1 of this section must also be calculated using the same value of [phi]tAXIAL-WELD used for the axial weld. The [Delta]T30 values in Equation 1 shall be calculated for the weld itself and each adjoining plate. The [Delta]T30 value for each plate or forging calculated as specified by Equations 2 and 3 of this section must be calculated for the maximum fluence ([phi]tMAX) occurring at the clad-to-base metal interface over the entire area of each plate or forging. In Equation 4, the fluence ([phi]tWELD-CIRC) value used for calculating the plate, forging, and circumferential weld [Delta]T30 value is the maximum fluence occurring for each material along the circumferential weld at the clad-to-base metal interface. The [Delta]T30 values in Equation 4 shall be calculated for the circumferential weld and for the adjoining plates or forgings. If the conditions specified in paragraph (f)(6)(v) of this section are not met, licensees must propose [Delta]T30 and RTMAX-X values in accordance with paragraph (f)(6)(vi) of this section.
(3) The values of Cu, Mn, P, and Ni in Equations 6 and 7 of this section must represent the best estimate values for the material. For a plate or forging, the best estimate value is normally the mean of the measured values for that plate or forging. For a weld, the best estimate value is normally the mean of the measured values for a weld deposit made using the same weld wire heat number as the critical vessel weld. If these values are not available, either the upper limiting values given in the material specifications to which the vessel material was fabricated, or conservative estimates (i.e., mean plus one standard deviation) based on generic data \4\ as shown in Table 4 of this section for P and Mn, must be used.---------------------------------------------------------------------------
\4\ Data from reactor vessels fabricated to the same material specification in the same shop as the vessel in question and in the same time is an example of ``generic data.''---------------------------------------------------------------------------
(4) The values of RTNDT(U) must be evaluated according to the procedures in the ASME Code, Section III, paragraph NB-2331. If any other method is used for this evaluation, the licensee shall submit the proposed method for review and approval by the Director along with the calculation of RTMAX-X values required in paragraph (c)(1) of this section.
(i) If a measured value of RTNDT(U) is not available, a generic mean value of RTNDT(U) for the class \5\ of material must be used if there are sufficient test results to establish a mean.---------------------------------------------------------------------------
\5\ The class of material for estimating RTNDT(U) must be determined by the type of welding flux (Linde 80, or other) for welds or by the material specification for base metal.---------------------------------------------------------------------------
(ii) The following generic mean values of RTNDT(U) must be used unless justification for different values is provided: 0 [deg]F for welds made with Linde 80 weld flux; and -56 [deg]F for welds made with Linde 0091, 1092, and 124 and ARCOS B-5 weld fluxes.
(5) The value of TC in Equation 6 of this section must represent the time-weighted average of the reactor cold leg temperature under normal operating full power conditions from the beginning of full power operation through the end of licensed operation.
(6) The licensee shall verify that an appropriate RTMAX-X value has been calculated for each reactor vessel beltline material by considering plant-specific information that could affect the use of the model (i.e., Equations 5, 6 and 7) of this section for the determination of a material's [Delta]T30 value.
(i) The licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data satisfy the criteria described in paragraphs (f)(6)(i)(A) and (f)(6)(i)(B) of this section:
(A) The surveillance material must be a heat-specific match for one or more of the materials for which RTMAX-X is being calculated. The 30-foot-pound transition temperature must be determined as specified by the requirements of 10 CFR part 50, Appendix H.
(B) If three or more surveillance data points measured at three or more different neutron fluences exist for a specific material, the licensee shall determine if the surveillance data show a significantly different trend than the embrittlement model predicts. This must be achieved by evaluating the surveillance data for consistency with the embrittlement model by following the procedures specified by paragraphs (f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of this section. If fewer than three surveillance data points exist for a specific material, then the embrittlement model must be used without performing the consistency check.
(ii) The licensee shall estimate the mean deviation from the embrittlement model for the specific data set (i.e., a group of surveillance data points representative of a given material). The mean deviation from the embrittlement model for a given data set must be calculated using Equations 8 and 9 of this section. The mean deviation for the data set must be compared to the maximum heat-average residual given in Table 5 or derived using Equation 10 of this section. The maximum heat-average residual is based on the material group into which the surveillance material falls and the number of surveillance data points. For surveillance data sets with greater than 8 data points, the maximum credible heat-average residual must be calculated using Equation 10 of this section. The value of [sigma] used in Equation 10 of this section must be obtained from Table 5 of this section.
(iii) The licensee shall estimate the slope of the embrittlement model residuals (estimated using Equation 8) plotted as a function of the base 10 logarithm of neutron fluence for the specific data set. The licensee shall estimate the T-statistic for this slope (TSURV) using Equation 11 and compare this value to the maximum permissible T-statistic (TMAX) in Table 6. For surveillance data sets with greater than 15 data points, the TMAX value must be calculated using Student's T distribution with a significance level ([alpha]) of 1 percent for a one-tailed test.
(iv) The licensee shall estimate the two largest positive deviations (i.e., outliers) from the embrittlement model for the specific data set using Equations 8 and 12. The licensee shall compare the largest normalized residual (r *) to the appropriate allowable value from the third column in Table 7 and the second largest normalized residual to the appropriate allowable value from the second column in Table 7.
(v) The [Delta]T30 value must be determined using Equations 5, 6, and 7 of this section if all three of the following criteria are satisfied:
(A) The mean deviation from the embrittlement model for the data set is equal to or less than the value in Table 5 or the value derived using Equation 10 of this section;
(B) The T-statistic for the slope (TSURV) estimated using Equation 11 is equal to or less than the Maximum permissible T-statistic (TMAX) in Table 6; and
(C) The largest normalized residual value is equal to or less than the appropriate allowable value from the third column in Table 7 and the second largest normalized residual value is equal to or less than the appropriate allowable value from the second column in Table 7. If any of these criteria is not satisfied, the licensee must propose [Delta]T30 and RTMAX-X values in accordance with paragraph (f)(6)(vi) of this section.
(vi) If any of the criteria described in paragraph (f)(6)(v) of this section are not satisfied, the licensee shall review the data base for that heat in detail, including all parameters used in Equations 5, 6, and 7 of this section and the data used to determine the baseline Charpy V-notch curve for the material in an unirradiated condition. The licensee shall submit an evaluation of the surveillance data to the NRC and shall propose [Delta]T30 and RTMAX-X values, considering their plant-specific surveillance data, to be used for evaluation relative to the acceptance criteria of this rule. These evaluations must be submitted for review and approval by the Director in the form of a license amendment in accordance with the requirements of paragraphs (c)(1) and (d)(1) of this section.
(7) The licensee shall report any information that significantly influences the RTMAX-X value to the Director in accordance with the requirements of paragraphs (c)(1) and (d)(1) of this section.
(g) Equations and variables used in this section.
[GRAPHIC] [TIFF OMITTED] TR03FE10.000
[GRAPHIC] [TIFF OMITTED] TR03FE10.001
[GRAPHIC] [TIFF OMITTED] TR03FE10.002
[GRAPHIC] [TIFF OMITTED] TR03FE10.003
[GRAPHIC] [TIFF OMITTED] TR03FE10.004 [GRAPHIC] [TIFF OMITTED] TR03FE10.005 [GRAPHIC] [TIFF OMITTED] TR03FE10.006 Where: P [wt-&%] = phosphorus contentMn [wt-%] = manganese contentNi [wt-%] = nickel contentCu [wt-%] = copper contentA = 1.140 x 10-7 for forgingsA = 1.561 x 10-7 for platesA = 1.417 x 10-7 for weldsB = 102.3 for forgingsB = 102.5 for plates in non-Combustion Engineering manufactured vesselsB = 135.2 for plates in Combustion Engineering vesselsB = 155.0 for welds [phis]te = [phis]t for [phis] =4.39 x 10\10\ n/cm\2\/sec[phis]te = [phis]t x (4.39 x 10\10\/[phis])\0.2595\ for [phis] <4.39 x 10\10\ n/cm\2\/sec Where: [phis] [n/cm\2\/sec] = average neutron fluxt [sec] = time that the reactor has been in full power operation[phis]t [n/cm\2\] = [phis] x t f(Cue,P) = 0 for Cu <=0.072f(Cue,P) = [Cue-0.072]\0.668\ for Cu 0.072 and P <=0.008f(Cue,P) = [Cue-0.072 + 1.359 x (P-0.008)]\0.668\ for Cu 0.072 and P 0.008 Where: Cue = 0 for Cu <=0.072Cue = MIN (Cu, maximum Cue) for Cu
0.072maximum Cue = 0.243 for Linde 80 weldsmaximum Cue = 0.301 for all other materials g(Cue,Ni,[phis]te) = 0.5 + (0.5 x tanh {[log10([phis]te) + (1.1390 x Cue)-(0.448 x Ni)-18.120]/0.629{time} Equation 8: Residual (r) = measured [Delta]T30-predicted [Delta]T30 (by Equations 5, 6 and 7)[GRAPHIC] [TIFF OMITTED] TR03FE10.007 Equation 10: Maximum credible heat-average residual = 2.33[sigma]/n\0.5\ Where: n = number of surveillance data points (sample size) in the specific
data set[sigma] = standard deviation of the residuals about the model for a
relevant material group given in Table 5.
[GRAPHIC] [TIFF OMITTED] TR03FE10.008
Where: m is the slope of a plot of all of the r values (estimated using
Equation 8) versus the base 10 logarithm of the neutron
fluence for each r value. The slope shall be estimated using
the method of least squares.se(m) is the least squares estimate of the standard-error associated
with the estimated slope value m. [GRAPHIC] [TIFF OMITTED] TR03FE10.009 Where: r is defined using Equation 8 and [sigma] is given in Table 5
Table 1--PTS Screening Criteria----------------------------------------------------------------------------------------------------------------
RTMAX X limits [ [deg]F] for different vessel wall
thicknesses \6\ (TWALL)
Product form and RTMAX X values --------------------------------------------------------
9.5 in. 0.072..... 26.4 35.5 30.8 27.5 25.1 23.2 21.7Plates, for Cu 0.072.... 21.2 28.5 24.7 22.1 20.2 18.7 17.5Forgings, for Cu 0.072.. 19.6 26.4 22.8 20.4 18.6 17.3 16.1Weld, Plate or Forging, for Cu 18.6 25.0 21.7 19.4 17.7 16.4 15.3
<=0.072...........................----------------------------------------------------------------------------------------------------------------
Table 6--TMAX Values for the Slope Deviation Test (Significance Level =
1%)------------------------------------------------------------------------
Number of available data points (n) TMAX------------------------------------------------------------------------3....................................................... 31.824....................................................... 6.965....................................................... 4.546....................................................... 3.757....................................................... 3.368....................................................... 3.149....................................................... 3.0010...................................................... 2.9011...................................................... 2.8212...................................................... 2.7613...................................................... 2.7214...................................................... 2.6815...................................................... 2.65------------------------------------------------------------------------
Table 7--Threshold Values for the Outlier Deviation Test (Significance
Level = 1%)------------------------------------------------------------------------
Second largest Largest
allowable allowable
Number of available data points (n) normalized normalized
residual value residual value
(r*) (r*)------------------------------------------------------------------------3....................................... 1.55 2.714....................................... 1.73 2.815....................................... 1.84 2.886....................................... 1.93 2.937....................................... 2.00 2.988....................................... 2.05 3.029....................................... 2.11 3.0610...................................... 2.16 3.0911...................................... 2.19 3.1212...................................... 2.23 3.1413...................................... 2.26 3.1714...................................... 2.29 3.1915...................................... 2.32 3.21------------------------------------------------------------------------ [75 FR 23, Jan. 4, 2010, as amended at 75 FR 5495, Feb. 3, 2010; 75 FR 10411, Mar. 8, 2010; 75 FR 72653, Nov. 26, 2010]