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Created February 15, 2016 05:11
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10 CFR 50 patch
Raw
2015.patch
--- .eregs_index/annual/10/50/2015 2016-02-15 05:04:10.250647377 +0000
+++ /vagrant/2015 2016-02-15 05:03:51.000000000 +0000
@@ -322,10 +322,8 @@
<E T="03">Common defense and security</E> means the common defense and security of the United States.</P>
<P>
<E T="03">Construction</E> or <E T="03">constructing</E> means, for the purposes of &#167; 50.55(e), the analysis, design, manufacture, fabrication, quality assurance, placement, erection, installation, modification, inspection, or testing of a facility or activity which is subject to the regulations in this part and consulting services related to the facility or activity that are safety related.</P>
- <P>
- <E T="03">Controls</E> when used with respect to nuclear reactors means apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor.</P>
- <P>
- <E T="03">Controls</E> when used with respect to any other facility means apparatus and mechanisms, the manipulation of which could affect the chemical, physical, metallurgical, or nuclear process of the facility in such a manner as to affect the protection of health and safety against radiation.</P>
+ <P>Controls when used with respect to nuclear reactors means apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor.</P>
+ <P>Controls when used with respect to any other facility means apparatus and mechanisms, the manipulation of which could affect the chemical, physical, metallurgical, or nuclear process of the facility in such a manner as to affect the protection of health and safety against radiation.</P>
<P>
<E T="03">Cost of service regulation</E> means the traditional system of rate regulation, or similar regulation, including &#8220;price cap&#8221; or &#8220;incentive&#8221; regulation, in which a rate regulatory authority generally allows an electric utility to charge its customers the reasonable and prudent costs of providing electricity services, including capital, operations, maintenance, fuel, decommissioning, and other costs required to provide such services.</P>
<P>
@@ -990,7 +988,7 @@
<SUBJECT>Technical specifications.</SUBJECT>
<P>(a)(1) Each applicant for a license authorizing operation of a production or utilization facility shall include in his <PRTPAGE P="866"/>application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.</P>
- <P>(a)(2) Each applicant for a design certification or manufacturing license under part 52 of this chapter shall include in its application proposed generic technical specifications in accordance with the requirements of this section for the portion of the plant that is within the scope of the design certification or manufacturing license application.</P>
+ <P>(2) Each applicant for a design certification or manufacturing license under part 52 of this chapter shall include in its application proposed generic technical specifications in accordance with the requirements of this section for the portion of the plant that is within the scope of the design certification or manufacturing license application.</P>
<P>(b) Each license authorizing operation of a production or utilization facility of a type described in &#167; 50.21 or &#167; 50.22 will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &#167; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.</P>
<P>(c) Technical specifications will include items in the following categories:</P>
<P>(1) <E T="03">Safety limits, limiting safety system settings, and limiting control settings.</E> (i)(A) Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission. The licensee shall retain the record of the results of each review until the Commission terminates the license for the reactor, except for nuclear power reactors licensed under &#167; 50.21(b) or &#167; 50.22 of this part. For these reactors, the licensee shall notify the Commission as required by &#167; 50.72 and submit a Licensee Event Report to the Commission as required by &#167; 50.73. Licensees in these cases shall retain the records of the review for a period of three years following issuance of a Licensee Event Report.</P>
@@ -1381,7 +1379,7 @@
<SECTION>
<SECTNO>&#167; 50.54</SECTNO>
<SUBJECT>Conditions of licenses.</SUBJECT>
- <P>The following paragraphs of this section, with the exception of paragraphs (r) and (gg), and the applicable requirements of 10 CFR 50.55a, are conditions in every nuclear power reactor operating license issued under this part. The following paragraphs with the exception of paragraph (r), (s), and (u) of this section are conditions in every combined license issued under part 52 of this chapter, provided, however, that paragraphs (i) introductory text, (i)(1), (j), (k), (l), (m), (n), (q), (w), (x), (y), (z), and (hh) of this section are only applicable after the Commission makes the finding under &#167; 52.103(g) of this chapter.</P>
+ <P>The following paragraphs of this section, with the exception of paragraphs (r) and (gg), and the applicable requirements of 10 CFR 50.55a, are conditions in every nuclear power reactor operating license issued under this part. The following paragraphs with the exception of paragraph (r), (s), and (u) of this section are conditions in every combined license issued under part 52 of this chapter, provided, however, that paragraphs (i), (i)(1), (j), (k), (l), (m), (n), (q), (w), (x), (y), (z), and (hh) of this section are only applicable after the Commission makes the finding under &#167; 52.103(g) of this chapter.</P>
<P>(a)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality assurance criteria in appendix B of this part shall implement, under &#167; 50.34(b)(6)(ii) or &#167; 52.79 of this chapter, the quality assurance program <PRTPAGE P="887"/>described or referenced in the safety analysis report, including changes to that report. However, a holder of a combined license under part 52 of this chapter shall implement the quality assurance program described or referenced in the safety analysis report applicable to operation 30 days prior to the scheduled date for the initial loading of fuel.</P>
<P>(2) Each licensee described in paragraph (a)(1) of this section shall, by June 10, 1983, submit to the appropriate NRC Regional Office shown in appendix D of part 20 of this chapter the current description of the quality assurance program it is implementing for inclusion in the Safety Analysis Report, unless there are no changes to the description previously accepted by NRC. This submittal must identify changes made to the quality assurance program description since the description was submitted to NRC. (Should a licensee need additional time beyond June 10, 1983 to submit its current quality assurance program description to NRC, it shall notify the appropriate NRC Regional Office in writing, explain why additional time is needed, and provide a schedule for NRC approval showing when its current quality assurance program description will be submitted.)</P>
@@ -1616,7 +1614,6 @@
<P>(ii) Operations to mitigate fuel damage; and</P>
<P>(iii) Actions to minimize radiological release.</P>
<P>(3) This section does not apply to a nuclear power plant for which the certifications required under &#167; 50.82(a) or &#167; 52.110(a)(1) of this chapter have been submitted.</P>
- <P>(ii) [Reserved]</P>
<P>(jj) Structures, systems, and components subject to the codes and standards in 10 CFR 50.55a must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.</P>
<CITA>[21 FR 355, Jan. 19, 1956]</CITA>
<EDNOTE>
@@ -1705,9 +1702,8 @@
<SUBJECT>Codes and standards.</SUBJECT>
<P>(a) <E T="03">Documents approved for incorporation by reference.</E> The standards listed in this paragraph have been approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51. The standards are available for inspection at the NRC Technical Library, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-6239; or at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 202-741-6030 or go to <E T="03">http://www.archives.gov/federal-register/cfr/ibr-locations.html.</E>
</P>
- <P>(1) <E T="03">American Society of Mechanical Engineers</E> (<E T="03">ASME),</E> Three Park Avenue, New York, NY 10016; telephone: </P>
- <FP>1-800-843-2763; <E T="03">http://www.asme.org/Codes/.</E>
- </FP>
+ <P>(1) <E T="03">American Society of Mechanical Engineers</E> (<E T="03">ASME),</E> Three Park Avenue, New York, NY 10016; telephone: 1-800-843-2763; <E T="03">http://www.asme.org/Codes/.</E>
+ </P>
<P>(i) <E T="03">ASME Boiler and Pressure Vessel Code, Section III.</E> The editions and addenda for Section III of the ASME Boiler and Pressure Vessel Code are listed below, but limited to those provisions identified in paragraph (b)(1) of this section.</P>
<P>(A) &#8220;Rules for Construction of Nuclear Vessels:&#8221;</P>
<P>(<E T="03">1</E>) 1963 Edition,</P>
@@ -2008,31 +2004,6 @@
<P>(<E T="03">i</E>) The clad-to-base-metal-interface and the adjacent metal to a depth of 15 percent T (measured from the clad-to-base-metal-interface) must be examined from one radial and two opposing circumferential directions using a procedure and personnel qualified in accordance with Supplement 4 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(B) and (C) of this section, for examinations performed in the radial direction, and Supplement 5 to Appendix VIII, as conditioned by paragraph (b)(2)(xv)(J) of this section, for examinations performed in the circumferential direction.</P>
<P>(<E T="03">ii</E>) The examination volume not addressed by paragraph (b)(2)(xv)(K)(<E T="03">3</E>)(<E T="03">i</E>) of this section must be examined in a minimum of one radial direction using a procedure and personnel qualified for single sided examination in accordance with Supplement 6 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (G) of this section.</P>
<P>(<E T="03">4</E>) Table VIII-S7-1, &#8220;Flaw Locations and Orientations,&#8221; Supplement 7 to Appendix VIII, is conditioned as follows:</P>
- <GPOTABLE CDEF="s75,20C,20C" COLS="3" OPTS="L2">
- <TTITLE>Table VIII&#8212;S7-1&#8212;Modified</TTITLE>
- <TDESC>[Flaw locations and orientations]</TDESC>
- <BOXHD>
- <CHED H="1"/>
- <CHED H="1">Parallel<LI>to weld</LI>
- </CHED>
- <CHED H="1">Perpendicular<LI>to weld</LI>
- </CHED>
- </BOXHD>
- <ROW>
- <ENT I="01">Inner 15 percent</ENT>
- <ENT>X</ENT>
- <ENT>X</ENT>
- </ROW>
- <ROW>
- <ENT I="01">Outside Diameter Surface</ENT>
- <ENT>X</ENT>
- </ROW>
- <ROW>
- <ENT I="01">Subsurface</ENT>
- <ENT>X</ENT>
- <ENT/>
- </ROW>
- </GPOTABLE>
<P>(L) <E T="03">Specimen set and qualification: Twelfth provision.</E> As a condition to the requirements of Supplement 8, Subparagraph 1.1(c), to Appendix VIII, notches may be located within one diameter of each end of the bolt or stud.</P>
<P>(M) <E T="03">Specimen set and qualification: Thirteenth provision.</E> When implementing Supplement 12 to Appendix VIII, only the provisions related to the coordinated implementation of Supplement 3 to Supplement 2 performance demonstrations are to be applied.</P>
<P>(xvi) <E T="03">Section XI condition: Appendix VIII single side ferritic vessel and piping and stainless steel piping examinations.</E> When applying editions and addenda prior to the 2007 Edition of Section XI, the following conditions apply.</P>
@@ -2055,10 +2026,6 @@
<P>(xxvii) <E T="03">Section XI condition: Removal of insulation.</E> When performing visual examination in accordance with IWA-5242 of Section XI of the ASME BPV Code, 2003 Addenda through the 2006 Addenda, or IWA-5241 of the 2007 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, insulation must be removed from 17-4 PH or 410 stainless steel studs or bolts aged at a temperature below 1100 &#176;F or having a Rockwell Method C hardness value above 30, and from A-286 stainless steel studs or bolts preloaded to 100,000 pounds per square inch or higher.</P>
<P>(xxviii) <E T="03">Section XI condition: Analysis of flaws.</E> Licensees using ASME BPV Code, Section XI, Appendix A, must use the following conditions when implementing Equation (2) in A-4300(b)(1):
</P>
- <EXTRACT>
- <P>For R &lt; 0, &#916;K<E T="52">I</E> depends on the crack depth (a), and the flow stress (&#963;<E T="52">f</E>). The flow stress is defined by &#963;<E T="52">f</E> = 1/2(&#963;<E T="52">ys</E> + &#963;<E T="52">ult</E>), where &#963;<E T="52">ys</E> is the yield strength and &#963;<E T="52">ult</E> is the ultimate tensilestrength in units ksi (MPa) and (a) is in units in. (mm). For &#8722;2 &#8804; R &#8804; 0 and K<E T="52">max</E>&#8722; K<E T="52">min</E> &#8804; 0.8 &#215; 1.12 &#963;<E T="52">f</E>&#8730;(&#960;a), S = 1 and &#916;K<E T="52">I</E> = K<E T="52">max.</E> For R &lt; &#8722;2 and K<E T="52">max</E>&#8722; K<E T="52">min</E> &#8804; 0.8&#215; 1.12 &#963;<E T="52">f</E>&#8730;(&#960;a), S = 1 and &#916;K<E T="52">I</E> = (1 &#8722; R) K<E T="52">max</E>/3. For R &lt; 0 and K<E T="52">max</E> &#8722; K<E T="52">min</E> &gt; 0.8 &#215; 1.12&#963;<E T="52">f</E>&#8730;(&#960;a), S = 1 and &#916;K<E T="52">I</E> = K<E T="52">max</E>&#8722;K<E T="52">min.</E></P>
- </EXTRACT>
-
<P>(xxix) <E T="03">Section XI condition: Nonmandatory Appendix R.</E> Nonmandatory Appendix R, &#8220;Risk-Informed Inspection Requirements for Piping,&#8221; of Section XI, 2005 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, may not be implemented without prior NRC authorization of the proposed alternative in accordance with paragraph (z) of this section.</P>
<P>(3) <E T="03">Conditions on ASME OM Code.</E> As used in this section, references to the OM Code refer to the ASME Code for Operation and Maintenance of Nuclear Power Plants, Subsections ISTA, ISTB, ISTC, ISTD, Mandatory Appendices I and II, and Nonmandatory Appendices A through H and J, including the 1995 Edition through the 2006 Addenda, subject to the following conditions:</P>
<P>(i) <E T="03">OM condition: Quality assurance.</E> When applying editions and addenda of the OM Code, the requirements of NQA-1, &#8220;Quality Assurance Requirements for Nuclear Facilities,&#8221; 1979 Addenda, are acceptable as permitted by ISTA 1.4 of the 1995 Edition through 1997 Addenda or ISTA-1500 of the 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(iv) of this section, provided the licensee uses its 10 CFR part 50, Appendix B, quality assurance program in conjunction with the OM Code requirements. Commitments contained <PRTPAGE P="915"/>in the licensee's quality assurance program description that are more stringent than those contained in NQA-1 govern OM Code activities. If NQA-1 and the OM Code do not address the commitments contained in the licensee's Appendix B quality assurance program description, the commitments must be applied to OM Code activities.</P>
@@ -2188,27 +2155,10 @@
<P>(1) [Reserved]</P>
<P>(2) <E T="03">Protection systems.</E> For nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements stated in either IEEE Std. 279, &#8220;Criteria for Protection Systems for Nuclear Power Generating Stations,&#8221; or in IEEE Std. 603-1991, &#8220;Criteria for Safety Systems for Nuclear Power Generating Stations,&#8221; and the correction sheet dated January 30, 1995. For nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.</P>
<P>(3) <E T="03">Safety systems.</E> Applications filed on or after May 13, 1999, for construction permits and operating licenses under this part, and for design approvals, design certifications, and combined licenses under part 52 of this chapter, must meet the requirements for safety systems in IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.</P>
- <P>(i)-(y) [Reserved]</P>
- <P>(z) <E T="03">Alternatives to codes and standards requirements.</E> Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor <PRTPAGE P="928"/>Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:</P>
+ <P>(i) <E T="03">Alternatives to codes and standards requirements.</E> Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor <PRTPAGE P="928"/>Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:</P>
<P>(1) <E T="03">Acceptable level of quality and safety.</E> The proposed alternative would provide an acceptable level of quality and safety; or</P>
<P>(2) <E T="03">Hardship without a compensating increase in quality and safety.</E> Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Footnotes to &#167; 50.55a:
</P>
- <EXTRACT>
- <P>
- <SU>1</SU>USAS and ASME Code addenda issued prior to the winter 1977 Addenda are considered to be &#8220;in effect&#8221; or &#8220;effective&#8221; 6 months after their date of issuance and after they are incorporated by reference in paragraph (a) of this section. Addenda to the ASME Code issued after the summer 1977 Addenda are considered to be &#8220;in effect&#8221; or &#8220;effective&#8221; after the date of publication of the addenda and after they are incorporated by reference in paragraph (a) of this section.</P>
- <P>
- <E T="51">2-3</E> [Reserved].</P>
- <P>
- <SU>4</SU>For ASME Code editions and addenda issued prior to the winter 1977 Addenda, the Code edition and addenda applicable to the component is governed by the order or contract date for the component, not the contract date for the nuclear energy system. For the winter 1977 Addenda and subsequent editions and addenda the method for determining the applicable Code editions and addenda is contained in Paragraph NCA 1140 of Section III of the ASME Code.</P>
- <P>
- <E T="51">5-6</E> [Reserved].</P>
- <P>
- <SU>7</SU>Guidance for quality group classifications of components that are to be included in the safety analysis reports pursuant to &#167; 50.34(a) and &#167; 50.34(b) may be found in Regulatory Guide 1.26, &#8220;Quality Group Classifications and Standards for Water-, Steam-, and Radiological-Waste-Containing Components of Nuclear Power Plants,&#8221; and in Section 3.2.2 of NUREG-0800, &#8220;Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants.&#8221;</P>
- <P>
- <E T="51">8-9</E> [Reserved].</P>
- <P>
- <SU>10</SU>For inspections to be conducted once per interval, the inspections must be performed in accordance with the schedule in Section XI, paragraph IWB-2400, except for plants with inservice inspection programs based on a Section XI edition or addenda prior to the 1994 Addenda. For plants with inservice inspection programs based on a Section XI edition or addenda prior to the 1994 Addenda, the inspection must be performed in accordance with the schedule in Section XI, paragraph IWB-2400, of the 1994 Addenda.</P>
- </EXTRACT>
<CITA>[79 FR 65798, Nov. 5, 2014, as amended at 79 FR 66603, Nov. 10, 2014; 79 FR 73462, Dec. 11, 2014]</CITA>
</SECTION>
<SECTION>
@@ -2314,11 +2264,6 @@
<E T="54">PTS</E> means the reference temperature, RT<E T="52">NDT</E>, evaluated for the EOL Fluence for each of the vessel beltline materials, using the procedures of paragraph (c) of this section.</P>
<P>(8) <E T="03">PTS Screening Criterion</E> means the value of RT<E T="52">PTS</E> for the vessel beltline material above which the plant cannot continue to operate without justification.</P>
<P>(b) <E T="03">Requirements.</E> (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 &#167; 50.82(a)(1) has been submitted, the licensee shall have projected values of RT<E T="52">PTS</E> or RT<E T="52">MAX-X</E>, 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 &#167; 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, <PRTPAGE P="933"/>the assessment of RT<E T="52">PTS</E> 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 RT<E T="52">PTS</E> 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 <SU footnote="Changes to RT_{PTS} 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</SU><FTREF/> change in projected values of RT<E T="52">PTS</E>, or upon request for a change in the expiration date for operation of the facility.</P>
- <FTNT>
- <P>
- <SU>2</SU> Changes to RT<E T="52">PTS</E> 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.</P>
- </FTNT>
-
<P>(2) The pressurized thermal shock (PTS) screening criterion is 270 &#176;F for plates, forgings, and axial weld materials, and 300 &#176;F for circumferential weld materials. For the purpose of comparison with this criterion, the value of RT<E T="52">PTS</E> 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. RT<E T="52">PTS</E> must be determined for each vessel beltline material using the EOL fluence for that material.</P>
<P>(3) For each pressurized water nuclear power reactor for which the value of RT<E T="52">PTS</E> 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 RT<E T="52">PTS</E> above the screening limit due to plant modifications, new information or new analysis techniques.</P>
@@ -2332,44 +2277,25 @@
<P>(c) <E T="03">Calculation of RT</E>
<E T="54">PTS.</E> RT<E T="52">PTS</E> must be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for the material. RT<E T="52">PTS</E> must be evaluated using the same procedures used to calculate RT<E T="52">NDT</E>, as indicated in paragraph (c)(1) of this section, and as provided in paragraphs (c)(2) and (c)(3) of this section.</P>
<P>(1) Equation 1 must be used to calculate values of RT<E T="52">NDT</E> for each weld and plate, or forging, in the reactor vessel beltline.
+ Equation 1: RT<E T="52">NDT</E> = RT<E T="52">NDT(U)</E> + M + &#916;RT<E T="52">NDT</E>
</P>
- <FP SOURCE="FP-2">Equation 1: RT<E T="52">NDT</E> = RT<E T="52">NDT(U)</E> + M + &#916;RT<E T="52">NDT</E>
- </FP>
<P>(i) If a measured value of RT<E T="52">NDT(U)</E> is not available, a generic mean value for the class <SU footnote="The class of material for estimating RT_{NDT(U)} is generally determined for welds by the type of welding flux (Linde 80, or other), and for base metal by the material specification.">3</SU><FTREF/> of material may be used if there are sufficient test results to establish a mean and a standard deviation for the class.</P>
- <FTNT>
- <P>
- <SU>3</SU> The class of material for estimating RT<E T="52">NDT(U)</E> is generally determined for welds by the type of welding flux (Linde 80, or other), and for base metal by the material specification.</P>
- </FTNT>
<P>(ii) For generic values of weld metal, the following generic mean values must be used unless justification for different values is provided: 0 &#176;F for welds made with Linde 80 flux, and &#8722;56 &#176;F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.</P>
<P>(iii) <E T="03">M</E> means the margin to be added to account for uncertainties in the values of RT<E T="52">NDT(U)</E>, copper and nickel contents, fluence and the calculational procedures. M is evaluated from Equation 2.</P>
- <MATH DEEP="19" SPAN="1">
- <MID>ER19DE95.003</MID>
- </MATH>
<P>(A) In Equation 2, &#963;<E T="52">U</E> is the standard deviation for RT<E T="52">NDT(U).</E> If a measured value of RT<E T="52">NDT(U)</E> is used, then &#963;<E T="52">U</E> is determined from the precision of the test method. If a measured value of RT<E T="52">NDT(U)</E> is not available and a generic mean value for that class of materials is used, then &#963;<E T="52">U</E> 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 &#963;<E T="52">U</E> is 17 &#176;F.</P>
<P>(B) In Equation 2, &#963;<E T="8064">&#916;</E> is the standard deviation for &#916;RT<E T="52">NDT.</E> The value of &#963;<E T="8064">&#916;</E> to be used is 28 &#176;F for welds and 17 &#176;F for base metal; the value of &#963;<E T="8064">&#916;</E> need not exceed one-half of &#916;RT<E T="52">NDT.</E></P>
<P>(iv) &#916;RT<E T="52">NDT</E> is the mean value of the transition temperature shift, or change in RT<E T="52">NDT</E>, due to irradiation, and must be calculated using Equation 3.
+ Equation 3: &#916;RT<E T="52">NDT</E> = (CF)f<E T="51">(0.28&#8722;0.10 log f)</E>
</P>
- <FP SOURCE="FP-2">Equation 3: &#916;RT<E T="52">NDT</E> = (CF)f<E T="51">(0.28&#8722;0.10 log f)</E>
- </FP>
<P>(A) <E T="03">CF</E> (&#176;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, &#8220;Wt &#8722; % copper&#8221; and &#8220;Wt &#8722; % nickel&#8221; 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 <SU footnote="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 &#8220;generic data.&#8221;">4</SU><FTREF/> may be used if justification is provided. If none of these alternatives are available, <PRTPAGE P="935"/>0.35% copper and 1.0% nickel must be assumed.</P>
- <FTNT>
- <P>
- <SU>4</SU> 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 &#8220;generic data.&#8221;</P>
- </FTNT>
<P>(B) <E T="03">f</E> is the best estimate neutron fluence, in units of 10<SU>19</SU>n/cm<SU>2</SU>(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 KRT<E T="52">PTS.</E></P>
<P>(v) Equation 4 must be used for determining RT<E T="52">PTS</E> using equation 3 with EOL fluence values for determining &#916;RT<E T="52">PTS.</E>
+ Equation 4: RT<E T="52">PTS</E> = RT<E T="52">NDT(U)</E> + M + &#916;RT<E T="52">PTS</E>
</P>
- <FP SOURCE="FP-2">Equation 4: RT<E T="52">PTS</E> = RT<E T="52">NDT(U)</E> + M + &#916;RT<E T="52">PTS</E>
- </FP>
<P>(2) To verify that RT<E T="52">NDT</E> 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 <SU footnote="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.">5</SU><FTREF/> results.</P>
- <FTNT>
- <P>
- <SU>5</SU> 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.</P>
- </FTNT>
-
<P>(i) Results from the plant-specific surveillance program must be integrated into the RT<E T="52">NDT</E> estimate if the plant-specific surveillance data has been deemed credible as judged by the following criteria:</P>
<P>(A) The materials in the surveillance capsules must be those which are the controlling materials with regard to radiation embrittlement.</P>
<P>(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.</P>
@@ -2378,10 +2304,6 @@
<P>(D) The irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the cladding/base metal interface within &#177;25 &#176;F.</P>
<P>(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.</P>
<P>(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.</P>
- <MATH DEEP="61" SPAN="2">
- <MID>ER19DE95.004</MID>
- </MATH>
-
<P>(B) In Equation 5, &#8220;n&#8221; is the number of surveillance data points, &#8220;A<E T="52">i</E>&#8221; is the measured value of &#916;RT<E T="52">NDT</E> and &#8220;f<E T="52">i</E>&#8221; 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, <E T="03">i.e.</E>, differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of &#916;RT<E T="52">NDT</E> must be adjusted for differences in copper and nickel content by multiplying them by the ratio of <PRTPAGE P="936"/>the chemistry factor for the vessel material to that for the surveillance weld.</P>
<P>(iii) For cases in which the results from a credible plant-specific surveillance program are used, the value of &#963;<E T="8064">&#916;</E> to be used is 14 &#176;F for welds and 8.5 &#176;F for base metal; the value of &#963;<E T="8064">&#916;</E> need not exceed one-half of &#916;RT<E T="52">NDT.</E></P>
@@ -2389,854 +2311,6 @@
<P>(iv) The use of results from the plant-specific surveillance program may result in an RT<E T="52">NDT</E> that is higher or lower than those determined in paragraph (c)(1).</P>
<P>(3) Any information that is believed to improve the accuracy of the RT<E T="52">PTS</E> value significantly must be reported to the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate. Any value of RT<E T="52">PTS</E> 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.</P>
- <GPOTABLE CDEF="s10,4,4,4,4,4,4,4" COLS="8" OPTS="L2">
- <TTITLE>Table 1&#8212;Chemistry Factor for Weld Metals, &#176;F</TTITLE>
- <BOXHD>
- <CHED H="1">Copper, wt-%</CHED>
- <CHED H="1">Nickel, wt-%</CHED>
- <CHED H="2">0</CHED>
- <CHED H="2">0.20</CHED>
- <CHED H="2">0.40</CHED>
- <CHED H="2">0.60</CHED>
- <CHED H="2">0.80</CHED>
- <CHED H="2">1.00</CHED>
- <CHED H="2">1.20</CHED>
- </BOXHD>
- <ROW>
- <ENT I="01">0</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.01</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.02</ENT>
- <ENT>21</ENT>
- <ENT>26</ENT>
- <ENT>27</ENT>
- <ENT>27</ENT>
- <ENT>27</ENT>
- <ENT>27</ENT>
- <ENT>27</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.03</ENT>
- <ENT>22</ENT>
- <ENT>35</ENT>
- <ENT>41</ENT>
- <ENT>41</ENT>
- <ENT>41</ENT>
- <ENT>41</ENT>
- <ENT>41</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.04</ENT>
- <ENT>24</ENT>
- <ENT>43</ENT>
- <ENT>54</ENT>
- <ENT>54</ENT>
- <ENT>54</ENT>
- <ENT>54</ENT>
- <ENT>54</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.05</ENT>
- <ENT>26</ENT>
- <ENT>49</ENT>
- <ENT>67</ENT>
- <ENT>68</ENT>
- <ENT>68</ENT>
- <ENT>68</ENT>
- <ENT>68</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.06</ENT>
- <ENT>29</ENT>
- <ENT>52</ENT>
- <ENT>77</ENT>
- <ENT>82</ENT>
- <ENT>82</ENT>
- <ENT>82</ENT>
- <ENT>82</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.07</ENT>
- <ENT>32</ENT>
- <ENT>55</ENT>
- <ENT>85</ENT>
- <ENT>95</ENT>
- <ENT>95</ENT>
- <ENT>95</ENT>
- <ENT>95</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.08</ENT>
- <ENT>36</ENT>
- <ENT>58</ENT>
- <ENT>90</ENT>
- <ENT>106</ENT>
- <ENT>108</ENT>
- <ENT>108</ENT>
- <ENT>108</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.09</ENT>
- <ENT>40</ENT>
- <ENT>61</ENT>
- <ENT>94</ENT>
- <ENT>115</ENT>
- <ENT>122</ENT>
- <ENT>122</ENT>
- <ENT>122</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.10</ENT>
- <ENT>44</ENT>
- <ENT>65</ENT>
- <ENT>97</ENT>
- <ENT>122</ENT>
- <ENT>133</ENT>
- <ENT>135</ENT>
- <ENT>135</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.11</ENT>
- <ENT>49</ENT>
- <ENT>68</ENT>
- <ENT>101</ENT>
- <ENT>130</ENT>
- <ENT>144</ENT>
- <ENT>148</ENT>
- <ENT>148</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.12</ENT>
- <ENT>52</ENT>
- <ENT>72</ENT>
- <ENT>103</ENT>
- <ENT>135</ENT>
- <ENT>153</ENT>
- <ENT>161</ENT>
- <ENT>161</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.13</ENT>
- <ENT>58</ENT>
- <ENT>76</ENT>
- <ENT>106</ENT>
- <ENT>139</ENT>
- <ENT>162</ENT>
- <ENT>172</ENT>
- <ENT>176</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.14</ENT>
- <ENT>61</ENT>
- <ENT>79</ENT>
- <ENT>109</ENT>
- <ENT>142</ENT>
- <ENT>168</ENT>
- <ENT>182</ENT>
- <ENT>188</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.15</ENT>
- <ENT>66</ENT>
- <ENT>84</ENT>
- <ENT>112</ENT>
- <ENT>146</ENT>
- <ENT>175</ENT>
- <ENT>191</ENT>
- <ENT>200</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.16</ENT>
- <ENT>70</ENT>
- <ENT>88</ENT>
- <ENT>115</ENT>
- <ENT>149</ENT>
- <ENT>178</ENT>
- <ENT>199</ENT>
- <ENT>211</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.17</ENT>
- <ENT>75</ENT>
- <ENT>92</ENT>
- <ENT>119</ENT>
- <ENT>151</ENT>
- <ENT>184</ENT>
- <ENT>207</ENT>
- <ENT>221</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.18</ENT>
- <ENT>79</ENT>
- <ENT>95</ENT>
- <ENT>122</ENT>
- <ENT>154</ENT>
- <ENT>187</ENT>
- <ENT>214</ENT>
- <ENT>230</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.19</ENT>
- <ENT>83</ENT>
- <ENT>100</ENT>
- <ENT>126</ENT>
- <ENT>157</ENT>
- <ENT>191</ENT>
- <ENT>220</ENT>
- <ENT>238</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.20</ENT>
- <ENT>88</ENT>
- <ENT>104</ENT>
- <ENT>129</ENT>
- <ENT>160</ENT>
- <ENT>194</ENT>
- <ENT>223</ENT>
- <ENT>245</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.21</ENT>
- <ENT>92</ENT>
- <ENT>108</ENT>
- <ENT>133</ENT>
- <ENT>164</ENT>
- <ENT>197</ENT>
- <ENT>229</ENT>
- <ENT>252</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.22</ENT>
- <ENT>97</ENT>
- <ENT>112</ENT>
- <ENT>137</ENT>
- <ENT>167</ENT>
- <ENT>200</ENT>
- <ENT>232</ENT>
- <ENT>257</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.23</ENT>
- <ENT>101</ENT>
- <ENT>117</ENT>
- <ENT>140</ENT>
- <ENT>169</ENT>
- <ENT>203</ENT>
- <ENT>236</ENT>
- <ENT>263</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.24</ENT>
- <ENT>105</ENT>
- <ENT>121</ENT>
- <ENT>144</ENT>
- <ENT>173</ENT>
- <ENT>206</ENT>
- <ENT>239</ENT>
- <ENT>268</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.25</ENT>
- <ENT>110</ENT>
- <ENT>126</ENT>
- <ENT>148</ENT>
- <ENT>176</ENT>
- <ENT>209</ENT>
- <ENT>243</ENT>
- <ENT>272</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.26</ENT>
- <ENT>113</ENT>
- <ENT>130</ENT>
- <ENT>151</ENT>
- <ENT>180</ENT>
- <ENT>212</ENT>
- <ENT>246</ENT>
- <ENT>276</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.27</ENT>
- <ENT>119</ENT>
- <ENT>134</ENT>
- <ENT>155</ENT>
- <ENT>184</ENT>
- <ENT>216</ENT>
- <ENT>249</ENT>
- <ENT>280</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.28</ENT>
- <ENT>122</ENT>
- <ENT>138</ENT>
- <ENT>160</ENT>
- <ENT>187</ENT>
- <ENT>218</ENT>
- <ENT>251</ENT>
- <ENT>284</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.29</ENT>
- <ENT>128</ENT>
- <ENT>142</ENT>
- <ENT>164</ENT>
- <ENT>191</ENT>
- <ENT>222</ENT>
- <ENT>254</ENT>
- <ENT>287</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.30</ENT>
- <ENT>131</ENT>
- <ENT>146</ENT>
- <ENT>167</ENT>
- <ENT>194</ENT>
- <ENT>225</ENT>
- <ENT>257</ENT>
- <ENT>290</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.31</ENT>
- <ENT>136</ENT>
- <ENT>151</ENT>
- <ENT>172</ENT>
- <ENT>198</ENT>
- <ENT>228</ENT>
- <ENT>260</ENT>
- <ENT>293</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.32</ENT>
- <ENT>140</ENT>
- <ENT>155</ENT>
- <ENT>175</ENT>
- <ENT>202</ENT>
- <ENT>231</ENT>
- <ENT>263</ENT>
- <ENT>296</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.33</ENT>
- <ENT>144</ENT>
- <ENT>160</ENT>
- <ENT>180</ENT>
- <ENT>205</ENT>
- <ENT>234</ENT>
- <ENT>266</ENT>
- <ENT>299</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.34</ENT>
- <ENT>149</ENT>
- <ENT>164</ENT>
- <ENT>184</ENT>
- <ENT>209</ENT>
- <ENT>238</ENT>
- <ENT>269</ENT>
- <ENT>302</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.35</ENT>
- <ENT>153</ENT>
- <ENT>168</ENT>
- <ENT>187</ENT>
- <ENT>212</ENT>
- <ENT>241</ENT>
- <ENT>272</ENT>
- <ENT>305</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.36</ENT>
- <ENT>158</ENT>
- <ENT>172</ENT>
- <ENT>191</ENT>
- <ENT>216</ENT>
- <ENT>245</ENT>
- <ENT>275</ENT>
- <ENT>308</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.37</ENT>
- <ENT>162</ENT>
- <ENT>177</ENT>
- <ENT>196</ENT>
- <ENT>220</ENT>
- <ENT>248</ENT>
- <ENT>278</ENT>
- <ENT>311</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.38</ENT>
- <ENT>166</ENT>
- <ENT>182</ENT>
- <ENT>200</ENT>
- <ENT>223</ENT>
- <ENT>250</ENT>
- <ENT>281</ENT>
- <ENT>314</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.39</ENT>
- <ENT>171</ENT>
- <ENT>185</ENT>
- <ENT>203</ENT>
- <ENT>227</ENT>
- <ENT>254</ENT>
- <ENT>285</ENT>
- <ENT>317</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.40</ENT>
- <ENT>175</ENT>
- <ENT>189</ENT>
- <ENT>207</ENT>
- <ENT>231</ENT>
- <ENT>257</ENT>
- <ENT>288</ENT>
- <ENT>320</ENT>
- </ROW>
- </GPOTABLE>
- <GPOTABLE CDEF="s10,4,4,4,4,4,4,4" COLS="8" OPTS="L2">
- <TTITLE>Table 2&#8212;Chemistry Factor for Base Metals, &#176;F</TTITLE>
- <BOXHD>
- <CHED H="1">Copper, wt-%</CHED>
- <CHED H="1">Nickel, wt-%</CHED>
- <CHED H="2">0</CHED>
- <CHED H="2">0.20</CHED>
- <CHED H="2">0.40</CHED>
- <CHED H="2">0.60</CHED>
- <CHED H="2">0.80</CHED>
- <CHED H="2">1.00</CHED>
- <CHED H="2">1.20</CHED>
- </BOXHD>
- <ROW>
- <ENT I="01">0</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.01</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.02</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.03</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- <ENT>20</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.04</ENT>
- <ENT>22</ENT>
- <ENT>26</ENT>
- <ENT>26</ENT>
- <ENT>26</ENT>
- <ENT>26</ENT>
- <ENT>26</ENT>
- <ENT>26</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.05</ENT>
- <ENT>25</ENT>
- <ENT>31</ENT>
- <ENT>31</ENT>
- <ENT>31</ENT>
- <ENT>31</ENT>
- <ENT>31</ENT>
- <ENT>31</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.06</ENT>
- <ENT>28</ENT>
- <ENT>37</ENT>
- <ENT>37</ENT>
- <ENT>37</ENT>
- <ENT>37</ENT>
- <ENT>37</ENT>
- <ENT>37</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.07</ENT>
- <ENT>31</ENT>
- <ENT>43</ENT>
- <ENT>44</ENT>
- <ENT>44</ENT>
- <ENT>44</ENT>
- <ENT>44</ENT>
- <ENT>44</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.08</ENT>
- <ENT>34</ENT>
- <ENT>48</ENT>
- <ENT>51</ENT>
- <ENT>51</ENT>
- <ENT>51</ENT>
- <ENT>51</ENT>
- <ENT>51</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.09</ENT>
- <ENT>37</ENT>
- <ENT>53</ENT>
- <ENT>58</ENT>
- <ENT>58</ENT>
- <ENT>58</ENT>
- <ENT>58</ENT>
- <ENT>58</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.10</ENT>
- <ENT>41</ENT>
- <ENT>58</ENT>
- <ENT>65</ENT>
- <ENT>65</ENT>
- <ENT>67</ENT>
- <ENT>67</ENT>
- <ENT>67</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.11</ENT>
- <ENT>45</ENT>
- <ENT>62</ENT>
- <ENT>72</ENT>
- <ENT>74</ENT>
- <ENT>77</ENT>
- <ENT>77</ENT>
- <ENT>77</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.12</ENT>
- <ENT>49</ENT>
- <ENT>67</ENT>
- <ENT>79</ENT>
- <ENT>83</ENT>
- <ENT>86</ENT>
- <ENT>86</ENT>
- <ENT>86</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.13</ENT>
- <ENT>53</ENT>
- <ENT>71</ENT>
- <ENT>85</ENT>
- <ENT>91</ENT>
- <ENT>96</ENT>
- <ENT>96</ENT>
- <ENT>96</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.14</ENT>
- <ENT>57</ENT>
- <ENT>75</ENT>
- <ENT>91</ENT>
- <ENT>100</ENT>
- <ENT>105</ENT>
- <ENT>106</ENT>
- <ENT>106</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.15</ENT>
- <ENT>61</ENT>
- <ENT>80</ENT>
- <ENT>99</ENT>
- <ENT>110</ENT>
- <ENT>115</ENT>
- <ENT>117</ENT>
- <ENT>117</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.16</ENT>
- <ENT>65</ENT>
- <ENT>84</ENT>
- <ENT>104</ENT>
- <ENT>118</ENT>
- <ENT>123</ENT>
- <ENT>125</ENT>
- <ENT>125</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.17</ENT>
- <ENT>69</ENT>
- <ENT>88</ENT>
- <ENT>110</ENT>
- <ENT>127</ENT>
- <ENT>132</ENT>
- <ENT>135</ENT>
- <ENT>135</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.18</ENT>
- <ENT>73</ENT>
- <ENT>92</ENT>
- <ENT>115</ENT>
- <ENT>134</ENT>
- <ENT>141</ENT>
- <ENT>144</ENT>
- <ENT>144</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.19</ENT>
- <ENT>78</ENT>
- <ENT>97</ENT>
- <ENT>120</ENT>
- <ENT>142</ENT>
- <ENT>150</ENT>
- <ENT>154</ENT>
- <ENT>154</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.20</ENT>
- <ENT>82</ENT>
- <ENT>102</ENT>
- <ENT>125</ENT>
- <ENT>149</ENT>
- <ENT>159</ENT>
- <ENT>164</ENT>
- <ENT>165</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.21</ENT>
- <ENT>86</ENT>
- <ENT>107</ENT>
- <ENT>129</ENT>
- <ENT>155</ENT>
- <ENT>167</ENT>
- <ENT>172</ENT>
- <ENT>174</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.22</ENT>
- <ENT>91</ENT>
- <ENT>112</ENT>
- <ENT>134</ENT>
- <ENT>161</ENT>
- <ENT>176</ENT>
- <ENT>181</ENT>
- <ENT>184</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.23</ENT>
- <ENT>95</ENT>
- <ENT>117</ENT>
- <ENT>138</ENT>
- <ENT>167</ENT>
- <ENT>184</ENT>
- <ENT>190</ENT>
- <ENT>194</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.24</ENT>
- <ENT>100</ENT>
- <ENT>121</ENT>
- <ENT>143</ENT>
- <ENT>172</ENT>
- <ENT>191</ENT>
- <ENT>199</ENT>
- <ENT>204</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.25</ENT>
- <ENT>104</ENT>
- <ENT>126</ENT>
- <ENT>148</ENT>
- <ENT>176</ENT>
- <ENT>199</ENT>
- <ENT>208</ENT>
- <ENT>214</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.26</ENT>
- <ENT>109</ENT>
- <ENT>130</ENT>
- <ENT>151</ENT>
- <ENT>180</ENT>
- <ENT>205</ENT>
- <ENT>216</ENT>
- <ENT>221</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.27</ENT>
- <ENT>114</ENT>
- <ENT>134</ENT>
- <ENT>155</ENT>
- <ENT>184</ENT>
- <ENT>211</ENT>
- <ENT>225</ENT>
- <ENT>230</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.28</ENT>
- <ENT>119</ENT>
- <ENT>138</ENT>
- <ENT>160</ENT>
- <ENT>187</ENT>
- <ENT>216</ENT>
- <ENT>233</ENT>
- <ENT>239</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.29</ENT>
- <ENT>124</ENT>
- <ENT>142</ENT>
- <ENT>164</ENT>
- <ENT>191</ENT>
- <ENT>221</ENT>
- <ENT>241</ENT>
- <ENT>248</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.30</ENT>
- <ENT>129</ENT>
- <ENT>146</ENT>
- <ENT>167</ENT>
- <ENT>194</ENT>
- <ENT>225</ENT>
- <ENT>249</ENT>
- <ENT>257</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.31</ENT>
- <ENT>134</ENT>
- <ENT>151</ENT>
- <ENT>172</ENT>
- <ENT>198</ENT>
- <ENT>228</ENT>
- <ENT>255</ENT>
- <ENT>266</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.32</ENT>
- <ENT>139</ENT>
- <ENT>155</ENT>
- <ENT>175</ENT>
- <ENT>202</ENT>
- <ENT>231</ENT>
- <ENT>260</ENT>
- <ENT>274</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.33</ENT>
- <ENT>144</ENT>
- <ENT>160</ENT>
- <ENT>180</ENT>
- <ENT>205</ENT>
- <ENT>234</ENT>
- <ENT>264</ENT>
- <ENT>282</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.34</ENT>
- <ENT>149</ENT>
- <ENT>164</ENT>
- <ENT>184</ENT>
- <ENT>209</ENT>
- <ENT>238</ENT>
- <ENT>268</ENT>
- <ENT>290</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.35</ENT>
- <ENT>153</ENT>
- <ENT>168</ENT>
- <ENT>187</ENT>
- <ENT>212</ENT>
- <ENT>241</ENT>
- <ENT>272</ENT>
- <ENT>298</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.36</ENT>
- <ENT>158</ENT>
- <ENT>173</ENT>
- <ENT>191</ENT>
- <ENT>216</ENT>
- <ENT>245</ENT>
- <ENT>275</ENT>
- <ENT>303</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.37</ENT>
- <ENT>162</ENT>
- <ENT>177</ENT>
- <ENT>196</ENT>
- <ENT>220</ENT>
- <ENT>248</ENT>
- <ENT>278</ENT>
- <ENT>308</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.38</ENT>
- <ENT>166</ENT>
- <ENT>182</ENT>
- <ENT>200</ENT>
- <ENT>223</ENT>
- <ENT>250</ENT>
- <ENT>281</ENT>
- <ENT>313</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.39</ENT>
- <ENT>171</ENT>
- <ENT>185</ENT>
- <ENT>203</ENT>
- <ENT>227</ENT>
- <ENT>254</ENT>
- <ENT>285</ENT>
- <ENT>317</ENT>
- </ROW>
- <ROW>
- <ENT I="01">0.40</ENT>
- <ENT>175</ENT>
- <ENT>189</ENT>
- <ENT>207</ENT>
- <ENT>231</ENT>
- <ENT>257</ENT>
- <ENT>288</ENT>
- <ENT>320</ENT>
- </ROW>
- </GPOTABLE>
<CITA>[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]</CITA>
</SECTION>
<SECTION>
@@ -4175,7 +3249,8 @@
<P>(ii)-(iii) [Reserved]</P>
<P>(iv)(A) Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.</P>
<P>(B) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.</P>
- <P>(v)-(x) [Reserved]</P>
+ <P>(v)-(vii) [Reserved]</P>
+ <P>(viii)-(x) [Reserved]</P>
<P>(xi) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.</P>
<P>(3) <E T="03">Eight-hour reports.</E> If not reported under paragraphs (a), (b)(1) or (b)(2) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following:</P>
<P>(i) [Reserved]</P>
@@ -4206,7 +3281,8 @@
<P>(C) Control the release of radioactive material; or</P>
<P>(D) Mitigate the consequences of an accident.</P>
<P>(vi) Events covered in paragraph (b)(3)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (b)(3)(v) of this section if redundant equipment in the same system was operable and available to perform the required safety function.</P>
- <P>(vii)-(xi) [Reserved]</P>
+ <P>(vii)-(ix) [Reserved]</P>
+ <P>(x)-(xi) [Reserved]</P>
<P>(xii) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.</P>
<P>(xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).</P>
<P>(c) <E T="03">Followup notification.</E> With respect to the telephone notifications made under paragraphs (a) and (b) of this section, in addition to making the required initial notification, each licensee, shall during the course of the event:</P>
@@ -4325,42 +3401,9 @@
<P>(4) The amount stated in the applicant's or licensee's certification may be based on a cost estimate for decommissioning the facility. As part of the certification, a copy of the financial instrument obtained to satisfy the requirements of paragraph (e) of this section must be submitted to NRC; <E T="03">provided, however,</E> that an applicant for or holder of a combined license need not obtain such financial instrument or submit a copy to the Commission except as provided in paragraph (e)(3) of this section.<PRTPAGE P="969"/>
</P>
- <P>(c) Table of minimum amounts (January 1986 dollars) required to demonstrate reasonable assurance of funds for decommissioning by reactor type and power level, P (in MWt); adjustment factor. <SU footnote="Amounts are based on activities related to the definition of &#8220;Decommission&#8221; in &#167; 50.2 of this part and do not include the cost of removal and disposal of spent fuel or of nonradioactive structures and materials beyond that necessary to terminate the license.">1</SU><FTREF/>
+ <P>(c) Table of minimum amounts (January 1986 dollars) required to demonstrate reasonable assurance of funds for decommissioning by reactor type and power level, P (in MWt); adjustment factor.
</P>
- <FTNT>
- <P>
- <SU>1</SU> Amounts are based on activities related to the definition of &#8220;Decommission&#8221; in &#167; 50.2 of this part and do not include the cost of removal and disposal of spent fuel or of nonradioactive structures and materials beyond that necessary to terminate the license.</P>
- </FTNT>
- <GPOTABLE CDEF="s25,14" COLS="2" OPTS="L0,p7,7/8,g1">
- <BOXHD>
- <CHED H="1"/>
- <CHED H="1">
- <E T="03">Millions</E>
- </CHED>
- </BOXHD>
- <ROW>
- <ENT I="11">(1)(i) For a PWR:</ENT>
- </ROW>
- <ROW>
- <ENT I="02">greater than or equal to 3400 MWt</ENT>
- <ENT>$105</ENT>
- </ROW>
- <ROW>
- <ENT I="02">between 1200 MWt and 3400 MWt (For a PWR of less than 1200 MWt, use P=1200 MWt)</ENT>
- <ENT>$(75+0.0088P)</ENT>
- </ROW>
- <ROW>
- <ENT I="11">(ii) For a BWR:</ENT>
- </ROW>
- <ROW>
- <ENT I="02">greater than or equal to 3400 MWt</ENT>
- <ENT>$135</ENT>
- </ROW>
- <ROW>
- <ENT I="02">between 1200 MWt and 3400 MWt (For a BWR of less than 1200 MWt, use P=1200 MWt)</ENT>
- <ENT>$(104+0.009P)</ENT>
- </ROW>
- </GPOTABLE>
+ <P>(1) Amounts are based on activities related to the definition of &#8220;Decommission&#8221; in &#167; 50.2 of this part and do not include the cost of removal and disposal of spent fuel or of nonradioactive structures and materials beyond that necessary to terminate the license.</P>
<P>(2) An adjustment factor at least equal to 0.65 L + 0.13 E + 0.22 B is to be used where L and E are escalation factors for labor and energy, respectively, and are to be taken from regional data of U.S. Department of Labor Bureau of Labor Statistics and B is an escalation factor for waste burial and is to be taken from NRC report NUREG-1307, &#8220;Report on Waste Burial Charges.&#8221;</P>
<P>(d)(1) Each non-power reactor applicant for or holder of an operating license for a production or utilization facility shall submit a decommissioning report as required by &#167; 50.33(k) of this part.</P>
<P>(2) The report must:</P>
@@ -5242,11 +4285,11 @@
<P>Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.</P>
<P>
<E T="03">Criterion 56&#8212;Primary containment isolation.</E> Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:</P>
- <P>(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or</P>
- <P>(2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or</P>
- <P>(3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or</P>
+ <P>1 One locked closed isolation valve inside and one locked closed isolation valve outside containment; or</P>
+ <P>2 One automatic isolation valve inside and one locked closed isolation valve outside containment; or</P>
+ <P>3 One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or</P>
- <P>(4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
+ <P>4 One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
</P>
<FP>Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.</FP>
<P>
@@ -5460,7 +4503,8 @@
<P>5. Arrangements for medical service providers qualified to handle radiological emergencies onsite;</P>
<P>6. Arrangements for transportation of contaminated injured individuals from the site to specifically identified treatment facilities outside the site boundary;</P>
<P>7. Arrangements for treatment of individuals injured in support of licensed activities on the site at treatment facilities outside the site boundary;</P>
- <P>8.a. (i) A licensee onsite technical support center and an emergency operations facility from which effective direction can be given and effective control can be exercised during an emergency;</P>
+ <P>8.</P>
+ <P>a. (i) A licensee onsite technical support center and an emergency operations facility from which effective direction can be given and effective control can be exercised during an emergency;</P>
<P>(ii) For nuclear power reactor licensees, a licensee onsite operational support center;</P>
<P>b. For a nuclear power reactor licensee's emergency operations facility required by paragraph 8.a of this section, either a facility located between 10 miles and 25 miles of the nuclear power reactor site(s), or a primary facility located less than 10 miles from the nuclear power reactor site(s) and a backup facility located between 10 miles and 25 miles of the nuclear power reactor site(s). An emergency operations facility may serve more than one nuclear power reactor site. A licensee desiring to locate an emergency operations facility more than 25 miles from a <PRTPAGE P="1010"/>nuclear power reactor site shall request prior Commission approval by submitting an application for an amendment to its license. For an emergency operations facility located more than 25 miles from a nuclear power reactor site, provisions must be made for locating NRC and offsite responders closer to the nuclear power reactor site so that NRC and offsite responders can interact face-to-face with emergency response personnel entering and leaving the nuclear power reactor site. Provisions for locating NRC and offsite responders closer to a nuclear power reactor site that is more than 25 miles from the emergency operations facility must include the following:</P>
@@ -5654,7 +4698,7 @@
<P>c. The minimum temperature requirements given in table 1 pertain to the controlling material, which is either the material in the closure flange or the material in the beltline region with the highest reference temperature. As specified in table 1, the minimum temperature requirements and the controlling material depend on the operating condition (i.e., hydrostatic pressure and leak tests, or normal operation including anticipated operational occurrences), the vessel pressure, whether fuel is in the vessel, and whether the core is critical. The metal temperature of the controlling material, in the region of the controlling material which has the least favorable combination of stress and temperature must exceed the appropriate minimum temperature requirement for the condition and pressure of the vessel specified in table 1.</P>
<P>d. Pressure tests and leak tests of the reactor vessel that are required by Section XI of the ASME Code must be completed before the core is critical.</P>
- <P>B. If the procedures of section IV.A. of this appendix do not indicate the existence of an equivalent safety margin, the reactor vessel beltline may be given a thermal annealing treatment to recover the fracture toughness of the material, subject to the requirements of &#167; 50.66. The reactor vessel may continue to be operated only for that service period within which the predicted fracture toughness of the beltline region materials satisfies the requirements of section IV.A. of this appendix using the values of RT<E T="52">NDT</E> and Charpy upper-shelf energy that include the effects of annealing and subsequent irradiation.<PRTPAGE P="1018"/>
+ <P>e. If the procedures of section IV.A. of this appendix do not indicate the existence of an equivalent safety margin, the reactor vessel beltline may be given a thermal annealing treatment to recover the fracture toughness of the material, subject to the requirements of &#167; 50.66. The reactor vessel may continue to be operated only for that service period within which the predicted fracture toughness of the beltline region materials satisfies the requirements of section IV.A. of this appendix using the values of RT<E T="52">NDT</E> and Charpy upper-shelf energy that include the effects of annealing and subsequent irradiation.<PRTPAGE P="1018"/>
</P>
<GPOTABLE CDEF="s15,5,xls75,xs85" COLS="4" OPTS="L2">
<TTITLE>Table 1&#8212;Pressure and Temperature Requirements for the Reactor Pressure Vessel</TTITLE>
@@ -5771,27 +4815,34 @@
<APPENDIX>
<EAR>Pt. 50, App. I</EAR>
<HD SOURCE="HED">Appendix I to Part 50&#8212;Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the Criterion &#8220;As Low as is Reasonably Achievable&#8221; for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents</HD>
- <P>SECTION I. <E T="03">Introduction.</E> Section 50.34a provides that an application for a construction permit shall include a description of the preliminary design of equipment to be installed to maintain control over radioactive materials in gaseous and liquid effluents produced during normal conditions, including expected occurrences. In the case of an application filed on or after January 2, 1971, the application must also identify the design objectives, and the means to be employed, for keeping levels of radioactive material in effluents to unrestricted areas as low as practicable. Sections 52.47, 52.79, 52.137, and 52.157 of this chapter provide that applications for design certification, combined license, design approval, or manufacturing license, respectively, shall include a description of the equipment and procedures for the control of gaseous and liquid effluents and <PRTPAGE P="1020"/>for the maintenance and use of equipment installed in radioactive waste systems.</P>
+ <HD SOURCE="HD3">SECTION I.</HD>
+ <P><E T="03">Introduction.</E> Section 50.34a provides that an application for a construction permit shall include a description of the preliminary design of equipment to be installed to maintain control over radioactive materials in gaseous and liquid effluents produced during normal conditions, including expected occurrences. In the case of an application filed on or after January 2, 1971, the application must also identify the design objectives, and the means to be employed, for keeping levels of radioactive material in effluents to unrestricted areas as low as practicable. Sections 52.47, 52.79, 52.137, and 52.157 of this chapter provide that applications for design certification, combined license, design approval, or manufacturing license, respectively, shall include a description of the equipment and procedures for the control of gaseous and liquid effluents and <PRTPAGE P="1020"/>for the maintenance and use of equipment installed in radioactive waste systems.</P>
<P>Section 50.36a contains provisions designed to assure that releases of radioactive material from nuclear power reactors to unrestricted areas during normal conditions, including expected occurrences, are kept as low as practicable.</P>
- <P>SECTION II. <E T="03">Guides on design objectives for light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter.</E> The guides on design objectives set forth in this section may be used by an applicant for a construction permit as guidance in meeting the requirements of &#167; 50.34a(a), or by an applicant for a combined license under part 52 of this chapter as guidance in meeting the requirements of &#167; 50.34a(d), or by an applicant for a design approval, a design certification, or a manufacturing license as guidance in meeting the requirements of &#167; 50.34a(e). The applicant shall provide reasonable assurance that the following design objectives will be met.</P>
+ <HD SOURCE="HD3">SECTION II.</HD>
+ <P><E T="03">Guides on design objectives for light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter.</E> The guides on design objectives set forth in this section may be used by an applicant for a construction permit as guidance in meeting the requirements of &#167; 50.34a(a), or by an applicant for a combined license under part 52 of this chapter as guidance in meeting the requirements of &#167; 50.34a(d), or by an applicant for a design approval, a design certification, or a manufacturing license as guidance in meeting the requirements of &#167; 50.34a(e). The applicant shall provide reasonable assurance that the following design objectives will be met.</P>
<P>A. The calculated annual total quantity of all radioactive material above background <SU footnote="Here and elsewhere in this appendix background means radioactive materials in the environment and in the effluents from light-water-cooled power reactors not generated in, or attributable to, the reactors of which specific account is required in determining design objectives.">1</SU><FTREF/> to be released from each light-water-cooled nuclear power reactor to unrestricted areas will not result in an estimated annual dose or dose commitment from liquid effluents for any individual in an unrestricted area from all pathways of exposure in excess of 3 millirems to the total body or 10 millirems to any organ.</P>
<FTNT>
<P>
<SU>1</SU> Here and elsewhere in this appendix background means radioactive materials in the environment and in the effluents from light-water-cooled power reactors not generated in, or attributable to, the reactors of which specific account is required in determining design objectives.</P>
</FTNT>
- <P>B.1. The calculated annual total quantity of all radioactive material above background to be released from each light-water-cooled nuclear power reactor to the atmosphere will not result in an estimated annual air dose from gaseous effluents at any location near ground level which could be occupied by individuals in unrestricted areas in excess of 10 millirads for gamma radiation or 20 millirads for beta radiation.</P>
+ <P>B.</P>
+ <P>1. The calculated annual total quantity of all radioactive material above background to be released from each light-water-cooled nuclear power reactor to the atmosphere will not result in an estimated annual air dose from gaseous effluents at any location near ground level which could be occupied by individuals in unrestricted areas in excess of 10 millirads for gamma radiation or 20 millirads for beta radiation.</P>
<P>2. Notwithstanding the guidance of paragraph B.1:</P>
<P>(a) The Commission may specify, as guidance on design objectives, a lower quantity of radioactive material above background to be released to the atmosphere if it appears that the use of the design objectives in paragraph B.1 is likely to result in an estimated annual external dose from gaseous effluents to any individual in an unrestricted area in excess of 5 millirems to the total body; and</P>
<P>(b) Design objectives based upon a higher quantity of radioactive material above background to be released to the atmosphere than the quantity specified in paragraph B.1 will be deemed to meet the requirements for keeping levels of radioactive material in gaseous effluents as low as is reasonably achievable if the applicant provides reasonable assurance that the proposed higher quantity will not result in an estimated annual external dose from gaseous effluents to any individual in unrestricted areas in excess of 5 millirems to the total body or 15 millirems to the skin.</P>
<P>C. The calculated annual total quantity of all radioactive iodine and radioactive material in particulate form above background to be released from each light-water-cooled nuclear power reactor in effluents to the atmosphere will not result in an estimated annual dose or dose commitment from such radioactive iodine and radioactive material in particulate form for any individual in an unrestricted area from all pathways of exposure in excess of 15 millirems to any organ.</P>
<P>D. In addition to the provisions of paragraphs A, B, and C above, the applicant shall include in the radwaste system all items of reasonably demonstrated technology that, when added to the system sequentially and in order of diminishing cost-benefit return, can for a favorable cost-benefit ratio effect reductions in dose to the population reasonably expected to be within 50 miles of the reactor. As an interim measure and until establishment and adoption of better values (or other appropriate criteria), the values $1000 per total body man-rem and $1000 per man-thyroid-rem (or such lesser values as may be demonstrated to be suitable in a particular case) shall be used in this cost-benefit analysis. The requirements of this paragraph D need not be complied with by persons who have filed applications for construction permits which were docketed on or after January 2, 1971, and prior to June 4, 1976, if the radwaste systems and equipment described in the preliminary or final safety analysis report and amendments thereto satisfy the Guides on Design Objectives for Light-Water-Cooled Nuclear Power Reactors proposed in the Concluding Statement of Position of the Regulatory Staff in Docket-RM-50-2 dated February 20, 1974, pp. 25-30, reproduced in the annex to this appendix I.</P>
- <P>SECTION III. <E T="03">Implementation.</E> A.1. Conformity with the guides on design objectives of Section II shall be demonstrated by calculational procedures based upon models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated, all uncertainties being considered together. Account shall be taken of the cumulative effect of all sources and pathways within the plant contributing to the particular type of effluent being considered. For determination of design objectives in accordance with the <PRTPAGE P="1021"/>guides of Section II, the estimations of exposure shall be made with respect to such potential land and water usage and food pathways as could actually exist during the term of plant operation: <E T="03">Provided,</E> That, if the requirements of paragraph B of Section III are fulfilled, the applicant shall be deemed to have complied with the requirements of paragraph C of Section II with respect to radioactive iodine if estimations of exposure are made on the basis of such food pathways and individual receptors as actually exist at the time the plant is licensed.</P>
+ <HD SOURCE="HD3">SECTION III.</HD>
+ <P><E T="03">Implementation.</E></P>
+ <P>A.</P>
+ <P>1. Conformity with the guides on design objectives of Section II shall be demonstrated by calculational procedures based upon models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated, all uncertainties being considered together. Account shall be taken of the cumulative effect of all sources and pathways within the plant contributing to the particular type of effluent being considered. For determination of design objectives in accordance with the <PRTPAGE P="1021"/>guides of Section II, the estimations of exposure shall be made with respect to such potential land and water usage and food pathways as could actually exist during the term of plant operation: <E T="03">Provided,</E> That, if the requirements of paragraph B of Section III are fulfilled, the applicant shall be deemed to have complied with the requirements of paragraph C of Section II with respect to radioactive iodine if estimations of exposure are made on the basis of such food pathways and individual receptors as actually exist at the time the plant is licensed.</P>
<P>2. The characteristics attributed to a hypothetical receptor for the purpose of estimating internal dose commitment shall take into account reasonable deviations of individual habits from the average. The applicant may take account of any real phenomenon or factors actually affecting the estimate of radiation exposure, including the characteristics of the plant, modes of discharge of radioactive materials, physical processes tending to attenuate the quantity of radioactive material to which an individual would be exposed, and the effects of averaging exposures over times during which determining factors may fluctuate.</P>
<P>B. If the applicant determines design objectives with respect to radioactive iodine on the basis of existing conditions and if potential changes in land and water usage and food pathways could result in exposures in excess of the guideline values of paragraph C of Section II, the applicant shall provide reasonable assurance that a monitoring and surveillance program will be performed to determine:</P>
<P>1. The quantities of radioactive iodine actually released to the atmosphere and deposited relative to those estimated in the determination of design objectives;</P>
<P>2. Whether changes in land and water usage and food pathways which would result in individual exposures greater than originally estimated have occurred; and</P>
<P>3. The content of radioactive iodine and foods involved in the changes, if and when they occur.</P>
- <P>SECTION IV. <E T="03">Guides on technical specifications for limiting conditions for operation for light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter.</E> The guides on limiting conditions for operation for light-water-cooled nuclear power reactors set forth below may be used by an applicant for an operating license under this part or a design certification or combined license under part 52 of this chapter, or a licensee who has submitted a certification of permanent cessation of operations under &#167; 50.82(a)(1) or &#167; 52.110 of this chapter as guidance in developing technical specifications under &#167; 50.36a(a) to keep levels of radioactive materials in effluents to unrestricted areas as low as is reasonably achievable.</P>
+ <HD SOURCE="HD3">SECTION IV.</HD>
+ <P><E T="03">Guides on technical specifications for limiting conditions for operation for light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter.</E> The guides on limiting conditions for operation for light-water-cooled nuclear power reactors set forth below may be used by an applicant for an operating license under this part or a design certification or combined license under part 52 of this chapter, or a licensee who has submitted a certification of permanent cessation of operations under &#167; 50.82(a)(1) or &#167; 52.110 of this chapter as guidance in developing technical specifications under &#167; 50.36a(a) to keep levels of radioactive materials in effluents to unrestricted areas as low as is reasonably achievable.</P>
<P>Section 50.36a(b) provides that licensees shall be guided by certain considerations in establishing and implementing operating procedures specified in technical specifications that take into account the need for operating flexibility and at the same time assure that the licensee will exert his best effort to keep levels of radioactive material in effluents as low as is reasonably achievable. The guidance set forth below provides additional and more specific guidance to licensees in this respect.</P>
<P>Through the use of the guides set forth in this section it is expected that the annual release of radioactive material in effluents from light-water-cooled nuclear power reactors can generally be maintained within the levels set forth as numerical guides for design objectives in Section II.</P>
<P>At the same time, the licensee is permitted the flexibility of operations, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power even under unusual conditions which may temporarily result in releases higher than numerical guides for design objectives but still within levels that assure that the average population exposure is equivalent to small fractions of doses from natural background radiation. It is expected that in using this operational flexibility under unusual conditions, the licensee will exert his best efforts to keep levels of radioactive material in effluents within the numerical guides for design objectives.</P>
@@ -5811,7 +4862,8 @@
<P>2. Provide data on measurable levels of radiation and radioactive materials in the environment to evaluate the relationship between quantities of radioactive material released in effluents and resultant radiation doses to individuals from principal pathways of exposure; and</P>
<P>3. Identify changes in the use of unrestricted areas (e.g., for agricultural purposes) to permit modifications in monitoring programs for evaluating doses to individuals from principal pathways of exposure.</P>
<P>C. If the data developed in the surveillance and monitoring program described in paragraph B of Section III or from other monitoring programs show that the relationship between the quantities of radioactive material released in liquid and gaseous effluents and the dose to individuals in unrestricted areas is significantly different from that assumed in the calculations used to determine design objectives pursuant to Sections II and III, the Commission may modify the quantities in the technical specifications defining the limiting conditions in a license to operate a light-water-cooled nuclear power reactor or a license whose holder has submitted a certification of permanent cessation of operations under &#167; 50.82(a)(1).</P>
- <P>SECTION V. <E T="03">Effective dates.</E> A. The guides for limiting conditions for operation set forth in this appendix shall be applicable in any case in which an application was filed on or after January 2, 1971, for a construction permit for a light-water-cooled nuclear power reactor under this part, or a design certification, a combined license, or a manufacturing license for a light-water-cooled nuclear power reactor under part 52 of this chapter.</P>
+ <HD SOURCE="HD3">SECTION V.</HD>
+ <P><E T="03">Effective dates.</E> A. The guides for limiting conditions for operation set forth in this appendix shall be applicable in any case in which an application was filed on or after January 2, 1971, for a construction permit for a light-water-cooled nuclear power reactor under this part, or a design certification, a combined license, or a manufacturing license for a light-water-cooled nuclear power reactor under part 52 of this chapter.</P>
<P>B. For each light-water-cooled nuclear power reactor constructed pursuant to a permit for which application was filed prior to January 2, 1971, the holder of the permit or a license, authorizing operation of the reactor shall, within a period of twelve months from June 4, 1975, file with the Commission:</P>
<P>1. Such information as is necessary to evaluate the means employed for keeping levels of radioactivity in effluents to unrestricted areas as low as is reasonably achievable, including all such information as is required by &#167; 50.34a (b) and (c) not already contained in his application; and</P>
<P>2. Plans and proposed technical specifications developed for the purpose of keeping releases of radioactive materials to unrestricted areas during normal reactor operations, including expected operational occurrences, as low as is reasonably achievable.</P>
@@ -5900,7 +4952,8 @@
<HD SOURCE="HD1">III. Leakage Testing Requirements</HD>
<P>A program consisting of a schedule for conducting Type A, B, and C tests shall be developed for leak testing the primary reactor containment and related systems and components penetrating primary containment pressure boundary.</P>
<P>Upon completion of construction of the primary reactor containment, including installation of all portions of mechanical, fluid, electrical, and instrumentation systems penetrating the primary reactor containment pressure boundary, and prior to any reactor operating period, preoperational and periodic leakage rate tests, as applicable, shall be conducted in accordance with the following:</P>
- <P>A. <E T="03">Type A test&#8212;</E>1. <E T="03">Pretest requirements.</E> (a) Containment inspection in accordance with V. A. shall be performed as a prerequisite to the performance of Type A tests. During the period between the initiation of the containment inspection and the performance of the Type A test, no repairs or adjustments shall be made so that the containment can be tested in as close to the &#8220;as is&#8221; condition as practical. During the period between the completion of one Type A test and the initiation of the containment inspection for the subsequent Type A test, repairs or adjustments shall be made to components whose leakage exceeds that specified in the technical specification as soon as practical after identification. If during a Type A test, including the supplemental test specified in III.A.3.(b), potentially excessive leakage paths are identified which will interfere with satisfactory completion of the test, or which result in the Type A test not meeting the acceptance criteria III.A.4.(b) or III.A.5.(b), the Type A test shall be terminated and the leakage through such paths shall be measured using local leakage testing methods. Repairs and/or adjustments to equipment shall be made and Type A test performed. The corrective action taken and the change in leakage rate determined from the tests and overall integrated leakage determined from local leak and Type A tests shall be included in the summary report required by V.B.</P>
+ <P>A. <E T="03">Type A test&#8212;</E></P>
+ <P>1. <E T="03">Pretest requirements.</E> (a) Containment inspection in accordance with V. A. shall be performed as a prerequisite to the performance of Type A tests. During the period between the initiation of the containment inspection and the performance of the Type A test, no repairs or adjustments shall be made so that the containment can be tested in as close to the &#8220;as is&#8221; condition as practical. During the period between the completion of one Type A test and the initiation of the containment inspection for the subsequent Type A test, repairs or adjustments shall be made to components whose leakage exceeds that specified in the technical specification as soon as practical after identification. If during a Type A test, including the supplemental test specified in III.A.3.(b), potentially excessive leakage paths are identified which will interfere with satisfactory completion of the test, or which result in the Type A test not meeting the acceptance criteria III.A.4.(b) or III.A.5.(b), the Type A test shall be terminated and the leakage through such paths shall be measured using local leakage testing methods. Repairs and/or adjustments to equipment shall be made and Type A test performed. The corrective action taken and the change in leakage rate determined from the tests and overall integrated leakage determined from local leak and Type A tests shall be included in the summary report required by V.B.</P>
<P>(b) Closure of containment isolation valves for the Type A test shall be accomplished by normal operation and without any preliminary exercising or adjustments (e.g., no tightening of valve after closure by valve motor). Repairs of maloperating or leaking valves shall be made as necessary. Information on any valve closure malfunction or valve leakage that require corrective action before the test, shall be included in the summary report required by V.B.</P>
<P>(c) The containment test conditions shall stabilize for a period of about 4 hours prior to the start of a leakage rate test.</P>
<P>(d) Those portions of the fluid systems that are part of the reactor coolant pressure boundary and are open directly to the containment atmosphere under post-accident conditions and become an extension of the boundary of the containment shall be opened or vented to the containment atmosphere prior to and during the test. Portions of closed systems inside containment that penetrate containment and rupture as a result of a loss of coolant accident shall be vented to the containment atmosphere. All vented systems shall be drained of water or other fluids to the extent necessary to assure exposure of the system containment isolation valves to containment air test pressure and to assure they will be subjected to the post accident differential pressure. Systems that are required to maintain the plant in a safe condition during the test shall be operable in their normal mode, and need not be vented. Systems that are normally filled with water and operating under post-accident conditions, such as the containment heat removal system, need not be vented. However, the containment isolation valves in the systems defined in III.A.1.(d) shall be tested in accordance with III.C. The measured leakage rate from these tests shall be included in the summary report required by V.B.</P>
@@ -5926,21 +4979,24 @@
<P>(2) <E T="03">Peak pressure tests.</E> The leakage rate Lam shall be less than 0.75 La. If local leakage measurements are taken to effect repairs in order to meet the acceptance criteria, these measurements shall be taken at a test pressure Pa.</P>
<P>6. <E T="03">Additional requirements.</E> (a) If any periodic Type A test fails to meet the applicable acceptance criteria in III.A.5.(b), the test schedule applicable to subsequent Type A tests will be reviewed and approved by the Commission.</P>
<P>(b) If two consecutive periodic Type A tests fail to meet the applicable acceptance criteria in III.A.5(b), notwithstanding the periodic retest schedule of III.D., a Type A test shall be performed at each plant shutdown for refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the acceptance criteria in III.A.5(b), after which time the retest schedule specified in III.D. may be resumed.</P>
- <P>B. <E T="03">Type B tests&#8212;</E>1. <E T="03">Test methods.</E> Acceptable means of performing preoperation and periodic Type B tests include:</P>
+ <P>B. <E T="03">Type B tests&#8212;</E></P>
+ <P>1. <E T="03">Test methods.</E> Acceptable means of performing preoperation and periodic Type B tests include:</P>
<P>(a) Examination by halide leak-detection method (or by other equivalent test methods such as mass spectrometer) of a test chamber, pressurized with air, nitrogen, or pneumatic fluid specified in the technical specifications or associated bases and constructed as part of individual containment penetrations.</P>
<P>(b) Measurement of the rate of pressure loss of the test chamber of the containment penetration pressurized with air, nitrogen, or pneumatic fluid specified in the technical specifications or associated bases.</P>
<P>(c) Leakage surveillance by means of a permanently installed system with provisions for continuous or intermittent pressurization of individual or groups of containment penetrations and measurement of rate of pressure loss of air, nitrogen, or pneumatic fluid specified in the technical specification or associated bases through the leak paths.</P>
<P>2. <E T="03">Test pressure.</E> All preoperational and periodic Type B tests shall be performed by local pneumatic pressurization of the containment penetrations, either individually or in groups, at a pressure not less than Pa.</P>
<P>3. <E T="03">Acceptance criteria.</E> (See also Type C tests.) (a) The combined leakage rate of all penetrations and valves subject to Type B and C tests shall be less than 0.60 La, with the exception of the valves specified in III.C.3.</P>
<P>(b) Leakage measurements obtained through component leakage surveillance systems (e.g., continuous pressurization of individual containment components) that maintains a pressure not less than Pa at individual test chambers of containment penetrations during normal reactor operation, are acceptable in lieu of Type B tests.</P>
- <P>C. <E T="03">Type C tests&#8212;</E>1. <E T="03">Test method.</E> Type C tests shall be performed by local pressurization. The pressure shall be applied in the same direction as that when the value would be required to perform its safety function, unless it can be determined that the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results. The test methods in III.B.1 may be substituted where appropriate. Each valve to be tested shall be closed by normal operation and without any preliminary exercising or adjustments (e.g., no tightening of valve after closure by valve motor).</P>
+ <P>C. <E T="03">Type C tests&#8212;</E></P>
+ <P>1. <E T="03">Test method.</E> Type C tests shall be performed by local pressurization. The pressure shall be applied in the same direction as that when the value would be required to perform its safety function, unless it can be determined that the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results. The test methods in III.B.1 may be substituted where appropriate. Each valve to be tested shall be closed by normal operation and without any preliminary exercising or adjustments (e.g., no tightening of valve after closure by valve motor).</P>
<P>2. <E T="03">Test pressure.</E> (a) Valves, unless pressurized with fluid (e.g., water, nitrogen) from a seal system, shall be pressurized with air or nitrogen at a pressure of Pa.<PRTPAGE P="1027"/>
</P>
<P>(b) Valves, which are sealed with fluid from a seal system shall be pressurized with that fluid to a pressure not less than 1.10 Pa.</P>
<P>3. <E T="03">Acceptance criterion.</E> The combined leakage rate for all penetrations and valves subject to Type B and C tests shall be less than 0.60 La. Leakage from containment isolation valves that are sealed with fluid from a seal system may be excluded when determining the combined leakage rate: <E T="03">Provided,</E> That;</P>
<P>(a) Such valves have been demonstrated to have fluid leakage rates that do not exceed those specified in the technical specifications or associated bases, and</P>
<P>(b) The installed isolation valve seal-water system fluid inventory is sufficient to assure the sealing function for at least 30 days at a pressure of 1.10 Pa.</P>
- <P>D. <E T="03">Periodic retest schedule&#8212;</E>1. <E T="03">Type A test.</E> (a) After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections. <SU footnote="Such inservice inspections are required by &#167; 50.55a.">2</SU><FTREF/>
+ <P>D. <E T="03">Periodic retest schedule&#8212;</E></P>
+ <P>1. <E T="03">Type A test.</E> (a) After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections. <SU footnote="Such inservice inspections are required by &#167; 50.55a.">2</SU><FTREF/>
</P>
<FTNT>
<P>
@@ -5961,7 +5017,8 @@
<P>B. <E T="03">Multiple leakage barrier or subatmospheric containments.</E> The primary reactor containment barrier of a multiple barrier or subatmospheric containment shall be subjected to Type A tests to verify that its leakage rate meets the requirements of this appendix. Other structures of multiple barrier or subatmospheric containments (e.g., secondary containments for boiling water reactors and shield buildings for pressurized water reactors that enclose the entire primary reactor containment or portions thereof) shall be subject to individual tests in accordance with the procedures specified in the technical specifications, or associated bases.</P>
<HD SOURCE="HD1">V. Inspection and Reporting of Tests</HD>
<P>A. <E T="03">Containment inspection.</E> A general inspection of the accessible interior and exterior surfaces of the containment structures and components shall be performed prior to any Type A test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak-tightness. If there is evidence of structural deterioration, Type A tests shall not be performed until corrective action is taken in <PRTPAGE P="1028"/>accordance with repair procedures, non destructive examinations, and tests as specified in the applicable code specified in &#167; 50.55a at the commencement of repair work. Such structural deterioration and corrective actions taken shall be included in the summary report required by V.B.</P>
- <P>B. <E T="03">Recordkeeping of test results.</E> 1. The preoperational and periodic tests must be documented in a readily available summary report that will be made available for inspection, upon request, at the nuclear power plant. The summary report shall include a schematic arrangement of the leakage rate measurement system, the instrumentation used, the supplemental test method, and the test program selected as applicable to the preoperational test, and all the subsequent periodic tests. The report shall contain an analysis and interpretation of the leakage rate test data for the Type A test results to the extent necessary to demonstrate the acceptability of the containment's leakage rate in meeting acceptance criteria.</P>
+ <P>B. <E T="03">Recordkeeping of test results.</E></P>
+ <P>1. The preoperational and periodic tests must be documented in a readily available summary report that will be made available for inspection, upon request, at the nuclear power plant. The summary report shall include a schematic arrangement of the leakage rate measurement system, the instrumentation used, the supplemental test method, and the test program selected as applicable to the preoperational test, and all the subsequent periodic tests. The report shall contain an analysis and interpretation of the leakage rate test data for the Type A test results to the extent necessary to demonstrate the acceptability of the containment's leakage rate in meeting acceptance criteria.</P>
<P>2. For each periodic test, leakage test results from Type A, B, and C tests shall be included in the summary report. The summary report shall contain an analysis and interpretation of the Type A test results and a summary analysis of periodic Type B and Type C tests that were performed since the last type A test. Leakage test results from type A, B, and C tests that failed to meet the acceptance criteria of III.A.5(b), III.B.3, and III.C.3, respectively, shall be included in a separate accompanying summary report that includes an analysis and interpretation of the test data, the least squares fit analysis of the test data, the instrumentation error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria. Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included.</P>
<HD SOURCE="HD1">Option B&#8212;Performance-Based Requirements</HD>
<HD SOURCE="HD2">Table of Contents</HD>
@@ -6029,13 +5086,15 @@
<P>Each evaluation model shall include a provision for predicting cladding swelling and rupture from consideration of the axial temperature distribution of the cladding and from the difference in pressure between the inside and outside of the cladding, both as functions of time. To be acceptable the swelling and rupture calculations shall be based on applicable data in such a way that <PRTPAGE P="1031"/>the degree of swelling and incidence of rupture are not underestimated. The degree of swelling and rupture shall be taken into account in calculations of gap conductance, cladding oxidation and embrittlement, and hydrogen generation.</P>
<P>The calculations of fuel and cladding temperatures as a function of time shall use values for gap conductance and other thermal parameters as functions of temperature and other applicable time-dependent variables. The gap conductance shall be varied in accordance with changes in gap dimensions and any other applicable variables.</P>
<HD SOURCE="HD2">C. Blowdown Phenomena</HD>
- <P>1. <E T="03">Break Characteristics and Flow.</E> a. In analyses of hypothetical loss-of-coolant accidents, a spectrum of possible pipe breaks shall be considered. This spectrum shall include instantaneous double-ended breaks ranging in cross-sectional area up to and including that of the largest pipe in the primary coolant system. The analysis shall also include the effects of longitudinal splits in the largest pipes, with the split area equal to the cross-sectional area of the pipe.</P>
+ <P>1. <E T="03">Break Characteristics and Flow.</E></P>
+ <P>a. In analyses of hypothetical loss-of-coolant accidents, a spectrum of possible pipe breaks shall be considered. This spectrum shall include instantaneous double-ended breaks ranging in cross-sectional area up to and including that of the largest pipe in the primary coolant system. The analysis shall also include the effects of longitudinal splits in the largest pipes, with the split area equal to the cross-sectional area of the pipe.</P>
<P>b. <E T="03">Discharge Model.</E> For all times after the discharging fluid has been calculated to be two-phase in composition, the discharge rate shall be calculated by use of the Moody model (F.J. Moody, &#8220;Maximum Flow Rate of a Single Component, Two-Phase Mixture,&#8221; Journal of Heat Transfer, Trans American Society of Mechanical Engineers, 87, No. 1, February, 1965). This publication has been approved for incorporation by reference by the Director of the Federal Register. A copy of this publication is available for inspection at the NRC Library, 11545 Rockville Pike, Rockville, Maryland 20852-2738. The calculation shall be conducted with at least three values of a discharge coefficient applied to the postulated break area, these values spanning the range from 0.6 to 1.0. If the results indicate that the maximum clad temperature for the hypothetical accident is to be found at an even lower value of the discharge coefficient, the range of discharge coefficients shall be extended until the maximum clad temperatures calculated by this variation has been achieved.</P>
<P>c. <E T="03">End of Blowdown.</E> (Applies Only to Pressurized Water Reactors.) For postulated cold leg breaks, all emergency cooling water injected into the inlet lines or the reactor vessel during the bypass period shall in the calculations be subtracted from the reactor vessel calculated inventory. This may be executed in the calculation during the bypass period, or as an alternative the amount of emergency core cooling water calculated to be injected during the bypass period may be subtracted later in the calculation from the water remaining in the inlet lines, downcomer, and reactor vessel lower plenum after the bypass period. This bypassing shall end in the calculation at a time designated as the &#8220;end of bypass,&#8221; after which the expulsion or entrainment mechanisms responsible for the bypassing are calculated not to be effective. The end-of-bypass definition used in the calculation shall be justified by a suitable combination of analysis and experimental data. Acceptable methods for defining &#8220;end of bypass&#8221; include, but are not limited to, the following: (1) Prediction of the blowdown calculation of downward flow in the downcomer for the remainder of the blowdown period; (2) Prediction of a threshold for droplet entrainment in the upward velocity, using local fluid conditions and a conservative critical Weber number.</P>
<P>d. <E T="03">Noding Near the Break and the ECCS Injection Points.</E> The noding in the vicinity of and including the broken or split sections of pipe and the points of ECCS injection shall be chosen to permit a reliable analysis of the thermodynamic history in these regions during blowdown.</P>
<P>2. <E T="03">Frictional Pressure Drops.</E> The frictional losses in pipes and other components including the reactor core shall be calculated using models that include realistic variation of friction factor with Reynolds number, and realistic two-phase friction multipliers that have been adequately verified by comparison with experimental data, or models that prove at least equally conservative with respect to maximum clad temperature calculated during the hypothetical accident. The modified Baroczy correlation (Baroczy, C. J., &#8220;A Systematic Correlation for Two-Phase Pressure Drop,&#8221; <E T="03">Chem. Enging. Prog. Symp. Series,</E> No. 64, Vol. 62, 1965) or a combination of the Thom correlation (Thom, J.R.S., &#8220;Prediction of Pressure Drop During Forced Circulation Boiling of Water,&#8221; <E T="03">Int. J. of Heat &amp; Mass Transfer,</E> 7, 709-724, 1964) for pressures equal to or greater than 250 psia and the Martinelli-Nelson correlation (Martinelli, R. C. Nelson, D.B., &#8220;Prediction of Pressure Drop During Forced Circulation Boiling of Water,&#8221; <E T="03">Transactions of ASME,</E> 695-702, 1948) for pressures lower than 250 psia is acceptable as a basis for calculating realistic two-phase friction multipliers.</P>
<P>3. <E T="03">Momentum Equation.</E> The following effects shall be taken into account in the conservation of momentum equation: (1) temporal change of momentum, (2) momentum convection, (3) area change momentum flux, (4) momentum change due to compressibility, (5) pressure loss resulting from wall friction, (6) pressure loss resulting from area change, and (7) gravitational acceleration. Any omission of one or more of these terms under stated circumstances shall be justified by comparative analyses or by experimental data.</P>
- <P>4. <E T="03">Critical Heat Flux.</E> a. Correlations developed from appropriate steady-state and transient-state experimental data are acceptable for use in predicting the critical heat flux <PRTPAGE P="1032"/>(CHF) during LOCA transients. The computer programs in which these correlations are used shall contain suitable checks to assure that the physical parameters are within the range of parameters specified for use of the correlations by their respective authors.</P>
+ <P>4. <E T="03">Critical Heat Flux.</E></P>
+ <P>a. Correlations developed from appropriate steady-state and transient-state experimental data are acceptable for use in predicting the critical heat flux <PRTPAGE P="1032"/>(CHF) during LOCA transients. The computer programs in which these correlations are used shall contain suitable checks to assure that the physical parameters are within the range of parameters specified for use of the correlations by their respective authors.</P>
<P>b. Steady-state CHF correlations acceptable for use in LOCA transients include, but are not limited to, the following:</P>
<P>(1) <E T="03">W 3.</E> L. S. Tong, &#8220;Prediction of Departure from Nucleate Boiling for an Axially Non-uniform Heat Flux Distribution,&#8221; <E T="03">Journal of Nuclear Energy,</E> Vol. 21, 241-248, 1967.</P>
<P>(2) <E T="03">B&amp;W-2.</E> J. S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, L. J. Stanek, &#8220;Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water,&#8221; <E T="03">Two-Phase Flow and Heat Transfer in Rod Bundles,</E> ASME, New York, 1969.</P>
@@ -6047,7 +5106,8 @@
<P>d. Transient CHF correlations acceptable for use in LOCA transients include, but are not limited to, the following:</P>
<P>(1) <E T="03">GE transient CHF.</E> B. C. Slifer, J. E. Hench, &#8220;Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors,&#8221; NEDO-10329, General Electric Company, Equation C-32, April 1971.</P>
<P>e. After CHF is first predicted at an axial fuel rod location during blowdown, the calculation shall not use nucleate boiling heat transfer correlations at that location subsequently during the blowdown even if the calculated local fluid and surface conditions would apparently justify the reestablishment of nucleate boiling. Heat transfer assumptions characteristic of return to nucleate boiling (rewetting) shall be permitted when justified by the calculated local fluid and surface conditions during the reflood portion of a LOCA.</P>
- <P>5. <E T="03">Post-CHF Heat Transfer Correlations.</E> a. Correlations of heat transfer from the fuel cladding to the surrounding fluid in the post-CHF regimes of transition and film boiling shall be compared to applicable steady-state and transient-state data using statistical correlation and uncertainty analyses. Such comparison shall demonstrate that the correlations predict values of heat transfer co-efficient equal to or less than the mean value of the applicable experimental heat transfer data throughout the range of parameters for which the correlations are to be used. The comparisons shall quantify the relation of the correlations to the statistical uncertainty of the applicable data.</P>
+ <P>5. <E T="03">Post-CHF Heat Transfer Correlations.</E></P>
+ <P>a. Correlations of heat transfer from the fuel cladding to the surrounding fluid in the post-CHF regimes of transition and film boiling shall be compared to applicable steady-state and transient-state data using statistical correlation and uncertainty analyses. Such comparison shall demonstrate that the correlations predict values of heat transfer co-efficient equal to or less than the mean value of the applicable experimental heat transfer data throughout the range of parameters for which the correlations are to be used. The comparisons shall quantify the relation of the correlations to the statistical uncertainty of the applicable data.</P>
<P>b. The Groeneveld flow film boiling correlation (equation 5.7 of D.C. Groeneveld, &#8220;An Investigation of Heat Transfer in the Liquid Deficient Regime,&#8221; AECL-3281, revised December 1969) and the Westinghouse correlation of steady-state transition boiling (&#8220;Proprietary Redirect/Rebuttal Testimony of Westinghouse Electric Corporation,&#8221; USNRC Docket RM-50-1, page 25-1, October 26, 1972) are acceptable for use in the post-CHF boiling regimes. In addition, the transition boiling correlation of McDonough, Milich, and King (J.B. McDonough, W. Milich, E.C. King, &#8220;An Experimental Study of Partial Film Boiling Region with Water at Elevated Pressures in a Round Vertical Tube,&#8221; Chemical Engineering Progress Symposium Series, Vol. 57, No. 32, pages 197-208, (1961) is suitable for use between nucleate and film boiling. Use of all these correlations is restricted as follows:</P>
<P>(1) The Groeneveld correlation shall not be used in the region near its low-pressure singularity,</P>
<P>(2) The first term (nucleate) of the Westinghouse correlation and the entire McDonough, Milich, and King correlation shall not be used during the blowdown after the temperature difference between the clad and the saturated fluid first exceeds 300 &#176;F,</P>
@@ -6064,7 +5124,8 @@
<P>3. <E T="03">Calculation of Reflood Rate for Pressurized Water Reactors.</E> The refilling of the reactor vessel and the time and rate of reflooding of the core shall be calculated by an acceptable model that takes into consideration the thermal and hydraulic characteristics of the core and of the reactor system. The primary system coolant pumps shall be assumed to have locked impellers if this assumption leads to the maximum calculated cladding temperature; otherwise the pump rotor shall be assumed to be running free. The ratio of the total fluid flow at the core exit plane to the total liquid flow at the core inlet plane (carryover fraction) shall be used to determine the core exit flow and shall be determined in accordance with applicable experimental data (for example, &#8220;PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Final Report,&#8221; Westinghouse Report WCAP-7665, April 1971; &#8220;PWR Full Length Emergency Cooling Heat Transfer (FLECHT) Group I Test Report,&#8221; Westinghouse Report WCAP-7435, January 1970; &#8220;PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Group II Test Report,&#8221; Westinghouse Report WCAP-7544, September 1970; &#8220;PWR FLECHT Final Report Supplement,&#8221; Westinghouse Report WCAP-7931, October 1972).</P>
<P>The effects on reflooding rate of the compressed gas in the accumulator which is discharged following accumulator water discharge shall also be taken into account.</P>
<P>4. <E T="03">Steam Interaction with Emergency Core Cooling Water in Pressurized Water Reactors.</E> The thermal-hydraulic interaction between steam and all emergency core cooling water shall be taken into account in calculating the core reflooding rate. During refill and reflood, the calculated steam flow in unbroken reactor coolant pipes shall be taken to be zero during the time that accumulators are discharging water into those pipes unless experimental evidence is available regarding the realistic thermal-hydraulic interaction between the steam and the liquid. In this case, the experimental data may be used to support an alternate assumption.</P>
- <P>5. <E T="03">Refill and Reflood Heat Transfer for Pressurized Water Reactors.</E> a. For reflood rates of one inch per second or higher, reflood heat transfer coefficients shall be based on applicable experimental data for unblocked cores including FLECHT results (&#8220;PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Final Report,&#8221; Westinghouse Report <PRTPAGE P="1034"/>WCAP-7665, April 1971). The use of a correlation derived from FLECHT data shall be demonstrated to be conservative for the transient to which it is applied; presently available FLECHT heat transfer correlations (&#8220;PWR Full Length Emergency Cooling Heat Transfer (FLECHT) Group I Test Report,&#8221; Westinghouse Report WCAP-7544, September 1970; &#8220;PWR FLECHT Final Report Supplement,&#8221; Westinghouse Report WCAP-7931, October 1972) are not acceptable. Westinghouse Report WCAP-7665 has been approved for incorporation by reference by the Director of the Federal Register. A copy of this report is available for inspection at the NRC Library, 11545 Rockville Pike, Rockville, Maryland 20852-2738. New correlations or modifications to the FLECHT heat transfer correlations are acceptable only after they are demonstrated to be conservative, by comparison with FLECHT data, for a range of parameters consistent with the transient to which they are applied.</P>
+ <P>5. <E T="03">Refill and Reflood Heat Transfer for Pressurized Water Reactors.</E></P>
+ <P>a. For reflood rates of one inch per second or higher, reflood heat transfer coefficients shall be based on applicable experimental data for unblocked cores including FLECHT results (&#8220;PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Final Report,&#8221; Westinghouse Report <PRTPAGE P="1034"/>WCAP-7665, April 1971). The use of a correlation derived from FLECHT data shall be demonstrated to be conservative for the transient to which it is applied; presently available FLECHT heat transfer correlations (&#8220;PWR Full Length Emergency Cooling Heat Transfer (FLECHT) Group I Test Report,&#8221; Westinghouse Report WCAP-7544, September 1970; &#8220;PWR FLECHT Final Report Supplement,&#8221; Westinghouse Report WCAP-7931, October 1972) are not acceptable. Westinghouse Report WCAP-7665 has been approved for incorporation by reference by the Director of the Federal Register. A copy of this report is available for inspection at the NRC Library, 11545 Rockville Pike, Rockville, Maryland 20852-2738. New correlations or modifications to the FLECHT heat transfer correlations are acceptable only after they are demonstrated to be conservative, by comparison with FLECHT data, for a range of parameters consistent with the transient to which they are applied.</P>
<P>b. During refill and during reflood when reflood rates are less than one inch per second, heat transfer calculations shall be based on the assumption that cooling is only by steam, and shall take into account any flow blockage calculated to occur as a result of cladding swelling or rupture as such blockage might affect both local steam flow and heat transfer.</P>
<P>6. <E T="03">Convective Heat Transfer Coefficients for Boiling Water Reactor Fuel Rods Under Spray Cooling.</E> Following the blowdown period, convective heat transfer shall be calculated using coefficients based on appropriate experimental data. For reactors with jet pumps and having fuel rods in a 7&#215;7 fuel assembly array, the following convective coefficients are acceptable:</P>
<P>a. During the period following lower plenum flashing but prior to the core spray reaching rated flow, a convective heat transfer coefficient of zero shall be applied to all fuel rods.</P>
@@ -6078,7 +5139,8 @@
<P>b. During the period after core spray reaches rated flow, but prior to wetting of the channel, a convective heat transfer coefficient of 5 Btu-hr <E T="51">&#8722;1</E>-ft <E T="51">&#8722;2</E>- &#176;F <E T="51">&#8722;1</E> shall be applied to both sides of the channel box.</P>
<P>c. Wetting of the channel box shall be assumed to occur 60 seconds after the time determined using the correlation based on the Yamanouchi analysis (&#8220;Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors,&#8221; General Electric Company Report NEDO-10329, April 1971). This report was approved for incorporation by reference by the Director of the Federal Register. A copy of the report is available for inspection at the NRC Library, 11545 Rockville Pike, Rockville, Maryland 20852-2738.</P>
<HD SOURCE="HD1">II. Required Documentation</HD>
- <P>1. a. A description of each evaluation model shall be furnished. The description shall be sufficiently complete to permit technical review of the analytical approach including the equations used, their approximations in difference form, the assumptions made, and the values of all parameters or the procedure for their selection, as for example, in accordance with a specified physical law or empirical correlation.</P>
+ <P>1.</P>
+ <P>a. A description of each evaluation model shall be furnished. The description shall be sufficiently complete to permit technical review of the analytical approach including the equations used, their approximations in difference form, the assumptions made, and the values of all parameters or the procedure for their selection, as for example, in accordance with a specified physical law or empirical correlation.</P>
<P>b. A complete listing of each computer program, in the same form as used in the evaluation model, must be furnished to the Nuclear Regulatory Commission upon request.</P>
<P>2. For each computer program, solution convergence shall be demonstrated by studies of system modeling or noding and calculational time steps.</P>
<P>3. Appropriate sensitivity studies shall be performed for each evaluation model, to evaluate the effect on the calculated results of variations in noding, phenomena assumed in the calculation to predominate, including pump operation or locking, and values of parameters over their applicable ranges. For items to which results are shown to be sensitive, the choices made shall be justified.</P>
@@ -6088,9 +5150,6 @@
<CITA>[39 FR 1003, Jan. 4, 1974, as amended at 51 FR 40311, Nov. 6, 1986; 53 FR 36005, Sept. 16, 1988; 57 FR 61786, Dec. 29, 1992; 59 FR 50689, Oct. 5, 1994; 60 FR 24552, May 9, 1995; 65 FR 34921, June 1, 2000]</CITA>
</APPENDIX>
<APPENDIX>
- <RESERVED>Appendixes L-M to Part 50 [Reserved]</RESERVED>
- </APPENDIX>
- <APPENDIX>
<EAR>Pt. 50, App. N</EAR>
<HD SOURCE="HED">Appendix N to Part 50&#8212;Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites</HD>
<P>Section 101 of the Atomic Energy Act of 1954, as amended, and &#167; 50.10 of this part require a Commission license to transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, use, import or export any production or utilization facility. The regulations in this part require the issuance of a construction permit by the Commission before commencement of construction of a production or utilization facility, except as provided in &#167; 50.10(e), and the issuance of an operating license before operation of the facility.</P>
@@ -6107,9 +5166,6 @@
<CITA>[40 FR 2977, Jan. 17, 1975, as amended at 49 FR 9405, Mar. 12, 1984; 51 FR 40311, Nov. 6, 1986; 70 FR 61888, Oct. 27, 2005]</CITA>
</APPENDIX>
<APPENDIX>
- <RESERVED>Appendixes O-P to Part 50 [Reserved]</RESERVED>
- </APPENDIX>
- <APPENDIX>
<EAR>Pt. 50, App. Q</EAR>
<HD SOURCE="HED">Appendix Q to Part 50&#8212;Pre-Application Early Review of Site Suitability Issues</HD>
<P>This appendix sets out procedures for the filing, Staff review, and referral to the Advisory Committee on Reactor Safeguards of requests for early review of one or more site suitability issues relating to the construction and operation of certain utilization facilities separately from and prior to the submittal of applications for construction permits for the facilities. The appendix also sets out procedures for the preparation and issuance of Staff Site Reports and for their incorporation by reference in applications for the construction and operation of certain utilization facilities. The utilization facilities are those which are subject to &#167; 51.20(b) of this chapter and are of the type specified in &#167; 50.21(b) (2) or (3) or &#167; 50.22 or are testing facilities. This appendix does not apply to proceedings conducted pursuant to subpart F of part 2 of this chapter.</P>
@@ -6187,7 +5243,8 @@
<P>Standpipe and hose stations shall be inside PWR containments and BWR containments that are not inerted. Standpipe and hose stations inside containment may be connected to a high quality water supply of sufficient quantity and pressure other than the fire main loop if plant-specific features prevent extending the fire main supply inside containment. For BWR drywells, standpipe and hose stations shall be placed outside the dry well with adequate lengths of hose to reach any location inside the dry well with an effective hose stream.</P>
<P>E. <E T="03">Hydrostatic hose tests.</E> Fire hose shall be hydrostatically tested at a pressure of 150 psi or 50 psi above maximum fire main operating pressure, whichever is greater. Hose stored in outside hose houses shall be tested annually. Interior standpipe hose shall be tested every three years.</P>
<P>F. <E T="03">Automatic fire detection.</E> Automatic fire detection systems shall be installed in all areas of the plant that contain or present an <PRTPAGE P="1039"/>exposure fire hazard to safe shutdown or safety-related systems or components. These fire detection systems shall be capable of operating with or without offsite power.</P>
- <P>G. <E T="03">Fire protection of safe shutdown capability.</E> 1. Fire protection features shall be provided for structures, systems, and components important to safe shutdown. These features shall be capable of limiting fire damage so that:</P>
+ <P>G. <E T="03">Fire protection of safe shutdown capability.</E></P>
+ <P>1. Fire protection features shall be provided for structures, systems, and components important to safe shutdown. These features shall be capable of limiting fire damage so that:</P>
<P>a. One train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) is free of fire damage; and</P>
<P>b. Systems necessary to achieve and maintain cold shutdown from either the control room or emergency control station(s) can be repaired within 72 hours.</P>
<P>2. Except as provided for in paragraph G.3 of this section, where cables or equipment, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:</P>
@@ -6278,7 +5335,8 @@
<P>h. Ventilation system operation that ensures desired plant air distribution when the ventilation flow is modified for fire containment or smoke clearing operations.</P>
<P>i. Operations requiring control room and shift engineer coordination or authorization.</P>
<P>j. Instructions for plant operators and general plant personnel during fire.</P>
- <P>L. <E T="03">Alternative and dedicated shutdown capability.</E> 1. Alternative or dedicated shutdown capability provided for a specific fire area shall be able to (a) achieve and maintain subcritical reactivity conditions in the reactor; (b) maintain reactor coolant inventory; (c) achieve and maintain hot standby <SU footnote="As defined in the Standard Technical Specifications.">2</SU><FTREF/> conditions for a PWR (hot shutdown <SU footnote="As defined in the Standard Technical Specifications.">2</SU>for a BWR); (d) achieve cold shutdown conditions within 72 hours; and (e) maintain cold shutdown conditions thereafter. During the postfire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of normal a.c. power, and the fission product boundary integrity shall not be affected; i.e., there shall be no fuel clad damage, rupture of any primary coolant boundary, of rupture of the containment boundary.</P>
+ <P>L. <E T="03">Alternative and dedicated shutdown capability.</E></P>
+ <P>1. Alternative or dedicated shutdown capability provided for a specific fire area shall be able to (a) achieve and maintain subcritical reactivity conditions in the reactor; (b) maintain reactor coolant inventory; (c) achieve and maintain hot standby <SU footnote="As defined in the Standard Technical Specifications.">2</SU><FTREF/> conditions for a PWR (hot shutdown <SU footnote="As defined in the Standard Technical Specifications.">2</SU>for a BWR); (d) achieve cold shutdown conditions within 72 hours; and (e) maintain cold shutdown conditions thereafter. During the postfire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of normal a.c. power, and the fission product boundary integrity shall not be affected; i.e., there shall be no fuel clad damage, rupture of any primary coolant boundary, of rupture of the containment boundary.</P>
<FTNT>
<P>
<SU>2</SU> As defined in the Standard Technical Specifications.</P>
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