Blowout Protection and 2200 Fahrenheit

I've already posted a lot on this matter.

Following is an e-mail I sent to each of the great Commissioners of the NRC on June 10, 2010.

Rescinding the Denial of PRM-50-76

To the Commissioners:

You are urged to direct the NRC staff to promptly write a document for your approval that rescinds the denial of PRM-50-76 that was published in the Federal Register on September 6, 2005 (NRC-2002-0019-00130). The attachment to this e-mail, Rebuttal SECY-05-0113 Leyse R, has Leyse’s rebuttal in 14 point type interspersed within the NRC’s draft notice, SECY-03-0113, dated June 29, 2005.

Rescinding the denial of PRM-50-76 is vital. Under current regulations at 1OCFR50.46(b)(1) and Appendix K to 1OCFR Part 50, our nuclear power plants lack sufficient measures to prevent runaway to meltdown when blowout occurs during a loss-of-coolant accident. The Baker-Just equation does not accurately and conservatively indicate the conditions under which an autocatalytic (runaway) oxidation reaction of Zircaloy begins.

Reference: Appendix K to Part 50--ECCS Evaluation Models

I. Required and Acceptable Features of the Evaluation Models

5. Metal--Water Reaction Rate. The rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation (Baker, L., Just, L.C., "Studies of Metal Water Reactions at High Temperatures, III. Experimental and Theoretical Studies of the Zirconium-Water Reaction," ANL-6548, page 7, May 1962).

Draft Final Rule 50.46a Note 2200 Fahrenheit

§ 50.46a DRAFT FINAL RULE LANGUAGE

Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements
(ADAMS Accession no. ML101250271)
In August 2009, the Nuclear Regulatory Commission (NRC) published for public comment a
supplemental proposed § 50.46a risk-informed emergency core cooling system (ECCS) rule
(74 FR 40006). The comment period ended on January 22, 2010. The NRC has evaluated the
public comments and has prepared draft final rule language. To facilitate public involvement in
this rulemaking, the NRC is making this draft rule language public. The NRC has tentative
plans to hold a public meeting in early June 2010 to discuss resolution of public comments and
the associated draft final rule language. In the future, if significant changes are made to this
rule, the NRC may periodically update the publicly available rule language.
NOTE: The availability of this draft rule language is intended to inform
stakeholders of the current status of the NRC’s activities to
provide a risk-informed alternative to the current ECCS
requirements. This draft rule language may be incomplete or
in error in one or more respects and may be subject to further
revisions during the rulemaking process. The NRC is not
requesting formal public comments on this draft rule
language.
Any questions on the requirements may be addressed to the NRC rulemaking project manager,
Richard Dudley (301-415-1116; richard.dudley@nrc.gov).
1 Revised: April 30, 2010
1 List of Subjects
2
3
4 10 CFR Part 50
5 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental
6 relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria,
7 Reporting and recordkeeping requirements.
8 10 CFR Part 52
9 Administrative practice and procedure, Antitrust, Backfitting, Combined license, Early
10 site permit, Emergency planning, Fees, Inspection, Limited work authorization, Nuclear power
11 plants and reactors, Probabilistic risk assessment, Prototype, Reactor siting criteria, Redress of
12 site, Reporting and recordkeeping requirements, Standard design, Standard design certification.
13 For the reasons set out in the preamble and under the authority of the Atomic Energy
14 Act of 1954, as amended; the Energy Reorganization Act of 1974; and 5 U.S.C. 553; the NRC is
15 adopting the following amendments to Title 10 of the Code of Federal Regulations (10 CFR)
16 Parts 50 and 52.
17 PART 50 -- DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES
18 1. The authority citation for part 50 continues to read as follows:
19 Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938,
20 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2132,
21 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
22 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750
23 (44 U.S.C. 3504 note); Energy policy Act of 2005, Pub. L. No. 109-58, 119 Stat. 194 (2005).
24 Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L.
25 102-486, sec. 2902, 106 Stat. 3123 (42 U.S.C. 5841). Section 50.10 also issued under secs.
26 101, 185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-190, 83 Stat.
2 Revised: April 30, 2010
853 (42 U.S.C. 4332). 1 Sections 50.13, 50.54(dd), and 50.103 also issued under sec. 108, 68
2 Stat. 939, as amended (42 U.S.C. 2138).
3 Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42
4 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-
5 190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88
6 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97-
7 415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42
8 U.S.C. 2152). Sections 50.80 - 50.81 also issued under sec. 184, 68 Stat. 954, as amended (42
9 U.S.C. 2234). Appendix F also issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237)
10 2. In § 50.34, paragraphs (a)(4) and (b)(4) are revised to read as follows:
11 § 50.34 Contents of application; technical information.
12
13 (a) * * *
14
15 (4) A preliminary analysis and evaluation of the design and performance of structures,
16 systems, and components of the facility with the objective of assessing the risk to public health
17 and safety resulting from operation of the facility and including determination of the margins of
18 safety during normal operations and transient conditions anticipated during the life of the facility,
19 and the adequacy of structures, systems, and components provided for the prevention of
20 accidents and the mitigation of the consequences of accidents.
21 (i) Analysis and evaluation of emergency core cooling system (ECCS) cooling
22 performance and the need for high point vents following postulated loss-of-coolant accidents
23 must be performed under the requirements of either § 50.46 or § 50.46a, and § 50.46b for
24 facilities whose operating licenses were issued after December 28, 1974, but before
25 [EFFECTIVE DATE OF RULE], and for facilities for which
3 Revised: April 30, 2010
construction permits may be issued after [EFFECTIVE DAT 1 E OF RULE] and are demonstrated
2 under § 50.46a(c)(2) to have designs that are similar to the designs of reactors licensed before
3 [EFFECTIVE DATE OF RULE].
4 (ii) Analysis and evaluation of ECCS cooling performance and the need for high point
5 vents following postulated loss-of-coolant accidents (LOCAs) must be performed under the
6 requirements of § 50.46 and § 50.46b for facilities for which construction permits may be issued
7 after [EFFECTIVE DATE OF RULE] and are not demonstrated under § 50.46a(c)(2) to have
8 designs that are similar to the designs of reactors licensed before [EFFECTIVE DATE OF
9 RULE].
10 * * * * *
11
12 (b) * * *
13
14 (4) A final analysis and evaluation of the design and performance of structures, systems,
15 and components with the objective stated in paragraph (a)(4) of this section and taking into
16 account any pertinent information developed since the submittal of the preliminary safety
17 analysis report.
18 (i) Analysis and evaluation of ECCS cooling performance following postulated LOCAs
19 must be performed under the requirements of either § 50.46 or § 50.46a, and § 50.46b for
20 facilities whose operating licenses were issued after December 28, 1974, but before
21 [EFFECTIVE DATE OF RULE], and for facilities whose operating licenses are issued after
22 [EFFECTIVE DATE OF RULE] and are demonstrated under § 50.46a(c)(2) to have designs that
23 are similar to the designs of reactors licensed before [EFFECTIVE DATE OF RULE].
24 (ii) Analysis and evaluation of ECCS cooling performance following postulated LOCAs
25
4 Revised: April 30, 2010
must be performed under the requirements of 1 §§ 50.46 and 50.46b for facilities whose operating
2 licenses are issued after [EFFECTIVE DATE OF RULE] and are not demonstrated under
3 § 50.46a(c)(2) to have designs that are similar to the designs of reactors licensed before
4 [EFFECTIVE DATE OF RULE].
5
6 * * * * *
7
8 3. In § 50.46, paragraph (a) is amended by adding an introductory paragraph and
9 revising paragraph (a)(1)(i) to read as follows:
10 § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear
11 power plants.
12 (a) Each boiling or pressurized light-water nuclear power reactor fueled with uranium
13 oxide pellets within cylindrical zircalloy or ZIRLO cladding must be provided with an ECCS. The
14 ECCS system must be designed under the requirements of this section or § 50.46a for facilities
15 whose operating licenses were issued before [EFFECTIVE DATE OF RULE]; for facilities
16 whose operating licenses, combined licenses under part 52 of this chapter, or manufacturing
17 licenses under part 52 of this chapter are issued after [EFFECTIVE DATE OF RULE] and are
18 demonstrated under § 50.46a(c)(2) to have designs that are similar to the designs of reactors
19 licensed before [EFFECTIVE DATE OF RULE]; and for design approvals and design
20 certifications under part 52 of this chapter issued after [EFFECTIVE DATE OF RULE] that are
21 demonstrated under § 50.46a(c)(2) to have designs that are similar to the designs of reactors
22 licensed before [EFFECTIVE DATE OF RULE]. The ECCS system must be designed under the
23 requirements of this section for facilities whose operating licenses, combined licenses under
24 part 52 of this chapter, or manufacturing licenses under part 52 of this chapter are issued after
25
5 Revised: April 30, 2010
1 [EFFECTIVE DATE OF RULE] and are not
2 demonstrated under § 50.46a(c)(2) to have designs that are similar to the designs of reactors
3 licensed before [EFFECTIVE DATE OF RULE]; and for design approvals and design
4 certifications under part 52 of this chapter that are not demonstrated under § 50.46a(c)(2) to
5 have designs that are similar to the designs of reactors licensed before [EFFECTIVE DATE OF
6 RULE].
7 (1)(i) The ECCS system must be designed so that its calculated cooling performance
8 following postulated LOCAs conforms to the criteria set forth in paragraph (b) of this section.
9 ECCS cooling performance must be calculated in accordance with an acceptable evaluation
10 model and must be calculated for a number of postulated LOCAs of different sizes, locations,
11 and other properties sufficient to provide assurance that the most severe postulated LOCAs are
12 calculated. Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must
13 include sufficient supporting justification to show that the analytical technique realistically
14 describes the behavior of the reactor system during a LOCA. Comparisons to applicable
15 experimental data must be made and uncertainties in the analysis method and inputs must be
16 identified and assessed so that the uncertainty in the calculated results can be estimated. This
17 uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is
18 compared to the criteria set forth in paragraph (b) of this section, there is a high level of
19 probability that the criteria would not be exceeded. Appendix K, Part II Required
20 Documentation, sets forth the documentation requirements for each evaluation model. This
21 section does not apply to a nuclear power reactor facility for which the certifications required
22 under § 50.82(a)(1) have been submitted.
23
24 * * * * *
25
6 Revised: April 30, 2010
4. Section 50.46a 1 is redesignated as § 50.46b, and a new § 50.46a is added to read as
2 follows:
3 § 50.46a Alternative acceptance criteria for emergency core cooling systems for light4
water nuclear power reactors.
5 (a) Definitions. For the purposes of this section:
6 (1) Changes enabled by this section means changes to the facility, technical
7 specifications, and procedures that satisfy the alternative ECCS analysis requirements under
8 this section but do not satisfy the ECCS requirements under 10 CFR 50.46.
9 (2) Evaluation model means the calculational framework for evaluating the behavior of
10 the reactor system during a postulated design-basis loss-of-coolant accident (LOCA). It
11 includes one or more computer programs and all other information necessary for application of
12 the calculational framework to a specific LOCA, such as mathematical models used,
13 assumptions included in the programs, procedure for treating the program input and output
14 information, specification of those portions of analysis not included in computer programs,
15 values of parameters, and all other information necessary to specify the calculational procedure.
16 (3) Loss-of-coolant accidents (LOCAs) means the hypothetical accidents that would
17 result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant
18 makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and
19 including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor
20 coolant system. LOCAs involving breaks at or below the transition break size (TBS) are
21 design-basis accidents. LOCAs involving breaks larger than the TBS are beyond design-basis
22 accidents.
23 (4) Operating configuration means those plant characteristics, such as power level,
24 equipment unavailability (including unavailability caused by corrective and preventive
25
7 Revised: April 30, 2010
maintenance), and equipment capabilit 1 y that affect plant response to a LOCA.
2 (5) Transition break size (TBS) for reactors licensed before [EFFECTIVE DATE OF
3 RULE] is a break area equal to the cross-sectional flow area of the inside diameter of the largest
4 piping attached to the reactor coolant system for a pressurized water reactor, or the inside
5 diameter of the larger of the feedwater line inside containment or the residual heat removal line
6 inside containment for a boiling water reactor. For reactors licensed after [EFFECTIVE DATE
7 OF RULE], and for design certifications, design approvals, and and manufacturing licenses
8 approved or issued after [EFFECTIVE DATE OF RULE], the TBS will be determined on a plant9
specific basis.
10 (b) Applicability and scope.
11 (1) The requirements of this section may be applied to each boiling or pressurized
12 light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircalloy or
13 ZIRLO cladding whose operating license was issued prior to [EFFECTIVE DATE OF RULE]; to
14 each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets
15 within cylindrical zircalloy or ZIRLO cladding whose operating license, combined license under
16 part 52 of this chapter or manufacturing license under part 52 of this chapter is issued after
17 [EFFECTIVE DATE OF RULE] and whose design is demonstrated under § 50.46a(c)(2) to be
18 similar to the designs of reactors licensed before [EFFECTIVE DATE OF RULE]; and to each
19 boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within
20 cylindrical zircalloy or ZIRLO cladding whose design approval or design certification under part
21 52 of this chapter is demonstrated under § 50.46a(c)(2) to be similar to the designs of reactors
22 licensed before [EFFECTIVE DATE OF RULE]. The requirements of this section do not apply
23 to a reactor for which the certification required under § 50.82(a)(1) has been submitted.
24 (2) The requirements of this section are in addition to any other requirements applicable
8 Revised: April 30, 2010
to ECCS set forth in this part, with the exception of § 50.46. The 1 criteria set forth in paragraphs
2 (e)(3) and (e)(4) of this section, with cooling performance calculated in accordance with an
3 acceptable evaluation model or analysis method under paragraphs (e)(1) and (e)(2) of this
4 section, are in implementation of the general requirements with respect to ECCS cooling
5 performance design set forth in this part, including in particular Criterion 35 of Appendix A to this
6 part.
7 (c) Application. (1) An applicant, permit holder, or licensee of a facility, or other entity
8 seeking to implement this section shall submit an application for a license amendment under
9 § 50.90 that contains the following information:
10 (i) A written evaluation demonstrating applicability of the results in NUREG-1829,
11 “Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the Elicitation Process”;
12 March 2008 and NUREG-1903, “Seismic Considerations for the Transition Break Size”;
13 February 2008” to the licensee’s facility. As part of this evaluation, the application must contain
14 a plant specific analysis demonstrating that the risk of seismically-induced LOCAs larger than
15 the TBS is comparable to or less than the seismically-induced LOCA risk associated with the
16 NUREG-1903 results.
17 (ii) Identification of the approved analysis method(s) for demonstrating compliance with
18 the ECCS criteria in paragraph (e) of this section.
19 (iii) A description of the risk-informed evaluation process used in evaluating whether
20 proposed changes to the facility meet the requirements in paragraphs (d)(5) and (f), of this
21 section.
22 (iv) An applicant, permit holder, or licensee of a facility or other entity who wishes to
23 make changes enabled by this section without prior NRC review and approval must submit for
24 NRC approval a process to be used for evaluating the acceptability of these changes; including:
9 Revised: April 30, 2010
(A) A description of the approach, methods, 1 and decision making process to be used for
2 evaluating compliance with the acceptance criteria in paragraphs (f)(1), (f)(2), and (f)(3) of this
3 section, and
4 (B) A description of the PRA model and non-PRA risk assessment methods to be used
5 for demonstrating compliance with paragraphs (f)(4) and (f)(5) of this section.
6 (v) A description of non safety equipment that is credited for demonstrating compliance
7 with the ECCS acceptance criteria in paragraph (e) of this section.
8 (2) An applicant for a construction permit, operating license, design approval, design
9 certification, manufacturing license, or combined license seeking to implement the requirements
10 of this section shall, in addition to the information required by paragraph (c)(1) of this section,
11 submit an analysis demonstrating why the proposed reactor design is similar to the designs of
12 reactors licensed before [EFFECTIVE DATE OF RULE] such that the provisions of this section
13 may properly apply. The analysis must also include a recommendation for an appropriate TBS
14 and a justification that the recommended TBS is consistent with the technical basis for this
15 section.
16 (3) Acceptance criteria. The NRC may approve an application to use this section if:
17 (i) The evaluation submitted under paragraph (c)(1)(i) of this section demonstrates that
18 the NUREG-1829 results are applicable to the facility, and the risk of seismically-induced
19 LOCAs larger than the TBS is comparable to or less than the seismically-induced LOCA risk
20 associated with the NUREG-1903 results;
21 (ii) The method(s) for demonstrating compliance with the ECCS acceptance criteria in
22 paragraphs (e)(3) and (e)(4) of this section meet the requirements in paragraphs (e)(1) and
23 (e)(2) of this section;
10 Revised: April 30, 2010
(iii) The risk-informed evaluation process 1 the licensee proposes to use for making
2 changes enabled by this section is adequate for determining whether the acceptance criteria in
3 paragraphs (d)(5) and (f) of this section, have been met; and
4 (iv) Non safety equipment that is credited for demonstrating compliance with the ECCS
5 acceptance criteria in paragraph (e) of this section is identified in plant Technical Specifications.
6 (v) For all applicants other than those holding operating licenses issued before
7 [EFFECTIVE DATE OF RULE], the proposed reactor design is similar to the designs of reactors
8 licensed before [EFFECTIVE DATE OF RULE] and the applicant’s proposed TBS is consistent
9 with the technical basis of this section.
10 (d) Requirements during operation. A licensee whose application under paragraph (c) of
11 this section is approved by the NRC shall comply with the following requirements as long as the
12 facility is subject to the requirements in this section until the licensee submits the certifications
13 required by § 50.82(a):
14 (1) The licensee shall maintain ECCS model(s) and/or analysis method(s) meeting the
15 requirements in paragraphs (e)(1) and (e)(2) of this section;
16 (2) The licensee shall have leak detection systems available at the facility and shall
17 implement actions as necessary to identify, monitor and quantify leakage to ensure that adverse
18 safety consequences do not result from primary pressure boundary leakage from piping and
19 components that are larger than the transition break size.
20 (3) A change enabled by this section must, in addition to meeting other applicable NRC
21 requirements, be evaluated by a risk-informed evaluation demonstrating that the acceptance
22 criteria in paragraph (f) of this section are met.
23 (4) The licensee shall periodically maintain and upgrade, as necessary, its risk
24 assessments to meet the requirements in paragraph (f)(4) and (f)(5) of this section. The
11 Revised: April 30, 2010
maintenance and upgrading 1 shall be consistent with NRC-endorsed consensus standards on
2 PRA and must be completed in a timely manner, but no less often than once every four years.
3 Based upon a re-evaluation of the risk assessments after the periodic maintenance and
4 upgrading are completed, the licensee shall take appropriate action to ensure that the
5 acceptance criteria in paragraph (f) of this section, as applicable, are met. The PRA
6 maintenance and upgrading required by this section, and any necessary changes to the facility,
7 technical specifications and procedures as a result of this re-evaluation, shall not be deemed to
8 be backfitting under any provision of this chapter.
9 (5) For LOCAs larger than the TBS, operation in a plant operating configuration not
10 demonstrated to meet the acceptance criteria in paragraph (e)(4) of this section may not exceed
11 a short time. A short time is either a total of fourteen (14) days in any 12 month period or an
12 NRC-approved alternative time proposed by the licensee that is commensurate with the
13 mitigation capability available and has been demonstrated to be acceptable by a risk-informed
14 evaluation of the configuration-specific risk, defense-in-depth, and safety margins.
15 (6) The licensee shall perform an evaluation to determine the effect of all planned facility
16 changes and shall not implement any facility change that would invalidate the evaluation
17 performed pursuant to § 50.46a(c)(1)(i) demonstrating the applicability to the licensee’s facility
18 of the generic results in NUREG-1829 and NUREG-1903.
19 (e) ECCS Performance. Each nuclear power reactor or nuclear power reactor design
20 subject to this section must be provided with an ECCS that must be designed so that its
21 calculated cooling performance following postulated LOCAs conforms to the criteria set forth in
22 this section. The evaluation models for LOCAs must meet the criteria in this paragraph, and
23 must be approved for use by the NRC. Appendix K, Part II, to 10 CFR Part 50, sets forth the
24 documentation requirements for evaluation models.
12 Revised: April 30, 2010
(1) ECCS evaluation f 1 or LOCAs involving breaks at or below the TBS. ECCS cooling
2 performance at or below the TBS must be calculated in accordance with an evaluation model
3 that meets the requirements of either section I to Appendix K of this part, or the following
4 requirements, and must demonstrate that the acceptance criteria in paragraph (e)(3) of this
5 section are satisfied. The evaluation model must be used for a number of postulated LOCAs of
6 different sizes, locations, and other properties sufficient to provide assurance that the most
7 severe postulated LOCAs involving breaks at or below the TBS are analyzed. The evaluation
8 model must include sufficient supporting justification to show that the analytical technique
9 realistically describes the behavior of the reactor system during a LOCA. Comparisons to
10 applicable experimental data must be made and uncertainties in the analysis method and inputs
11 must be identified and assessed so that the uncertainty in the calculated results can be
12 estimated. This uncertainty must be accounted for, so that when the calculated ECCS cooling
13 performance is compared to the criteria set forth in paragraph (e)(3) of this section, there is a
14 high level of probability that the criteria would not be exceeded.
15 (2) ECCS analyses for LOCAs involving breaks larger than the TBS. ECCS cooling
16 performance for LOCAs involving breaks larger than the TBS must be calculated in accordance
17 with an evaluation model that meets the requirements of either section I to Appendix K of this
18 part, or the following requirements, and must demonstrate that the acceptance criteria in
19 paragraph (e)(4) of this section are satisfied. The evaluation model must include sufficient
20 supporting justification to show that the analytical technique realistically describes the behavior
21 of the reactor system during a LOCA. Comparisons to applicable experimental data must be
22 made and uncertainties in the analysis method and inputs must be identified and assessed so
23 that the uncertainty in the calculated results can be estimated. This uncertainty must be
24 accounted for, so that when the calculated ECCS cooling performance is compared to the
13 Revised: April 30, 2010
criteria set forth in paragraph (e)(4) of this section, there is a 1 high level of probability that the
2 criteria would not be exceeded. The evaluation model must be used for a number of postulated
3 LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the
4 most severe postulated LOCAs larger than the TBS up to the double-ended rupture of the
5 largest pipe in the reactor coolant system are analyzed. These calculations may take credit for
6 the availability of offsite power and do not require the assumption of a single failure. Realistic
7 initial conditions and availability of safety-related or non safety-related equipment may be
8 assumed if supported by plant-specific data or analysis, and provided that onsite power can be
9 readily provided through simple manual actions to equipment that is credited in the analysis.
10 (3) Acceptance criteria for LOCAs involving breaks at or below the TBS. The following
11 acceptance criteria must be used in determining the acceptability of ECCS cooling performance:
12 (i) Peak cladding temperature. The calculated maximum fuel element cladding
13 temperature must not exceed 2200°F.
14 (ii) Maximum cladding oxidation. The calculated total oxidation of the cladding must not
15 at any location exceed 0.17 times the total cladding thickness before oxidation. As used in this
16 paragraph, total oxidation means the total thickness of cladding metal that would be locally
17 converted to oxide if all the oxygen absorbed by and reacted with the cladding locally were
18 converted to stoichiometric zirconium dioxide. If cladding rupture is calculated to occur, the
19 inside surfaces of the cladding must be included in the oxidation, beginning at the calculated
20 time of rupture. Cladding thickness before oxidation means the radial distance from inside to
21 outside the cladding, after any calculated rupture or swelling has occurred but before significant
22 oxidation. Where the calculated conditions of transient pressure and temperature lead to a
23 prediction of cladding swelling, with or without cladding rupture, the unoxidized cladding
24 thickness must be defined as the cladding cross-sectional area, taken at a horizontal plane at
14 Revised: April 30, 2010
the elevation of the rupture, if it occurs, or at the elevation of the 1 highest cladding temperature if
2 no rupture is calculated to occur, divided by the average circumference at that elevation. For
3 ruptured cladding the circumference does not include the rupture opening.
4 (iii) Maximum hydrogen generation. The calculated total amount of hydrogen generated
5 from the chemical reaction of the cladding with water or steam must not exceed 0.01 times the
6 hypothetical amount that would be generated if all of the metal in the cladding cylinders
7 surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
8 (iv) Coolable geometry. Calculated changes in core geometry must be such that the
9 core remains amenable to cooling.
10 (v) Long term cooling. After any calculated successful initial operation of the ECCS, the
11 calculated core temperature must be maintained at an acceptably low value and decay heat
12 must be removed for the extended period of time required by the long-lived radioactivity
13 remaining in the core.
14 (4) Acceptance criteria for LOCAs involving breaks larger than the TBS. The following
15 acceptance criteria must be used in determining the acceptability of ECCS cooling performance:
16 (i) Coolable geometry. Calculated changes in core geometry must be such that the core
17 remains amenable to cooling.
18 (ii) Long term cooling. After any calculated successful initial operation of the ECCS, the
19 calculated core temperature must be maintained at an acceptably low value and decay heat
20 must be removed for the extended period of time required by the long-lived radioactivity
21 remaining in the core.
22 (5) Imposition of restrictions. The Director of the Office of Nuclear Reactor Regulation or
23 the Office of New Reactors may impose restrictions on reactor operation if it is found that the
24 evaluations of ECCS cooling performance submitted are not consistent with paragraph (e) of
25 this section.
15 Revised: April 30, 2010
(f) Changes to facility, technical specifications, or procedures. 1 An applicant, permit
2 holder, or licensee or other entity who wishes to make changes enabled by this rule, to the
3 facility, facility design, or procedures or to the technical specifications shall perform a
4 risk-informed evaluation.
5 (1) The licensee may make changes, other than changes to the technical specifications,
6 without prior NRC approval if:
7 (i) The change is permitted under § 50.59 for holders of operating licenses, combined
8 licenses that do not reference a design certification, design approval, or manufacturing license
9 (per § 52.98(b)), or combined licenses that reference a design approval; permitted under
10 § 52.98(c) for holders of combined licenses that reference a design certification; or permitted
11 under § 52.98(d) for holders of combined licenses that reference a manufacturing license,
12 (ii) The risk informed evaluation process described in paragraph (c)(1)(iv) of this section
13 demonstrates that any increases in the estimated risk are minimal and the criteria in paragraph
14 (f)(3) of this section are met, and
15 (iii) The change does not invalidate the evaluation performed pursuant to paragraph
16 (c)(1)(i) of the applicability of the results in NUREG-1829 and NUREG-1903 to the licensee’s
17 facility.
18 (2) For implementing changes which are not permitted under paragraph (f)(1) of this
19 section, the permit holder, or licensee must submit an application for license amendment under
20 § 50.90. The application must contain:
21 (i) The information required under § 50.90;
22 (ii) Information from the risk-informed evaluation demonstrating that the total increases in
23 core damage frequency and large early release frequency are very small and the overall risk
24 remains small, and the criteria in paragraph (f)(3) of this section are met;
16 Revised: April 30, 2010
(iii) If previous changes 1 have been made under § 50.46a, information from the
2 risk-informed evaluation on the cumulative effect on risk of the proposed change and all
3 previous changes made under this section. If more than one plant change is combined;
4 including plant changes not enabled by this section, into a group for the purposes of evaluating
5 acceptable risk increases; the evaluation of each individual change shall be performed along
6 with the evaluation of combined changes; and
7 (iv) Information demonstrating that the criteria in paragraphs (e)(3) and (e)(4) of this
8 section are met.
9 (v) Information demonstrating that the proposed change will not increase the LOCA
10 frequency of the facility (including the frequency of seismically-induced LOCAs) by an amount
11 that would invalidate the applicability to the facility of the generic studies
12 (NUREG 1829,”Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the
13 Elicitation Process”, March 2008 and NUREG-1903, “Seismic Considerations for the Transition
14 Break Size”, February 2008”) that support the technical basis for this section.
15 (3) All changes enabled by this rule must meet the following criteria:
16 (i) Adequate defense in depth is maintained;
17 (ii) Adequate safety margins are retained to account for uncertainties; and
18 (iii) Adequate performance-measurement programs are implemented to ensure the
19 risk-informed evaluation continues to reflect actual plant design and operation. These programs
20 shall be designed to detect degradation of the system, structure or component before plant
21 safety is compromised, provide feedback of information and timely corrective actions, and
22 monitor systems, structures or components at a level commensurate with their safety
23 significance, and
24
25 (iv) For applicants or licensees referencing a certified design, will not result in a
17 Revised: April 30, 2010
significant decrease in the level of safety otherwise pr 1 ovided by the certified design.
2 (4) Requirements for risk assessment - PRA. Whenever a PRA is used in the
3 risk-informed evaluation, the PRA must, with respect to the area of evaluation which is the
4 subject of the PRA:
5 (i) Address initiating events from sources both internal and external to the plant and for
6 all modes of operation, including low power and shutdown modes, that would affect the
7 regulatory decision in a substantial manner;
8 (ii) Reasonably represent the current configuration and operating practices at the plant;
9 (iii) Have sufficient technical adequacy (including consideration of uncertainty) and level
10 of detail to provide confidence that the total risk estimate and the change in total risk estimate
11 adequately reflect the plant and the effect of the proposed change on risk; and
12 (iv) Be determined, through peer review, to meet industry standards for PRA quality that
13 have been endorsed by the NRC.
14 (5) Requirements for risk assessment other than PRA. Whenever risk assessment
15 methods other than PRAs are used to develop quantitative or qualitative estimates of changes
16 to risk in the risk-informed evaluation, an integrated, systematic process must be used. All
17 aspects of the analyses must reasonably reflect the current plant configuration and operating
18 practices, and applicable plant and industry operating experience.
19 (g) Reporting.
20 (1) Licensees. (i) Each licensee shall estimate the effect of any change to or error in
21 evaluation models or analysis methods or in the application of such models or methods to
22 determine if the change or error is significant. For each change to or error discovered in an
23 ECCS evaluation model or analysis method or in the application of such a model that affects the
24
18 Revised: April 30, 2010
calculated results, the licensee shall report the nature of the change 1 or error and its estimated
2 effect on the limiting ECCS analysis to the Commission at least annually as specified in §§ 50.4
3 or 52.3. If the change or error is significant, the licensee shall provide this report within 30 days
4 and include with the report a proposed schedule for providing a reanalysis or taking other action
5 as may be needed to show compliance with § 50.46a requirements. This schedule may be
6 developed using an integrated scheduling system previously approved for the facility by the
7 NRC. For those facilities not using an NRC-approved integrated scheduling system, a schedule
8 will be established by the NRC staff within 60 days of receipt of the proposed schedule. Any
9 change or error correction that results in a calculated ECCS performance that does not conform
10 to the criteria set forth in paragraphs (e)(3) or (e)(4) of this section is a reportable event as
11 described in §§ 50.55(e), 50.72 and 50.73. The licensee shall propose immediate steps to
12 demonstrate compliance or bring plant design or operation into compliance with § 50.46a
13 requirements. For the purpose of this paragraph, a significant change or error is:
14 (A) For LOCAs involving pipe breaks at or below the TBS, one which results either in a
15 calculated peak fuel cladding temperature different by more than 50°F from the temperature
16 calculated for the limiting transient using the last acceptable model, or is a cumulation of
17 changes and errors such that the sum of the absolute magnitudes of the respective temperature
18 changes is greater than 50°F; or
19 (B) For LOCAs involving pipe breaks larger than the TBS, one which results in a
20 significant reduction in the capability to meet the requirements of paragraph (e)(4) of this
21 section.
22 (ii) As part of the PRA maintenance and upgrading under paragraph (d)(4) of this
23 section, the licensee shall report to the NRC if the re-evaluation results in exceeding the
24 acceptance criteria in paragraph (f) of this section, as applicable. The report must be filed with
25
19 Revised: April 30, 2010
the NRC no more than 60 days after completing the PRA r 1 e-evaluation. The report must
2 describe and explain the changes in the PRA modeling, plant design, or plant operation that led
3 to the increase(s) in risk, and must include a description of and implementation schedule for any
4 corrective actions required under paragraph (d)(4) of this section.
5 (iii) Every 24 months, the licensee shall submit, as specified in §§ 50.4 or 52.3, a short
6 description of each change involving minimal changes in risk made under paragraph (f)(1) of
7 this section after the last report and a brief summary of the basis for the licensee’s
8 determination pursuant to § 50.46a(f)(2)(vi) that the change does not invalidate the applicability
9 evaluation made under § 50.46a(c)(1)(i).
10 (2) Design certifications; applicants for and holders of design approvals. Each design
11 certification applicant and each applicant for and holder of a design approval shall report to the
12 NRC any significant error in evaluation models or analysis methods or in the application of such
13 models or methods. A design certification applicant’s duty to report under this paragraph
14 continues until the later of either the termination or expiration of the design certification; or the
15 termination of the last license directly or indirectly referencing the design certification. For the
16 purpose of this paragraph, a significant change or error is:
17 (i) For LOCAs involving pipe breaks at or below the TBS, one which results either in a
18 calculated peak fuel cladding temperature different by more than 50 °F from the temperature
19 calculated for the limiting transient using the last acceptable model, or is a cumulation of
20 changes and errors such that the sum of the absolute magnitudes of the respective temperature
21 changes is greater than 50 °F; or
22 (ii) For LOCAs involving pipe breaks larger than the TBS, one which results in a
23 significant reduction in the capability to meet the requirements of paragraph (e)(4) of this
24 section.
20 Revised: April 30, 2010
(h) Documentation. Following implementation of 1 the § 50.46a requirements, each entity
2 subject to this section shall maintain records sufficient to demonstrate compliance with the
3 requirements in this section in accordance with § 50.71.
4 (i) through (l) - [RESERVED]
5 (m) Changes to TBS.
6 (1) Operating licenses under Part 50, combined licenses under Part 52, and
7 manufacturing licenses. If the NRC increases the TBS specified in this section, each licensee
8 subject to this section (other than a licensee referencing a design certification rule complying
9 with the requirements of this section) shall re-perform the evaluations required by paragraphs
10 (e)(1) and (e)(2) of this section and reconfirm compliance with the acceptance criteria in
11 paragraphs (e)(3) and (e)(4) of this section. If the licensee cannot demonstrate compliance with
12 the acceptance criteria, then the licensee shall change its facility, technical specifications or
13 procedures so that the acceptance criteria are met. The evaluation required by this paragraph,
14 and any necessary changes to the facility, technical specifications or procedures as the result of
15 this evaluation, are not to be deemed to be backfitting under any provision of this chapter or a
16 violation of any finality provision in Part 52.
17 (2) Design certifications and referencing combined licenses. Changes to a TBS for a
18 design certification must be accomplished by rulemaking, in accordance with 10 CFR 52.63(a).
19 Holders of combined licenses referencing a design certification rule shall re perform the
20 evaluations required by paragraphs (e)(1) and (e)(2) of this section and reconfirm compliance
21 with the acceptance criteria in paragraphs (e)(3) and (e)(4) of this section. If the licensee cannot
22 demonstrate compliance with the acceptance criteria, then the licensee shall change its facility,
23 technical specifications or procedures so that the acceptance criteria are met. These actions are
24 deemed to be in conformance with applicable finality provisions in Part 52.
25
21 Revised: April 30, 2010
5. In ' 50.109, paragraph (b) 1 is revised to read as follows:
2 ' 50.109 Backfitting.
3
4 * * * * *
5
6 (b) Paragraph (a)(3) of this section shall not apply to:
7 (1) Backfits imposed prior to October 21, 1985; and
8 (2) Any changes made to the TBS specified in ' 50.46a or as otherwise applied to a
9 licensee.
10 * * * * *
11 6. In Appendix A to 10 CFR Part 50, under the heading, ACRITERIA,@ Criterion 17, 35,
12 38, 41, 44, and 50 are revised to read as follows:
13 APPENDIX A TO PART 50 -GENERAL DESIGN CRITERIA FOR NUCLEAR POWER
14 PLANTS
15
16 * * * * *
17 CRITERIA
18 * * * * *
19
20 Criterion 17--Electrical power systems. An on-site electric power system and an offsite
21 electric power system shall be provided to permit functioning of structures, systems, and
22 components important to safety. The safety function for each system (assuming the other
23 system is not functioning) shall be to provide sufficient capacity and capability to assure that (1)
24 specified acceptable fuel design limits and design conditions of the reactor coolant pressure
22 Revised: April 30, 2010
boundary are not exceeded as a result of 1 anticipated operational occurrences and (2) the core
2 is cooled and containment integrity and other vital functions are maintained in the event of
3 postulated accidents.
4 The onsite electric power supplies, including the batteries, and the onsite electrical
5 distribution system, shall have sufficient independence, redundancy, and testability to perform
6 their safety functions assuming a single failure, except for loss of coolant accidents involving
7 pipe breaks larger than the transition break size under ' 50.46a, where a single failure of the
8 onsite power supplies and electrical distribution system need not be assumed for plants under
9 ' 50.46a. For those pipe breaks only, neither a single failure nor the unavailability of offsite
10 power need be assumed.
11 Electric power from the transmission network to the onsite electric distribution system
12 shall be supplied by two physically independent circuits (not necessarily on separate rights of
13 way) designed and located so as to minimize to the extent practical the likelihood of their
14 simultaneous failure under operating and postulated accident conditions. A switchyard common
15 to both circuits is acceptable. Each of these circuits shall be designed to be available in
16 sufficient time following a loss of all onsite alternating current power supplies and the other
17 offsite electric power circuit, to assure that specified acceptable fuel design limits and design
18 conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits
19 shall be designed to be available within a few seconds following a LOCA to assure that core
20 cooling, containment integrity, and other vital safety functions are maintained.
21 Provisions shall be included to minimize the probability of losing electric power from any
22 of the remaining supplies as a result of, or coincident with, the loss of power generated by the
23 nuclear power unit, the loss of power from the transmission network, or the loss of power from
24 the onsite electric power supplies.
25
23 Revised: April 30, 2010
1 * * * * *
2
3 Criterion 35--Emergency core cooling. A system to provide abundant emergency core
4 cooling shall be provided. The system safety function shall be to transfer heat from the reactor
5 core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could
6 interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is
7 limited to negligible amounts.
8 Suitable redundancy in components and features, and suitable interconnections, leak
9 detection, isolation, and containment capabilities shall be provided to assure that for onsite
10 electric power system operation (assuming offsite power is not available) and for offsite electric
11 power system operation (assuming onsite power is not available) the system safety function can
12 be accomplished, assuming a single failure, except for loss of coolant accidents involving pipe
13 breaks larger than the transition break size under ' 50.46a. For those pipe breaks only, neither
14 a single failure nor the unavailability of offsite power need be assumed.
15
16 * * * * *
17
18 Criterion 38--Containment heat removal. A system to remove heat from the reactor
19 containment shall be provided. The system safety function shall be to reduce rapidly, consistent
20 with the functioning of other associated systems, the containment pressure and temperature
21 following any LOCA and maintain them at acceptably low levels.
22 Suitable redundancy in components and features, and suitable interconnections, leak
23 detection, isolation, and containment capabilities shall be provided to assure that for onsite
24 electric power system operation (assuming offsite power is not available) and for offsite electric
24 Revised: April 30, 2010
power system operation (assuming onsite 1 power is not available) the system safety function can
2 be accomplished, assuming a single failure, except for analysis of loss of coolant accidents
3 involving pipe breaks larger than the transition break size under ' 50.46a. For those pipe
4 breaks only, neither a single failure nor the unavailability of offsite power need be assumed.
5
6 * * * * *
7
8 Criterion 41--Containment atmosphere cleanup. Systems to control fission products,
9 hydrogen, oxygen, and other substances which may be released into the reactor containment
10 shall be provided as necessary to reduce, consistent with the functioning of other associated
11 systems, the concentration and quality of fission products released to the environment following
12 postulated accidents, and to control the concentration of hydrogen or oxygen and other
13 substances in the containment atmosphere following postulated accidents to assure that
14 containment integrity is maintained.
15 Each system shall have suitable redundancy in components and features, and suitable
16 interconnections, leak detection, isolation, and containment capabilities to assure that for onsite
17 electric power system operation (assuming offsite power is not available) and for offsite electric
18 power system operation (assuming onsite power is not available) its safety function can be
19 accomplished, assuming a single failure, except for analysis of loss of coolant accidents
20 involving pipe breaks larger than the transition break size under ' 50.46a. For those pipe
21 breaks only, neither a single failure nor the unavailability of offsite power need be assumed.
22
23 * * * * *
24
25 Revised: April 30, 2010
Criterion 44--Cooling water. A system to transfer heat 1 from structures, systems, and
2 components important to safety, to an ultimate heat sink shall be provided. The system safety
3 function shall be to transfer the combined heat load of these structures, systems, and
4 components under normal operating and accident conditions.
5 Suitable redundancy in components and features, and suitable interconnections, leak
6 detection, and isolation capabilities shall be provided to assure that for onsite electric power
7 system operation (assuming offsite power is not available) and for offsite electric power system
8 operation (assuming onsite power is not available) the system safety function can be
9 accomplished, assuming a single failure, except for analysis of loss of coolant accidents
10 involving pipe breaks larger than the transition break size under ' 50.46a. For those pipe
11 breaks only, neither a single failure nor the unavailability of offsite power need be assumed.
12
13 * * * * *
14
15 Criterion 50--Containment design basis. The reactor containment structure, including
16 access openings, penetrations, and the containment heat removal system shall be designed so
17 that the containment structure and its internal compartments can accommodate, without
18 exceeding the design leakage rate and with sufficient margin, the calculated pressure and
19 temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect
20 consideration of (1) the effects of potential energy sources which have not been included in the
21 determination of the peak conditions, such as energy in steam generators and as required by
22 ' 50.44 energy from metal-water and other chemical reactions that may result from degradation
23 but not total failure of emergency core cooling functioning, (2) the limited experience and
24 experimental data available for defining accident phenomena and containment responses, and
25
26 Revised: April 30, 2010
(3) the conservatism 1 of the calculational model and input parameters.
2 For licensees voluntarily choosing to comply with ' 50.46a, the structural and leak tight
3 integrity of the reactor containment structure, including access openings, penetrations, and its
4 internal compartments, shall be maintained for realistically calculated pressure and temperature
5 conditions resulting from any loss of coolant accident larger than the transition break size.
6
7 * * * * *
8
9 PART 52 - LICENSES, CERTIFICATIONS AND APPROVALS FOR NUCLEAR POWER
10 PLANTS
11 7. The authority citation for part 52 continues to read as follows:
12 AUTHORITY: Secs. 103, 104, 161, 182, 183, 185, 186, 189, 68 Stat. 936, 948, 953, 954,
13 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2133, 2201, 2232, 2233,
14 2235, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended
15 (42U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); Energy Policy
16 Act of 2005, Pub. L. No. 109-58, 119 Stat. 594 (2005), secs. 147 and 149 of the Atomic Energy
17 Act.
18 8. In § 52.47, paragraph (a)(4) is revised to read as follows:
19 § 52.47 Contents of applications; technical information
20
21 (a) * * *
22
23 (4) An analysis and evaluation of the design and performance of structures, systems,
24 and components with the objective of assessing the risk to public health and safety resulting
27 Revised: April 30, 2010
from operation of the facility and including determination of the 1 margins of safety during normal
2 operations and transient conditions anticipated during the life of the facility, and the adequacy of
3 structures, systems, and components provided for the prevention of accidents and the mitigation
4 of the consequences of accidents.
5 (i) Analysis and evaluation of emergency core cooling system (ECCS) cooling
6 performance and the need for high-point vents following postulated loss-of-coolant accidents
7 may be performed under the requirements of either § 50.46 or § 50.46a and § 50.46b of this
8 chapter for designs certified after [EFFECTIVE DATE OF RULE] and demonstrated under
9 § 50.46a(c)(2) of this chapter to be similar to reactor designs licensed before [EFFECTIVE
10 DATE OF RULE], or
11 (ii) Analysis and evaluation of ECCS cooling performance and the need for high-point
12 vents following postulated loss-of-coolant accidents must be performed under the requirements
13 of §§ 50.46 and 50.46b of this chapter for designs that are not demonstrated under
14 § 50.46a(c)(2) of this chapter to be similar to reactor designs licensed before [EFFECTIVE
15 DATE OF RULE].
16
17 * * *
18
19 § 52.54 Issuance of standard design certification.
20
21 * * *
22
23 (b) The design certification rule must specify the site parameters, design
24 characteristics, and any additional requirements and restrictions of the design certification rule.
28 Revised: April 30, 2010
A design certification rule which was reviewed and appr 1 oved as meeting the requirements of
2 10 CFR 50.46a must specify the criteria governing departures that a referencing combined
3 license must meet. The criteria must ensure that the safety bases for the NRC’s approval of the
4 certified design’s compliance with § 50.46a (including applicability of the TBS) continue to apply
5 despite the departure.
6
7 * * * * *
8
9 9. In § 52.79, paragraph (a)(5) is revised to read as follows:
10 § 52.79 Contents of applications; technical information in final safety analysis report.
11
12 (a) * * *
13
14 (5) An analysis and evaluation of the design and performance of structures, systems,
15 and components with the objective of assessing the risk to public health and safety resulting
16 from operation of the facility and including determination of the margins of safety during normal
17 operations and transient conditions anticipated during the life of the facility, and the adequacy of
18 structures, systems, and components provided for the prevention of accidents and the mitigation
19 of the consequences of accidents.
20 (i) Analysis and evaluation of ECCS cooling performance and the need for high-point
21 vents following postulated loss-of-coolant accidents must be performed under the requirements
22 of either § 50.46 or § 50.46a and § 50.46b of this chapter for facilities licensed after
23 [EFFECTIVE DATE OF RULE] and demonstrated under § 50.46a(c)(2) of this chapter to be
24 similar to reactor designs licensed before [EFFECTIVE DATE OF RULE], or
29 Revised: April 30, 2010
(ii) Analysis and evaluation of 1 ECCS cooling performance and the need for high-point
2 vents following postulated loss-of-coolant accidents must be performed under the requirements
3 of §§ 50.46 and 50.46b of this chapter for facilities licensed after [EFFECTIVE DATE OF RULE]
4 and not demonstrated under § 50.46a(c)(2) of this chapter to be similar to reactor designs
5 licensed before [EFFECTIVE DATE OF RULE].
6
7 * * * * *
8
9 10. In § 52.137, paragraph (a)(4) is revised to read as follows:
10 § 52.137 Contents of applications; technical information.
11
12 (a) * * *
13
14 (4) An analysis and evaluation of the design and performance of SSCs with the objective
15 of assessing the risk to public health and safety resulting from operation of the facility and
16 including determination of the margins of safety during normal operations and transient
17 conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the
18 prevention of accidents and the mitigation of the consequences of accidents.
19 (i) Analysis and evaluation of ECCS cooling performance and the need for high-point
20 vents following postulated loss-of-coolant accidents must be performed under the requirements
21 of either § 50.46 or § 50.46a and § 50.46b of this chapter for designs approved after
22 [EFFECTIVE DATE OF RULE] and demonstrated under § 50.46a(c)(2) of this chapter to be
23 similar to reactor designs licensed before [EFFECTIVE DATE OF RULE], or
24 (ii) Analysis and evaluation of ECCS cooling performance and the need for high-point
30 Revised: April 30, 2010
vents following postulated loss-of-coolant accidents must be 1 performed under the requirements
2 of §§ 50.46 and 50.46b of this chapter for designs that are not demonstrated under
3 § 50.46a(c)(2) of this chapter to be similar to reactor designs licensed before [EFFECTIVE
4 DATE OF RULE].
5
6 * * * * *
7
8 11. In § 52.157, paragraph (f)(1) is revised to read as follows:
9 § 52.157 Contents of applications; technical information in final safety analysis report.
10
11 (f) * * *
12
13 (1) An analysis and evaluation of the design and performance of structures, systems,
14 and components with the objective of assessing the risk to public health and safety resulting
15 from operation of the facility and including determination of the margins of safety during normal
16 operations and transient conditions anticipated during the life of the facility, and the adequacy of
17 structures, systems, and components provided for the prevention of accidents and the mitigation
18 of the consequences of accidents.
19 (i) Analysis and evaluation of ECCS cooling performance and the need for high-point
20 vents following postulated loss-of-coolant accidents must be performed under the requirements
21 of either § 50.46 or § 50.46a and § 50.46b of this chapter for facilities licensed after
22 [EFFECTIVE DATE OF RULE] and demonstrated under § 50.46a(c)(2) to be similar to reactor
23 designs licensed before [EFFECTIVE DATE OF RULE], or
24 (ii) Analysis and evaluation of ECCS cooling performance and the need for high-point
31 Revised: April 30, 2010
vents following postulated loss-of-coolant accidents must be 1 performed under the requirements
2 of §§ 50.46 and 50.46b of this chapter for facilities licensed after [EFFECTIVE DATE OF RULE]
3 and not demonstrated under § 50.46a(c)(2) of this chapter to be similar to reactor designs
4 licensed before [EFFECTIVE DATE OF RULE].
5
6 * * * * *
7
8 Dated at Rockville, Maryland, this day of , 2010.
9 For the Nuclear Regulatory Commission.
10
11
12
13
14 R. W. Borchardt,
15 Executive Director
16 for Operations
17
18

10 CFR 50.46a – Risk-Informed ECCS Requirements Meeting June 4, 2010

Here is the agenda:

Public Meeting Agenda
10 CFR 50.46a – Risk-Informed ECCS Requirements
June 4, 2010 (8:30 am to 12:30 pm)
Two White Flint North Room T08 A01
Rockville, Md



Welcoming Remarks, Purpose, and Meeting Conduct 8:30 am – 8:40 am
Richard Dudley

Introductory Remarks 8:40 am – 8:50 am
William Ruland, Director
Division of Safety Systems

Discussion of Public Comment Resolution and Rule Changes 8:50 am – 10:20 am

Comments on Applying § 50.46a to New Reactors 8:50 am – 8:55 am
Richard Dudley, Stephen Downey

Comments Related to Risk-Assessment 8:55 am – 10:25 am
Stephen Dinsmore

Break 10:25 am – 10:40 am

Comments Related to the Transition Break Size 10:40 am – 10:50 am
Richard Dudley

Comments on Demonstrating the Applicability of Generic Studies 10:50 am – 11:05 am
Richard Dudley, Robert Tregoning

Comments on Enhanced Leak Detection 11:05 am – 11:15 am
Richard Dudley, Robert Hardies

Comments on Thermal-Hydraulic Analysis 11:15 am – 11:25 am
Timothy Collins

Comments Related to Petitions for Rulemaking 11:25 am – 11:45 am
Richard Dudley

Additional Stakeholder Questions or Discussion 11:45 am – 12:15 pm


Adjourn 12:15 pm

2200°F is alive, but it has never been well!

I have "parked the following documents here. Sorting and discussion will follow at a later date.May 21, 2010
MEMORANDUM TO: Robert D. Carlson, Chief
Financial Analysis Branch
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
FROM: Richard F. Dudley, Senior Project Manager /RA/
Financial Analysis Branch
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
SUBJECT: FORTHCOMING PUBLIC MEETING TO DISCUSS
NRC STAFF RESOLUTION OF PUBLIC COMMENTS ON
SUPPLEMENTAL PROPOSED RULE ON RISK-INFORMED CHANGES
TO LOSS-OF-COOLANT ACCIDENT TECHNICAL REQUIREMENTS
(10 CFR 50.46a) AND DRAFT FINAL RULE LANGUAGE
DATE & TIME: Friday, June 4, 2010
8:30 a.m. – 12:30 p.m.
LOCATION: U.S. Nuclear Regulatory Commission
Two White Flint North, Room T-2B3
11545 Rockville Pike
Rockville, Maryland 20852
PURPOSE: To discuss (1) NRC responses to major comments received on the
10 CFR 50.46a Supplemental Proposed Rule published on August 10,
2009 (74 FR 40006) and (2) NRC’s associated draft final rule
language.
Draft final rule language is available for review in preparation for this
meeting at ADAMS Accession Number ML101250271 and on
Regulations.gov under Docket ID NRC-2004-0006. Note that the NRC is
not soliciting or accepting formal public comments at this time. This
preliminary rule language has not been fully vetted through internal
NRC reviews, and therefore may be subject to significant revision before
formal publication.
MEETING CONTACT: Richard F. Dudley, NRR
301-415-1116
Richard.Dudley@nrc.gov
- 2 -
CATEGORY 3: * This is a Category 3 Meeting: The public is invited to participate in this
meeting by providing comments and asking questions throughout the
meeting.
The NRC provides reasonable accommodations to individuals with
disabilities where appropriate. If you need a reasonable accommodation
to participate in this meeting (e.g., sign language), or need this meeting
notice or other information from the meeting in another format, please
notify the NRC meeting contact by May 28, 2010, so that arrangements
can be made.
Teleconferencing: Interested members of the public unable to attend the
meeting may participate by telephone via a toll-free teleconference. For
details, please call the NRC meeting contact listed or the toll-free number,
1-800-368-5642, and ask the operator for a connection to the meeting
contact. Those interested in participating in this meeting by
teleconference should call or email the meeting contact as soon as
possible, but no later than May 28, 2010.
PARTICIPANTS: Participants from the NRC include members of the Office of Nuclear
Reactor Regulation (NRR), the Office of New Reactors (NRO), and the
Office of Nuclear Regulatory Research (RES).
NRC Industry
T. Collins, NRR B. Bradley, NEI
S. Dinsmore, NRR V. Anderson, NEI
R. Landry, NRO
Docket ID: NRC-2004-0006; RIN 3150-AH29
Enclosure:
Meeting Agenda
* Commission’s Policy Statement on “Enhancing Public Participation in NRC Meetings”
(67 FR 36920, May 28, 2002)
- 2 -
CATEGORY 3: * This is a Category 3 Meeting: The public is invited to participate in this
meeting by providing comments and asking questions throughout the
meeting.
The NRC provides reasonable accommodations to individuals with
disabilities where appropriate. If you need a reasonable accommodation
to participate in this meeting (e.g., sign language), or need this meeting
notice or other information from the meeting in another format, please
notify the NRC meeting contact by May 28, 2010, so that arrangements
can be made.
Teleconferencing: Interested members of the public unable to attend the
meeting may participate by telephone via a toll-free teleconference. For
details, please call the NRC meeting contact listed or the toll-free number,
1-800-368-5642, and ask the operator for a connection to the meeting
contact. Those interested in participating in this meeting by
teleconference should call or email the meeting contact as soon as
possible, but no later than May 28, 2010.
PARTICIPANTS: Participants from the NRC include members of the Office of Nuclear
Reactor Regulation (NRR), the Office of New Reactors (NRO), and the
Office of Nuclear Regulatory Research (RES).
NRC Industry
T. Collins, NRR B. Bradley, NEI
S. Dinsmore, NRR V. Anderson, NEI
R. Landry, NRO
Docket ID: NRC-2004-0006; RIN 3150-AH29
Enclosure:
Meeting Agenda
* Commission’s Policy Statement on “Enhancing Public Participation in NRC Meetings”
(67 FR 36920, May 28, 2002)
DISTRIBUTION:
See next page
ADAMS Accession No: ML101400381
OFFICE NRR/PFAB: PM NRR/PFAB: BC
NAME RDudley RCarlson
DATE 5/21 /10 5/21 /10
OFFICIAL RECORD COPY
Memo to Robert D. Carlson from Richard F. Dudley dated May , 2010
SUBJECT: FORTHCOMING CATEGORY 3 PUBLIC MEETING TO DISCUSS NRC STAFF
RESOLUTION OF PUBLIC COMMENTS ON SUPPLEMENTAL PROPOSED RULE ON RISKINFORMED
CHANGES TO LOSS-OF-COOLANT ACCIDENT TECHNICAL REQUIREMENTS (10
CFR 50.46a) AND DRAFT FINAL RULE LANGUAGE
DISTRIBUTION:
Public
PFAB r/f
PMNS
RidsNrrOd
RidsNrrDpr
RCarlson
RDudley
Enclosure
Public Meeting Agenda
10 CFR 50.46a – Risk-Informed ECCS Requirements
June 4, 2010 (8:30 am to 12:30 pm)
Two White Flint North Room T08 A01
Rockville, Md
Friday, June 4, 2010
Welcoming Remarks, Purpose and Meeting Conduct 8:30 am – 8:40 am
R. Dudley
Introductory Remarks 8:40 am – 8:50 am
Summary of Public Comment Resolution 8:50 am – 9:00 am
R. Dudley
Discussion of Public Comments and Rule Changes Related to 9:00 am – 10:30 am
Risk Assessment S. Dinsmore
Break
Response to Stakeholder Questions 10:45am – 11:15 am
R. Dudley/all
Additional Opportunity for Stakeholder Discussion 11:15 am – 12:15 pm
NRC Closing Remarks and Adjournment 12:15 pm – 12:30 pm

The telecom bridge no. for this meeting is 888-455-9746; pass code = 19442.
Richard Dudley 415-1115


1.
(80)
50.46a Draft Final Rule Language, "Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements".
ML101250271
2010-05-11
32

§ 50.46a DRAFT FINAL RULE LANGUAGE

Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements

(ADAMS Accession no. ML101250271)

3. In § 50.46, paragraph (a) is amended by adding an introductory paragraph and revising paragraph (a)(1)(i) to read as follows:

§ 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power plants.

(1)(i) The ECCS system must be designed so that its calculated cooling performance following postulated LOCAs conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated LOCAs of different sizes, locations,
and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated. Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must
include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a LOCA. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded. Appendix K, Part II Required Documentation, sets forth the documentation requirements for each evaluation model. This section does not apply to a nuclear power reactor facility for which the certifications required under § 50.82(a)(1) have been submitted.

(1) ECCS evaluation f 1 or LOCAs involving breaks at or below the TBS. ECCS cooling performance at or below the TBS must be calculated in accordance with an evaluation model that meets the requirements of either section I to Appendix K of this part, or the following requirements, and must demonstrate that the acceptance criteria in paragraph (e)(3) of this section are satisfied. The evaluation model must be used for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs involving breaks at or below the TBS are analyzed. The evaluation
model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a LOCA. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (e)(3) of this section, there is a high level of probability that the criteria would not be exceeded.

(3) Acceptance criteria for LOCAs involving breaks at or below the TBS. The following acceptance criteria must be used in determining the acceptability of ECCS cooling performance:

(i) Peak cladding temperature. The calculated maximum fuel element cladding temperature must not exceed 2200°F.

(ii) Maximum cladding oxidation. The calculated total oxidation of the cladding must not at any location exceed 0.17 times the total cladding thickness before oxidation. As used in this paragraph, total oxidation means the total thickness of cladding metal that would be locally converted to oxide if all the oxygen absorbed by and reacted with the cladding locally were converted to stoichiometric zirconium dioxide. If cladding rupture is calculated to occur, the inside surfaces of the cladding must be included in the oxidation, beginning at the calculated time of rupture. Cladding thickness before oxidation means the radial distance from inside to
outside the cladding, after any calculated rupture or swelling has occurred but before significant oxidation. Where the calculated conditions of transient pressure and temperature lead to a prediction of cladding swelling, with or without cladding rupture, the unoxidized cladding thickness must be defined as the cladding cross-sectional area, taken at a horizontal plane at
the elevation of the rupture, if it occurs, or at the elevation of the 1 highest cladding temperature if no rupture is calculated to occur, divided by the average circumference at that elevation. For ruptured cladding the circumference does not include the rupture opening.

(iii) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam must not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(iv) Coolable geometry. Calculated changes in core geometry must be such that the core remains amenable to cooling.

(v) Long term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature must be maintained at an acceptably low value and decay heat must be removed for the extended period of time required by the long-lived radioactivity remaining in the core.


APPENDIX A TO PART 50 -GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS

CRITERIA

Criterion 35--Emergency core cooling. A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.