# Economic Simplified Boiling-Water Reactor Design Certification > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/fr%3A2014-24362 ## Record - **Collection:** Federal Register - **Document type:** Rule - **Published:** October 15, 2014 - **Citation:** 79 FR 61944 ## Text Nuclear Regulatory Commission 10 CFR Part 52 [NRC-2010-0135] RIN 3150-AI85 Economic Simplified Boiling-Water Reactor Design Certification AGENCY: Nuclear Regulatory Commission. ACTION: Final rule. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is adopting a new rule certifying the Economic Simplified Boiling-Water Reactor (ESBWR) standard plant design. This action is necessary so that applicants or licensees intending to construct and operate an ESBWR design may do so by referencing this design certification rule (DCR). The applicant for certification of the ESBWR design is GE-Hitachi Nuclear Energy (GEH). DATES: This final rule is effective on November 14, 2014. The incorporation by reference of certain publications listed in this regulation is approved by the Director of the Office of the Federal Register (OFR) as of November 14, 2014. ADDRESSES: Please refer to Docket ID NRC-2010-0135 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods: • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2010-0135. Address questions about NRC dockets to Carol Gallagher, telephone: 301-287-3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individuals listed in the FOR FURTHER INFORMATION CONTACT section of this document. • NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. For the convenience of the reader, instructions about obtaining materials referenced in this document are provided in a table in Section VII, “Availability of Documents,” of this document. • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. FOR FURTHER INFORMATION CONTACT: George M. Tartal, Office of New Reactors, telephone: 301-415-0016, email: George.Tartal@nrc.gov; or David Misenhimer, Office of New Reactors, telephone: 301-415-6590, email: David.Misenhimer@nrc.gov; U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. SUPPLEMENTARY INFORMATION: Executive Summary A. Need for the Regulatory Action The NRC is amending its regulations related to licenses, certifications, and approvals for nuclear power plants. This final rule certifies the ESBWR standard plant design. This action is necessary so that applicants or licensees intending to construct and operate an ESBWR design may do so by referencing this DCR. B. Major Provisions Major provisions of the final rule include changes to: • specify which documents contain the requirements for the ESBWR design, • specify how a nuclear power plant license applicant can reference the ESBWR design, • describe how the NRC considers matters within the scope of the design to be resolved for proceedings involving a license or application referencing the ESBWR design, and • describe the processes for changes to and departures from the ESBWR design. C. Costs and Benefits The NRC did not prepare a regulatory analysis to determine the expected quantitative or qualitative costs and benefits of the final rule. The NRC prepares regulatory analyses for rulemakings that establish generic regulatory requirements applicable to all licensees. Design certifications are not generic rulemakings in the sense that design certifications do not establish standards or requirements with which all licensees must comply. Rather, design certifications are NRC approvals of specific nuclear power plant designs by rulemaking, which then may be voluntarily referenced by an applicant for a combined license (COL). Furthermore, design certification rulemakings are initiated by an applicant for a design certification, rather than the NRC. Preparation of a regulatory analysis in this circumstance would not be useful because the design to be certified is proposed by the applicant rather than the NRC. For these reasons, the NRC concludes that preparation of a regulatory analysis is neither required nor appropriate. Table of Contents I. Background II. Summary and Analysis of Public Comments on the ESBWR Proposed Rule and Supplemental Proposed Rule A. Overview of Public Comments B. Comments Regarding Technical Content in the Design Control Document C. Comments Regarding NRC's Response to Fukushima Dai-ichi Accident III. Regulatory and Policy Issues A. How the ESBWR Design Addresses Fukushima Near Term Task Force (NTTF) Recommendations B. Incorporation by Reference of Public Documents and Issue Resolution Associated With Non-Public Documents C. Changes to Tier 2* Information D. Change Control for Severe Accident Design Features E. Access to Safeguards Information (SGI) and Sensitive Unclassified Non-Safeguards Information (SUNSI) F. Human Factors Engineering (HFE) Operational Program Elements Exclusion From Finality G. Other Changes to the ESBWR Rule Language and Difference Between the ESBWR Rule and Other DCRs IV. Technical Issues A. Regulatory Treatment of Nonsafety Systems (RTNSS) B. Containment Performance C. Control Room Cooling D. Feedwater Temperature Operating Domain E. Steam Dryer Analysis Methodology F. Aircraft Impact Assessment (AIA) G. American Society of Mechanical Engineers (ASME) Code Case N-782 H. Exemption for the Safety Parameter Display System I. Hurricane-Generated Winds and Missiles J. Loss of One or More Phases of Offsite Power K. Spent Fuel Assembly Integrity in Spent Fuel Racks L. Turbine Building Offgas System Design Requirements M. ASME Boiler and Pressure Vessel Code (BPV Code) Statement in Chapter 1 of the ESBWR Design Control Document (DCD) N. Clarification of ASME Component Design Inspections, Tests, Analyses, and Acceptance Criteria (ITAACs) O. Corrections, Editorial, and Conforming Changes V. Rulemaking Procedure A. Exclusions From Issue Finality and Issue Resolution for Spent Fuel Pool Instrumentation B. Incorporation by Reference of Public Documents C. Changes to Tier 2* Information D. Other Changes to the ESBWR Rule Language and Difference From Other DCRs E. Exclusions From Issue Finality and Issue Resolution for Hurricane-Generated Winds and Missiles F. Loss of One or More Phases of Offsite Power G. Spent Fuel Assembly Integrity in Spent Fuel Racks H. Turbine Building Offgas System Design Requirements I. ASME BPV Code Statement in Chapter 1 of the ESBWR DCD J. Clarification of ASME Component Design Inspections, Tests, Analyses, and Acceptance Criteria (ITAACs) K. Changes to the Supplemental FSER After Publication of the Supplemental Proposed Rule L. Corrections, Editorial, and Conforming Changes VI. Planned Withdrawal of the ESBWR Standard Design Approval (SDA) VII. Section-by-Section Analysis A. Introduction (Section I) B. Definitions (Section II) C. Scope and Contents (Section III) D. Additional Requirements and Restrictions (Section IV) E. Applicable Regulations (Section V) F. Issue Resolution (Section VI) G. Duration of This Appendix (Section VII) H. Processes for Changes and Departures (Section VIII) I. Inspections, Tests, Analyses, and Acceptance Criteria (Section IX) J. Records and Reporting (Section X) VIII. Agreement State Compatibility IX. Availability of Documents X. Voluntary Consensus Standards XI. Finding of No Significant Environmental Impact: Availability XII. Paperwork Reduction Act XIII. Regulatory Analysis XIV. Regulatory Flexibility Certification XV. Backfitting and Issue Finality XVI. Congressional Review Act XVII. Plain Writing XVIII. Availability of Guidance I. Background Part 52 of Title 10 of the Code of Federal Regulations (10 CFR), “Licenses, Certifications, and Approvals for Nuclear Power Plants,” subpart B, presents the process for obtaining standard design certifications. On August 24, 2005, GEH tendered its application for certification of the ESBWR standard plant design (ADAMS Accession No. ML052450245) with the NRC. The NRC published a notice of receipt of the application in the Federal Register (70 FR 56745; September 28, 2005). GEH submitted this application in accordance with subpart B of 10 CFR part 52. On December 1, 2005, the NRC formally accepted the application as a docketed application for design certification (Docket No. 52-010) (70 FR 73311; December 9, 2005). The pre-application information submitted before the NRC formally accepted the application can be found in ADAMS under Docket No. PROJ0717 (Project No. 717). The NRC staff issued a final safety evaluation report (FSER) for the ESBWR design in March 2011. The FSER is available in ADAMS under Accession No. ML103470210. The NRC subsequently published the FSER in April 2014 as NUREG-1966, “Final Safety Evaluation Report Related to the Certification of the Economic Simplified Boiling-Water Reactor Standard Design” (ADAMS Accession No. ML14100A304). The NRC also published a proposed rule to certify the ESBWR design in the Federal Register on March 24, 2011 (76 FR 16549), and a supplemental proposed rule on May 6, 2014 (79 FR 25715). The FSER and the proposed rule were based on the NRC's review of Revision 9 of the ESBWR DCD. On April 17, 2014, the NRC issued an advanced supplemental safety evaluation report (SER) (ADAMS Accession No. ML14043A134) to address several matters identified by the NRC and revisions to the ESBWR DCD in Revision 10. The advanced supplemental SER was referenced in the supplemental proposed rule (79 FR 25715; May 6, 2014). The supplemental FSER will be published as Supplement No. 1 to NUREG-1966 before this final rule becomes effective. Because Revision 10 of the DCD was issued after the ESBWR proposed rule was published, all of the substantive changes in Revision 10 of the DCD are addressed in the SUPPLEMENTARY INFORMATION section of this document, including a discussion of why the change was or was not addressed in a supplemental proposed rule. In its application for design certification, GEH also requested the NRC to provide an SDA for the ESBWR design. An SDA for the ESBWR design was issued in March 2011 (ADAMS Accession No. ML110540310) following the NRC staff's issuance of the ESBWR FSER. On June 3, 2014, GEH requested that the NRC retire the SDA at the time of issuance of the final ESBWR design certification rule (ADAMS Accession No. ML14154A094). After this final rule is published, the NRC intends, as a separate action from this rulemaking, to withdraw the SDA. The application for design certification of the ESBWR design has been referenced in the following COL applications as of the date of this document: (1) Detroit Edison Company, Fermi Unit 3, Docket No. 52-033 (73 FR 73350; December 2, 2008); (2) Dominion Virginia Power, North Anna Unit 3, Docket No. 52-017 (73 FR 6528; February 4, 2008); (3) Entergy Operations, Inc., Grand Gulf Unit 3, Docket No. 52-024 (73 FR 22180; April 24, 2008) (APPLICATION SUSPENDED); (4) Entergy Operations, Inc., River Bend Unit 3, Docket No. 52-036 (73 FR 75141; December 10, 2008) (APPLICATION SUSPENDED); and (5) Exelon Nuclear Texas Holdings, LLC, Victoria County Station Units 1 and 2, Docket Nos. 52-031 and 52-032 (73 FR 66059; November 6, 2008) (APPLICATION WITHDRAWN). II. Summary and Analysis of Public Comments on the ESBWR Proposed Rule and Supplemental Proposed Rule A. Overview of Public Comments The NRC published a proposed rule to certify the ESBWR design in the Federal Register on March 24, 2011 (76 FR 16549). The period for submitting comments on the proposed DCR, ESBWR DCD, or draft environmental assessment (EA) closed on June 7, 2011. The NRC received a total of 10 public comments on the proposed rule. The types of comments, the organization of comments, the comment identification format, and comment responses follow. The NRC also published a supplemental proposed rule to request public comments on two specific topics regarding the ESBWR design certification. The supplemental proposed rule was published in the Federal Register on May 6, 2014 (79 FR 25715). The period for submitting comments on these specific topics closed on June 5, 2014. The NRC received no public comments on the supplemental proposed rule. Types of Comments The NRC received two types of comment submissions on the proposed rule for the ESBWR design certification. A comment submission means a communication or document, submitted to the NRC by an individual or entity, with one or more individual comments addressing a subject or an issue. The two types of comment submissions were: 1. Comment submissions that were not identical or similar in content (unique comment submissions); and 2. Comment submissions self-characterized as “petitions” or comment submissions related to such “petitions” (petitions). The NRC received four unique comment submissions, including three comment submissions from private citizens and one comment submission from a non-government organization. Table 1 provides summary information on the unique comment submissions and their ADAMS Accession numbers. In addition, in light of the Fukushima Dai-ichi accident and during the public comment period on the proposed rule, the NRC received a series of petitions to suspend adjudicatory, licensing, and rulemaking activities, including the ESBWR design certification rulemaking. The NRC subsequently authorized responsive and supplemental filings on these petitions. In its Memorandum and Order, CLI-11-05, September 9, 2011, 74 NRC 141 (2011) (this decision is available on the NRC Web site in Volume 74 at http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0750/ ), the Commission addressed the petitions and the responsive and supplemental filings and determined that the petitions should be denied in the relevant adjudicatory proceedings; and, on its own motion referred the petitions to the NRC staff for consideration as comments in the ESBWR rulemaking. The staff considered the petitions and the responsive and supplemental filings and identified six comment submissions applicable to the ESBWR rulemaking. Table 2 provides summary information on these “petition-related” comment submissions and their ADAMS Accession numbers. Four of those comment submissions were “petitions” filed during the public comment period. One of the comment submissions was a responsive filing to the “petitions.” The sixth of these comment submissions, self-characterized as a “petition” and referred to the NRC staff in CLI-11-05, was received on August 15, 2011, after the close of the public comment period. As stated in the proposed rule, comments received after June 7, 2011, “will be considered if it is practical to do so, but assurance of consideration cannot be given” to comments received after this date. The NRC determined that it was practical to consider this comment. This comment opposed issuance of the final ESBWR rule. Table 1—Unique Comment Submissions Comment submission No. Commenter ADAMS Accession No. 1 Paul Daugherty ML110880057 2 Farouk Baxter ML110880315 3 Patricia T. Birnie, Chairman, General Electric Stockholders' Alliance ML11158A088 4 Anonymous ML11187A303 Table 2—Comment Submissions Self-Characterized as Petitions and Responsive Filings Comment submission No. Commenter ADAMS Accession No. 1 (Note 1) Various organizations and individuals ML111040472 2 (Note 1) Various organizations and individuals ML111080855 3 Various organizations and individuals ML111100618 4 Jerald G. Head, Senior VP, Regulatory Affairs, GE Hitachi Nuclear Energy ML11124A103 5 Various organizations and individuals ML111260637 6 ESBWR Intervenors ML112430118 Note 1: Petition comment submission 2 was submitted as an amendment to petition comment submission 1. Therefore, the NRC is only addressing comments on petition comment submission 2 in this final rule and no further response is needed on petition comment submission 1. Organization of Comments and Responses Comments and the NRC's responses are organized into two categories: Comments on technical issues presented in the DCD, and comments regarding Fukushima lessons learned. Comments on technical issues include the inclusion of beyond-design-basis accidents into the design, design of the ancillary diesel generators, safety-related battery design, control rod drive design, and control room flood protection. Comments regarding Fukushima lessons learned include delaying certification of the ESBWR design until lessons learned have been incorporated and the NRC's obligation under the National Environmental Policy Act (NEPA) to evaluate new information (such as the NTTF report, ADAMS Accession No. ML111861807) relevant to the environmental impact of its actions prior to certifying the ESBWR design. The NRC received comments related to the draft EA for this rule but those comments did not include anything to suggest that: (i) A rule certifying the ESBWR standard design would be a major Federal action, or (ii) the severe accident mitigation design alternatives (SAMDA) evaluation omitted a design alternative that should have been considered or incorrectly considered the costs and benefits of the alternatives it did consider. Therefore, no change to the EA was warranted. The NRC received no comments on the two specific topics in the supplemental proposed rule. The detailed comment summaries and the NRC's responses are provided in Sections II.B and II.C of this document. Comment Identification Format All comments are identified uniquely by using the format [W][X]-[Y], where: [W] represents the comment submission type (S = unique comment submission, P = petition). [X] represents the comment submission identification number (refer to the comment submission tables). [Y] represents the comment number, which the NRC assigned to the comment. In some instances, lower-case alphabetic characters [Ya, Yb, Yc * * *] were added to a comment number after the initial designation of comments. The NRC has created a document (ADAMS Accession No. ML113130141) which compiles all comment submissions and annotates each comment submission with the comment number indicated in the right hand margin. B. Comments Regarding Technical Content in the DCD Design-Basis Accidents Comment: Beyond-Design-Basis Accidents (DBAs) should be included in the design, final safety analysis report (FSAR), and Technical Specifications (TS). (S1-1) NRC Response: The NRC agrees that beyond-DBAs should be considered in the ESBWR design and the FSAR. In its 1985 policy statement on severe accidents (50 FR 32138), the Commission defined the term “severe accident” as an event that is “beyond the substantial coverage of design basis events,” (DBE) including events in which there is substantial damage to the reactor core (whether or not there are serious offsite consequences). Consistent with the objectives of standardization and early resolution of design issues, 10 CFR 52.47(a)(23) requires applicants for design certification to include a description and analysis of severe accident prevention and mitigation features in the new reactor designs. These features are discussed in Chapter 19 of the DCD (equivalent to an FSAR), and the staff's evaluation of them is found in Chapter 19 of the FSER. The NRC disagrees that beyond-DBAs should be included in the TS. The TS prescribe safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, and administrative controls associated with DBEs, but need not prescribe limits or settings for conditions that could be experienced during a beyond-DBE. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The NRC's current regulatory scheme requires significant re-evaluation and revision in order to expand or upgrade the design-basis for reactor safety as recommended by its NTTF report. (P6-1) NRC Response: The NRC considers this comment to be outside the scope of the ESBWR design certification rulemaking. The comment deals with the adequacy of the NRC's overall regulatory scheme for nuclear power reactors and does not directly address the adequacy of the ESBWR design certification. Nonetheless, the NRC disagrees with the comment. The NRC's rules and regulations provide reasonable assurance of adequate protection of public health and safety and the common defense and security. However, the Commission has “initiated a comprehensive examination of the implications of the Fukushima accident. . . . As a result [of that examination], the NRC may implement changes to its regulations and regulatory processes.” CLI-11-05, 74 NRC at 168. If such changes are warranted, the NRC's “regulatory processes provide sufficient time and avenues to ensure that design certifications and COLs satisfy any Commission-directed changes before any new power plant commences operations. . . . Whether [the Commission] adopt[s] the Task Force recommendations or require[s] more, or different, actions associated with certified designs or COL applications, [the Commission has] the authority to ensure that certified designs and combined licenses include appropriate Commission-directed changes before operation.” Id. at 162-163. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The ESBWR environmental documents do not address the radiological consequences of DBAs or demonstrate that those reactors can be operated without undue risk to the health and safety of the public and conclude that any health effects resulting from the DBAs are negligible. This conclusion is based on a review of the DBAs considered in the ESBWR DCD (WEC 2008) and NUREG-0800, Standard Review Plan (SRP). The findings of the Fukushima NTTF report call into question whether this represents a full, accurate description and examination of all DBAs having the potential for releases to the environment. See Makhijani Declaration at 7. If the design-basis for the reactors does not incorporate accidents that should be considered in order to satisfy the adequate protection standard, then it is not possible to reach a conclusion that the design of the reactor adequately protects against accident risks. See Makhijani Declaration at 9. (P6-3) NRC Response: The NRC disagrees with this comment. The NRC notes that the Makhijani Declaration citations do not address DBAs as discussed in the comment, but rather the declaration specifically refers to beyond-DBEs. The NRC interprets the comment to be referring to the environmental report required to be provided by the design certification applicant per 10 CFR 52.47, “Contents of applications; technical information,” and 10 CFR 51.55, “Environmental report—standard design certification.” The environmental report (NEDO-33306; ADAMS Accession No. ML102990433) referenced in Chapter 19 of the ESBWR DCD and evaluated in Chapter 19 of the FSER, as well as the NRC's EA, addresses costs and benefits of severe accident mitigation design alternatives. Conversely, DBAs for the ESBWR, and their associated radiological consequences, are not addressed in the environmental report but rather are addressed in Chapter 15 of the ESBWR DCD and evaluated in Chapter 15 of the FSER. The environmental report addresses the costs and benefits of severe accident mitigation design alternatives but does not address the design basis accidents discussed in the comment. In any event, the Commission has stated that, if warranted and after “a comprehensive examination of the implications of the Fukushima accident . . ., the NRC may implement changes to its regulations and regulatory processes.” CLI-11-05, 74 NRC at 168. The NRC's “regulatory processes provide sufficient time and avenues to ensure that design certifications and COLs satisfy any Commission-directed changes before any new power plant commences operations. . . .” Id. at 162-163. No change was made to the rule, the DCD, or the EA as a result of this comment. Electrical Systems Comment: The ESBWR design is flawed because it has failed to comply with the requirements of Institute of Electrical and Electronics Engineers (IEEE) Standard 603, which requires the electrical portion of the safety systems that perform safety functions—specifically, alternating current (ac) power from the Ancillary Diesel Generators (ADGs)—be classified as Class 1E. The DCD acknowledges that ac power from the ADGs is not needed for the first 72 hours of an accident, but are needed to perform Class 1E functions (recharging the Class 1E direct current (dc) batteries that provide power during the first 72 hours of an accident) when no other sources of power are available. The ESBWR design has classified these ac power sources as commercial grade, nonsafety-related, and non-Class 1E (S2-1, referencing ADAMS Accession No. ML102350160). NRC Response: The NRC disagrees with the comment. The NRC's position remains as stated in the separate correspondence between the commenter and the NRC that is attached to the comment letter. Specifically, the NRC stated that the events described in the commenter's previous letters (no ac power available to the plant for 72 hours after initiation of the accident and all batteries are depleted) are not DBEs but are beyond design-basis, for which the requirements of IEEE Standard 603 do not apply. As stated in the staff requirements memorandum (SRM), dated January 15, 1997, concerning SECY-96-128, “Policy and Key Technical Issues Pertaining to the Westinghouse AP600 Standardized Passive Reactor Design,” dated June 12, 1996, the Commission approved Item IV—Post-72 Hour Actions. The approval specified that the post-72 hour systems, structures, and components (SSCs) are not required to be safety-related. In addition, as stated in NUREG-1242, Volume 3, Part 1, “NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document: Passive Plant Designs, Chapter 1,” August 1994, a passive advanced light-water reactor, such as the ESBWR design, need not include or rely upon an active safety-related ac power source to support safety system functions after 72 hours from the onset of an accident, but may rely on electrical power sources that are not safety-related after that time. Specifically, the ESBWR is designed so that safety-related passive systems are able to perform all safety functions for 72 hours after initiation of a DBE without the need for operator actions. The DBE is assumed to be resolved (except for long-term cooling) within 72 hours, and thus, the Class 1E batteries are designed for and need only function for 72 hours without being recharged. In the ESBWR, the ADGs, which are the subject of the commenter's concern, are not used to recharge the Class 1E batteries. Rather, the ADGs provide power directly to post accident monitoring instrumentation, main control room lighting, the reactor pressure vessel (RPV) makeup pump, and containment cooling systems, among others. After 72 hours, consistent with NUREG-1242, nonsafety-related systems other than the ADGs are used to replenish safety-related passive systems so that they will perform long-term core cooling and containment integrity functions. These nonsafety-related systems are designed in accordance with quality standards commensurate with the importance of these functions and that provide reasonable assurance they will function when needed. In the event that the ADGs are not available, the Seismic Category I firewater storage tanks and Seismic Category I diesel pump and fire protection piping can be used to provide post-accident makeup water to the Isolation Condenser and Passive Containment Cooling System (PCCS) pools and Spent Fuel Pool (SFP) using the Fuel and Auxiliary Plant Cooling System (FAPCS) for long-term cooling beyond 72 hours. The NRC also stated in its May 15, 2009, letter (in the referenced document) that the offsite power system, a nonsafety-related power source, is the preferred source of power for safety-related systems at all current plants. Further, the station blackout (SBO) rule, 10 CFR 50.63, “Loss of all alternating current power,” does not require the use of safety-related alternative ac power sources to cope with an SBO. Therefore, neither of these ac power sources—offsite power or alternate ac power source—is required to be safety-related or classified as Class 1E under IEEE 603. Thus, the ADGs need not be classified as Class 1E power sources as well. In summary, the design bases of the passive safety systems are centered on the 72-hour capability and these safety-related systems must remain functional to assure the integrity of the reactor coolant pressure boundary and the capability to shut down the reactor and maintain it in a safe shutdown condition without operator action or support from nonsafety systems for the first 72 hours following the initiation of a DBE. Beyond 72 hours, these systems must continue to remain functional to provide such assurance for the following 4 days, with allowance for operator actions and support from nonsafety SSCs consistent with NUREG-1242. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The NRC should require GEH to relocate the safety-related dc batteries and their related systems above grade level so that they are not subject to external flooding. This recommendation is supported by the following points: 1. There is a fair chance of a failure of the dc supply as safety-related battery banks (Class-1E grade batteries) are housed below grade in the reactor building, as well as their electrical penetration to primary containment. In a natural disaster they may not remain watertight, as water may enter through the doors and incapacitate the battery banks. 2. Water may also enter the battery rooms if those doors are open for maintenance, testing, or replacement of cells. 3. ESBWR emergency core cooling systems (ECCS) are dependent on this dc supply. If the dc supply is lost, emergency cooling and depressurization systems will fail. There is no diversity for the core cooling and depressurization systems if the dc supply fails. (S4-1) NRC Response: The NRC disagrees with the comment. The safety-related dc batteries and their related systems do not need to be relocated above grade level. The NRC has reviewed the ESBWR DCD and has determined that the ESBWR safety-related SSCs (including the reactor building, which houses the dc batteries) are designed to withstand the effects of external flooding. With the exception of loads due to hurricane winds and wind-generated missiles beyond those considered in the ESBWR DCD, the NRC concluded that the ESBWR DCD meets the requirements of 10 CFR part 50, appendix A, “General Design Criteria for Nuclear Power Plants,” (GDC) 2, which requires the design bases of SSCs important to safety to include protection against natural phenomena (including earthquakes, tornadoes, floods, hurricanes, and tsunami) such that these SSCs will not lose the capability to perform their safety functions as a result of such phenomena. This conclusion is documented in the NRC's FSER for the ESBWR design. In the following paragraphs, the NRC addresses each of the three supporting points for the comment. Supporting Point 1: The NRC agrees that safety-related batteries are located below grade per the ESBWR DCD, Tier 2, Figure 1.2-2. This is acceptable because all components of safety-related dc electric systems are housed in structures which provide protection against external flood damage. The structures that may be subjected to a design-basis flood are designed to withstand the flood level by locating the plant grade elevation 1 ft. (0.30 m) above the flood level and incorporating structural provisions into the plant design to protect the SSCs from the postulated flood conditions. GEH's application for design certification was submitted with proposed vendor-specified site parameters. These values are provided in Table 2.0-1 (Tier 2) and in Table 5.1-1 (Tier 1) of the DCD. For the ESBWR design, the maximum groundwater level is 2 ft. (0.61 m) below plant grade and the maximum flood level is 1 ft. (0.30 m) below plant grade. The ESBWR design was evaluated using the vendor-specified flood levels and found to be safe. All exterior access openings are above flood level. The flood design incorporates reinforced concrete walls designed to resist the static and dynamic forces of the design-basis flood and water stops at construction joints to prevent in-leakage. External surfaces below flood and ground water levels are waterproofed. Penetrations are sealed and also capable of withstanding the static and dynamic forces of the design-basis flood. Watertight doors provide physical separation of flood zones. In addition, the applicant has specified the site parameters, design characteristics, and any additional requirements and restrictions necessary for a COL applicant to ensure that safety-related SSCs will be adequately protected from the site-specific probable maximum flood conditions. Based on the evaluation in Section 3.4 of the FSER, the NRC concludes that the ESBWR design regarding flood protection provides reasonable assurance that safety-related SSCs (including the safety-related dc batteries and their related systems) will maintain their structural integrity or are located within structures that will maintain their integrity, and will perform their intended safety functions when subjected to a design-basis flood, and therefore, satisfy the requirements of GDC 2. Supporting Point 2: The comment stated that water may enter the battery rooms if the watertight doors are open for maintenance, testing, or replacement of the battery cells. The NRC agrees that this scenario is possible for one division of safety-related battery banks. The ESBWR TS, under limiting condition of operation 3.8.1, restricts maintenance, testing, or replacement of the battery cells during plant operation to only one required division of safety-related battery banks. In addition, the COL applicant is required to develop plant operating and maintenance procedures that provide control for activities that are important to the safe operation of the facility, including limiting conditions of operation. However, there are four divisions of safety-related battery banks, which are physically separated by concrete walls and watertight doors. Only two divisions of dc systems are required for safe shutdown of the plant. If one of the safety-related battery room doors is open during a flood, as suggested in the comment, the other batteries will still be adequately protected by design features for physical separation to ensure the safety-related SSCs can perform their functions. Supporting Point 3: The comment stated that the ESBWR ECCS is dependent on dc power, and if dc power is lost, emergency cooling and depressurization systems will fail. The ESBWR ECCS consists of the Gravity Driven Cooling System, the Isolation Condenser System, the Standby Liquid Control System, and the Automatic Depressurization System. The Gravity Driven Cooling System, Standby Liquid Control System, and the Automatic Depressurization System do rely on dc power for actuation (as pointed out in the comment). The four trains of Isolation Condenser System, on the other hand, automatically begin removal of decay heat and control RPV level above the top of active fuel upon loss of all ac and dc power because the only valve in the system relied upon to change position upon initiation of the system fails in the safe (open) position upon loss of power. Beginning 4 hours after the start of an accident, the Isolation Condenser System upper and lower header vent valves are opened periodically to remove non-condensable gases to maintain optimum heat removal and allow continued reactor cooldown. These valves are solenoid-operated valves and rely upon electric power to open. The comment also suggests that there is no diversity for several systems that rely on the dc power supply. The NRC agrees that the Automatic Depressurization System, Gravity Driven Cooling System, the Suppression Pool Equalization Line Valves, and the Standby Liquid Control System all require safety-related dc power in order to perform their safety functions and therefore lack diversity in that regard, but does not agree that the Basemat Internal Melt Arrest Coolability (BiMAC) cooling system requires safety-related dc power to perform its safety function. As discussed below, the BiMAC cooling system—a non-safety system—is designed to automatically fire squib valves and drain water to the area below the RPV upon sensing high temperatures in the BiMAC without dependence on any of the four safety-related power sources. Also, as discussed above, the four trains of the Isolation Condenser System automatically begin removal of decay heat and control RPV level above the top of active fuel upon loss of all ac and dc power because the only valve in the system relied upon to change position upon initiation of the system fails in the safe (open) position upon loss of power. Decay heat can be removed with the Isolation Condenser System for 72 hours without any additional action. The ESBWR is designed such that the Isolation Condenser System heat exchanger pool can be replenished after 72 hours with the diesel driven fire pump to allow continued cooling with the Isolation Condenser System. Safety-related dc power is not needed to operate this pump. In light of these facts, the NRC concludes that the capability of the ESBWR to remove decay heat from the reactor core following an accident is sufficiently diverse. It should also be noted that the ESBWR safety-related 120 volts ac uninterruptible power supply (UPS) input is normally supplied by offsite power or a nonsafety-related onsite power system. During a loss of offsite and nonsafety-related onsite power, the UPS gets its power from 250 volts dc batteries. The ESBWR design includes an offsite power system, nonsafety-related standby diesel generators, and ADGs, any of which can mitigate the consequences of an accident if available. Safety-related UPS systems are housed in seismic Category I structures and meet GDCs 2, 4, and 17. Common cause failure of the safety-related batteries in the ESBWR design would clearly be an event of substantial safety significance because dc power is used to power the distributed control and instrumentation system, which is used to actuate passive safety systems. However, the ESBWR design includes a number of defense-in-depth features for reducing the likelihood of losing all ability to accomplish key safety functions. As previously stated, the Isolation Condenser System automatically begins removal of decay heat and controls RPV level above the top of active fuel upon loss of all ac and dc power. All safety divisions (including concrete walls and watertight doors that separate the four safety-related battery banks) are physically separated. The ESBWR design also includes design features specifically for the purpose of injecting water into the containment to flood the containment floor and cover core debris. The BiMAC cooling system is designed to automatically fire squib valves and drain water to the area below the RPV upon sensing high temperatures in the BiMAC, indicating core debris below the RPV. This occurs without operator action and without dependence on any of the four safety-related power sources. No change was made to the rule, the DCD, or the EA as a result of this comment. Control Rod Drive System Comment: Two Control Rod Drives (CRD) are scrammed by one hydraulic control unit (HCU). A single failure of one HCU will affect the scram function of two CRDs. It is done for cost saving. This is not acceptable in a safety system. (S4-2) NRC Response: The NRC disagrees with the comment. In Section 4.6.3 of the FSER, the NRC stated that a single failure in an HCU may result in the failure of two control rods. The DCD describes that the control rods are assigned to HCUs in a manner such that no 4X4 array of rods contain both rods connected to the same HCU. This arrangement assures that shutdown is achieved (among other things) assuming a single failure of an HCU. The NRC reviewed the effects of an HCU failure and concluded in Section 4.3 of the FSER that sufficient shutdown margin exists in the case of an HCU failure. In addition, TS 3.1.5 requires that all control rod scram accumulators are operable during Modes 1 (Power Operation) and 2 (Start-Up). If an accumulator is inoperable, the associated control rod pair is declared inoperable and Limiting Condition of Operation (LCO) 3.1.3, Control Rod Operability, is entered. This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying the intended function in accordance with actions of LCO 3.1.3. If an accumulator is inoperable, TS require the affected control rod to be inserted and hence the scram function of two CRDs is satisfied. Finally, the ESBWR has a diverse method to scram the reactor. An electric motor is provided for each CRD for scram in addition to the hydraulic scram using the accumulator. Accordingly, the NRC has determined that the CRD system design is adequate. No change was made to the rule, the DCD, or the EA as a result of this comment. Control Room Comment: For safety reasons, the Control Room should be located at a sufficient height from the ground to prevent its flooding during a tsunami, tornado, hurricane, heavy rain, etc. (S4-3) NRC Response: The NRC agrees that the control room should be protected from flooding. GEH's application for SDA and design certification was submitted with proposed vendor-specified site parameters. The values for maximum groundwater is 2 feet (0.61 m) below plant grade as provided in Table 2.0-1 (Tier 2) of the DCD and the maximum flood level is 1 foot (0.30 m) below plant grade as provided in Table 5.1-1 (Tier 1) of the DCD. The ESBWR design was evaluated using the vendor-specified flood levels and found to be safe. As described in Chapter 3 of the DCD, the ESBWR construction incorporates several water proofing features: The external walls below groundwater and flood levels are designed to withstand hydrostatic loads, construction and expansion joints have water stops, external surfaces below groundwater and flood levels are waterproofed, penetrations below groundwater and flood levels are sealed, and there are no exterior openings below grade. If a COL application referencing the ESBWR design is submitted to the NRC, the COL applicant must demonstrate that the site-specific characteristics are bounded by the DCD site parameters. During the review of a COL application using this design, the staff will perform an independent analysis to verify that the flood levels and other relevant site characteristics are within the DCD parameters. No change was made to the rule, the DCD, or the EA as a result of this comment. Spent Fuel Pool Comment: The ESBWR design has an elevated SFP. This is a particularly troublesome feature in common with the Mark I BWR design, which is the design of the Fukushima reactors. (P2-2) NRC Response: The NRC disagrees with this comment. The ESBWR SFP design is different from the Mark I BWR design in that the ESBWR SFP is located entirely below grade. The ESBWR design does include an additional buffer pool located above grade in the reactor building. The buffer pool contains a small array of spent fuel racks that is used for temporary storage of spent fuel during refueling operations and also includes a location to store new fuel assemblies during power operations. GDC 2 requires that the ESBWR spent fuel storage facilities (SFP and buffer pool) and the structure within which they are housed, as SSCs important to safety, be protected against the effects of natural phenomena without loss of their safety function. In addition, GDC 61 requires that the design prevents drainage of coolant inventory below an adequate shielding depth, provides adequate coolant flow to the spent fuel racks, and provides a system for detecting and containing pool liner leakage. The reactor building and the concrete containment, which houses the SFP and additional buffer pool, are seismic Category I structures that are designed to meet the requirements of GDC 2 for protection against natural phenomena such as an earthquake, tornado, or hurricane in combination with normal and accident condition loads considering the effects due to the elevated location of the buffer pool. Information relating to the analysis and design of the reactor building is provided in DCD Sections 3.7 and 3.8 and Appendices 3A, 3B, 3F, and 3G. Through analysis and review of the design, the NRC determined that the reactor building and the concrete containment are structurally adequate to withstand all design-basis loads. The NRC concluded in the FSER that both pools are adequately protected from the effects of natural phenomena without loss of capability to perform their safety functions. The NRC also concluded in its FSER that, because the SFP and buffer pools have anti-siphoning devices on all submerged Fuel and Auxiliary Pools Cooling System (FAPCS) piping, and there are no other drainage paths by which the level in the SFP or buffer pool could be reduced, coolant will not drain below an adequate shielding depth in either pool. Cooling of spent fuel located in either the SFP or buffer pool is provided by the FAPCS. In the unlikely event that a loss of active cooling to the spent fuel assemblies occurs, there is enough water to keep the fuel assemblies cooled for a minimum of 72 hours before operator actions are needed. After 72 hours, additional water can be provided through safety-related connections to the fire protection system or another onsite or offsite water source. The NRC concluded in the FSER that cooling for both ESBWR SFP and buffer pools will be maintained. Finally, the NRC concluded in the FSER that, because the spent fuel pool and buffer pool are equipped with stainless steel liners, concrete walls, and leak detection drains, both detection and containment of pool liner leakage capability are provided. No change was made to the rule, the DCD, or the EA as a result of this comment. C. Comments Regarding the NRC's Response to Fukushima Dai-ichi Accident Some commenters favored delaying (in some fashion) the ESBWR rulemaking until lessons are learned from the Fukushima Dai-ichi Nuclear Power Plant (Fukushima) accident that occurred on March 11, 2011, and the NRC applies the lessons learned to United States (U.S.) nuclear power plants, including the ESBWR design. Background on how the Commission responded to the Fukushima accident and how the ESBWR design addresses Fukushima NTTF recommendations is discussed in Section III of the SUPPLEMENTARY INFORMATION section of this document. As discussed in Section III of the SUPPLEMENTARY INFORMATION section of this document, the NRC concludes that no changes to the ESBWR design are warranted at this time to provide reasonable assurance of adequate protection of public health and safety. Moreover, even if the Commission concludes at a later time that some additional action is needed for the ESBWR design, the NRC has ample opportunity and legal authority to modify the ESBWR DCR to implement design changes, as well as to take any necessary action to ensure that COLs that reference the ESBWR make any necessary design changes. Comment: The NRC should suspend the certification of the ESBWR reactor design and rescind the final design approval it granted on March 9, 2011. Based on the recent events at the Fukushima Dai-ichi site, the NRC should first undertake a far more rigorous, long-term review of the design and the regulatory implication of the events, implement new regulations to protect public health and safety, and revise the environmental analyses to evaluate the potential health, environmental and economic costs of reactor and SFP accidents. (S3-1, P3-1, P3-2) NRC Response: The NRC declines to suspend the ESBWR rulemaking. See Memorandum and Order, CLI-11-05, 74 NRC 141 (2011) (ADAMS Accession No. ML112521106). Background on how the Commission responded to the Fukushima accident and how the ESBWR design addresses Fukushima NTTF recommendations is discussed in Section III of the SUPPLEMENTARY INFORMATION section of this document. In that section, the NRC concludes that no changes to the ESBWR design are required at this time to provide reasonable assurance of adequate protection of public health and safety. If the Commission concludes at a later time that some additional action is needed for the ESBWR design, the NRC has ample opportunity and legal authority to modify the ESBWR DCR to implement design changes, as well as to take any necessary action to ensure that COLs that reference the ESBWR also make any necessary design changes. For these reasons the NRC does not regard delays in the ESBWR design certification process to be appropriate. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The Atomic Energy Act (AEA) and NEPA preclude the NRC from approving standardized plant designs until it has completed the investigation of the Fukushima accident and considered the safety and environmental implications of the accident with respect to its regulatory program. NEPA imposes on agencies a continuing obligation to gather and evaluate new information relevant to the environmental impact of its actions. The need to supplement under NEPA when there is new and significant information is also found throughout the NRC regulations, e.g., 10 CFR 51.92(a)(2), 51.50(c)(iii), 51.53(b), and 51.53(c)(3)(iv). The conclusions and recommendations presented in the NTTF report constitute “new and significant information” whose environmental implications must be considered before the NRC may certify the ESBWR design and operating procedures. (P2-2, P6-2) NRC Response: The NRC disagrees with this comment. The comment did not explain what particular provision of the AEA precludes the NRC from issuing a standard DCR. Furthermore, NEPA has no “continuing obligation” to gather and evaluate new information relevant to the environmental impact of its actions, because the Commission has determined that issuance of a standard DCR is not a major Federal action significantly affecting the quality of the human environment. See the EA at page 1 (ADAMS Accession No. ML111730382). No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The whole nuclear culture must be reviewed before any reactor designs are certified for potential construction, and that all licensing of new reactor designs be put on hold until the NRC's systems of regulations, oversight, and enforcement are thoroughly reviewed and, where required, are made more restrictive. (S3-2) NRC Response: The NRC considers this comment to be outside the scope of the ESBWR design certification rulemaking. The comment addresses overall nuclear industry safety culture and does not directly address the adequacy of the ESBWR design certification. Nonetheless, the NRC disagrees with the comment. The NRC considers that its regulatory framework and requirements provide a rigorous and comprehensive design certification and license review process that examines the full extent of siting, system design, and operations of nuclear power plants. The NRC will continue to process existing applications for new design certifications and licenses in accordance with the schedules that have been established. Background on how the Commission responded to the Fukushima accident and how the ESBWR design addresses Fukushima near-term task force recommendations is discussed in Section III of the SUPPLEMENTARY INFORMATION section of this document. In that section, the NRC concludes that no changes to the ESBWR design are warranted at this time to provide reasonable assurance of adequate protection of public health and safety. Moreover, even if the Commission concludes at a later time that some additional action is needed for the ESBWR design, the NRC has ample opportunity and legal authority to modify the ESBWR DCR to implement design changes, as well as to take any necessary action to ensure that COLs that reference the ESBWR also make any necessary design changes. For these reasons the NRC does not regard delays in the ESBWR design certification process to be appropriate. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The NRC should include a review of public health challenges worldwide from radiation in its decision-making process. (S3-3) NRC Response: The NRC considers this comment to be outside the scope of the ESBWR DCR. The comment addresses the NRC's generic process and criteria for regulatory decision making, and does not directly address the adequacy of the ESBWR design. Nonetheless, the NRC disagrees with the comment. The NRC interprets the comment's reference to the “decision-making process” to mean the Commission's decision whether to certify the ESBWR design. The NRC reviewed the design and has found that it complies with the NRC's regulations, which provide reasonable assurance of adequate protection of public health and safety, including protection of the public from radiation. The comment did not provide any data, analyses, or other technical information to suggest why the EBSWR design would be unable to provide adequate protection of the public from radiation. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The NTTF recommended that licensees reevaluate the seismic and flooding hazards at their sites and if necessary update the design-basis and SSCs important to safety to protect against the updated hazards. NTTF Report, page 30. The ESBWR environmental documents must be supplemented in light of this new and significant information. The NTTF's findings and recommendations are directly relevant to environmental concerns and have a bearing on the proposed action and its impacts. They demonstrate a need to reevaluate the seismic and flooding hazards on the ESBWR reactors, the environmental consequences such hazards could pose, and what, if any, design measures could be implemented (i.e., through NEPA's requisite “alternatives” analysis) to ensure that the public is adequately protected from these risks. (P6-4) NRC Response: The NRC disagrees with the comment. Recommendation 2 of the NTTF, which is the subject of the comment, was focused on licensees of nuclear power reactors and was addressed through site-specific evaluations of the adequacy of the design of the reactors as applied to the site-specific seismic and flooding characteristics. By contrast, the ESBWR design certification—as any other design certification—is not approved for use on any specific site. Rather, the ESBWR design specifies “design parameters,” including maximum flood levels and seismic ground motion frequencies and magnitudes, representing the values for which the NRC has determined the ESBWR may safely be placed. A nuclear power plant applicant intending to use the ESBWR must show that the actual site characteristics for the site that the applicant intends to use for the ESBWR fall within the ESBWR-specified design parameters. Thus, NTTF Recommendation 2 is not relevant to the adequacy of the ESBWR design certification. Rather, the NRC regards this NTTF recommendation as an issue relevant to the determination whether a referenced design certification has been adequately demonstrated to be appropriate at the COL applicant's designated site. In addition, the NRC does not agree that NTTF Recommendation 2 demonstrates that the NRC must “reevaluate the seismic and flooding hazards on the ESBWR reactors, the environmental consequences such hazards could pose, and what, if any, design measures could be implemented” through a NEPA “alternatives” analysis. Recommendation 2 of the NTTF can best be thought of as a determination to ensure that each site's seismic and flooding characteristics are adequately justified based upon current information. The recommendation does not concern the adequacy of the NRC's substantive regulatory requirements governing protection against seismic and flooding events or their application to any specific reactor design (such as the ESBWR). Thus, even if Recommendation 2 were adopted in full by the Commission and fully implemented, those implementing actions would be directed at licensees of existing nuclear power plants and applicants for new nuclear power plants. The NRC's implementing actions would not be directed at the ESBWR design certification. For these reasons, the NRC does not agree with the comment that ESBWR's EA must be supplemented to address the NTTF Recommendation 2 and implementing actions. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The NTTF report makes several significant findings when it comes to increasing and improving mitigation measures for new reactor designs and recommends a number of specific steps licensees could take in this regard. Accordingly, the ESBWR environmental report must be supplemented to consider the use of these additional mitigation measures to reduce the project's environmental impacts. See 40 CFR 1502.14(f), 1502.16, 1508.25(b)(3). (P6-5) NRC Response: The NRC disagrees with the comment. The NTTF report explicitly states that by the “nature of their passive designs and inherent 72-hour coping capability for core, containment, and SFP cooling with no operator action required, the ESBWR and AP1000 designs have many of the design features and attributes necessary to address the Task Force recommendations. The Task Force supports completing those design certification rulemaking activities without delay.” (see NTTF Report, pages 71-72). Specifically, the NTTF report does not recommend any actions for the ESBWR design in the near term. NEPA's obligation to evaluate new information relevant to the environmental impact does not attach unless and until the Commission determines whether “new and significant” information has arisen and there is a “major Federal action” being undertaken by the NRC for which the new information is relevant and material. The Commission has stated that “[a]lthough the Task Force completed its review and provided its recommendations to us, the agency continues to evaluate the accident and its implications for U.S. facilities and the full picture of what happened at Fukushima is still far from clear. In short, we do not know today the full implications of the Japan event for U.S. facilities. Therefore, any generic NEPA duty—if one were appropriate at all—does not accrue now. If, however, new and significant information comes to light that requires consideration as part of the ongoing preparation of application-specific NEPA documents, the agency will assess the significance of that information as appropriate.” CLI-11-05, 74 NRC at 167. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: Before certifying the ESBWR, the NRC must evaluate the relative costs and benefits of adopting all of the NTTF report recommendations, and specifically Recommendations 4 and 7, in light of the NRC's increased understanding regarding accident risks and the strength of its regulatory program to prevent or mitigate them . (P6-6) NRC Response: The NRC disagrees with the comment. The NTTF report explicitly states that by “nature of their passive designs and inherent 72-hour coping capability for core, containment, and SFP cooling with no operator action required, the ESBWR and AP1000 designs have many of the design features and attributes necessary to address the Task Force recommendations. The Task Force supports completing those design certification rulemaking activities without delay.” Id., at 71-72. Specifically, the NTTF report does not recommend any actions, to include Recommendations 4 and 7, for the ESBWR design in the near term. Any potential need to address these recommendations, by addressing “prestaging of any needed equipment for beyond 72 hours,” and the establishment of inspection, test, analysis, and acceptance criteria (ITAACs) “to confirm effective implementation of minimum and extended coping, as described in detailed Recommendation 4.1” of the NTTF report would be placed on COL applicants referencing the ESBWR design. Id., at 72. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The comment questions the summary conclusions in Section 7 of the NTTF report regarding Recommendations 4 and 7. Both of these recommendations are contrary to the certification process as currently followed by the NRC in which an applicant for a COL can incorporate by reference a certified reactor design. Directly contrary to this long-standing process, the process suggested in the NTTF report pushes the Fukushima lessons learned onto a COL applicant rather than resolved these issues during the design certification process. Each reactor then becomes a prototype as case-by-case review of potential design and operational changes are made after construction begins. If the phrase “completing those design certification rulemaking activities without delay” is an endorsement of the current rulemaking on the ESBWR DCD Revision 9 without consideration of the other Fukushima-driven recommendations (or the subsequent revision to the DCD), the comment questions the depth into which the NTTF analyzed the ESBWR reactor design . (P6-7) NRC Response: The NRC considers this comment to be outside the scope of the ESBWR design certification rulemaking. The comment presents the commenter's views on Recommendations 4 and 7 of the NTTF Report, but does not address the adequacy of the ESBWR design, the rule, or the EA. Nonetheless, the NRC disagrees with the comment. The NTTF suggestions that COL applicants or holders address Recommendations 4 and 7, rather than the design certification applicant during the certification process, would not necessitate those COLs to be considered “prototypes.” The Commission has stated that “the agency continues to evaluate the accident and its implications for U.S, facilities and the full picture of what happened at Fukushima is still far from clear. In short, we do not know today the full implications of the Japan event for U.S. facilities.” CLI-11-05, 74 NRC at 167. Should changes need to be made to the ESBWR design as a result of the evaluation of the Fukushima event, the Commission has stated that “we have the authority to ensure that certified designs and combined licenses include appropriate Commission-directed changes before operation.” Id. at 163. Further, it is not contrary to the certification process to require changes resulting from Fukushima lessons learned on COLs. The NRC may, under 10 CFR 52.97(c), place conditions upon the COL that the “Commission deems necessary and appropriate.” Further, the requirements under 10 CFR 52.63(a)(1) provide a mechanism for the NRC to modify certified designs. Such design changes would be applied to all COL holders referencing this design under 10 CFR 52.63(a)(3). As a result, all COL holders referencing the certified design would be required to make such changes. Moreover, in appropriate (but relatively limited) circumstances the NRC could also impose changes as an “administrative exemption” to the issue finality provisions of 10 CFR 52.63 and the ESBWR analogous to what the NRC did in the aircraft impact assessment (AIA) final rule, 10 CFR 50.150 (72 FR 56287; October 3, 2007). No change was made to the rule, the DCD, or the EA as a result of this comment. Emergency Petition NRC Note: The Emergency Petition is comment submissions P1 and P2 in this ESBWR design certification rulemaking proceeding. Comment: The emergency petition is out of process and should be dismissed on that basis alone. However, if this petition is not so dismissed, the NRC should treat this petition, for aspects related to the single issue specifically regarding the ESBWR design certification rulemaking, as a public comment on the proposed rule. (P4-1) NRC Response: The NRC need not address, in this rulemaking, the comment's suggestion that the emergency petition is out of process because the Commission considered the merits of it and related filings in its Memorandum and Order, CLI-11-05, 74 NRC at 141 (2011) (ADAMS Accession No. ML112521106). The Commission determined that the Emergency Petition should be denied in the relevant adjudicatory proceedings and, on its own motion referred the emergency petition to the NRC staff for consideration as comments in the ESBWR rulemaking. To the extent that it is relevant to the ESBWR design certification rulemaking, the NRC agrees that the Emergency Petition should be treated as a public comment on the proposed rule. Comments in the Emergency Petition are addressed in this comment response portion of this statement of considerations for the final ESBWR DCR. No change was made to the rule, the DCD, or the EA as a result of this comment. Comment: The responses, filed by various industry representatives and COL applicants in accordance with an April 19, 2011, Commission Order (ADAMS Accession No. ML111101277) and setting forth those representatives' and applicants' views on an “Emergency Petition” (ADAMS Accession No. ML111080855), were based on mischaracterizations of the Emergency Petition, incorrect representations regarding the NRC's response to the Three Mile Island accident, and incorrect interpretations of the law. Therefore, the responses should be rejected and the Emergency Petition should be granted. (P5-1) NRC Response: On September 9, 2011, the Commission issued a Memorandum and Order on the Emergency Petition, CLI-11-05, 74 NRC 141 (ADAMS Accession No. ML112521106), which referred both the Emergency Petition and certain documents filed with the NRC to the NRC staff for “consideration as comments” in the applicable design certification rulemaking. CLI-11-05, 74 NRC at 176. Comment submission P5 was one of the documents referred by the Commission to the staff for consideration as comments. In accordance with the Commission's direction in CLI-11-05, comment submission P5 has been considered in the ESBWR rulemaking in a manner consistent with other comment submissions filed in the ESBWR rulemaking. Thus, the NRC reviewed the submission to determine the nature of the comments within this comment submission, if it is within the scope of the ESBWR rulemaking, and if so, what substantive response is appropriate. Based upon that review, the NRC determined that comment submission P5 is essentially a procedural reply to responses filed by other entities on the Emergency Petition. The NRC has determined that the reply does not contain any new substantive comments on the adequacy of the ESBWR design that were not already presented in the Emergency Petition and, therefore, has concluded that no further response is needed. No change was made to the rule, the DCD, or the EA as a result of this comment. III. Regulatory and Policy Issues This document addresses the regulatory and policy issues that were addressed in the March 2011 proposed rule, the May 2014 supplemental proposed rule, and those not addressed in either the proposed rule or the supplemental proposed rule. The regulatory and policy issues addressed in the March 2011 proposed rule are: (1) Access to safeguards information (SGI) and sensitive unclassified non-safeguards information (SUNSI), and (2) human factors engineering (HFE) operational program elements exclusion from finality. An additional regulatory and policy issue addressed in the May 2014 supplemental proposed rule is incorporation by reference of public documents and issue resolution associated with non-public documents. The NRC provided an opportunity for public comment in the supplemental proposed rule on the issue resolution associated with non-public documents, but not for incorporation by reference of public documents. A number of regulatory and policy issues were not included in either the March 2011 proposed rule or the May 2014 supplemental proposed rule. These are: (1) How the ESBWR design addresses Fukushima NTTF recommendations, (2) changes to Tier 2* information, (3) change control for severe accident design features, and (4) other changes to the ESBWR rule language and difference between the ESBWR rule and other DCRs. Each of these issues identified above is discussed below. 1 1 Some of the regulatory and policy issues discussed below arose after the close of the public comment period on the March 24, 2011, proposed rule. The public was afforded an opportunity to comment on some of these issues in the May 16, 2014, supplemental proposed rule. Section V of the SUPPLEMENTARY INFORMATION section of this document describes the NRC's bases for not offering a comment opportunity for some of the regulatory and policy issues that arose after the close of the public comment period on the proposed rule. A. How the ESBWR Design Addresses Fukushima NTTF Recommendations The application for certification of the ESBWR design was prepared and submitted, and the NRC staff's review of the application was completed, before the March 11, 2011, Great Tohoku earthquake and tsunami and subsequent events at the Fukushima Dai-ichi Nuclear Power Plant in Japan. In response to the events at Fukushima, the NRC established the NTTF to conduct a systematic and methodical review of NRC processes and regulations to: (1) Determine whether the agency should make additional improvements to its regulatory system; and (2) make recommendations to the Commission for policy directions. On July 12, 2011, the NTTF issued a 90-day report, SECY-11-0093 (ADAMS Accession Number ML11186A950), “Near Term Report and Recommendations for Agency Actions Following the Events in Japan,” identifying 12 recommendations. Among other recommendations, the NTTF supported completing the ESBWR design certification rulemaking activity without delay (see NTTF Report, pages 71-72). On September 9, 2011, in SECY-11-0124, “Recommended Actions to Be Taken Without Delay from NTTF Report,” (ADAMS Accession No. ML11245A144) the NRC staff submitted to the Commission for its consideration NTTF recommendations that should be partially or entirely initiated without delay. In SECY-11-0124, the NRC staff concluded that the following subset of actions would provide the greatest potential for improving safety in the near term: (1) Recommendation 2.1: Seismic and Flood Hazard Reevaluations (2) Recommendation 2.3: Seismic and Flood Walkdowns (3) Recommendation 4.1: Station Blackout Regulatory Actions (4) Recommendation 4.2: Equipment Covered under 10 CFR 50.54(hh)(2) (subsequently renamed “Mitigation Strategies for Beyond-Design-Basis External Events” with the issuance of Order EA-12-049) (5) Recommendation 5.1: Reliable Hardened Vents for Mark I Containments (6) Recommendation 8: Strengthening and Integration of Emergency Operating Procedures, Severe Accidents Management Guidelines, and Extensive Damage Mitigation Guidelines (7) Recommendation 9.3: Emergency Preparedness Regulatory Actions (staffing and communications). On October 3, 2011, in SECY-11-0137, “Prioritization of Recommended Actions To Be Taken in Response to Fukushima Lessons Learned” (ADAMS Accession No. ML11272A203), the NRC staff identified two additional actions that would have the greatest potential for improving safety in the near term. The additional actions are: (1) Inclusion of Mark II containments in the staff's recommendation for reliable hardened vents associated with NTTF Recommendation 5.1 and (2) the implementation of SFP instrumentation proposed in Recommendation 7.1. The NRC staff determined that the following two near term recommendations are applicable and should be considered for the ESBWR design certification: (1) Recommendation 4.2, Mitigation Strategies for Beyond-Design-Basis External Events (onsite equipment and connections only) and (2) Recommendation 7.1, SFP Instrumentation. The remaining Commission-approved near term recommendations are applicable only to COLs and existing plants (Recommendations 2.1 and 9.3), only to existing plants (Recommendations 2.3 and 5.1), or are planned to be addressed through rulemaking (Recommendations 4.1, 4.2, 7.1, 8, and 9.3). On February 17, 2012, in SECY-12-0025, “Proposed Orders and Requests for Information in Response to Lessons Learned from Japan's March 11, 2011, Great Tohoku Earthquake and Tsunami,” (ADAMS Accession No. ML12039A103) the NRC staff provided the Commission with proposed orders and requests for information to be issued to all power reactor licensees and holders of construction permits. In SECY-12-0025, the staff indicated its intent to address similar requirements in its reviews of pending and future design certification and COL applications. On March 9, 2012, in the SRM to SECY-12-0025, the Commission approved issuing the proposed orders with some modifications. On March 12, 2012, the NRC issued Order EA-12-049, “Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events”; and Order EA 12-051, “Order Modifying Licenses With Regard to Reliable Spent Fuel Pool Instrumentation” to the appropriate licensees and permit holders (ADAMS Accession Nos. ML12054A735 and ML12054A679, respectively). The NRC staff provides 6-month updates to the Commission on all Fukushima-related activities, including the NTTF recommendations that will be addressed in the longer term. The latest update is provided in SECY-14-0046, “Fifth 6-Month Status Update on Response to Lessons Learned from Japan's March 11, 2011, Great Tōhoku Earthquake and Subsequent Tsunami,” dated April 17, 2014 (ADAMS Accession No. ML14064A523). The NRC considered Recommendation 4.2, as modified by SRM-SECY-12-0025, using the requirements in Order EA-12-049. SECY-12-0025 outlines a three-phase approach to developing the strategies. The initial phase requires the use of installed equipment and resources to maintain or restore core cooling, containment, and SFP cooling without alternating current power or loss of normal access to the ultimate heat sink. The transition phase requires providing sufficient, portable, onsite equipment and consumables to maintain or restore these functions until they can be accomplished with resources brought from offsite. The final phase requires obtaining sufficient offsite resources to sustain those functions indefinitely. As discussed in multiple sections of the DCD, and in the FSER, the ESBWR is designed such that the reactor core and associated coolant, control, and protection systems, including station batteries and other necessary support systems, provide sufficient capacity and capability to ensure that the core will be cooled and there will be appropriate containment integrity and adequate cooling for the spent fuel for 72 hours in the event of an SBO—loss of all normal and emergency ac power. The ESBWR design credits the isolation condenser system for the first 72 hours of an event in which all ac power sources are lost. Beyond the first 72 hours, the isolation condenser system pool and SFP need to be refilled. The ESBWR design includes provisions to refill the isolation condenser system pool and SFP with onsite equipment without reliance on ac power, such as by the diesel-driven fire pump. In addition, after the first 72 hours of an event, accident mitigation is achieved through the ancillary diesel, which supplies ac power to various components such as: PCCS vent fans, motor driven fire pump, control room habitability area ventilation system air handling units, and emergency lighting. The standby diesels are also needed to support FAPCS operations. Both the ancillary and standby diesels supply short-term and long-term safety loads. For the reasons set forth in Section 22.5 of the FSER, the NRC found that the applicant has included sufficient nonsafety-related equipment in the RTNSS program to ensure that safety functions relied upon in the post-72-hour period are successful. Emergency procedures are to be developed by the COL applicant to support emergencies, which includes the period after 72 hours from the onset of the loss of all ac power. Further, the nonsafety-related equipment relied upon in the post-72-hour period has been designed in accordance with Commission policy (as described in Section 22.5.6.2 of the FSER) for use of augmented design standards for protection from external hazards and the NRC is engaging with COL applicants to ensure they have established appropriate availability controls for this equipment. Availability controls will be addressed in connection with a COL application referencing the ESBWR standard design. The ESBWR design supports a COL applicant refilling the pools with offsite equipment, such as local fire pumpers. In the period beyond seven days from the onset of the event, the COL applicant will be responsible for describing how it will make available offsite sources, such as diesel fuel oil for the ancillary and standby diesel generators and water makeup to support long term cooling. The COL applicant must address the ability of offsite support to sustain these functions indefinitely, including procedures, guidance, training and acquisition, staging or installing needed equipment. Therefore, the NRC concludes that the ESBWR design, as described in the DCD, satisfies the underlying purpose of Order EA-12-049 insofar as it includes additional equipment to maintain or restore core and spent fuel pool cooling and containment function in the event of the loss of all ac power. While the ESBWR design includes all of the necessary design features in this respect, the COL applicant must address the programmatic aspects of Order EA-12-049. The NRC staff has already engaged with COL applicants on these arrangements. To the extent a COL applicant proposes to rely on additional equipment to perform required functions in the event of a loss of all ac power, that equipment is outside the scope of the standard ESBWR design and the NRC staff will evaluate it in connection with the COL application. The NRC considered Recommendation 7.1, as modified by SRM-SECY-12-0025, using the requirements in Order EA-12-051, which describes the key parameters to be used to determine that a level instrument is considered reliable. JLD-ISG-2012-03, Revision 0, “Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation,” (ADAMS Accession No. ML12221A339) endorses with exceptions and clarifications the methodologies described in the industry guidance document NEI 12-02, Revision 1, “Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation,” (ADAMS Accession No. ML122400399) and provides an acceptable approach for satisfying the applicable requirements. The NRC finds that the ESBWR design has design features that satisfy the underlying purpose of Order EA-12-051 for reliable SFP level instrumentation, except for two matters. The exceptions are whether the safety-related level instrumentation: (1) Are designed to allow the connection of an independent power source, and (2) will maintain its design accuracy following a power interruption or change in power source without recalibration. While the ESBWR design includes all of the necessary design features in this respect, the DCD did not include any information addressing these two matters. In addition, the NRC is currently developing a rulemaking which would address spent fuel pool instrumentation for beyond design basis events/accidents. This rulemaking may adopt different requirements than what is currently considered acceptable to meet the underlying purpose of order EA-12-051 and its related guidance. For these reasons, the NRC is excluding from issue finality and issue resolution these two aspects of the ESBWR spent fuel pool instrumentation design features. The exclusions have two consequences. First, any combined license applicant referencing the ESBWR design certification rule will have to provide information demonstrating that the NRC's requirements on these two matters are met. Second, the NRC need not address the factors of 10 CFR 52.63 either when it reviews the combined license application for adequacy with respect to these two matters, or in connection with any amendment of the ESBWR design certification rule imposing requirements to govern those matters. B. Incorporation by Reference of Public Documents and Issue Resolution Associated With Non-Public Documents In Section III, “Scope and Contents,” of the proposed ESBWR DCR (76 FR 16549; March 24, 2011), the only document for which the NRC proposed to obtain approval from the Office of the Federal Register (OFR) for incorporation by reference into the ESBWR design certification rule was the ESBWR DCD, Revision 9 (DCD Revision 9). Such approval would make DCD Revision 9 a legally-binding requirement on any referencing combined license applicant and holder by virtue of publication in the Federal Register as a final rule. This was based upon the assumption that the DCD specified all necessary requirements in Tier 1 and Tier 2 (with the exception of non-public documents containing proprietary information, 2 security-related information, 3 and SGI). 2 For purposes of this discussion, “proprietary information” constitutes trade secrets or commercial or financial information that are privileged or confidential, as those terms are used under the Freedom of Information Act and the NRC's implementing regulation at 10 CFR part 9. 3 For purposes of this discussion, “security-related information” means information subject to non-disclosure under 10 CFR 2.390(a)(7)(vi). After the close of the public comment period, the NRC recognized that Tier 2, Section 1.6, “Material Incorporated by Reference and General Reference Material,” of the ESBWR DCD states that a number of documents are “incorporated by reference” into Tier 2 of the ESBWR DCD, and which contain information intended to be requirements. These documents were listed in Tables 1.6-1, “Referenced GE/GEH Reports,” and 1.6-2, “Referenced non-GE/GEH Topical Reports,” of the DCD Revision 9. Although some of the documents contain information which is intended to be requirements (based on the text of the DCD), neither Tables 1.6-1 and 1.6-2 of the DCD nor Section III of the proposed ESBWR design certification rule clearly stated which of these documents were intended as requirements. Documents intended as requirements (and which are publicly available) should have been listed in Section III of the ESBWR design certification rule as being approved for incorporation by reference by the Director of the OFR. Tables 1.6-1 and 1.6-2 also included documents that, although “incorporated by reference” into DCD Revision 9, were not intended to be requirements, but were references “for information only.” Thus, the ESBWR proposed rule did not clearly differentiate between these two different classes of documents. Finally, Tables 1.6-1 and 1.6-2 of DCD Revision 9 included both publicly-available documents and non-publicly available documents, 4 but for some of the documents which were not publicly available, GEH had not created a publicly-available version of that document to support the public comment process. The creation of publicly-available versions of non-public documents to support the public commenting process and transparency has been a long-standing practice for both design certification rulemakings and licensing actions. 4 The non-publicly available documents contain proprietary, security-related, and/or safeguards information. To address the NRC's concerns, for those non-public documents which include information intended to be treated as requirements and for which publicly-available versions were not previously created, GEH created publicly-available versions of those non-public documents. GEH also submitted Revision 10 to the DCD (DCD Revision 10), which included three tables in Section 1.6 that superseded Tables 1.6-1 and 1.6-2 in DCD Revision 9. These three tables—Tables 1.6-1, “GE/GEH Reports Incorporated by Reference,” 1.6-2, “Non-GE/GEH Reports Incorporated by Reference,” and 1.6-3, “Referenced Reports (not Incorporated by Reference,”—collectively clarify which documents are intended to be requirements and which documents are references only. The supplemental proposed rule (79 FR 25715; May 6, 2014): (1) Announced the availability of DCD Revision 10; (2) described the distinction between those documents intended as requirements versus those which were for information only; (3) requested public comments on the NRC's intent to treat 50 non-public, referenced documents in DCD Revision 10 (listed in Table 2 of the supplemental proposed rule) as requirements and matters resolved in subsequent licensing and enforcement actions for plants referencing the ESBWR design certification; and (4) clarified, but did not request public comments on, the NRC's intent to obtain approval for incorporation by reference from the Director of the OFR for both DCD Revision 10 and the 20 publicly-available documents referenced in DCD Revision 10 (listed in Table 3 of the supplemental proposed rule), which are intended by the NRC to be requirements. The 50 non-publicly available documents listed in Table 3 below are considered by the NRC to be requirements applicable to any combined license applicant or holder of a combined license referencing the ESBWR design certification rule, where the language of DCD Revision 10 makes clear that any one of those documents is intended to be a requirement. In addition, the 50 non-public documents are within the scope of issue resolution under Section VI of Appendix E, and are accorded issue finality protection under that Section VI and 10 CFR 52.63. Table 3—50 Non-Public Documents Which the NRC Regards as Requirements, Are Matters Resolved Under Paragraph VI, ISSUE RESOLUTION, of the ESBWR Design Certification Rule, and Are Accorded Issue Finality Protection Document No. Document title Publicly- available ADAMS Accession No. Non-publicly available ADAMS Accession No. NEDE-33391, NEDO-33391 GE Hitachi Nuclear Energy, “ESBWR Safeguards Assessment Report,” NEDE-33391, Class III (Safeguards, Security-Related, and Proprietary), Revision 3, March 2010, and NEDO-33391, Class I (Non-safeguards, Non-security related, and Non-proprietary), Revision 3, March 2014 ML14093A138 N/A (Safeguards information cannot be placed in ADAMS) NEDC-31959P, NEDO-31959 GE Nuclear Energy, “Fuel Rod Thermal-Mechanical Analysis Methodology (GSTRM),” NEDC-31959P (Proprietary), April 1991, and NEDO-31959 (Non-proprietary), April 1991 ML14093A145 ML14093A146 NEDC-32992P-A, NEDO-32992-A GE Nuclear Energy, J.S. Post and A.K. Chung, “ODYSY Application for Stability Licensing Calculations,” NEDC-32992P-A, Class III (Proprietary), July 2001, and NEDO-32992-A, Class I (Non-proprietary), July 2001 ML14093A250 ML012610605 NEDC-33139P-A, NEDO-33139-A Global Nuclear Fuel, “Cladding Creep Collapse,” NEDC-33139P-A, Class III (Proprietary), July 2005, and NEDO-33139-A, Class I (Non-proprietary), July 2005 ML14094A227 ML14094A228 NEDE-31758P-A, NEDO-31758-A GE Nuclear Energy, “GE Marathon Control Rod Assembly,” NEDE-31758P-A (Proprietary), October 1991, and NEDO-31758-A (Non-proprietary), October 1991 ML14093A142 ML14093A143 NEDC-32084P-A, NEDO-32084-A GE Nuclear Energy, “TASC-03A, A Computer Program for Transient Analysis of a Single Channel,” NEDC-32084P-A, Revision 2, Class III (Proprietary), July 2002, and NEDO-32084-A, Class 1 (Non-proprietary), Revision 2, September 2002 ML100220484 ML100220485 NEDC-32601 P-A, NEDO-32601-A GE Nuclear Energy, “Methodology and Uncertainties for Safety Limit MCPR Evaluations,” NEDC-32601P-A, Class III (Proprietary), and NEDO-32601-A, Class I (Non-proprietary), August 1999 ML14093A216 ML003740145 NEDC-32983P-A, NEDO-32983-A GE Nuclear Energy, “GE Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations,” Licensing Topical Report NEDC-32983P-A, Class III (Proprietary), Revision 2, January 2006, and NEDO-32983-A, Class I (Non-proprietary), Revision 2, January 2006 ML072480121 ML072480125 NEDC-33075P-A, NEDO-33075-A GE Hitachi Nuclear Energy, “General Electric Boiling Water Reactor Detect and Suppress Solution—Confirmation Density,” NEDC-33075P-A, Class III (Proprietary), and NEDO-33075-A, Class I (Non-proprietary), Revision 6, January 2008 ML080310396 ML080310402 NEDC-33079P, NEDO-33079 GE Nuclear Energy, “ESBWR Test and Analysis Program Description,” NEDC-33079P, Class III (Proprietary), Revision 1, March 2005, and NEDO-33079, Class I (Non-proprietary), Revision 1, November 2005 ML053460471 ML051390233 NEDC-33083P-A, NEDO-33083-A GE Nuclear Energy, “TRACG Application for ESBWR,” NEDC-33083P-A, Revision 1, Class III (Proprietary), September 2010, and NEDO-33083-A, Revision 1, Class I (Non-proprietary), September 2010 ML102770606 ML102770608 NEDC-33237P-A, NEDO-33237-A Global Nuclear Fuel, “GE14 for ESBWR—Critical Power Correlation, Uncertainty, and OLMCPR Development,” NEDC-33237P-A, Revision 5, Class III (Proprietary), and NEDO-33237-A, Revision 5, Class I (Non-proprietary), September 2010 ML102770246 ML102770244 NEDC-33238P, NEDO-33238 Global Nuclear Fuel, “GE14 Pressure Drop Characteristics,” NEDC-33238P, Class III (Proprietary), and NEDO-33238, Class I (Non-proprietary), December 2005 ML060050328 ML060050330 NEDC-33239P-A, NEDO-33239P-A Global Nuclear Fuel, “GE14 for ESBWR Nuclear Design Report,” NEDC-33239P-A, Class III (Proprietary), and NEDO-33239-A, Class I (Non-proprietary), Revision 5, October 2010 ML102800405 ML102800408 (part 1) ML102800425 (part 2) NEDC-33240P-A, NEDO-33240-A Global Nuclear Fuel, “GE14E Fuel Assembly Mechanical Design Report,” NEDC-33240P-A, Revision 1, Class III (Proprietary), and NEDO-33240-A, Revision 1, Class I (Non-proprietary), September 2010 ML102770060 ML102770061 NEDC-33242P-A, NEDO-33242-A Global Nuclear Fuel, “GE14 for ESBWR Fuel Rod Thermal-Mechanical Design Report,” NEDC-33242P-A, Revision 2, Class III (Proprietary), and NEDO-33242-A, Revision 2, Class I (Non-proprietary), September 2010 ML102730885 ML102730886 NEDC-33326P-A, NEDO-33326-A Global Nuclear Fuel, “GE14E for ESBWR Initial Core Nuclear Design Report,” NEDC-33326P-A, Revision 1, Class III (Proprietary), and NEDO-33326-A, Revision 1, Class I (Non-proprietary), September 2010 ML102740191 ML102740193 (part 1) ML102740194 (part 2) NEDC-33374P-A, NEDO-33374-A GE-Hitachi Nuclear Energy, “Safety Analysis Report for Fuel Storage Racks Criticality Analysis for ESBWR Plants,” NEDC-33374P-A, Revision 4, Class III (Proprietary), September 2010, and NEDO-33374-A, Revision 4, Class I (Non-proprietary), September 2010 ML102860687 ML102860688 NEDC-33456P, NEDO-33456 Global Nuclear Fuel, “Full-Scale Pressure Drop Testing for a Simulated GE14E Fuel Bundle,” NEDC-33456P, Class III (Proprietary), and NEDO-33456, Class I (Non-proprietary), Revision 0, March 2009 ML090920867 ML090920868 NEDE-10958-PA, NEDO-10958-A General Electric Company, “General Electric Thermal Analysis Basis Data, Correlation and Design Application,” NEDE-10958-PA, Class III (Proprietary), and “General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application,” NEDO-10958-A, Class I (Non-proprietary), January 1977 ML102290144 ML092820214 NEDE-24011-P-A-16, NEDO-24011-A-16 Global Nuclear Fuel, “GESTAR II General Electric Standard Application for Reactor Fuel,” NEDE-24011-P-A-16, Class III (Proprietary), and NEDO-24011-A-16, Class I (Non-proprietary), Revision 16, October 2007 ML091340077 ML091340081 NEDE-24011-P-A-US-16, NEDO-24011-A-US-16 Global Nuclear Fuel, “GESTAR II General Electric Standard Application for Reactor Fuel, Supplement for United States,” NEDE-24011-P-A-US-16, Class III (Proprietary), and NEDO-24011-A-US-16, Class I (Non-proprietary), Revision 16, October 2007 ML091340080 ML091340082 NEDE-30130-P-A, NEDO-30130-A General Electric Company, “Steady State Nuclear Methods,” NEDE-30130-P-A, Class III (Proprietary), April 1985, and NEDO-30130-A, Class I (Non-proprietary), May 1985 ML14104A064 ML070400570 NEDE-31152P, NEDO-31152 Global Nuclear Fuel, “Global Nuclear Fuels Fuel Bundle Designs,” NEDE-31152P, Revision 9, Class III (Proprietary), May 2007, and NEDO-33152, Revision 9, Class I (Non-proprietary), May 2007 ML071510287 ML071510289 NEDE-32176P, NEDO-32176 GE Hitachi Nuclear Energy, J.G.M. Andersen, et al., “TRACG Model Description,” NEDE-32176P, Revision 4, Class III (Proprietary), January 2008, and NEDO-32176, Class I (Non-proprietary), Revision 4, January 2008 ML080370271 ML080370276 NEDE-33083 Supplement 1P-A, NEDO-33083 Supplement 1-A GE Hitachi Nuclear Energy, B.S. Shiralkar, et al, “TRACG Application for ESBWR Stability Analysis,” NEDE-33083, Supplement 1P-A, Revision 2, Class III (Proprietary), September 2010, and NEDO-33083, Supplement 1-A, Revision 2, Class I (Non-proprietary), September 2010 ML102770552 ML102770550 NEDE-33083 Supplement 2P-A, NEDO-33083 Supplement 2-A GE Hitachi Nuclear Energy, “TRACG Application for ESBWR Anticipated Transient Without Scram Analyses,” NEDE-33083, Supplement 2P-A, Revision 2, Class III (Proprietary), October 2010 and NEDO-33083, Supplement 2-A, Revision 2, Class I (Non-proprietary), October 2010 ML103000353 ML103000355 NEDE-33083 Supplement 3P-A, NEDO-33083 Supplement 3-A GE Hitachi Nuclear Energy, “TRACG Application for ESBWR Transient Analysis,” NEDE-33083, Supplement 3P-A, Revision 1, Class III (Proprietary), and NEDO-33083, Supplement 3-A, Revision 1, Class I (Non-proprietary), September 2010 ML102770606 ML102770608 NEDE-33197P-A, NEDO-33197-A GE Hitachi Nuclear Energy, “Gamma Thermometer System for LPRM Calibration and Power Shape Monitoring,” NEDE-33197P-A, Revision 3, Class III (Proprietary), and NEDO-33197-A, Revision 3, Class I, (Non-proprietary), October 2010 ML102810320 ML102810341 NEDE-33217P, NEDO-33217 GE Hitachi Nuclear Energy, “ESBWR Man-Machine Interface System and Human Factors Engineering Implementation Plan,” NEDE-33217P, Class III (Proprietary), and NEDO-33217, Class I (Non-proprietary), Revision 6, February 2010 ML100480284 ML100480285 NEDE-33220P, NEDO-33220 GE Hitachi Nuclear Energy, “ESBWR Human Factors Engineering Allocation of Function Implementation Plan,” NEDE-33220P, Class III (Proprietary), and NEDO-33220, Class I (Non-proprietary), Revision 4, February 2010 ML100480209 ML100480202 NEDE-33221P, NEDO-33221 GE Hitachi Nuclear Energy, “ESBWR Human Factors Engineering Task Analysis Implementation Plan,” NEDE-33221P, Class III (Proprietary), and NEDO-33221, Class I (Non-proprietary), Revision 4, February 2010 ML100480212 ML100480213 NEDE-33226P, NEDO-33226 GE Hitachi Nuclear Energy, “ESBWR—Software Management Program Manual,” NEDE-33226P, Class III (Proprietary), Revision 5, February 2010, and NEDO-33226, Class I (Non-proprietary), Revision 5, February 2010 ML100550837 ML100550844 NEDE-33243P-A, NEDO-33243-A GE Hitachi Nuclear Energy, “ESBWR Control Rod Nuclear Design,” NEDE-33243P-A, Revision 2, Class III (Proprietary), September 2010, and NEDO-33243-A, Revision 2, Class I (Non-proprietary), September 2010 ML102740171 ML102740178 NEDE-33244P-A, NEDO-33244-A GE Hitachi Nuclear Energy, “ESBWR Marathon Control Rod Mechanical Design Report,” NEDE-33244P-A, Class III (Proprietary), Revision 2, September 2010, and NEDO-33244-A, Revision 2, Class I (Non-proprietary), September 2010 ML102770208 ML102770209 NEDE-33245P, NEDO-33245 GE Hitachi Nuclear Energy, “ESBWR—Software Quality Assurance Program Manual,” NEDE-33245P, Class III (Proprietary), Revision 5, February 2010, and NEDO-33245, Class I (Non-proprietary), Revision 5, February 2010 ML100550839 ML100550847 NEDE-33259P-A, NEDO-33259-A GE Hitachi Nuclear Energy, “Reactor Internals Flow Induced Vibration Program,” NEDE-33259P-A, Class III (Proprietary), Revision 3, October 2010, and NEDO-33259-A, Class I (Non-proprietary), Revision 3, October 2010 ML102920241 ML102920248 NEDE-33261P, NEDO-33261 GE Hitachi Nuclear Energy, “ESBWR Containment Load Definition,” NEDE-33261P, Class III (Proprietary), and NEDO-33261, Class I (Non-proprietary), Revision 2, June 2008 ML082600720 ML082600721 NEDE-33268P, NEDO-33268 GE Hitachi Nuclear Energy, “ESBWR Human Factors Engineering Human-System Interface Design Implementation Plan,” NEDE-33268P, Class III (Proprietary), and NEDO-33268, Class I (Non-proprietary), Revision 5, February 2010 ML100480179 ML100480180 NEDE-33276P, NEDO-33276 GE Hitachi Nuclear Energy, “ESBWR Human Factors Engineering Verification and Validation Implementation Plan,” NEDE-33276P, Class III (Proprietary), and NEDO-33276, Class I (Non-proprietary), Revision 4, February 2010 ML100480182 ML100480183 NEDE-33295P, NEDO-33295 GE Hitachi Nuclear Energy, “ESBWR Cyber Security Program Plan,” NEDE-33295P, Class III (Proprietary), Revision 2, September 2010, and NEDO-33295, Class I (Non-proprietary), Revision 2, September 2010 ML102880103 ML102880104 NEDE-33304P, NEDO-33304 GE Hitachi Nuclear Energy, “GEH ESBWR Setpoint Methodology,” NEDE-33304P, Class III (Proprietary), and NEDO-33304, Class I (Non-proprietary), Revision 4, May 2010 ML101450251 ML101450253 NEDE-33312P, NEDO-33312 GE Hitachi Nuclear Energy, “ESBWR Steam Dryer Acoustic Load Definition,” NEDE-33312P, Class III (Proprietary), Revision 5, December 2013, and NEDO-33312, Class I (Non-proprietary), Revision 5, December 2013 ML13344B157 ML13344B163 NEDE-33313P, NEDO-33313 GE Hitachi Nuclear Energy, “ESBWR Steam Dryer Structural Evaluation,” NEDE-33313P, Class III (Proprietary), Revision 5, December 2013, and NEDO-33313, Class I (Non-proprietary), Revision 5, December 2013 ML13344B158 ML13344B164 NEDE-33408P, NEDO-33408 GE Hitachi Nuclear Energy, “ESBWR Steam Dryer—Plant Based Load Evaluation Methodology, PBLE01 Model Description,” NEDE-33408P, Class III (Proprietary), Revision 5, December 2013, and NEDO-33408, Class I (Non-proprietary), Revision 5, December 2013 ML13344B159 ML13344B176 (part 1) ML13344B175 (part 2) NEDE-33440P, NEDO-33440 GE Hitachi Nuclear Energy “ESBWR Safety Analysis—Additional Information,” NEDE-33440P, Class III (Proprietary), and NEDO-33440, Class I (Non-proprietary), Revision 2, March 2010 ML100920316 ML100920317 (part 1) ML100920318 (part 2) NEDE-33516P-A, NEDO-33516-A GE Hitachi Nuclear Energy, “ESBWR Qualification Plan Requirements for a 72-Hour Duty Cycle Battery,” NEDE-33516P-A, Revision 2, Class III (Proprietary), September 2010, and NEDO-33516-A, Revision 2, Class I (Non-proprietary), September 2010 ML102880499 ML102880500 NEDE-33536P, NEDO-33536 GE Hitachi Nuclear Energy, “Control Building and Reactor Building Environmental Temperature Analysis for ESBWR,” NEDE-33536P, Class III (Security-Related and Proprietary), Revision 1, October 2010, and NEDO-33536, Class I (Non-security Related and Non-proprietary), Revision 1, October 2010 ML102780329 ML102780330 NEDE-33572P, NEDO-33572 GE Hitachi Nuclear Energy, “ESBWR ICS and PCCS Condenser Combustible Gas Mitigation and Structural Evaluation,” NEDE-33572P, Class II (Proprietary), Revision 3, September 2010, and NEDO-33572, Revision 3, Class I (Non-proprietary), September 2010 ML102740579 ML102740566 Letter w/attachment Letter from R.J. Reda (GE) to R.C. Jones, Jr. (NRC), MFN 098-96, “Implementation of Improved Steady-State Nuclear Methods,” Class III (Proprietary), July 2, 1996, and Letter from J.G. Head (GEH) to NRC Document Control Desk, MFN 098-96 Supplement 1, Class I (Non-proprietary), March 31, 2014 ML14093A140 ML14094A240 Table 3 Note: Documents whose document number contains “NEDC” or “NEDE” are non-public and documents whose document number contains “NEDO” are public. C. Changes to Tier 2* Information The NRC is making three changes from the proposed rule regarding Tier 2* matters under Section VIII, “Processes for Changes and Departures,” of the ESBWR rule language. These changes are described below. First, paragraph VIII.B.6.c(1) is changed from “ASME Boiler and Pressure Vessel Code, Section III” to “ASME Boiler and Pressure Vessel Code, Section III, Subsections NE (Division 1) and CC (Division 2) for containment vessel design.” This re-designation of Tier 2* information in paragraph VIII.B.6.c.(1) applies only to the ASME BPV Code, Section III, Subsections NE (Division 1) and CC (Division 2) for the design of ASME BPV Code Class MC (metal containment) and CC (concrete containment) pressure-retaining components (e.g., the containment vessel). This change does not apply to the design and construction of mechanical pressure-boundary components because they are required to meet the design and construction requirements in Section III for ASME BPV Code Class 1, 2, and 3 mechanical pressure-boundary components, which are incorporated by reference into 10 CFR 50.55a. The regulations in 10 CFR 50.55a include provisions in paragraphs 50.55a(c)(3), (d)(2) and (e)(2) for reactor coolant pressure boundary, Quality Group B, and Quality Group C (i.e., ASME BPV Code Classes 1, 2, and 3 components, respectively. These paragraphs provide the necessary regulatory controls on the use of later edition and addenda to the ASME BPV Code, Section III through the conditions the NRC established on the use of paragraph NCA-1140 of the ASME BPV Code, Section III. As a result, these rule requirements adequately control the ability of a licensee to use later editions or addenda of the ASME BPV Code, Section III such that a Tier 2* designation is not necessary. Second, paragraph VIII.B.6.c(3) is changed from “Motor-operated valves” to “Power-operated valves.” This change is necessary to correct an error in the proposed rule text. Consistent with Revisions 9 and 10 of the ESBWR DCD, which were the versions of the DCD available for public comment, the only valves that are described in Tier 2* information in an ESBWR nuclear power plant are air-operated rather than motor-operated. Third, the NRC discussed in the supplemental proposed rule its proposal to designate the revised ESBWR steam dryer analysis methodology as Tier 2* information throughout the life of any license referencing the ESBWR DCR. This is a change from Revision 9 of the ESBWR DCD, which identified much of this information (in its earlier form before the revisions reflected in Revision 10) as Tier 2. Therefore, the ESBWR steam dryer analysis methodology was not identified as Tier 2* information in the proposed rule. In the supplemental proposed rule, the NRC proposed to designate the revised ESBWR steam dryer pressure load analysis methodology as Tier 2* for two reasons. First, the NRC's experience with other applications using this methodology highlights the importance of the proper application of the steam dryer pressure load analysis methodology. Therefore, it is necessary for the NRC to review any changes a referencing applicant or licensee proposes to the methodology from that which the NRC previously reviewed and approved. Second, in Revision 10 to the ESBWR DCD, GEH revised the designation of this methodology to Tier 2* and, therefore, the rule's designation is consistent with GEH's designation in the DCD. The supplemental proposed rule provided an opportunity for public comment on the proposed designation as Tier 2* of certain information related to the pressure load analysis methodology supporting the ESBWR steam dryer design. The NRC staff did not receive any public comments on the proposal to designate information related to the ESBWR steam dryer pressure load analysis methodology as Tier 2* information. Therefore, the final rule designates the revised ESBWR steam dryer pressure load analysis methodology as Tier 2* information throughout the life of any license referencing the ESBWR DCR. D. Change Control for Severe Accident Design Features The SUPPLEMENTARY INFORMATION section of the amendment to 10 CFR part 52 (72 FR 49392, at 49394; August 28, 2007), states that the Commission codified separate criteria in paragraph B.5.c of Section VIII of each DCR for determining if a departure from design information that resolves these severe accident issues would require a license amendment. Originally, the final rule was applied specifically to changes to ex-vessel severe accident design features. In the SRM to SECY-12-0081, “Risk-Informed Regulatory Framework for New Reactors,” dated October 22, 2012, the Commission directed the staff to make the change process in paragraph B.5.c of Section VIII applicable to severe accident design features, both ex-vessel and non-ex-vessel, that are described in the plant-specific DCD. This policy was changed after issuance of the proposed ESBWR rule. The policy was changed to ensure that, for changes to Tier 2 information, the effects on all severe accident design features—and not just ex-vessel severe accident design features—are considered. However, the NRC has not changed the rule language in paragraph B.5.c of Section VIII for the ESBWR rulemaking because all of the relevant severe accident design features (i.e., those that are non-ex-vessel) are described in Tier 1 information. Tier 1 information, by definition, includes change controls in Section VIII of the rule text that meet the underlying purpose of the Commission's direction. Therefore, this change was not necessary for the ESBWR design certification. E. Access to Safeguards Information (SGI) and Sensitive Unclassified Non-Safeguards Information (SUNSI) In the four currently approved design certifications (10 CFR part 52, appendices A through D), paragraph VI.E sets forth specific directions on how to obtain access to proprietary information and SGI on the design certification in connection with a license application proceeding referencing that DCR. These provisions were developed before the events of September 11, 2001. After September 11, 2001, Congress changed the statutory requirements governing access to SGI and the NRC has revised its rules, procedures, and practices governing control of and access to SGI and SUNSI. The NRC has determined that generic direction on obtaining access to SGI and SUNSI is no longer appropriate for newly approved DCRs. Accordingly, the specific requirements governing access to SGI and SUNSI contained in paragraph VI.E of the four currently approved DCRs are not included in the DCR for the ESBWR. Instead, the NRC will specify the procedures to be used for obtaining access at an appropriate time in the COL proceeding referencing the ESBWR DCR. F. Human Factors Engineering (HFE) Operational Program Elements Exclusion From Finality In the December 6, 1996, SRM (ADAMS Accession No. ML003754873) to SECY-96-077, “Certification of Two Evolutionary Designs,” dated April 15, 1996, the Commission set forth a policy that operational programs should be excluded from finality except where necessary to find design elements acceptable. For HFE programs for the ESBWR standard design, the Commission is implementing this policy in a manner different than for other existing DCRs. The difference in treatment of HFE for the ESBWR design arises from the level of detail of HFE review for the ESBWR as compared to earlier certified standard designs. For the earlier designs, the NRC staff reviewed the HFE programs at a “programmatic” level of design, while for the ESBWR, the staff reviewed the HFE programs at a more detailed “implementation plan” level of design. In providing this additional detail, GEH addressed existing NRC guidelines in NUREG-0711, Revision 2, “Human Factors Engineering Program Review Model,” which are comprehensive and go beyond the operational program information needed as input to the HFE design. Therefore, GEH included, in the DCD, details on two HFE operational program elements (procedures and training) that are not used to determine the adequacy of the HFE design. In keeping with the established Commission policy of not approving operational program elements through design certification except where necessary to find design elements acceptable, the NRC is excluding these two HFE operational program elements in the ESBWR DCD from the scope of the design approved in the rule. This is done explicitly in Section VI, Issue Resolution, of the ESBWR rule, by excluding the two HFE operational program elements from the issue finality and issue resolution accorded to the design. In addition, the procedures and training elements included in the HFE program are redundant to what is reviewed as part of the operational programs described in Chapter 13, “Conduct of Operations,” of the SRP. Accordingly, the NRC is revising the HFE regulatory guidance in NUREG-0711, Revision 3, “Human Factors Engineering Program Review Model,” to address this overlap, but the corresponding revision to the SRP has not yet been completed. This exclusion is unique to the ESBWR design because all other DCDs for the previously certified designs do not include operational program descriptions of HFE procedures and training and the respective DCRs did not include specific exclusions from finality for them. G. Other Changes to the ESBWR Rule Language and Differences Between the ESBWR Rule and Other DCRs The language of the ESBWR design certification rule differs from the rule language of other DCRs in two substantive areas. First, paragraph IX was reserved for future use because the substantive requirements in this paragraph (for other DCRs) has since been incorporated into 10 CFR part 52 in a 2007 rulemaking (72 FR 49352; August 28, 2007) and thus are no longer needed in the four existing DCR appendices. The NRC intends to remove these requirements from Section IX of the four existing DCR appendices in future amendment(s) separate from this rulemaking. The second difference involves documents incorporated by reference into the ESBWR design certification rule. In the first four DCRs, the DCD is the only document identified in Section III of the rule language as being approved by the Office of the Federal Register for incorporation by reference. However, the ESBWR final rule identifies the ESBWR DCD and 20 publicly-available documents referenced in the DCD, Tier 2, Section 1.6 as approved for incorporation by reference. These 20 documents, which are intended by the NRC and GEH to be requirements, are listed in a table in Section III of the ESBWR final rule language. By being approved by the Office of the Federal Register for incorporation by reference, Revision 10 of the DCD and the 20 publicly-available documents are considered to be requirements as if they had been published in the Federal Register . IV. Technical Issues The NRC issued an FSER for the ESBWR design in March 2011, and subsequently published the FSER as NUREG-1966 in April 2014. The NRC issued an advanced supplemental SER in April 2014 (ADAMS Accession No. ML14043A134) and plans to publish Supplement No. 1 to NUREG-1966, as described in Section III of the SUPPLEMENTARY INFORMATION section of this document, before this final rule becomes effective. The FSER and its supplement provide the basis for issuance of a design certification under subpart B to 10 CFR part 52. The significant technical issues that were resolved during the initial review of the ESBWR design (i.e., the NRC staff's review of Revision 9 of the ESBWR DCD and development of an FSER) are: (1) Regulatory treatment of nonsafety systems (RTNSS), (2) containment performance, (3) control room cooling, (4) feedwater temperature operating domain, (5) steam dryer analysis methodology, (6) aircraft impact assessment, (7) the use of ASME Code Case N-782, and (8) an exemption for the safety parameter display system. These issues were discussed in the March 2011 proposed rule. No public comments were received on these issues. After publishing the proposed rule, the NRC addressed several issues that were changed in Revision 10 of the DCD or required a change to the FSER. The NRC staff reviewed these changes and developed an advanced supplemental SER as described above. The issues that were resolved in the advanced supplemental SER are: (1) Steam dryer analysis methodology, (2) loss of one or more phases of offsite power, (3) spent fuel assembly integrity in spent fuel racks, (4) Turbine Building Offgas System design requirements, (5) ASME Code statement in Chapter 1 of the ESBWR DCD, and (6) clarification of ASME component design ITAACs. The NRC also made changes to the advanced supplemental SER after the publication of the supplemental proposed rule. After publication of the proposed rule, the NRC addressed two issues that were not addressed in Revision 10 of the DCD or in the advanced supplemental FSER. These issues are: (1) Hurricane-generated winds and missiles, and (2) changes to Tier 2* information. Each of these issues identified above is discussed below. The public was afforded an opportunity to comment on some of these issues in the May 6, 2014 supplemental proposed rule. Section V of the SUPPLEMENTARY INFORMATION section of this document describes the NRC's bases for not offering a supplemental comment opportunity for any of the other technical issues that arose after the close of the public comment period on the proposed rule. A. Regulatory Treatment of Nonsafety Systems (RTNSS) The ESBWR safety analysis credits passive systems to perform safety functions for 72 hours following an initiating event. After 72 hours, nonsafety systems, either passive or active, replenish the passive systems in order to keep them operating or perform post-accident recovery functions directly. The ESBWR design also uses nonsafety-related active systems to provide defense-in-depth capabilities for key safety functions provided by passive systems. The challenge during the review was to identify the nonsafety SSCs that should receive enhanced regulatory treatment and to identify the appropriate regulatory treatment to be applied to these SSCs. Such SSCs are denoted as “RTNSS SSCs” in the context of the ESBWR design. As a result of the NRC's review, the applicant added Appendix 19A to the DCD to identify the nonsafety systems that perform these post-72 hour or defense-in-depth functions and the basis for their selection. The applicant's selection process was based on the guidance in SECY-94-084, “Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs.” To provide reasonable assurance that RTNSS SSCs will be available if called upon to function, the applicant established availability controls in DCD Tier 2, Appendix 19ACM, and TS in DCD Tier 2, Chapter 16, when required by 10 CFR 50.36, “Technical specifications.” The applicant also included all RTNSS SSCs in the reliability assurance program described in Chapter 17 of DCD Tier 2 and applied augmented design standards as described in DCD Tier 2, Section 19A.8.3. For the reasons set forth in Section 22.5 of the FSER, the NRC finds the applicant's treatment of the RTNSS SSCs, as described in the DCD, acceptable. B. Containment Performance The PCCS maintains the containment within its design pressure and temperature limits for DBAs. The system is passive and does not rely upon moving components or external power for initiation or operation for 72 hours following a loss-of-coolant accident (LOCA). The PCCS and its design basis are described in detail in Section 6.2.2 of the DCD Tier 2. The NRC identified a concern regarding the PCCS long-term cooling capability for the period from 72 hours to 30 days following a LOCA. To address this concern, the applicant proposed additional design features credited after 72 hours to reduce the long-term containment pressure. The features are the PCCS vent fans and passive autocatalytic hydrogen recombiners as described in DCD Tier 2, Section 6.2.1. These SSCs have been identified in DCD Appendix 19A as RTNSS SSCs. The NRC staff's review of the PCCS design is documented in Section 6.2.2 of the FSER. The following is a summary of key points of that review. The applicant provided calculation results to demonstrate that the long-term containment pressure would be acceptable and that the design complies with GDC 38. The NRC's independent calculations confirmed the applicant's conclusion and the NRC accepts the proposed design and licensing basis. The NRC also raised a concern regarding the potential accumulation of high concentrations of hydrogen and oxygen in the PCCS and Isolation Condenser System, which could lead to combustion following a LOCA. The applicant modified the design of the PCCS and Isolation Condenser System heat exchangers to withstand potential hydrogen detonations. Accordingly, the NRC concludes that the design changes to the PCCS and Isolation Condenser System are acceptable and meet the applicable requirements. C. Control Room Cooling The ESBWR primarily relies on the mass and structure of the control building to maintain acceptable temperatures for human and equipment performance for up to 72 hours on loss of normal cooling. The NRC had not previously approved this approach for maintaining acceptable temperatures in the control building. The applicant proposed acceptance criteria for the evaluation of the control building structure's thermal performance based on industry and NRC guidelines. The applicant incorporated by reference an analysis of the control building structure's thermal performance as described in Tier 2, Sections 3H, 6.4, and 9.4. The applicant also proposed ITAACs to confirm that an updated analysis of the as-built structure continues to meet the thermal performance acceptance criteria. For the reasons set forth in Section 6.4.3 of the FSER, the NRC finds that the applicant's acceptance criteria are consistent with the advanced light water reactor control room envelope atmosphere temperature limits in NUREG-1242, “NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document,” and the use of the wet bulb globe temperature index in evaluation of heat stress conditions as described in NUREG-0700, “Human-System Interface Design Review Guidelines.” For the reasons set forth in Section 9.4.1 of the FSER, the NRC finds the control building structure thermal performance analysis and ITAACs acceptable based on the analysis using bounding environmental assumptions. Accordingly, the NRC finds that the acceptance criteria, control building structure thermal performance analysis, and the ITAACs, provide reasonable assurance that acceptable temperatures will be maintained in the control building for 72 hours. Therefore, the NRC finds that the control building design in regard to thermal performance conforms to the guidelines of SRP Section 6.4 and complies with the requirements of the GDC 19. D. Feedwater Temperature Operating Domain In operating BWRs, the recirculation pumps are used in combination with the control rods to control and maneuver reactor power level during normal power operation. The ESB [Text truncated at 120,000 characters. The full text is on the page linked above.] --- Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A2014-24362. Public record. Not legal advice.