Becht Nuclear Services Newsletter Spring...

2009 NJ BusiNess of the Year

IntroductionThis article is written based on experience with ASME

N-Stamp Surveys using a Demonstration Component in the last 20 years, and particularly in the last 5 years. This article presents observations that have resulted in successful ASME N-Stamp surveys. The article is not meant to endorse or promote any particular approach. Reference to Section III documents are to the 2010 Edition, 2011 Addenda. Reference to NQA-1 is to the 2008 Edition and 2009 Addenda.

Abbreviations: CH is Certificate Holder; DC is demonstration component; dDS is demonstration Design Specification; dDR is demonstration Design Report; ANI and ANIS are Authorized Nuclear Inspector and ANI Supervisor, respectively.

Appendix XXIII Specialized Professional Engineer Requirements. ASME Section III requires certification of Owner Design Specifications and Certificate Holder Design Reports. Such certification is required to be done by one or more “Specialized Professional Engineers” qualified in accordance with Section III Division 1 Appendix XXIII. NOTE: Although that is labeled as an Appendix for Division 1, Divisions 2 and 3 are included in the Appendix. Currently, the “entry qualification” for being an Appendix XXIII Specialized Professional Engineer is being a Registered Professional Engineer (RPE) in one of the States of the US or in one of the Provinces of Canada. This is expected to be broadened beyond the USA and Canada shortly, by issue of a Code Case that allows persons registered in accordance with the International Engineers Mobility Forum. Nonetheless, to perform certification of Owner’s Design Specifications and Certificate Holder’s Design Reports, the requirements of Appendix XXIII remain.

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Becht Nuclear Services www.BechtNS.comSpring 2013Newsletter

ASME Section III Nuclear RPE Requirements and Demonstration Design Reports

You may be wondering what this is all about. It is about assuring a safe lift by having someone other than the lift contractor review engineered lifts. As result of many crane related accident

investigations it is evident that in most cases a single error does not result in a catastrophic accident. More often it is a combination of errors that come together at one time or in series. In every case, initially, the focus is on crane operator error and although that is sometimes true, it is usually not the case. We see equipment, rigging or ground support failure as the root cause in most cases.

In order to achieve safe lift and prevent accidents, Becht Engineering’s Heavy Lift Division routinely reviews heavy and super lift plans for accuracy and code compliance. Re-views have included lifts as high as 1800-ton with a 5,000 ton crane.

A thorough review can include structural analysis of the lift structure, host structure, and or the ground bearing ca-pacities as they relate to the reaction produced by the crane or lifting structure. At a minimum, the crane assembly process/

Heavy Lift; Cold Eyes Review and Lift Monitoring

In this Newsletter:• ASME Section III Nuclear RPE Requirements• Heavy Lift; Cold Eyes Review and Lift Monitoring• Fitness-for-Service Evaluation of Bulged Vessel• Proposed Enhancements to SRP Sections 3.7 and 3.8• Nuclear Industry Response to the 2011 Tohoku Earthquake and Tsunami in Japan

2009 NJ BusiNess of the Year2

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The purpose of Fitness-for-Service (FFS) is to evaluate the integrity of in-service pressure equipment given a certain degraded condition and a rate for future damage. A degraded condition can take forms other than

corrosion; for example, sometimes a vessel can experience a large accidental permanent distortion. When such a large distortion occurs it is important to first inspect the vessel to make sure no crack occurred during the deformation event. If it is found that the surface is defect-free then the next step is to determine if the vessel can operate in the deformed shape.

One example of a non-corrosion related degraded condition is a bulged tank. Figure 1 shows an example of a tank that bulged due to an over-pressure event. This tank was modeled using the FEA program Abaqus, the general bulged shape in the model can be seen in Figure 2. An elastic-plastic analysis was performed with the mesh shown in Figure 3. The global model used shell elements to save computational time while the sub-model on the critical nozzle was created using 3D brick elements. This analysis indicated that the vessel was qualified for continued operation, even with an additional future corrosion allowance.

In this case, the fitness-for-service analysis resulted in significant schedule and cost savings by establishing the adequacy of a vessel that otherwise would be replaced.

FFS Evaluation of Bulged Vessel

Figure 3Mesh in Abaqus,

including Sub-Model

Figure 2Abaqus Model

Figure 1 Bulged Shape of the Vessel (rotated 90º off vertical)

The US Nuclear Regulatory Commission (NRC) provides Standard Review Plans (SRPs) to aid licensees and applicants of licenses for Nuclear Power Plants in developing and interpreting design criteria for plants.

The SRP, published as NUREG-0800, consists of 19 Chapters covering a wide range of subjects relevant to Nuclear Power Plant design. Sections 3.7 and 3.8 of the SRP provide acceptance criteria for seismic analysis, and design of containment and other Seismic Category I structures and foundations. Sections 3.7 and 3.8 were last revised in 2007. Since then the NRC and license applicants have identified a number of lessons learned during license application reviews. On April 10, 2013, the NRC held a public meeting to present proposed enhancements to the SRPs. A summary of the proposed enhancements is provided in this article.

Eleven technical issues were identified that have caused difficulty for license applicants and for NRC staff during the license review process. To address these lessons learned the NRC is revising Sections 3.7 and 3.8 of the SRP, with the new revisions scheduled to be issued in the Summer of 2013. The goal of the revisions is not to change NRC staff positions on technical issues, but rather to clarify and provide enhanced guidance for license applicants and licensees.

The 11 technical issues are addressed by proposing enhanced SRP acceptance criteria and developing technical rationale to support the proposed acceptance criteria. The 11 technical issues identified are shown below with a brief summary of the current status and proposed enhancements:

1. Seismic Uplift in Soil-Structure Interaction (SSI) Analysis• Current Status – SRP 3.7.2 provides qualitative guidance for

studying separation and sliding of soil from sidewalls. It does not provide criteria for evaluating building uplift.

• Proposed Enhancements – the revision will provide for calculating a ground contact ratio and will provide for a minimum ground contact ratio for use with linear analysis. Nonlinear analyses will be addressed on a case-by-case basis.

2. Seismic Stability Evaluation• Current Status - SRP Section 3.8.5 on foundations of Seismic

Category I structures is geared toward conservative methods, which can cause difficulties in meeting current criteria because of conservative assumptions for capacity analysis in generic plant designs.

• Proposed Enhancements – provide for calculating a factor of safety at each time step when performing linear analysis. For nonlinear analysis, develop multiple time histories, increase the input by 10%, and include contact interfaces for modeling

Proposed Enhancements to NRC Standard Review Plan Sections 3.7 and 3.8


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Heavy Lift, Cold Eyes Review(Continued from Page 1)

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Nuclear Industry Response to the 2011 Tohoku Earthquake and Tsunami in Japan

Prompt Early Response by Regulators and Industry - Shortly after the March 11, 2011 magnitude 9.0 earthquake and tsunami struck Japan, and damaged the six nuclear power plants at the Fukushima Dai-Ichi site, the US NRC began to follow the situation. The NRC also established a Task Force to develop lessons learned. The Near-Term Task Force (NTTF) Review Insights from the Fukushima Dai-Ichi Accident were published on July 12, 2011, and issued through a Request for Information under 10 CFR 50.54(f ) on March 12, 2012.

The earthquake and tsunami also struck the three nuclear power plants at the Onagawa site, which is nearer to the epicenter than Fukushima Dai-Ichi, but which was safely shutdown and underwent little damage. The lessons learned from this success are also investigated, but have received less attention than the lessons to be learned from the accident at Fukushima Dai-Ichi.

The Industry Organizes its Actions – In response to the accident, to gather lessons learned, develop action plans, follow implementation, and optimize response to regulation, the nuclear power industry formed a well-structured “Fukushima Response Steering Committee” which combines the expertise from the plants (the Owners Group), EPRI, INPO, and NEI.

A Confluence of Seismic-Related Activity – The NTTF recommendations arrive at a time when other seismic-related activities were underway. First, there was the 1990s seismic retrofit at the older plants under the Seismic Qualification Utility Group (SQUG) program, then the seismic Individual Plant Evaluation for External Events (IPEEE) assessment in the mid to late-1990s. More recently, the general consensus had developed that the seismic input to plants in the central and eastern US (CEUS) may have more high-frequency content than estimated in the past. This was the crux of NRC Generic Issue 199. Then, on August 23, 2011 the North Anna Power Station in Virginia was struck by a magnitude 5.8 earthquake centered approximately 10 miles from the plant. The two reactors automatically shutdown, and the plant was brought to a safe shutdown with little damage. Therefore, for engineers, the NTTF recommendations are at a confluence of these three issues: Fukushima Dai-Ichi, Generic Issue 199, and the North Anna earthquake.

inspection, crane capacities/limitations, rigging configura-tion capacities, operator qualification and compliance with federal, state, and facility owner codes and policies should be analyzed and confirmed. Other items of high importance in-clude analysis of the wind sail area of the load and the effect of these lateral forces on the crane safe working limits, clear-ance of the crane boom or load to obstructions, and clearance of the load to the crane boom or lift structure.

Becht highly recommends a load test of the crane founda-tion on super lifts to simulate the reactions that will occur during the heaviest lift. It is imperative to be aware of all potential safety concerns, such the crane or load swinging too close to power lines.

For heavy lift operations, in addition to lift review, a highly experienced third party heavy lift supervisor can be provided to oversee the crane/lift structure erection for safety and compliance with manufacture’s recommendations. Note that it is important to include the crane erection in this over-sight process, as major accidents have occurred during erec-tion, as well as disassembly. As the process continues, they confirm the dimensions called out on the plan and check that all rigging is certified and applied according to the ap-proved plan. They take an active role in pre-lift and safety meetings. Finally, the supervisor is present during the lift to assure the plan is carefully followed and no deviation from the plan is allowed.


Recommendations - The recommendations of the NTTF are subdivided into three tiers. The Tier 1 recommendations are to be started without delay and are the recommendations discussed herein. The following two Tier 1 recommendations focus on seismic re-assessment of US nuclear power plants:

• Recommendation 2.1 - Seismic and flood hazard reevaluations

• Recommendation 2.3 - Seismic and flood walkdowns

Recommendation 2.3 First - Recommendation 2.3 was implemented first. The guidelines for implementation of Recommendation 2.3 were developed by industry as Seismic Walkdown Guidance: For Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic.

Following reviews, comments, and updates, the Guidance was published as EPRI report 1025286 in June 2012, and approved by the NRC. Implementation of this Recommendation 2.3 was completed first, ahead of Recommendation 2.1, and completed in late 2012. The work completed consists of several activities, including review of the 1990s IPEEE, and sample walkdowns. At writing of this report, January 2013; the Recommendation 2.3 reports have been submitted by each plant to the NRC.

Longer Term Recommendation 2.1 - Recommendation 2.1 is an exhaustive re-evaluation of the seismic hazard of each plant and an identification of plant-specific vulnerabilities. The implementation guide for this Recommendation 2.1 is contained in a document referred to as the SPID which stands for Guidance for Screening and Prioritization Implementation Details. The SPID contains a flowchart to help plants determine the level of re-evaluation required. For most plants, the following process will address Recommendation 2.1:

• Development of site-specific probabilistic seismic hazard assessments (PSHA), using state-of-the-art techniques for defining the seismic source terms, attenuation models, and site amplification factors

• Development of site-specific seismic ground motions response spectra (GMRS).

• Comparison of the design basis safe shutdown earthquake (SSE) to the new GMRS to determine the extent and method of re-evaluation.

• In many cases a seismic probabilistic risk analysis (SPRA) will be necessary, as outlined in the following bullets. But the above bullets are requested to be completed by September 2013.

• Development of the systems risk analysis, in the form of event trees and fault trees for the seismic event, as it challenges the reactor and the spent fuel pool (ASME/NS SA-2009, and NRC RG 1.200)

• Calculation of margins and fragilities for structures, systems and components (SSCs) called-out through the system risk analysis

• Convolve the PSHA (the probability of earthquakes of different magnitudes), the system risk analysis, and the fragilities analysis to obtain a core damage frequency and a probability of large early release (i.e. a release before mitigative actions are implemented)

• For CEUS plants, the complete effort is requested to be completed by 2016 or 2018 depending on priority

FLEX Component - The broader lessons learned from the Fukushima Dai-Ichi go beyond earthquake and tsunami response. They are being viewed as response to an extended station blackout, when little support can be expected from the surrounding devastated communities. This broader concern is being addressed in the form of procurement and deployment of Diverse and Flexible Coping Strategies (referred to as FLEX). These are primarily portable equipment (emergency generators, pumps, hoses, etc.) and hook-ups to mitigate an extended loss of offsite power, and a loss of emergency power (station blackout). With this equipment comes the necessary training for its operation under difficult circumstances, which is also being implemented.

Summary. The events of the Great “3/11” Earthquake and Tsunami in Japan continue to significantly affect the nuclear industry worldwide. Regulators are reviewing the status of their plants, particularly concerning their vulnerability to beyond-design-basis events such as the magnitude 9 earthquake and the resulting tsunami. There are many lessons-learned activities, most notable of which are seismic and flooding hazards reevaluations and corresponding walkdowns at operating plants. Broader lessons learned go beyond earthquake and tsunami response, and they focus on mitigation after extended station blackout with coping strategies for broad aspects of an accident occurrence, regardless of cause.

Nuclear Industry Response (Continued from Page 3)


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Proposed Enhancements (Continued from Page 2)

sliding and uplift. Achieve zero or minimal sliding and no overturning.

3. Seismic Category II/I Analysis• Current Status - SRP Acceptance Criterion 3.7.2.8.C states

that “The non-Category I structure will be analyzed and designed to prevent its failure under SSE conditions, such that the margin of safety is equivalent to that of Category I structures”.

• Proposed Enhancements – remove the requirement that the margin of safety for the non-Category I structure is equivalent to that of Category I structures and instead enforce a criterion based on no physical interaction non-Category I structures on Category I structures.

4. Seismic Soil Pressure on Embedded Walls• Current Status – SRP Section 3.8.4 provides two acceptable

methods of calculating lateral soil pressures on embedded walls. i. Lateral pressure equal to the sum of the static earth pressure plus the dynamic earth pressure calculated in accordance with ASCE 4-98, Section 3.5.3.2. ii. Lateral earth pressure equal to the passive pressure.• Proposed Enhancements – seismic soil pressures can vary

substantially depending on many factors and the uncertainties need to be addressed in soil pressure calculations. New acceptance criteria will state that the dynamic lateral soil pressures are to be evaluated for the governing of the following three cases:

i. Sum of the static and dynamic pressures calculated in accordance with ASCE 4-98. ii. Sum of the static and dynamic pressures calculated using an embedded soil-structure interaction finite element model. iii. Fraction of the passive earth pressure that it effectively mobilized.

5. Seismic Wave Incoherency Effects• Current Status – SRP Section 3.7.2 contains a brief discussion

of wave incoherency effects, but no review guidance is provided regarding acceptable response reduction in the analysis.

• Proposed Enhancements – comparisons between coherent and incoherent seismic demands should be shown. Specific guidance is provided for the maximum acceptable reduction in in-structure response spectra as a function of frequency.

6. The Effect of Concrete Cracking on the Seismic Analysis of Structures

• Current Status – SRP Section 3.7.2 provides some guidance on addressing the effects of parameter variations on the development of floor response spectra and states that the effects of potential concrete cracking should be specifically addressed.

• Proposed Enhancements – concrete cracking affects the stiffness and damping in the structural models and specific guidance will be provided for treating these effects.

7. Differential Settlement and Construction Sequence Considerations in Foundation Design

• Current Status – SRP Section 3.8.5 on foundations of Seismic Category I structures provides for a case-by-case review of design and analysis procedures for differential settlements, construction sequence, mat flexibility, and settlement monitoring programs.

• Proposed Enhancements – the revision of SRP 3.8.5 will i. Provide for identification of loads due to construction sequence and differential settlement. ii. Provide specific guidance on modeling of soil stiffness, and implementation of differential settlement and construction sequence issues in the Standard Plant Design. iii. Provide guidance on the interface between the Design Certification (DC) and the Combined Operating License Application (COLA), that is, verification of the Standard Design in the COLA. iv. Provide guidance on short-term and long-term settlement monitoring programs.

8. Artificial Time History Development• Current Status – SRP Section 3.7.1 states that artificial time

histories should be based on seed recorded earthquakes. Option 1, Approach 2 of SRP 3.7.1 allows for use of a single set of time histories with spectral matching at only 5% damping as long as the spectrum does not exceed 130% of the target spectrum at any point, or if an acceptable power spectrum density check (PSD) is performed. Limited guidance is provided for nonlinear time history analysis.

• Proposed Enhancements i. If seeds are based on recorded ground motions, the response spectra of the seed should be similar in shape to the target spectra ii. If target spectra include multiple characteristic events, multiple time histories may be needed. iii. Option 1, Approach 2 will now include satisfaction of both the 1.3 factor and the PSD check. iv. If multiple linear analyses are performed, guidance will be provided for the number of time histories needed and the use of averaging or enveloping of responses. v. The use of artificial time histories based on random generation routines (i.e. not based on recorded ground motions) will be allowed. vi. Additional guidance will be provided for nonlinear analyses.

9. Considerations for the Seismic Design Basis• Current Status – SRP 3.7.2 does not address the relation

between the Certified Seismic Design Response Spectrum (CSDRS) and the generic site profiles.

• Proposed Enhancements - The CSDRS should be shown to be consistent with generic site profiles for assumed sites to ensure that the soil profiles will be able to propagate the CSDRS motions into the structural response. CSDRS are appropriate

Continued on Page 6


Proposed Enhancements (Continued from Page 5)

With the increase in new construction plans worldwide, the need for more Appendix XXIII qualified RPEs has become evident. ASME Survey Teams are focusing on qualifications offered to them during surveys to ensure that qualifications are being handled adequately. An N-Type Certificate Holder who constructs a Section III component for use in a nuclear facility is the entity responsible for ensuring that the Design Report for that component has been certified by an adequately qualified RPE in accordance with Appendix XXIII. Similarly, the Owner of a nuclear facility is responsible to ensure that the Design Specifications for nuclear components to be installed at the Owner’s facility are certified by an adequately qualified RPE in accordance with Appendix XXIII. Mandatory Appendix XXIII provides various methods for an RPE to demonstrate the appropriate Appendix XXIII qualifications. Appendix XXIII

contains two non-mandatory guides providing suggested portions of Section III that should be used to provide guidance to the RPE for preparing qualification documentation.

Need for Explicit Documentation. Whatever Appendix XXIII methods an RPE uses to demonstrate qualification, the methods must be documented, including the state(s) or province(s) of certification, the expiration dates of RPE registration, a copy of the registration document, e.g. PE card, proof of continuing education records appropriate to the “specialty field” of the RPE. In today’s certification environment, the more directly applicable documentation of the RPE’s Section III Specialty Field, the better.

Periodic Updates of Qualifications (Maintenance of Qualification). It is expected that once qualified, the RPE maintains records showing that years following original

Continued on Page 8

Figure 1Example Demonstration Component 1. Internal Pressure (Design, Service, Test)2. Deadweight and seismic 3. External Pressure (Design, Service) as desired beyond NB, NC, ND-3133 Requirements4. Nozzle Restrained Free-End Displacement Load(s)5. Simple heat-up/cool-down transient such as shown below

ASME (Continued from Page 1)

for postulated site profiles if there is a reasonable distribution of the site profile column frequencies over the amplified portion of the CSDRS.

10. Issues with the SASSI Subtraction Method• Current Status – SRP Section 3.7.2 provides generic guidance

for SSI analysis using the “substructure approach”.• Proposed Enhancements – SSI analysis of embedded structures

using the subtraction method (SM) in the SSI analysis program SASSI can result in erroneous and unconservative seismic response compared to the SASSI direct method (DM). The revised SRP will include a brief description of the SM and the DM and specific guidance will be provided for the finite element mesh of the “excavated volume”. The DM should be used to the extent possible. Specific guidance will be provided

for implementation of the modified subtraction method (MSM).

11. Guidance for Assessing Spent Fuel Rack Integrity• Current Status – SRP Section 3.8.4 does not provide specific

acceptance criteria for spent fuel assemblies. Spent fuel racks are Seismic Category I and identified as having a safety function.

• Proposed Enhancements i. Incorporate recommendations of NRC spent fuel working group (SFWG) for fuel assemblies. ii. Specify that spent fuel racks be treated as safety related for QA and condition monitoring. iii. Update language in SRP Section 3.8.4, Appendix D to include new plants.


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BECHt ENgINEERINg StAFF IS ACtIvE IN NAtIoNAL CodES ANd StANdARdS CoMMIttEES

ASME CoMMIttEESBoard of Governors - Current & Former MembersBoard on New Development - Current ChairBoard on Hearings and Appeals - Former ChairBoard on Nuclear Codes and Standards

- Current Member, Former ChairBoard on Pressure Technology Codes and Standards

- Former Chair, VP & MembersBoard on International Standards (Inactive Committee)

- Former MemberBoard on Council Operations - Former ChairsCouncil on Standards and Certification

- Current Member, Former Chairs & Sr. VP’s

ASME PoSt CoNStRUCtIoN StANdARdSPost Construction Standards Committee

- Former Chairs & Vice-ChairsPost Construction Executive Committee

- Current Members, Former Chairs & Vice-ChairsSubcommittee on Repair and Testing - Founding Chair,

Current Member, Former Chair & MembersSubcommittee on Inspection Planning

- Current MemberSubcommittee on Flange Joint Assembly

- Current ChairSubcommittee on Flaw Evaluation - Former Member

ASME BoILER ANd PRESSURE vESSEL CodE CoMMIttEES

Committee on Pressure Vessels (BPV VIII) - Current Chair & Members, Former Members

Subcommittee VIII, Pressure Vessels (Inactive Committee) - Former Vice-Chair & Members

Committee on Transport Tanks - Former MemberCommittee on Construction of Nuclear Facility

Components - BPV III - Current Vice-ChairDesign and Analysis Committee (Inactive Committee)

- Former Chairs & MembersSubgroup on Design Analysis (BPV III & VIII)

- Former Chairs & MembersExecutive Committee on Strategy and Project

Management (BPV III) - Current ChairProject Team on Hydrogen Tanks - Current MemberSpecial Committee on Interpretations (BPV III)

- Current ChairSpecial Interpretations Committee (SC-VIII)

- Current MemberSpecial Working Group on High Density Polyethylene

Piping (SC III) - Current MemberSpecial Working Group on High Pressure Vessels,

Section VIII, Div 3 (Inactive Committee) - Former Chair

Standards Committee on Qualification of Mechanical Equipment - Current Member

Subcommittee on Accreditation - Former Member-Alternate

Subcommittee on Design - (Inactive Committee) - Former Members

Subgroup on General Requirements (BPV III) - Current & Former Members

Subgroup on High Pressure Vessels (Section VIII, Div 3) - Current Member, Former Chair

Subgroup on Design (BPV VIII) - Current Member, Former Chair & Members

Subgroup on Design SC-VIII (Inactive Committee) - Former Chair & Members

Subgroup on on Design/Analysis (SC-D) (Inactive Committee) - Former Chair & Members

Subgroup on General Requirements (BPV VIII) - Current Member, Former Chair

Subgroup on Heat Transfer Equipment (BPV VIII) - Current Member, Former Chair

Subgroup on Toughness (BPV II & VIII) - Current Member

Subgroup on Component Design - Current MemberSubgroup on Elevated Temperature Design

- Current Member, Former Chair

Task Group on Impulsively Loaded Vessels (SC VIII) - Current Members

Technical Oversight Committee (former Main Committee) - Current & Former Members

Working Group on Piping (BPV III) - Current Members, Former Chairs & Vice-Chair

Working Group on Piping (SG-D) (BPV III) - Current Member

Operation and Maintenance of Nuclear Power Plants – Piping Group - Current Member

Task Group on Buried Pipe - Current MemberWorking Group on Supports (BPV III) - Former MemberWorking Group on Design-By-Analysis (BPV VIII)

- Current MemberSubgroup on External Pressure (BPV II)

- Current Vice-Chair

ASME PIPINg ANd CoMPoNENt CodES ANd StANdARdSB31, Code for Pressure Piping, Executive Committee

- Current & Former MembersB31, Code for Pressure Piping, Mechanical Design

Committee - Current Vice-Chair & Member, Former Vice-Chair

B31, Code for Pressure Piping, Standards Committee - Current Members, Former Chair

B31.3, Process Piping Section Committee - Current Members, Former Chairs

B31.12 Hydrogen Piping and Pipelines Section Committee - Current Members

B31.3 Subgroup on High Pressure Piping - Former ChairB31.3 Subgroup on Design - Current & Former MembersB16 Standards Committee for National & International

Standardization of Valves, Flanges, Fittings,Gaskets - Current Member

B16 Subcommittee on Steel Flanges and Flanged Fittings - Current Member, Former Chair

B16 Subcommittees on fittings, gaskets, copper components and valves - Current Member

ASME/APIASME/API Joint Fitness-for-Service Committee (API

579/ASME FFS-1) - Current Member, Former Vice-Chairs & Members

ASCEWorking Group on Revision of ASCE Standard 4 Seismic

Analysis of Safety-Related Nuclear Structures -Voting MemberWorking Group on Revision of ASCE Standard 43

Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities

-Voting MemberASCE Dynamic Analysis of Nuclear Structures (DANS)

Committee -Current MemberAPICommittee on Risk Based Inspection - Former MemberCommittee on Materials and Corrosion

- Former MemberCommittee on Refining Equipment - Former MemberSubcommittee on Heat Transfer Equipment

- Former ChairSubcommittee on Above Ground Storage Tanks

- Former MemberSubcommittee on Piping - Former Vice-ChairTask Group on Materials for Heavy Wall Vessels

- Former ChairTask Group on API 580 - Former MemberTask Group (934F) for Development of a Guidance

Document for Establishing Minimum Pressurization Temperatures for Heavy Wall Reactors in Hydrogen Service - Current Member

Task Group (934G) for Coking Drums in Delayed Coking Units - Current Member

Task Group (941) for High Temperature Hydrogen Attack - Current Member

Task Group on Shell & Tube Heat Exchangers - Former Member

Task Group on Air Cooled Heat Exchangers - Former Member

API 582 – Welding Guidelines for the Chemical, Oil and Gas Industries - Former Chair

AddItIoNAL CoMMIttEE INvoLvEMENtMTI - Board of Directors - Current MemberMTI – Knowledge Management Project Development

Committee - Current ChairAWS A5 (Filler Metals) Committee - Source for ASME

IIC - Current MemberASTM F-17, Plastic Piping Systems Main Committee

- Current MemberASTM Standards and Certification Board of Directors

- Current Member, Former ChairASTM Committees CO3, C-16, and D-20.23

- Current MemberISA – The International Society for Measurement &

Control, member of SP 93, Sealing Technologies Committee - Former Member

Materials Technology Institute’s Technical Advisory Council Committee - Current Member

Materials Technology Institute, Evaluation of Gaskets Committee - Current Member

US TAG for ISO/TC 153/SC 1 – Valve Design, Manufacturing, Marking and Testing - Current Member

US TAG for ISO/TC 5/SC 10 – Metallic Flanges and Their Joints - Current Member

US TAG for ISO TC 197 – Hydrogen Technologies - Current Member

US TAG for ISO TC 11 - Boilers and Pressure Vessels - Current Member

Joint ASME – IEEE Seismic Qualification Committee - Former Chair

ASME – ASCE Joint Task Group on Buried Piping - Founding & Current Chair

Process/Industry Practices (PIP): Coatings and Insulation - Former Function Team Leader

Process Industry Practices (PIP): Piping - Former Member

Society for Protective Coatings (SSPC): Standards Review Committee - Former Member

US DOE Hydrogen Safety Panel - Current MemberThe Materials Properties Council - Former MemberSection III, SG on General Requirements - Current

MemberSection III, SWG on Polyethylene Pipe (Inactive

Committee) - Former MemberSection III, Working Group on Piping Design (Inactive

Committee) - Former MemberPVRC Committee on Continuing Operation of

Equipment - Former ChairASME Hydrogen Steering Committee (Inactive

Committee) - Former ChairASME High Pressure Systems Committee

- Former Vice-Chair & MemberASME Task Group on Risk Analysis for the Critical

Assets Protection Initiative - Former Chair & MemberASME BPTCS Ad Hoc Task Group on Incorporation of

Risk Based Principles into Pressure Technology Codes and Standards - Former Chair & Member

PVP Design and Analysis Committee - Current ChairPVP Executive Committee - Current MemberWorking Group Duties and Responsibilities (SGGR-

BPVIII) - Current MemberQAI-1 Qualification for Authorized Inspection

- Former ChairSubcommittee on Accreditation of Authorized

Inspection Agencies - Current Member-AlternateNational Commission for the Certification of Crane

Operators (NCCCO) - Vice President, Board of Directors


qualification, that the qualification is maintained. This includes maintenance of a valid RPE license (yearly or otherwise), appropriate continuing education activities, certification of Section III design specifications or design reports, etc. Copies of diplomas, seminar completion certificates, and Continuing Education Unit (CEU) or Professional Development Hour (PDH) certificates for example, are of particular interest. Merely using list of activities without documentation is not typically adequate. It is important that the continuing education be applicable to the RPE’s Section III Specialty Field.

Records Demonstrating Qualification. Appendix XXIII Guide B “Nonmandatory Guidelines for Establishing ASME Code Knowledge” is valuable, and the Tables B1 through B8 can be used to create a documentation outline of an RPE’s working and general knowledge of various aspects of Section III. The RPE need only address the Tables B1 through B8 that bear on the types of DSs and DRs being certified. For example, Tables B1 and B2 would be used for documenting working and general knowledge for certifying Division 1 Design Specifications and Design Reports. Appendix XXIII Guide C “Nonmandatory Guidelines for Demonstrating PE Qualifications” is similarly a useful resource for creating a checklist for a PE’s qualifications, referencing explicit documents that support the items in the checklist. NOTE: Just having a checklist is typically not adequate; the supporting documents are needed and the more detail, the better. A qualified Appendix XXIII RPE must maintain certification, typically documenting maintenance of that qualification every three years.

Approvals and Appointment within Responsible Organization (NQA-1 program). Once an RPE is qualified using the requirements of Appendix XXIII and ready to perform work for an N-Certificate Holder or Owner, the RPE is typically added to the Certificate Holder or Owner’s QA system, such as being added to the CH’s or Owner’s Approved Supplier List. Doing so is likely to require an audit, review or surveillance of the RPE’s qualifications by the CH or the Owner. Auditing is becoming typical.

demonstration ComponentsMost CHs use a demonstration component, although, if the

CH has current Section III work, that work is open to review. A demonstration component is devised to provide a reasonable demonstration of the CH’s capabilities relative to how those capabilities are covered in the CH’s QA Program. The CH’s ANI and ANIS are helpful in the choice of a demonstration component or components so that the DC(s) adequately cover the scope of the CHs capabilities, and then demonstration of software and hardware aspects of the CH’s Program. The demonstration component(s) should demonstrate how design, fabrication, inspection, testing and certification are accomplished by the CH to meet Section III requirements.

The DC may be simple such as that shown in Figure 1. The

CH’s ANI and ANIS typically help define the demonstration component(s). The chosen geometry is typically a small pressure retaining component such as a vessel, e.g. a 16 inch diameter vessel with hemispherical, elliptical or flat heads or a capped-end piping subassembly. The vessel or piping usually includes one or more nozzles. The material is typically that for which the CH’s fabrication procedures are adequately developed, e.g. SA-240 316, SA-312 304, SA-516 Gr 70. The loading includes internal design pressure, service pressure and test pressure, external pressure, external loads such as nozzle restrained free-end displacement reaction loads, seismic loads, and thermal loading. One convenient thermal loading is a slow heat-up and a slow cool-down transient that is simple to evaluate using conservative methods; closed-form or otherwise. The service level load combinations ideally include one Level A, B, C and D load combination. These combinations can be conservatively evaluated, e.g. showing that Levels B, C and D meet design condition allowable stresses, etc. It is also useful to evaluate them, each with their own allowable stresses.

demonstration design SpecificationThe DC needs a demonstration Design Specification (dDS),

as required by Subsubarticle NCA-3250. NCA-3250 defines the required content, and the demonstration component design specification must address all applicable aspects of NCA-3250. The dDS must provide “a complete basis” for construction of the DC. This typically means that geometry, materials, loadings, load combinations, fabrication and inspection details related to design (known as “design interfaces”) are unambiguously defined. It is, however, recognized that the CH may need additional information during construction, and that in a real situation, that information would be documented in the demonstration Design Report (dDR) or in an eventual revision of the dDS. Usually, it is not

Table 1Topics Covered in NQA-1 Requirement 3

Design InputDesign ProcessDesign Analyses Use of Computer Programs Documentation of Design AnalysesDesign Verification Design Reviews Alternate Calculations Qualification TestsChange ControlInterface ControlSoftware Design ControlDocumentation and Records


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necessary to carry the dDS or dDR to this level, since the DC is not taken to completion for the survey, i.e., fabrication is held at a point where internal features such as welds are visible to the ASME Survey Team. The dDS is typically shown to have been certified by an Appendix XXIII qualified RPE, either a real one or an imaginary one, and either with a real PE stamp or with an imaginary one. The key aspect is that whoever that person is, that same person cannot be the RPE to certify the dDR. NOTE: It seems preferred to use a real RPE and the RPE’s actual PE stamp for the dDR.

demonstration design Report and Associated Engineering drawings

The dDR documents satisfaction of the dDS and Section III requirements for the DC. The Section III requirements include those from the appropriate subsection, i.e., Subsection NB for Class 1, NC for Class 2, ND for Class 3, etc. The dDR typically focuses on the design portion of the applicable subsection, e.g., NB-3000. However, the dDR should include mention of the -1000 (Introduction), -2000 (Materials), -4000 (Fabrication), -5000 (Inspection), -6000 (Testing), -7000 (Overpressure) and -8000 (Stamping and Data Reports) articles. The “design interfaces” between -3000 and -4000, -5000, -6000 are of particular importance. For example, -3000 refers to -4000 for fabrication details of nozzles that affect the structural evaluation of the nozzle/shell intersection. As a further more specific example, using the DC depicted in Figure 1, a Class 1 dDR should clearly refer to the design, fabrication and inspection requirements for a corner-welded nozzle in accordance with NB-3352.4(b), NB-4244(b) and NB-5243. The dDR and its associated drawings need to document those design interfaces and address the appropriate requirements in the design evaluation. Engineering

drawings should provide a complete basis for construction, providing adequate information for geometry, material, fabrication, welding, inspection, testing, overpressure protection and stamping supported by the evaluation reported in the dDR. The engineering drawings must also provide the CH adequate basis for generating fabrication drawings and procedures that are supported by the design calculations. NCA-4134.3 “Design Control” and its reference to NQA-1 Requirement 3 “Design Control” are important to demonstrate. The important topics covered in NQA-1 Requirement 3 are listed in Table 1, and these are important to demonstrate, along with the purely technical adequacy of the DC and the dDR. A critical element is the Design Verification; design review versus alternate calculations versus qualification tests. At least one of these design verification methods needs to be demonstrated. All methodology must be clearly documented, such as sources for closed-form solutions, important Code paragraph citation, computer program references, computer IDs used for the calculations, drawings, specifications. If in doubt, document it.

One important consideration is to evaluate the effect of supports on the DC. Subsection NCA requires that the Owner provide for adequate support. NB, NC and ND also require that the CH account for the effects of the supports on the component within the NB, NC, or ND jurisdictional boundaries in NB, NC, ND-1130; see Figure 1. This typically means that the Engineering Drawings provide enough detail of the supports and their connection to the component such that the effects on the component caused by reaction(s) from the attached support(s) can be adequately evaluated by the CH in the dDR. If the CH is demonstrating capabilities for Supports, Subsection NF, then the evaluation of the supports can be effectively included in the dDR or in a separate demonstration document. The important aspect

Figure 2Demonstration Component – Simple Axisymmetric Finite Element Analysis

Simple axisymmetric solid element model of vessel shell, head and nozzle subjected to simple axisymmetric loadings, such as:1. Internal Pressure (Design, Service, Test)2. Deadweight and seismic in deadweight direction3. Seismic, static equivalent symmetric longitudinal g-load4. Nozzle Restrained Free-End Displacement Load(s) (Design, Service); total or partitioned as deadweight, seismic, thermal as desired.5. External Pressure (Design, Service, Test) as desired beyond NB, NC, ND-3133 Requirements6. Simple heat-up/cool-down transient


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is to pay attention to the jurisdictional boundaries specified in NB, NC and ND-1130, and to apply them correctly.

The dDR must conform to the CH’s QA Program, which itself is governed by the eighteen parts of NCA-4134 and their reference to NQA-1. How a CH implements NCA-4134.3 (Design Control) is of critical importance along with how the CH implements the technical design requirements in NB, NC, ND-3000.

NB, NC, Nd-3000, “design”NB, NC, ND-3000 is arranged in the following articles:-3100 General Rules (generally applicable to vessels, pumps,

valves and piping), especially external pressure in NB, NC, ND-3133.

-3200 Design by Analysis (NB and NC and not ND). NOTE: NB-3200 applies in differing ways for vessels, pumps, valves and piping. Vessels make the most extensive use of NB-3200 in that a Class 1 vessel must be evaluated using NB-3200. The other three components – pumps, valves and piping – treat the use of NB-3200 in a “permissive manner.”

-3300 Vessel Design (controls over -3100 and -3200)-3400 Pump Design (controls over -3100 and -3200)-3500 Valve Design (controls over -3100 and -3200)-3600 Piping (controls over -3100 and -3200)Therefore, if the DC is a Class 1 vessel, NB-3100, NB-3200

and NB-3300 are used in the hierarchical order stated in NB-3100 and NB-3300 and as shown above. This is often not evident is dDRs. If the DC is Class 2 piping, NC-3100, NC-3200 and NC-3600 are used, etc.

design Methodology: verification Considerations of Software

Section III neither requires nor prohibits the use of software such as finite element analysis (FEA). However, most nuclear components are evaluated using FEA. Figure 2 provides an example FEA model of a portion of the component shown in Figure 1. Nowadays, Survey Teams seem to “expect” that a DC be evaluated fully or partially with FEA. When FEA is used, then the requirements of NB-3200 apply for Class 1 components, and Appendix XIII for Class 2 components. When FEA, or any other software is used in evaluation of a Section III component, NCA-4134.3 and its reference to NQA-1 require verification of the software. This has become a focused aspect of N-Stamp surveys. The dDR should address monotonic loading strength criteria, cyclic loading criteria (fatigue), toughness and fracture control and buckling. These are often done simply and conservatively.

RPE Certification Stamp on demonstration documents

As mentioned earlier, the dDR is certified by an Appendix XXIII RPE, and it must be a different person than the person who certifies the dDS. Actually stamping the dDR using the qualified RPE’s PE stamp or seal seems to be a preference by the CH’s

ANI, ANIS and the Survey Team. NOTE: As far as ASME is concerned, the only certification by an RPE that matters is that certification done at the time the component is completed, and the design report has reconciled the actual component characteristics with the calculations as stated in NCA-3555(a). In other words, versions of a design report prior to completion may be certified by an RPE, but those certifications are not the certifications that satisfy NCA-3555(a) for the CH’s N-Stamp application. Such certifications might satisfy customer contractual requirements, but they cannot be considered to satisfy Section III requirements.

SummaryThe ASME N-Stamp certification process has evolved and

the focus is on showing that a CH can correctly use Section III within the confines of the CH’s approved QA Program.

The qualifications of RPEs for certification of Section III DSs and DRs require explicit and documented demonstration to Section III Division 1 Mandatory Appendix XXIII.

Currently, only PEs from a State of the US or a Province of Canada may be used. A Code Case is being developed to permit engineers from other countries though use of the International Engineers Mobility Forum.

Detailed, up-to-date documentation providing evidence of qualifications and directly related experience is key. An original qualification is typical with periodic (3 year) updates and updates for renewal of PE licenses.

Demonstration components can be simple and the associated dDS and dDR can be simple including enough geometric features and fabrication processes, loadings and evaluation techniques that demonstrate the CH’s “Code Knowledge” in a simple, but non-trivial manner. The CH’s ANI and ANIS usually provide guidance on the demonstration component.

The DC should be simple and the dDR should be simple, but cover adequately represented loads (beyond internal pressure), load combinations, design conditions, service conditions and test conditions.

The details of supports applicable to the DC should be clearly defined in the dDS and adequately treated in the dDR and Engineering Drawings within the appropriate jurisdictional boundaries of NB, NC, ND-1130.

The demonstration of a CH’s capabilities is more than just “Design.” It is the all-encompassing “Construction” which includes design, materials, fabrication, inspection, testing, overpressure protection and stamping and data reports.

A successful demonstration adequately addresses these aspects, and particular attention to the CH’s QA program for Design Control for the how the dDR is developed and used.

When software is used in design activities, verification in accordance with NCA-4341.3 and its reference to NQA-1 is important.

These are some of the key aspects evident from our experience in Section XXIII RPE and Certificate Holder N-Stamp Surveys in the past 20 years and in the past 5 years, particularly.


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