2191 R0 XI.S1

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XI.S1 ASME SECTION XI, SUBSECTION IWE

Program Description

Title 10 of the Code of Federal Regulations (10 CFR) 50.55a imposes the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Subsection IWE, for steel containments (Class MC) and steel liners for concrete containments (Class CC). The scope of Subsection IWE includes steel containment shells and their integral attachments, steel liners for concrete containments and their integral attachments, containment penetrations, hatches, airlocks, moisture barriers, and pressure-retaining bolting. The requirements of ASME Code, Section XI, Subsection IWE, with the additional requirements specified in 10 CFR 50.55a(b)(2), are supplemented herein to augment an existing program applicable to managing aging of steel containments, steel liners of concrete containments, and other containment components for the subsequent period of extended operation.

The primary ISI method specified in IWE is visual examination (general visual, VT-3, VT-1). Limited volumetric examination (ultrasonic thickness measurement) and surface examination (e.g., liquid penetrant) may also be necessary in some instances to detect aging effects. IWE specifies acceptance criteria, corrective actions, and expansion of the inspection scope when degradation exceeding the acceptance criteria are found.

Subsection IWE requires examination of coatings that are intended to prevent corrosion. Aging management program (AMP) XI.S8 is a protective coating monitoring and maintenance program that is recommended to provide reasonable assurance of emergency core cooling system (ECCS) operability, whether or not the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report AMP XI.S8 is credited in the GALL-SLR Report AMP XI.S1.

The program attributes are supplemented to incorporate aging management activities, recommended in the Final License Renewal Interim Staff Guidance (LR-ISG)-2006-01, needed to address the potential loss of material due to corrosion in the inaccessible areas of the boiling water reactor (BWR) Mark I steel containment.

The program attributes are supplemented to consider the operating experience (OE) of two-ply bellows for detection of cracking described in the U.S. Nuclear Regulatory Commission (US NRC) Information Notice (IN) 92-20, “Inadequate Local Leak Rate Testing,” and to also include preventive actions to provide reasonable assurance that bolting integrity is maintained. The program is also supplemented to perform surface examination’ of pressure-retaining components that are subject to cyclic loading but have no current licensing basis (CLB) fatigue analysis; and, based on plant-specific OE, a one-time volumetric examination of metal shell or liner surfaces that are inaccessible from one side.

Evaluation and Technical Basis

1. Scope of Program: The scope of this program addresses the pressure-retaining components of steel containments and steel liners of concrete containments specified in Subsection IWE-1000 and are supplemented to address aging management of potential corrosion in inaccessible areas of the drywell shell exterior of BWR Mark I steel containments. The components within the scope of Subsection IWE are Class Metal Containment (MC) pressure-retaining components (steel containments) and their integral attachments, metallic shell and penetration liners of Class CC containments and their integral attachments, containment moisture barriers, containment pressure-retaining bolting, and metal containment surface areas, including welds and base metal. The concrete portions of containments are inspected in accordance with Subsection IWL. Subsection IWE requires examination of coatings that are intended to prevent corrosion, including those inside BWR suppression chambers. The GALL-SLR Report AMP XI.S8 is a protective coating monitoring and maintenance program that is recommended to provide reasonable assurance of ECCS operability, whether or not the GALL-SLR Report AMP XI.S8 is credited in GALL-SLR Report AMP XI.S1.

Subsection IWE exempts the following from examination:

1. Components that are outside the boundaries of the containment, as defined in the plant-specific design specification;
2. Embedded or inaccessible portions of containment components that met the requirements of the original construction code of record;
3. Components that become embedded or inaccessible as a result of containment structure [i.e., steel containments (Class MC) and steel liners of concrete containments (Class CC)] repair or replacement, provided the requirements of IWE-1232 and IWE-5220 are met; and
4. Piping, pumps, and valves that are part of the containment system or that penetrate or are attached to the containment vessel (governed by IWB or IWC).

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10 CFR 50.55a(b)(2)(ix) and IWE-2420 (2006 and later editions/addenda) specify additional requirements for inaccessible areas. It states that the licensee is to evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. Examination requirements for containment supports are not within the scope of Subsection IWE.

2. Preventive Action: This is a condition monitoring program. The program is supplemented to include preventive actions that provide reasonable assurance that moisture levels associated with an accelerated corrosion rate do not exist in the exterior portion of the BWR Mark I steel containment drywell shell. The actions consist of ensuring that the sand pocket area drains and/or the refueling seal drains are clear. The program is also supplemented to include preventive actions to provide reasonable assurance that bolting integrity is maintained, as discussed in Electric Power Research Institute (EPRI) documents (such as EPRI NP-5067(Archived) and TR-104213(Archived)), American Society for Testing and Materials (ASTM) standards, and American Institute of Steel Construction specifications, as applicable. The preventive actions should emphasize proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting. If the structural bolting consists of ASTM A325 and/or ASTM A490 bolts (including respective equivalent twist-off type ASTM F1852 and/or ASTM F2280 bolts), the preventive actions for storage, lubricant selection, and bolting and coating material selection discussed in Section 2 of Research Council for Structural Connections publication “Specification for Structural Joints Using High-Strength Bolts,” need to be considered.

3. Parameters Monitored or Inspected: Table IWE-2500-1 references the applicable sections in IWE-2300 and IWE-3500 that identify the parameters examined or monitored. Noncoated surfaces are examined for evidence of cracking, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, and other signs of surface irregularities including discernible liner plate bulges. Painted or coated surfaces, including those inside BWR suppression chambers, are examined for evidence of flaking, blistering, peeling, discoloration, and other signs of potential distress of the underlying metal shell or liner system, including discernible liner plate bulges. Steel, stainless steel (SS), and dissimilar metal weld pressure-retaining components that are subject to cyclic loading but have no CLB fatigue analysis (i.e., components covered by Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR) Table 3.5-1, items 27 and 40, and corresponding GALL-SLR items; as applicable), are monitored for cracking. The moisture barriers are examined for wear, damage, erosion, tear, surface cracks, or other defects that permit intrusion of moisture in the inaccessible areas of the pressure retaining surfaces of the metal containment shell or liner. Pressure-retaining bolting is examined for loosening and material conditions that cause the bolted connection to affect either containment leak-tightness or structural integrity.

Subsequent license renewal applicants with BWR Mark I steel containments should periodically monitor the sand pocket area drains and/or the refueling seal drains for water leakage. The applicants should also ensure the drains are clear to prevent moisture levels associated with accelerated corrosion rates in the exterior portion of the drywell shell.

4. Detection of Aging Effects: The examination methods, frequency, and scope of examination specified in 10 CFR 50.55a and Subsection IWE provide reasonable assurance that aging effects are detected before they compromise the design-basis requirements. IWE-2500-1 and the requirements of 10 CFR 50.55a provide information regarding the examination categories, parts examined, and examination methods to be used to detect aging.

Regarding the extent of examination, all accessible surfaces receive at least a general visual examination as specified in Table IWE-2500-1 and the requirements of 10 CFR 50.55a with results evaluated in accordance with IWE-3100. The acceptability of inaccessible areas of the steel containment shell or concrete containment steel liner is evaluated when conditions are found in accessible areas that could indicate the presence of, or could result in, flaws or degradation in such inaccessible areas. IWE-1240 requires augmented examinations (Examination Category E-C) of containment surface areas subject to accelerated degradation and aging. A VT-1 visual examination is performed for areas accessible from both sides, and volumetric (ultrasonic thickness measurement) examination is performed for areas accessible from only one side.

The requirements of ASME Code Section XI, Subsection IWE and 10 CFR 50.55a are supplemented to perform surface examination (or other applicable technique) in addition to visual examinations, to detect cracking in steel, SS, and dissimilar metal weld pressure-retaining components that are subject to cyclic loading but have no CLB fatigue analysis (i.e., components covered by SRP-SLR Table 3.5-1, items 27 and 40, and corresponding GALL-SLR items; as applicable to the plant). Where feasible, appropriate Appendix J leak rate tests ( GALL-SLR Report AMP XI.S4) capable of detection of cracking may be performed or credited in lieu of the supplemental surface examination; the type of leak test determined to be appropriate is identified with the basis for components for which this option is used.

The requirements of ASME Code Section XI, Subsection IWE and 10 CFR 50.55a are further supplemented to require a one-time volumetric examination of metal shell or liner surfaces that are inaccessible from one side, only if triggered by plant-specific OE. The trigger for this supplemental examination is plant-specific occurrence or recurrence of measurable metal shell or liner corrosion (base metal material loss exceeding 10 percent of nominal plate thickness) initiated on the inaccessible side or areas, identified since the date of issuance of the first renewed license. This supplemental volumetric examination consists of a sample of one-foot square locations that include both randomly-selected and focused areas most likely to experience degradation based on OE and/or other relevant considerations such as environment. Any identified degradation is addressed in accordance with the applicable provisions of the AMP. The sample size, locations, and any needed scope expansion (based on findings) for this one-time set of volumetric examinations should be determined on a plant-specific basis to demonstrate statistically with 95 percent confidence that 95 percent of the accessible portion of the containment liner is not experiencing corrosion degradation with greater than 10 percent loss of nominal thickness. Guidance provided in EPRI TR–107514 may be used for sampling considerations.

5. Monitoring and Trending: With the exception of inaccessible areas, all surfaces are monitored by virtue of the examination requirements on a scheduled basis.

IWE-2420 specifies that:

1. The sequence of component examinations established during the first inspection interval shall be repeated during successive intervals, to the extent practical.
2. When examination results require evaluation of flaws or areas of degradation in accordance with IWE-3000, and the component is acceptable for continued service, the areas containing such flaws or areas of degradation shall be reexamined during the next inspection period listed in the schedule of the inspection program of IWE-2411 or IWE-2412, in accordance with Table IWE-2500-1, Examination Category E-C.
3. When the reexaminations required by IWE-2420(b) reveal that the flaws or areas of degradation remain essentially unchanged for the next inspection period, these areas no longer require augmented examination in accordance with Table IWE-2500-1 and the regular inspection schedule is continued.

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IWE-3120 requires examination results to be compared with recorded results of prior inservice examinations and evaluated for acceptance.

Applicants for subsequent license renewal (SLR) for plants with BWR Mark I containment should augment IWE monitoring and trending requirements to address inaccessible areas of the drywell. The applicant should consider the following recommended actions based on plant-specific design and OE.

1. Develop a corrosion rate that can be inferred from past ultrasonic testing (UT) examinations or establish a corrosion rate using representative samples in similar operating conditions, materials, and environments. If degradation has occurred, provide a technical basis using the developed or established corrosion rate to demonstrate that the drywell shell will have sufficient wall thickness to perform its intended function through the subsequent period of extended operation.
2. Demonstrate that UT measurements performed in response to US NRC Generic Letter (GL) 87-05, “Request for Additional Information Assessment of Licensee Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells” did not show degradation inconsistent with the developed or established corrosion rate.

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6. Acceptance Criteria: IWE-3000 provides acceptance standards for components of steel containments and liners of concrete containments. IWE-3410 refers to criteria to evaluate the acceptability of the containment components for service following the preservice examination and each inservice examination. Most of the acceptance standards rely on visual examinations. Areas identified with damage or degradation that exceed acceptance standards require an engineering evaluation or require correction by repair or replacement. For some examinations, such as augmented examinations, numerical values are specified for the acceptance standards. For the containment steel shell or liner, material loss locally exceeding 10 percent of the nominal containment wall thickness or material loss that is projected to locally exceed 10 percent of the nominal containment wall thickness before the next examination are documented. Such areas are corrected by repair or replacement in accordance with IWE-3122 or accepted by engineering evaluation. Cracking of steel, SS, and dissimilar metal weld pressure-retaining components that are subject to cyclic loading but have no CLB fatigue analysis (i.e., components covered by SRP-SLR Table 3.5-1, items 27 & 40, and corresponding GALL-SLR items; as applicable) is corrected by repair or replacement or accepted by engineering evaluation.

7. Corrective Actions: Results that do not meet the acceptance criteria are addressed in the applicant’s corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR 50, Appendix B, QA program to fulfill the corrective actions element of this AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

Subsection IWE states that components whose examination results indicate flaws or areas of degradation that do not meet the acceptance standards listed in IWE-3500 are acceptable if an engineering evaluation indicates that the flaw or area of degradation is nonstructural in nature or has no effect on the structural integrity of the containment. Components that do not meet the acceptance standards are subject to additional examination requirements, and the components are repaired or replaced to the extent necessary to meet the acceptance standards of IWE-3000. For repair of components within the scope of Subsection IWE, IWE-3124 states that repairs and re-examinations are to comply with IWA-4000. IWA-4000 provides repair specifications for pressure retaining components, including metal containments and metallic liners of concrete containments.

For BWR Mark I steel containments, if moisture has been detected or suspected in the inaccessible area on the exterior of the containment drywell shell or the source of moisture cannot be determined subsequent to root cause analysis, then:

1. Include in the scope of SLR any components that are identified as a source of moisture, if applicable, such as the refueling seal or cracks in the SS liners of the refueling cavity pool walls, and perform an aging management review.
2. Pursuant to Subsection IWE-1240, identify in the inspection program affected drywell surfaces requiring augmented examination for the subsequent period of extended operation in accordance with Table IWE-2500-1, Examination Category E-C.
3. Conduct augmented inspections of the identified drywell surfaces using examination methods that are in accordance with Subsection IWE-2500.
4. Demonstrate, through use of augmented inspections performed in accordance with Subsection IWE, that corrosion is not occurring or that corrosion is progressing so slowly that the age-related degradation will not jeopardize the intended function of the drywell shell through the subsequent period of extended operation.

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8. Confirmation Process: The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

When areas of degradation are identified, an evaluation is performed to determine whether repair or replacement is necessary. If the evaluation determines that repair or replacement is necessary, Subsection IWE specifies confirmation that appropriate corrective actions have been completed and are effective. Subsection IWE states that repairs and re-examinations are to comply with the requirements of IWA-4000. Re-examinations are conducted in accordance with the requirements of IWA-2200, and the recorded results are to demonstrate that the repair meets the acceptance standards set forth in IWE-3500.

9. Administrative Controls: Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR 50, Appendix B, associated with managing the effects of aging. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

IWA-6000 provides specifications for the preparation, submittal, and retention of records and reports.

10. Operating Experience: ASME Code Section XI, Subsection IWE, was incorporated into 10 CFR 50.55a in 1996. Prior to this time, OE pertaining to degradation of steel components of containment was gained through the inspections required by 10 CFR Part 50, Appendix J and adhoc inspections conducted by licensees and the US NRC. US NRC Information Notice (IN) 86-99, IN 88-82, IN 89-79, IN 2004-09, IN 2010-12, NUREG–1522, and NUREG/CR-7111 described occurrences of corrosion in steel containment shells, containment liners, and tori. US NRC GL 87-05 addressed the potential for corrosion of BWR Mark I steel drywells in the “sand pocket region.” IN 2011-15 described occurrences of corrosion in BWR Mark I steel containments, both inside the suppression chamber (torus) and outside the drywell. IN 2014-07 described OE concerning degradation of floor weld leak-chase channel systems of the steel containment shell and concrete containment steel liner that could affect leak tightness and aging management of containment structures.

US NRC IN 97-10 identified specific locations where concrete containments are susceptible to liner plate corrosion; IN 92-20 described instances of two-ply containment bellows cracking for which leak rate testing was inadequate for detection, resulting in loss of leak tightness. Based on occurrences of transgranular stress corrosion cracking, NUREG–1611 (Tables 1 and 2) recommends augmented examination on the surfaces of two-ply bellow bodies using qualified enhanced techniques so that cracking can be detected. Other OE indicates that foreign objects embedded in concrete have caused through-wall corrosion of the liner plate at a few plants with reinforced concrete containments. US NRC Technical Report, “Containment Liner Corrosion Operating Experience Summary” dated August 2, 2011, summarizes the industry OE related to containment liner corrosion and containment liner bulges. Some examples of OE related to liner bulges are noted in NUREG–1522 and Enclosure 2 to NRC Inspection Progress Report 05000302/2011009 dated May 12, 2011.

US NRC IN 2006-01 described through-wall cracking and its probable cause in the torus of a BWR Mark I containment. The cracking was identified by the licensee in the heat-affected zone at the high-pressure coolant injection (HPCI) turbine exhaust pipe torus penetration. The licensee concluded that the cracking was most likely initiated by cyclic loading due to condensation oscillation during HPCI operation. These condensation oscillations induced on the torus shell may have been excessive due to a lack of an HPCI turbine exhaust pipe sparger that many licensees have installed.

The program is to consider the liner plate and containment shell corrosion and cracking concerns described in these generic communications and technical report. Implementation of the ISI requirements of Subsection IWE, in accordance with 10 CFR 50.55a, augmented to consider OE, and as recommended in LR-ISG-2006-01, is a necessary element of aging management for steel components of steel and concrete containments through the subsequent period of extended operation.

Degradation of threaded bolting and fasteners in closures for the reactor coolant pressure boundary has occurred from boric acid corrosion, stress corrosion cracking (SCC), and fatigue loading (US NRC Inspection and Evaluation Bulletin (IEB) 82-02, US NRC GL 91-17).

SCC has occurred in high strength bolts used for nuclear steam supply system component supports (EPRI NP-5769(Archived)).

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry OE including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

References

10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR Part 50, Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.55a, “Codes and Standards.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWA, General Requirements.” New York, New York: The American Society of Mechanical Engineers. 2008.

_____. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWE, Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants.” New York, New York: The American Society of Mechanical Engineers. 2008.

_____. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWL, Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants.” New York, New York: The American Society of Mechanical Engineers. 2008.

EPRI. EPRI NP-5067(Archived), “Good Bolting Practices, A Reference Manual for Nuclear Power Plant Maintenance Personnel.” Volume 1: Large Bolt Manual, 1987; Volume 2: Small Bolts and Threaded Fasteners. Palo Alto, California: Electric Power Research Institute. 1990.

_____. EPRI NP-5769(Archived), “Degradation and Failure of Bolting in Nuclear Power Plants.” Volumes 1 and 2. Palo Alto, California: Electric Power Research Institute. April 1988.

_____. EPRI TR–104213(Archived), “Bolted Joint Maintenance & Application Guide.” Palo Alto, California: Electric Power Research Institute. December 1995.

_____. EPRI TR–107514, “Age-Related Degradation Inspection Method and Demonstration.” In Behalf of Calvert Cliffs Nuclear Power Plant License Renewal Application. Palo Alto, California: Electric Power Research Institute. April 1998.

US NRC Generic Letter 87-05, “Request for Additional Information Assessment of Licensee Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells.” Agencywide Documents Access and Management System (ADAMS) Accession No. ML031140335. Washington, DC: U.S. Nuclear Regulatory Commission. March 1987.

_____. Generic Letter 91-17, “Generic Safety Issue 79, Bolting Degradation or Failure in Nuclear Power Plants.” ADAMS Accession No. ML0311140534. Washington, DC: U.S. Nuclear Regulatory Commission. October 1991.

_____. IE Bulletin 82-02, “Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants.” ADAMS Accession No. ML03120720. Washington, DC: U.S. Nuclear Regulatory Commission. June 1982.

_____. Information Notice 86-99, “Degradation of Steel Containments.” ADAMS Accession Nos. ML031250248, ML 031250234. Washington, DC: U.S. Nuclear Regulatory Commission, December 8, 1986. Supplement 1 February 1991.

_____. Information Notice 88-82, “Torus Shells with Corrosion and Degraded Coatings in BWR Containments.” ADAMS Accession Nos. ML031150069, ML082910476. Washington, DC: U.S. Nuclear Regulatory Commission. October 1988. Supplement 1 May 1989.

_____. Information Notice 89-79, “Degraded Coatings and Corrosion of Steel Containment Vessels.” ADAMS Accession No. ML031190089. Washington, DC: U.S. Nuclear Regulatory Commission. December 1989. Supplement 1 June 1989.

_____. Information Notice 92-20, “Inadequate Local Leak Rate Testing.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1992.

_____. Information Notice 97-10, “Liner Plate Corrosion in Concrete Containment.” ADAMS Accession No. ML031050365. Washington, DC: U.S. Nuclear Regulatory Commission. March 1997.

_____. Information Notice 2004-09, “Corrosion of Steel Containment and Containment Liner.” ADAMS Accession No. ML041170030. Washington DC: U.S. Nuclear Regulatory Commission. April 2004.

_____. Information Notice 2006-01, “Torus Cracking in a BWR Mark I Containment.” ADAMS Accession No. ML053060311. Washington, DC: U.S. Nuclear Regulatory Commission. January 2006.

_____. Information Notice 2010-12, “Containment Liner Corrosion.” Washington, DC: U.S. Nuclear Regulatory Commission. June 2010.

_____. Information Notice 2011-15, “Steel Containment Degradation and Associated License Renewal Aging Management Issues.” ADAMS Accession No. ML111460369. Washington, DC: U.S. Nuclear Regulatory Commission. August 2011.

_____. Information Notice 2014-07, “Degradation of leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner.” ADAMS Accession No. ML14070A114. Washington, DC: U.S. Nuclear Regulatory Commission. May 2014.

_____. Inspection Report 05000302/2011009, Crystal River Nuclear Plant – Steam Generator Replacement Inspection Progress Report. ADAMS Accession No. ML111330350. Washington, DC: U.S. Nuclear Regulatory Commission. May 12, 2011.

_____. NUREG–1522, “Assessment of Inservice Conditions of Safety-Related Nuclear Plant Structures.” ADAMS Accession No. ML06510407. Washington, DC: U.S. Nuclear Regulatory Commission. June 1995.

_____. NUREG–1611, “Aging Management of Nuclear Power Plant Containments for License Renewal.” ADAMS Accession No. ML071650341. Washington, DC: U.S. Nuclear Regulatory Commission. September 1997.

_____. NUREG/CR–7111, “A Summary of Aging Effects and Their Management in Reactor Spent Fuel Pools, Refueling Cavities, Tori, and Safety-Related Concrete Structures.” ADAMS Accession No. ML12047A184. Washington, DC: U.S. Nuclear Regulatory Commission. January 2012.

_____. Staff Position and Rationale for the Final License Renewal Interim Staff Guidance LR-ISG-2006-01, “Plant-Specific Aging Management Program for Inaccessible Areas of Boiling Water Reactor (BWR) Mark I Steel Containments Drywell Shell.” ADAMS Accession No. ML063210074. Washington, DC: U.S. Nuclear Regulatory Commission. November 2006.

_____. Technical Report, “Containment Liner Corrosion Operating Experience Summary.” ADAMS Accession No. ML112070867. Revision 1. Washington, DC: U.S. Nuclear Regulatory Commission. August 2011.

RCSC. “Specification for Structural Joints Using High-Strength Bolts.” Chicago, Illinois: Research Council on Structural Connections. August 2014.

American Society of Mechanical Engineers

Aging Management Program

License Renewal Interim Staff Guidance

Boiling Water Reactor

Operating Experience

United States of America

Nuclear Regulatory Commission - The US federal government agency that regulates nuclear power plants

Current Licensing Basis - 10 CFR 54.3 under references

Electric Power Research Institute

Subsequent License Renewal - The extension of the original license from 40 years to 60 years is known as License Renewal. Extensions of the license beyond 60 years are known as Subsequent License Renewal.

Ultrasonic Test - A nondestructive test method using ultrasonic waves to determine the thickness of a component and to detect cracks or other degradation.

Quality Assurance

US NRC regulatory guide

Stress Corrosion Cracking

Agency wide Documents Access and Management System - The official recordkeeping system, through which the U.S. Nuclear Regulatory Commission (US NRC) provides access to the "libraries" or collections of publicly available documents