ML070310045

of Regulatory Guide 1.136, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments (Formerly Draft Regulatory Guide DG-1159)

ML070310045
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Issue date:	03/11/2007
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DG-1159 RG-1.136, Rev 3
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March 2007 U.S. NUCLEAR REGULATORY COMMISSION Revision 3 REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH REGULATORY GUIDE 1.136 (Draft was issued as DG-1159, dated October 2006)

DESIGN LIMITS, LOADING COMBINATIONS, MATERIALS, CONSTRUCTION, AND TESTING OF CONCRETE CONTAINMENTS A. INTRODUCTION This regulatory guide describes an approach that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for use in satisfying the requirements of General Design Criteria (GDC) 1, 2, 4, 16, and 50, as specified in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), Domestic Licensing of Production and Utilization Facilities (Ref. 1). Specifically, GDC 1, Quality standards and records, requires, in part, that structures, systems, and components (SSCs) important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

To augment those requirements, GDC 2, Design bases for protection against natural phenomena, requires that SSCs important to safety be designed to withstand the effects of expected natural phenomena when combined with the effects of normal and accident conditions. Similarly, GDC 4, Environmental and dynamic effects, design bases, requires that nuclear power plant SSCs important to safety be designed to accommodate the effects of and to be compatible with environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents (LOCAs).

The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agencys regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staff need in reviewing applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission.

This guide was issued after consideration of comments received from the public. The NRC staff encourages and welcomes comments and suggestions in connection with improvements to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed. The NRC staff will revise existing guides, as appropriate, to accommodate comments and to reflect new information or experience. Written comments may be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

Regulatory guides are issued in 10 broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review; and 10, General.

Requests for single copies of draft or active regulatory guides (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to Distribution@nrc.gov.

Electronic copies of this guide and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRCs Electronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML070310045.

In addition, GDC 16, Containment design, requires that a reactor containment and its associated systems be provided to establish an essentially leaktight barrier against uncontrolled release of radioactivity to the environment and to ensure that design conditions important to safety are not exceeded for as long as required for postulated accident conditions. Finally, GDC 50, Containment design basis, requires that the reactor containment structure (including access openings, penetrations, and containment heat removal systems) be designed so that the structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions caused by any LOCA.

10 CFR 50.44 provides the requirements for combustible gas control for currently licensed reactors and for future water-cooled reactor applicants and licensees. This regulatory guide describes an approach acceptable to the NRC staff to consider the structural loads involved and determine the containment response in order to demonstrate the containment structural integrity.

In addition, for certain reactors specified in 10 CFR 50.34, Contents of applications: technical information, requires that plant designs must accommodate loadings associated with hydrogen generation that results from metal-water reaction of the fuel cladding accompanied by hydrogen burning or the added pressure of inerting system actuation. In addition, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 requires, in part, that measures must be established to ensure materials control and control of special processes, such as welding, and proper testing must be performed.

This regulatory guide contains information collections covered by the requirements of 10 CFR Part 50, which the Office of Management and Budget (OMB) approved under OMB control number 3150-0011. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement unless the requesting document displays a currently valid OMB control number.

Rev. 3 of RG 1.136, Page 2

B. DISCUSSION The American Society of Mechanical Engineers (ASME) and the American Concrete Institute (ACI) have jointly published the Code for Concrete Containments, also known as either the ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 2, or ACI Standard 359-01, which this guide refers to as the Code (Ref. 2). This regulatory guide endorses the 2001 Edition of the Code with the 2003 addenda, with the exceptions discussed herein.

Significant advancement in technology, both in the nuclear industry and the Code, has prompted a need to revise this regulatory guide for concrete containments. The existing industry codes and standards are based on the current class of light-water reactors and, as such, may not adequately address design and construction features of the next generation of advanced light-water reactors and high-temperature gas-cooled reactors. Nonetheless, the NRC remains committed to the use of industry consensus codes and standards for the design, construction, and licensing of commercial nuclear power reactors facilities.

Toward that end, this guide describes methods that the NRC staff considers acceptable with regard to the materials, design, construction, and testing of reinforced and prestressed concrete containments. As such, the provisions of this guide may be used for the current light-water reactors, as well as future advanced reactors, such as the Advanced Pressurized-Water Reactor (AP1000) and the Economic Simplified Boiling-Water Reactor (ESBWR).

The NRC staff has evaluated the provisions contained in the Articles CC-1000 through CC-6000 of the Code, and is in the process of coordinating related updates to other regulatory guides, as well as NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants (Ref. 3). As a result, this regulatory guide endorses Articles CC-1000 through CC-6000 of the Code, with the exceptions noted herein. This regulatory guide also provides guidance on loads and load combinations, design and analysis, and a method for determining the ultimate capacity of a concrete containment. In addition, this guide reviews the quality control program proposed for the fabrication and construction of the containment, with emphasis on the extent of compliance with Articles CC-4000 and CC-5000 of the Code, including the following:

Examination of the materials, including tests to determine the physical properties of concrete, reinforcing steel, mechanical splices, the liner plate and its anchors, and the prestressing system, if any; placement of concrete; and erection tolerances of the liner plate, reinforcement, and prestressing systems.

10 CFR 50.44(b)(2)(i) requires that all currently licensed boiling-water reactors with Mark I or Mark II type containments must have an inerted atmosphere. 10 CFR 50.44(b)(2)(ii) requires that all currently licensed boiling-water reactors with Mark III type containments and all pressurized-water reactors with ice condenser containments must have the capability for controlling combustible gas generated from a metal-water reaction involving 75 percent of the fuel cladding surrounding the active fuel region so that there is no loss of containment structural integrity. 10 CFR 50.44(b)(5)(v)(B) requires that for all currently licensed boiling-water reactors with Mark III type containments and all pressurized-water reactors with ice condenser containments demonstrate that systems and components necessary to establish and maintain safe shutdown and to maintain containment integrity will be capable of performing their functions during and after exposure to the environmental conditions created by the burning of hydrogen, including local detonations, unless such detonations can be shown to be unlikely to occur.

Rev. 3 of RG 1.136, Page 3

10 CFR 50.44(c)(3) requires that future water-cooled reactors containments that do not rely upon an inerted atmosphere to control combustible gases must have the capability for controlling combustible gas generated from a metal-water reaction involving 100 percent of the fuel cladding surrounding the active fuel region so that there is no loss of containment structural integrity. 10 CFR 50.44(c)(5) requires that for future water-cooled reactors containments, an applicant must perform an analysis that demonstrates containment structural integrity. This demonstration must use an analytical technique that is accepted by the NRC and include sufficient supporting justification to show that the technique describes the containment response to the structural loads involved. The analysis must address an accident that releases hydrogen generated from 100 percent fuel clad-coolant reaction accompanied by hydrogen burning.

To address the requirements of 10 CFR 50.34(f) and 10 CFR 50.44(b) and (c), Regulatory Position C.5 provides loads and load combinations for pressure loads that result from a fuel-clad metal-water reaction, an uncontrolled hydrogen burn, and from a post-accident condition inerted by carbon dioxide.

10 CFR 50.55a imposes the examination requirements established in Section XI, Subsection IWL, of the Boiler and Pressure Vessel (B&PV) Code (Ref. 4) promulgated by the American Society of Mechanical Engineers (ASME), as they relate to reinforced and prestressed concrete (Class CC) containments.

In those areas where the provisions of the referenced Code are insufficient for licensing purposes, the staff has provided supplementary guidelines, as part of the regulatory position presented in Section C of this guide. The reasons for the supplementary guidance for each regulatory position are as follows.

CC-2243: Cement Grout for Grouted Tendon Systems1 Regulatory Position C.1 recommends using the guidance in Regulatory Guide 1.107, Qualification for Cement Grouting for Prestressing Tendons in Containment Structures (Ref. 5), rather than Article CC-2243 with respect to grouting of prestressing tendons. The staff believes that the recommendations presented herein provide needed assurance regarding the integrity of grouted tendons that cannot be directly inspected during the life of the containment.

CC-2433.2.3: Acceptance Standards Experience with the use of alloy steel materials for anchor blocks and wedge blocks (such as AISI 4140) indicates that a high degree of hardness of these materials is a factor in causing cracking (presumably stress-corrosion) under certain inevitable environments. Also, it is necessary to control the uniformity of the hardness of these materials. A thorough surface examination and proper protection before and after installation of these materials, together with close control of the amount and uniformity of hardness in these materials, may eliminate cracking. Regulatory Position C.2 provides guidance in addition to the Code specifications.

1 This alphanumeric citing identifies the article, and paragraph if applicable, of the Code for Concrete Reactor Vessels and Containments being discussed.

Rev. 3 of RG 1.136, Page 4

CC-2434: Wedges and Anchor Nuts The testing of prestressing materials to qualify them against loss of ductility in cold temperatures is needed; therefore, the guidance in Regulatory Position C.3 is recommended.

CC-2463.1: Static Tensile Test Various prestressing systems may require different numbers of tests for tendon systems to establish their adequacy for use. Variations within the tolerance limits of the construction specification for material properties and geometry of anchorages and tendons must be realistically and adequately represented in the system testing. Therefore, Regulatory Position C.4 recommends that any system of prestressing should be subjected to a sufficient number of tests to establish its adequacy before it is adopted for use.

CC-3000: Design To facilitate design and analysis procedures, some specific guidance in addition to the Code specification are included in Regulatory Position C.5 in order to be consistent with the current staff position. That is, the loads and loading combinations guidelines for the design and analysis of concrete containments are added in this regulatory guide. This method has been used by the staff in reviewing new reactor applications and is being considered for inclusion in the update to SRP Section 3.8.1, Concrete Containment.

CC-3542: Loss of Prestress Regulatory Position C.6 recommends using the guidance in Regulatory Guide 1.35.1, Determining Prestressing Forces for Inspection of Prestressed Concrete Containments (Ref. 6), rather than Article CC-3542 with respect to loss of prestress in tendons. The staff believes that the recommendations presented herein provide needed assurance regarding the loss tendon prestress.

CC-4240: Curing The Code does not provide explicit guidance for curing concrete at temperatures higher than 4.4 °C (40 °F). Consequently, Regulatory Position C.7 gives guidance for curing concrete.

CC-4352: Splices Welded splices and other mechanical connections are allowed as long as they conform to ACI-349-01, Section 12.14.3 (Ref. 7). Regulatory Position C.8 gives guidance for splices.

CC-4470: Permanent Corrosion Protection See discussion concerning CC-3542, Loss of Prestress, and Regulatory Position C.9.

Rev. 3 of RG 1.136, Page 5

CC-5210: General The locations of all major embedments, such as plates, embedded piping penetration sleeves, major structural framings, and anchor bolts, should be preplanned, identified on the design drawings, and documented on field changes thereto. This would permit verification that embedments have been placed with full consideration given to the resulting reduction in structural strengths, radiation shielding effectiveness, and hindrance to the placement and consolidation of concrete. In this regard, Regulatory Position C.10 provides guidance in addition to the Code.

Ultimate Capacity of Concrete Containment New guidelines for the ultimate capacity of concrete containments are added in this regulatory guide in order to be consistent with the current staff position. This method has been used by the staff in reviewing new reactor applications and is being considered for inclusion in the update to SRP 3.8.1, Concrete Containment.

Rev. 3 of RG 1.136, Page 6

C. REGULATORY POSITION The design, materials, fabrication, erection, inspection, testing, and inservice surveillance of concrete containments are covered by codes, standards, specifications, and guides that are applicable either in their entirety or in part. In addition to this regulatory guide, the following codes and guides are acceptable to the staff:

Code Title ASME,Section III, Division 2 Code for Concrete Containments ASME,Section III, Subsection NCA General Requirements for Division 1 and Division 2 For earthquake engineering criteria, 10 CFR Part 50, Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants, would be applicable for the operating-basis earthquake (OBE) and safe-shutdown earthquake (SSE). In this manner, the OBE serves the function as an inspection-level earthquake below which the effect on the health and safety of the public would be insignificant and above which the licensee would be required to shut down the plant and inspect for damage.

Articles CC-1000 through CC-6000 of the Code are acceptable to the NRC staff for the scope, material, design, construction, examination, and testing of concrete containments of nuclear power plants subject to the following regulatory positions.

1. CC-2243: Cement Grout for Grouted Tendon Systems Regulatory Guide 1.107, Qualifications for Cement Grouting for Prestressing Tendons in Containment Structures (Ref. 5), should be used for guidance on qualifying grout for grouted tendon systems.

2. CC-2433.2.3: Acceptance Standards In addition to CC-2433.2.3, Acceptance Standards, the following guidance should be used:

The maximum hardness for material of anchor head assemblies and wedge blocks shall not exceed that of Rockwell C40. To maintain uniformity in hardness, the tolerance on a designated hardness number shall not exceed +/-2.

3. CC-2434: Wedges and Anchor Nuts In addition to CC-2434, Wedges and Anchor Nuts, the following guidance should be used for protection of prestressing materials from low-temperature effects:

Materials for all load-bearing components of prestressing systems should be selected so that they can withstand the anticipated low-temperature effects without loss in their ductility.

Methods and procedures similar to those used for materials of liners in CC-2520, Fracture Toughness Requirements for Materials, are acceptable for qualifying the materials. Additionally, suitable tests should be conducted to demonstrate that with the maximum allowable flaw size (cracked buttonheads, wedges, and anchor nuts), the specific components will exhibit the required strength and ductility under the lowest anticipated temperatures.

Rev. 3 of RG 1.136, Page 7

4. CC-2463.1: Static Tensile Test In addition to CC-2463.1, Static Tensile Test, the following guidance should be used:

Any system of prestressing should be subjected to a sufficient number of tests to establish its adequacy. Justification that a sufficient number of tests have been performed, as well as a description of the test program, should be submitted to the NRC for review.

5. CC-3000: Design Design and analysis procedures for structural portions of the containment, and specified allowable limits for stresses and strains, should be in accordance with Article CC-3000 of the Code, with the following considerations:

A. The specified loads and load combinations are acceptable if found to be in accordance with Article CC-3000 of the Code, with the exceptions listed below to Table CC-3230-1.

(1) Hydrodynamic loads resulting from LOCA and/or safety/relief valve (SRV) actuation should be combined according to the approach contained in the appendix to SRP Section 3.8.1.

(2) Where post-LOCA flooding is a design consideration, the load combination in the Code containing LOCA flooding along with OBE should be considered. Where post-LOCA flooding is combined with the OBE set at one-third or less of the SSE for the plant, this load combination may be eliminated provided the load combination is shown to be less severe than one of the other load combinations.

B. Subarticle CC-3720 of the Code should be followed when the containment structure is exposed to the loading conditions listed below. These loading conditions should include the effect of temperature. For prestressed concrete containment the effects of prestress should also be considered.

(1) For the Factored Load Category:

D + Pg1 + [Pg2 or Pg3 ] where D = Dead load Pg1 = Pressure resulting from an accident that releases hydrogen generated from 100% fuel clad metal-water reaction Pg2 = Pressure resulting from uncontrolled hydrogen burning Pg3 = Pressure resulting from post-accident inerting, assuming carbon dioxide is the inerting agent See Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Containment (Ref. 8), for additional guidance about the pressure load Pg3 due to combustible gas concentration.

(2) For the Service Load Category, the strains in the containment liner should not exceed the limits set forth in Subarticle CC-3720 when exposed to pressure Pg3.

(3) As a minimum design condition for either condition 1 or 2 above, the following load combination should be satisfied:

D + 310 kPa (45 psig)

Rev. 3 of RG 1.136, Page 8

C. For the structural portions of the containment, the specified allowable limits for stresses and strains should be in accordance with Subsection CC-3400 of the Code, but with the following exceptions:

CC-3421.5 For existing plants, the tangential shear stress carried by the concrete in reinforced concrete containments should be limited to 276 kPa (40 psi) and 414 kPa (60 psi) for the load combinations of Table CC-3230-1, representing abnormal/severe environmental conditions and abnormal/extreme environmental conditions, respectively.

For new plants, tangential shear stress carried by the concrete in reinforced concrete containments should be zero in accordance with the Code. This position is also supported by the ASME Code and the test data contained in NUREG/CR-5209, Design Provisions for Tangential Shear in Containment Walls (Ref. 9).

The allowable upper limit of the tangential shear strength provided by orthogonal reinforcement should be limited to the following value:

26.25 f'c (kPa) [10 f'c (psi)] where f'c is in accordance with the Code, and is in kPa in the first expression and is in psi in the second expression.

For existing prestressed concrete containments, the principal tensile stress should not exceed the following value:

10.5 f'c (kPa) [4 f'c (psi)] where f'c is in accordance with the Code, and is in kPa in the first expression and is in psi in the second expression.

For new prestressed concrete containments, the principal tensile stress should be in accordance with the Code.

6. CC-3542: Loss of Prestress Regulatory Guide 1.35.1, Determining Prestressing Forces for Inspection of Prestressed Concrete Containments (Ref. 6), should be used for guidance in determining loss of prestress in tendons.

7. CC-4240: Curing In addition to the specifications for curing concrete in Subarticle CC- 4240, the following guidance should be used:

Curing and protection against physical and thermal damage from time of placement until end of minimum curing period should be in accordance with ACI 308.1, and ACI 305R-99 or ACI 306.1-90(R2002) as applicable.

8. CC-4352: Splices In addition to the specifications in paragraph CC-4352, the following guidance should be used:

Mechanical splices located in areas of high stresses (maximum computed tensile stress 0.5 Fy) should have alternate bars spliced or adjacent splices staggered. If tests for slip (or internal plastic deformation) of the splice demonstrate that the slip is low (i.e., not to exceed 50% of the elongation of the unspliced bar along the spliced length), at 0.9 Fy, the adjacent splices need not be staggered.

Rev. 3 of RG 1.136, Page 9

9. CC-4470: Permanent Corrosion Protection:

ISI of reinforced and prestressed containment with ungrouted tendons should be in accordance with the requirements of Subsection IWL of Section XI of the ASME Code. ISI of prestressed concrete containment with grouted tendons should be in accordance with Regulatory Guide 1.90, Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons (Ref. 10).

10. CC-5210: General The provisions of Article CC-5210 should be supplemented by an inspection to ensure that only those embedments shown on the design drawings (except minor embedments such as rebar supports and form ties), or covered by documented field changes and later placed on the as-built drawings, remain in the form after the concrete is placed. Additionally, the inspection should ensure that hollow tubes and pipe sections used as support systems or for other construction convenience, if left embedded in the concrete, are filled with concrete or grout as appropriate.

Ultimate Capacity of Concrete Containment A non-linear finite element analysis should be performed to determine the ultimate capacity of the containment. Additional information guidance is provided in the SRP 3.8.1.

Rev. 3 of RG 1.136, Page 10

D. IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staffs plans for using this regulatory guide. No backfitting is intended or approved in connection with its issuance.

Except in those cases in which an applicant or licensee proposes or has previously established an acceptable alternative method for complying with specified portions of the NRCs regulations, the methods to be described in the active guide will reflect public comments and will be used in evaluating (1) submittals in connection with applications for construction permits, standard plant design certifications, operating licenses, early site permits, and combined licenses; and (2) submittals from operating reactor licensees who voluntarily propose to initiate system modifications if there is a clear nexus between the proposed modifications and the subject for which guidance is provided herein.

REGULATORY ANALYSIS / BACKFIT ANALYSIS The regulatory analysis and backfit analysis for this regulatory guide are available in Draft Regulatory Guide DG-1159, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments (Ref. 11). The NRC issued DG-1159 in October 2006 to solicit public comment on the draft of this Revision 3 of Regulatory Guide 1.136.

Rev. 3 of RG 1.136, Page 11

REFERENCES

1. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities.2

2. ASME Boiler & Pressure Vessel Code,Section III, Rules for Construction of Nuclear Facility Components, Division 2, Code for Concrete Containments, 2001 Edition with 2003 Addenda, also known as ACI Standard 359-01, American Society of Mechanical Engineers, New York, New York.3

3. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC.4

4. ASME Boiler & Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2001 Edition with 2003 Addenda, American Society of Mechanical Engineers, New York, New York.3

5. Regulatory Guide 1.107, Qualifications for Cement Grouting for Prestressing Tendons in Containment Structures, U.S. Nuclear Regulatory Commission, Washington, DC.5

6. Regulatory Guide 1.35.1, Determining Prestressing Forces for Inspection of Prestressed Concrete Containments, U.S. Nuclear Regulatory Commission, Washington, DC.5 2

All NRC regulations listed herein are available electronically through the Electronic Reading Room on the NRCs public Web site, at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Copies are also available for inspection or copying for a fee from the NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; email PDR@nrc.gov.

3 Copies of the Code and addenda thereto may be obtained from the American Society of Mechanical Engineers, Three Park Avenue, New York, New York 10016-5990.

4 All NUREG-series reports listed herein were published by the U.S. Nuclear Regulatory Commission. Copies are available for inspection or copying for a fee from the NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; email PDR@nrc.gov. Copies are also available at current rates from the U.S.

Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328, telephone (202) 512-1800; or from the National Technical Information Service (NTIS) at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov, by telephone at (800) 553-NTIS (6847) or (703) 605-6000, or by fax to (703) 605-6900.

NUREG-0800 is also available electronically through the Electronic Reading Room on the NRCs public Web site, at http://www.nrc.gov/ reading-rm/doc-collections/nuregs/.

5 All regulatory guides listed herein were published by the U.S. Nuclear Regulatory Commission. Where an ADAMS accession number is identified, the specified regulatory guide is available electronically through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html.

All other regulatory guides are available electronically through the Electronic Reading Room on the NRCs public Web site, at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/. Single copies of regulatory guides may also be obtained free of charge by writing the Reproduction and Distribution Services Section, ADM, USNRC, Washington, DC 20555-0001, or by fax to (301)415-2289, or by email to DISTRIBUTION@nrc.gov. Active guides may also be purchased from the National Technical Information Service (NTIS)on a standing order basis. Details on this service may be obtained by contacting NTIS at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov, by telephone at (800) 553-NTIS (6847) or (703) 605-6000, or by fax to (703) 605-6900. Copies are also available for inspection or copying for a fee from the NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs mailing address is USNRC PDR, Washington, DC 20555-0001.

The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax at (301) 415-3548, and by email to PDR@nrc.gov.

Rev. 3 of RG 1.136, Page 12

7. ACI-349-01, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Farmington Hills, Michigan, 2001.6

8. Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Containment, U.S. Nuclear Regulatory Commission, Washington, DC.5

9. NUREG/CR-5209, Design Provisions for Tangential Shear in Containment Walls, U.S. Nuclear Regulatory Commission, Washington, DC, August 1988.4

10. Regulatory Guide 1.90, Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons, U.S. Nuclear Regulatory Commission, Washington, DC.5

11. Draft Regulatory Guide DG-1159, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments, U.S. Nuclear Regulatory Commission, Washington, DC.7 6

Copies of the Code may be obtained from the American Concrete Institute, 38800 Country Club Drive, Farmington Hills, MI 48331.

7 Draft Regulatory Guide DG-1159 is available electronically under Accession #ML063000430 in the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html.

Copies are also available for inspection or copying for a fee from the NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville Maryland; the PDRs mailing address is USNRC PDR, Washington, DC 20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax at (301) 415-3548, and by email to PDR@nrc.gov.

Rev. 3 of RG 1.136, Page 13

BIBLIOGRAPHY ACI-305R, Hot Weather Concreting, American Concrete Institute, Farmington Hills, Michigan, 1999.6 ACI-306.1, Standard Specification for Cold Weather Concreting, American Concrete Institute, Farmington Hills, Michigan, 1990 (R2002).6 ACI-308.1, Standard Specification for Curing Concrete, American Concrete Institute, Farmington Hills, Michigan, 1998.6 ACI-318, Building Code Requirements for Reinforced Concrete, American Concrete Institute, Farmington Hills, Michigan, 2005.6 NUREG/CR-6906, Containment Integrity Research at Sandia National Laboratories: An Overview, U.S. Nuclear Regulatory Commission, Washington, DC, June 2006.4 Regulatory Guide 1.35, Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments, U.S. Nuclear Regulatory Commission, Washington, DC.5 NRC Enforcement Policy, Policy Statement: Revision, U.S. Nuclear Regulatory Commission, Federal Register, Vol. 69, No. 115, June 16, 2004, pp. 33684-33685.8 NRC Enforcement Policy; Extension of Enforcement Discretion of Interim Policy, Policy Statement:

Revision, U.S. Nuclear Regulatory Commission, Federal Register, Vol. 70, No. 10, January 14, 2005, pp. 2662-2664.8 8

All Federal Register notices listed herein were issued by the U.S. Nuclear Regulatory Commission, and are available electronically through the Federal Register Main Page of the public GPOAccess Web site, which the U.S. Government Printing Office maintains at http://www.gpoaccess.gov/fr/index.html. Copies are also available for inspection or copying for a fee from the NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; email PDR@nrc.gov.

Rev. 3 of RG 1.136, Page 14