View all text of Subjgrp 108 [§ 50.50 - § 50.69]
§ 50.61a - Alternate fracture toughness requirements for protection against 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 § 50.55a.
(2) RT
(3) RT
(4) RT
(5) RT
(6) RT
(7) φt means fast neutron fluence for neutrons with energies greater than 1.0 MeV. φt is utilized under the provisions of paragraph (g) of this section and has units of n/cm 2.
(8) φ means average neutron flux for neutrons with energies greater than 1.0 MeV. φ is utilized under the provisions of paragraph (g) of this section and has units of n/cm 2/sec.
(9) ΔT
(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) T
(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 § 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 RT
(1) Each licensee shall have projected values of RT
(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 § 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 RT
(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 RT
(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 RT
(4) If the analysis required by paragraph (d)(3) of this section indicates that no reasonably practicable flux reduction program will prevent the RT
(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 RT
(6) If the Director concludes, under paragraph (d)(5) of this section, that operation of the facility with RT
(7) If the limiting RT
(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
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 × 10
(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
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 RT
3 Regulatory Guide 1.190 dated March 2001, establishes acceptable methods for determining neutron flux.
(1) The values of RT
(2) The values of ΔT
(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
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 RT
(i) If a measured value of RT
5 The class of material for estimating RT
(ii) The following generic mean values of RT
(5) The value of T
(6) The licensee shall verify that an appropriate RT
(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 RT
(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 σ 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 (T
(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 ΔT
(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 (T
(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 ΔT
(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 ΔT
(7) The licensee shall report any information that significantly influences the RT
(g) Equations and variables used in this section.
Where:Table 1—PTS Screening Criteria
Product form and RT values | RT | T | 9.5 in. <T ≤10.5 in. | 10.5 in. <T ≤11.5 in. | Axial Weld—RT | 269 | 230 | 222 | Plate—RT | 356 | 305 | 293 | Forging without underclad cracks—RT | 356 | 305 | 293 | Axial Weld and Plate—RT | 538 | 476 | 445 | Circumferential Weld—RT | 312 | 277 | 269 | Forging with underclad cracks—RT | 246 | 241 | 239 |
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6 Wall thickness is the beltline wall thickness including the clad thickness.
7 Forgings without underclad cracks apply to forgings for which no underclad cracks have been detected and that were fabricated in accordance with Regulatory Guide 1.43.
8 RT
9 Forgings with underclad cracks apply to forgings that have detected underclad cracking or were not fabricated in accordance with Regulatory Guide 1.43.
Table 2—Allowable Number of Flaws in Welds
Through-wall extent, TWE [in.] | Maximum number of flaws per 1,000-inches of weld length in the inspection volume that are greater than or equal to TWE | TWE | TWE | 0 | 0.075 | No Limit. | 0.075 | 0.475 | 166.70. | 0.125 | 0.475 | 90.80. | 0.175 | 0.475 | 22.82. | 0.225 | 0.475 | 8.66. | 0.275 | 0.475 | 4.01. | 0.325 | 0.475 | 3.01. | 0.375 | 0.475 | 1.49. | 0.425 | 0.475 | 1.00. | 0.475 | Infinite | 0.00. |
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Table 3—Allowable Number of Flaws in Plates and Forgings
Through-wall extent, TWE [in.] | Maximum number of flaws per 1,000 square-inches of inside surface area in the inspection volume that are greater than or equal to TWE | TWE | TWE | 0 | 0.075 | No Limit. | 0.075 | 0.375 | 8.05. | 0.125 | 0.375 | 3.15. | 0.175 | 0.375 | 0.85. | 0.225 | 0.375 | 0.29. | 0.275 | 0.375 | 0.08. | 0.325 | 0.375 | 0.01. | 0.375 | Infinite | 0.00. |
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Table 4—Conservative Estimates for Chemical Element Weight Percentages
Materials | P | Mn | Plates | 0.014 | 1.45 | Forgings | 0.016 | 1.11 | Welds | 0.019 | 1.63 |
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Table 5—Maximum Heat-Average Residual [ °F] for Relevant Material Groups by Number of Available Data Points (Significance Level = 1%)
Material group | σ [ °F] | Number of available data points | 3 | 4 | 5 | 6 | 7 | 8 | Welds, for Cu >0.072 | 26.4 | 35.5 | 30.8 | 27.5 | 25.1 | 23.2 | 21.7 | Plates, for Cu >0.072 | 21.2 | 28.5 | 24.7 | 22.1 | 20.2 | 18.7 | 17.5 | Forgings, for Cu >0.072 | 19.6 | 26.4 | 22.8 | 20.4 | 18.6 | 17.3 | 16.1 | Weld, Plate or Forging, for Cu ≤0.072 | 18.6 | 25.0 | 21.7 | 19.4 | 17.7 | 16.4 | 15.3 |
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Table 6—T
Number of available
data points (n) | T | 3 | 31.82 | 4 | 6.96 | 5 | 4.54 | 6 | 3.75 | 7 | 3.36 | 8 | 3.14 | 9 | 3.00 | 10 | 2.90 | 11 | 2.82 | 12 | 2.76 | 13 | 2.72 | 14 | 2.68 | 15 | 2.65 |
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Table 7—Threshold Values for the Outlier Deviation Test (Significance Level = 1%)
Number of available data points (n) | Second largest allowable normalized residual value (r*) | Largest allowable normalized residual value (r*) | 3 | 1.55 | 2.71 | 4 | 1.73 | 2.81 | 5 | 1.84 | 2.88 | 6 | 1.93 | 2.93 | 7 | 2.00 | 2.98 | 8 | 2.05 | 3.02 | 9 | 2.11 | 3.06 | 10 | 2.16 | 3.09 | 11 | 2.19 | 3.12 | 12 | 2.23 | 3.14 | 13 | 2.26 | 3.17 | 14 | 2.29 | 3.19 | 15 | 2.32 | 3.21 |
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