ML19161A374
| ML19161A374 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/10/2019 |
| From: | C-10 Research & Education Foundation, Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| 50-443-LA-2, ASLBP 17-953-02-LA-BD01, RAS 55032 | |
| Download: ML19161A374 (11) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No. 50-443-LA-2 NEXTERA ENERGY SEABROOK, LLC ASLBP No. 17-953-02-LA-BD01 (Seabrook Station, Unit 1)
Hearing Exhibit Exhibit Number: INT002 Exhibit
Title:
Summary of Testimony of Victor E. Saouma, Ph.D Regarding Scientific Evaluation of NextEra's Aging Management Program for Alkali-Silica Reaction at the Seabrook Nuclear Power Plant (Non-Proprietary Version)
A Introduction A.1 Please state your name and employment.
My name is Victor E. Saouma. I am Professor of Civil Engineering at the University of Colorado in Boulder. I am also the Managing Partner of XElastica, LLC, a consulting firm. And I am Professeur des Universités in France.
A.2 Please identify this document.
This is my testimony regarding my scientific evaluation of NextEras Aging Management Program for Alkali-Silica Reaction at the Seabrook nuclear power plant. My written testimony is submitted in two versions: EXHIBIT INT001 is my complete testimony, and includes some proprietary information. I am also submitting EXHIBIT INT002, which contains the introductory section and a summary of my conclusions. I also plan to submit a redacted version of Exhibit 1 as soon as possible.
A.3 Please describe your professional qualifications to give this testimony.
I am a leading international expert in the field of Alkali-Aggregate Reaction (AAR), which is also known as Alkali Silica Reaction (ASR). I am not aware of any other researcher (other than one in France) who has conducted the same breadth and depth of research on ASR: theoretical, numerical (deterministic/probabilistic, static/dynamic), experimental (material and structural). I have developed what is probably the most widely referenced and copied model for ASR, the Saouma Model. The Saouma Model is used by the Idaho National Laboratory in the Abaqus, Vector3, and Grizzly/Moose computer programs. It is also used by HydroQuebec for dam analysis. And it is used as well in China, Switzerland, and Canada.
I have conducted research for numerous government agencies, including the U.S. Nuclear Regulatory Commission (NRC), the U.S. Army Corps of Engineers, the U.S. Department of the Interiors Bureau of Reclamation, the U.S. Department of Energys Oak Ridge National Laboratory, the National Science Foundation, the Tokyo Electric Power Company (TEPCO), and the Swiss Federal Office for Water Management (dam safety). My research has encompassed material and structural testing, theoretical and computational modeling, fracture mechanics, risk-based numerical assessment of bridges, nuclear containment structures and dams, chloride diffusion, and experimental dynamics. I have written about 100 peer-reviewed articles on these topics, including approximately 30 articles on ASR and the related topics of chloride diffusion, seismic analysis and stochastic analysis. I have also written a book on numerical modeling of ASR, Numerical Modeling of Alkali Aggregate Reaction (CRC Press 2013).
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I have been a consultant (providing expertise in fracture mechanics) to Performance Improvement International (PII) investigating the root cause of Crystal River nuclear containment delamination.
In addition, I serve or have served on numerous scientific organizations, committees, and panels, including current chair of a RILEM (French acronym of International Meeting of Laboratories and Experts of Materials, Construction Systems and Structures) committee on Diagnosis and Prognosis of ASR affected Structures, RILEM TC 259-ISR. And I am the past president of the International Association of Fracture Mechanics for Concrete and Concrete Structures.
A copy of my curriculum vitae is attached to my testimony as EXHIBIT INT003.
A.4 Have you done any research or writing specifically related to ASR at Seabrook?
In 2014, I co-authored a journal article regarding aging management of ASR at Seabrook. The article presented a scholarly assessment of the gap between the reported methodology and the state-of-the-art, based on the limited amount of information that was publicly available at the time (Saouma, V. E., & Hariri-Ardebili, M. A., 2014).
In addition, in 2014, the NRC awarded me a three-year $703,000 contract to provide support for a project entitled "Experimental and Numerical Investigation of Alkali Silica Reaction in Nuclear Reactors." A copy of the grant award is attached as EXHIBIT INT004. As stated at page 4 of the grant award, the impetus for my proposed research stemmed from the apparent challenge confronting the NRC in assessing safety issues pertaining to the Seabrook nuclear power plant which suffers from Alkali Silica Reaction (ASR), and in particular NRCs request that the licensee determines the long term safety of the plant within the framework of [Seabrook Alkali Silica Reaction Issue Technical Team Charter (July 9, 2012)] ML121250588 (2012).
My research ended in December 2017 when I submitted a four-volume final report. A copy of the Final Summary Report is attached as EXHIBIT INT005.
To date, to the best of my knowledge, our study for the NRC is the most comprehensive on the eect of ASR on the shear strength of concrete. Sixteen large specimens were carefully prepared and tested using a unique apparatus designed for shear testing. It was determined that a 0.6% expansion reduces strength by 20%. We found that ASR of a relatively low 0.3%
reduced the resilience of an NCVS subjected to seismic excitation by approximately 20%. We also successfully demonstrated the applicability of a modern probabilistic based static/dynamic nonlinear methodology for evaluating ASR.
In addition, in 2018 I was retained by the C-10 Research and Education Foundation (C-10) to evaluate work done by NextEra, NextEras consultants, and the NRC technical staff regarding the presence of ASR in concrete at the Seabrook nuclear power plant; and the effect of ASR on Page l 2
the integrity of the concrete, including the containment. In the course of my evaluation, I reviewed both public and proprietary documents regarding NextEras investigations. I also applied the insights of my work under the NRC contract described above. C-10 submitted my declaration and report to the NRC Commissioners in support of an emergency petition to further address ASR at Seabrook before re-licensing the reactor. A copy of my declaration is attached as EXHIBIT INT006. A copy of my expert report is attached as EXHIBIT INT007 (PROPRIETARY). A publicly available summary of my expert report is attached as EXHIBIT INT008. And a copy of the Reply Declaration I submitted in support of my expert report is attached as EXHIBIT INT009.
A.5 What documents have you reviewed in preparing your testimony?
I have reviewed NextEras license amendment request (LAR), Seabrook, License Amendment Request 16 Revise Current Licensing Basis to Adopt a Methodology for the Analysis of Seismic Category I Structures with Concrete Affected by Alkali- Silica Reaction dated August 1, 2016 (Letter SBK-L-16071) (ML16216A240) (EXHIBIT INT010), including its subsequent revisions.
The major elements of the LAR and the consultant reports that I have evaluated are as follows:
NextEra Energys Evaluation of the Proposed Change Including Attachment 1 Markup of UFSAR Pages (Proprietary) (Enclosure 1 to Letter SBK-L-16071) (EXHIBIT INT011)
(Proprietary);
MPR-4288, Rev. 0, Seabrook Station: Impact of Alkali-Silica Reaction on Structural Design Evaluations (July 2016) (Non-proprietary version) (ML16216A241) (Enclosure 2 to Letter SBK-L-16071) (EXHIBIT INT012);
SG&H Report 160268-R-01, Rev. 0, Development of ASR Load Factors for Seismic Category I Structures (Including Containment) at Seabrook Station, Seabrook, NH (July 2016)
(ML16216A243) (Enclosure 4 to Letter SBK-L-16071) (EXHIBIT INT013);
MPR-4288, Rev. 0, Seabrook Station: Impact of Alkali-Silica Reaction on Structural Design Evaluations (July 2016) (Proprietary Version) (Enclosure 5 to Letter SBK-L-16071) (EXHIBIT INT014) (Proprietary);
Simpson Gumpertz & Heger, Inc., Evaluation and Design Confirmation of As-Deformed CEB, 150252-CA-02 Revision 0, July 2016. (ML16279A049) (2016) (Enclosure 2 to Letter SBK-L-16153, re: Seabrook Station (Sept. 30, 2016)) (EXHIBIT INT015).
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Revised Seabrook Station License Renewal Application Updated Final Safety Analysis Report Sections A.2.1.31 for Structures Monitoring, A.2.1.31A for Alkali-Silica Reaction and A.2.1.3b for Building Deformation (Enclosure 1 to Letter SBK-L-18072 re: Seabrook Station Revised Structures Monitoring Aging Management Program (May 18, 2018) (Letter SBK-L-18072)) (EXHIBIT INT016);
Revised Seabrook Station License Renewal Application Appendix B Sections B.2.1.31 for Structures Monitoring, B.2.1.31A for Alkali-Silica Reaction and B.2.1.3b for Building Deformation (Enclosure 2 to Letter SBK-18072), (EXHIBIT INT017);
MPR-4153, Revision 3, Seabrook Station-Approach for Determining Through-Thickness Expansion from Alkali-Silica Reaction (Sept. 2017) (Non-proprietary version) (ML16279A050)
(Enclosure 4 to Letter SBK-18072) (EXHIBIT INT018);
) MPR-4273, Rev. 1, Seabrook Station - Implications of Large-Scale Test Program Results on Reinforced Concrete Affected by Alkali-Silica Reaction (July 2016) (Non-proprietary version)
(ML18141A785) (Enclosure 5 to Letter SBK-18072) (EXHIBIT INT019);
MPR-4153, Revision 3, Seabrook Station-Approach for Determining Through-Thickness Expansion from Alkali-Silica Reaction (Sept. 2017) (Proprietary version) (Enclosure 6 to Letter SBK-18072) (EXHIBIT INT020) (Proprietary);
MPR-4273, Rev. 1, Seabrook Station - Implications of Large-Scale Test Program Results on Reinforced Concrete Affected by Alkali-Silica Reaction (March 2018) (Proprietary version)
(Enclosure 7 to Letter SBK-18072) (EXHIBIT INT021) (Proprietary);
Simpson Gumpertz & Heger Document No. 170444-MD-01, Rev. 1, "Methodology for the Analysis of Seismic Category I Structures with Concrete Affected by Alkali-Silica Reaction," for Seabrook Station (Enclosure 3 to Letter SBK-L-18074, re: Seabrook Station, Response to Request for Additional Information Regarding License Amendment Request 1603 (June 7, 2018) (Letter SBK-L-18074)) (EXHIBIT INT022)
Simpson Gumpertz & Heger Document No. 170444-L-003 Rev. 1, Response to RAI-D8- Example Calculation of Rebar Stress For a Section Subjected to Combined Effect of External Axial Moment and Internal ASR (Enclosure 4 to Letter SBK-L-18074) (EXHIBIT INT023);
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In addition, I have reviewed the publicly available and proprietary versions of NRCs Safety Evaluation Related to Amendment No. 159 to Facility Operating License No. NPF-86 (March 11, 2019) (publicly available version at ML18204A291) (EXHIBITS INT024 (public) and INT025 Proprietary).
I have reviewed applicable government and industry standards.
I have also reviewed the Licensing Board decision admitting C-10s contentions, LBP-17-07, 86 N.R.C. 59 (2017).
Finally, I have reviewed a large body of research reports and academic literature regarding the phenomenon of ASR.
A.6 Please describe the purpose of your testimony.
The purpose of my testimony is to provide technical support for C-10s assertion that the large-scale test program, undertaken for NextEra at the Ferguson Structural Engineering Laboratory (FSEL) of the University of Texas, has yielded data that are not representative of the progression of ASR at Seabrook; and that as a result, the proposed monitoring, acceptance criteria, and inspection intervals are not adequate. My testimony will also address C-10s particular concerns regarding the insufficiency of crack width indexing and extensometer deployment for determining the presence of ASR, NextEras misconception of the effects of ASR within a reinforced concrete structure, the need for continuous petrographic sampling and analysis, and the unacceptability of the proposed length of intervals between inspections.
A.7 Why are you providing this testimony?
I am providing my testimony to C-10 pro bono, because I am very concerned, both as a scientist and a citizen, about the inadequacy of the work that has been done on ASR at Seabrook. To address a problem as complex and potentially dangerous as ASR, it is essential to avail oneself of the best possible information and expertise. Therefore, it disturbs me that neither NextEra nor the NRC has sought to apply the current state of knowledge regarding ASR or to obtain independent review of their work. Instead, they have offered assurances of safety to the public that are based on simplistic analyses, erroneous assumptions, and data that are not representative of conditions at Seabrook. These analyses and data were far from adequate to give the NRC technical ground to continue to operate Seabrook during its current license term or to re-license Seabrook for another 30 years (i.e., until 2050).
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B
Background:
ASR B.1 Please describe the phenomenon of Alkali Silica Reaction (ASR)
Alkali Silica Reaction is a chemical reaction in concrete caused by a Ph imbalance. Cement and some aggregates are responsible for the alkalinity, and the silica inside aggregates provides acidity. Under conditions of high relative humidity (at least 80%), ASR results in the formation of a viscous gel (with calcium playing a major role in the viscosity of the gel). The expanding concrete first fills up voids, and then causes the concrete to expand. The kinetics of the reaction (that is the rate of expansion) is a function of time, temperature and concrete relative humidity.
ASR is almost never homogeneously spread over a large structure, because reactive concrete tends to occur in pockets where silica-rich aggregates may have been used. Heterogeneous distribution of ASR (as is the case of Seabrook) is more problematic than homogeneous distribution, because it will cause gradients of expansion (think of the Tower of Pisa with unequal settlement).
ASR progress depends very much on the geological nature of the aggregate and sand. In some cases, we have an early-expansion (such as rhyolitic aggregate), and in others a late-expansion (such as granite). Furthermore, sand will result in a rapid expansion, and aggregates will cause a slower, but larger, future expansion. Hence, it is nearly impossible to duplicate a reactive concrete unless one uses exactly the same concrete mix and ingredients.
If unimpeded, ASR expansion is volumetric and isotropic (i.e., the same amount of expansion occurs in three directions or planes). However, confinement of the concrete will inhibit ASR expansion in those directions and reorient it along the direction of least confinement.
Confinement in Seabrook and other nuclear plants is lateral due to geometry, and vertical due to geometry and weight of the reactor; hence expansion will be mostly out of plane, that is radial.
The ultimate effects of ASR include both expansion and degradation of the concrete mechanical properties. This combination of expansion and degradation affects tensile and shear strengths along with elastic modulus. Tensile strength will control the formation of (undesirable) cracking, and the elastic modulus degradation will result in larger deformation and potential cracking. The decrease in shear strength can compromise the integrity of a containment during an earthquake.
Many tests have shown an increase in structural shear strength in reinforced concrete beams (through the so-called prestressing effect) because of ASR. This is not to be confused with the inherent shear strength of plain concrete material, which is not strengthened by ASR but rather is degraded.
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B.2 What legal or industry standards are applicable to ASR?
ASR is a relatively new, complex and potentially dangerous problem. I am aware of no regulations or industry standards that have been developed to specifically address the presence of ASR and its implication on serviceability and strength.
The federal highway administration (FHWA) has published a number of reports (written by leading experts) addressing this problem and providing a road map on how to deal with ASR using modern tools. For example, see Fournier, B., Berube, M. A., Folliard, K. J., & Thomas, M.
(2010).
The U.S. government has invested significant resources into research on ASR, including my contract, a grant of approximately $7 million to the National Institute of Standards and Technology (NIST), and U.S. Department of Energy (DOE) research through the Oak Ridge National Laboratory. Given this investment, I would have thought that the NRC and DOE would develop similar guidelines for the nuclear industry as the FHWA. This did not happen, and for all practical purposes it was effectively left to NextEra to write their own guidelines through their License Amendment Request.
B.3 Can we treat safety assessment of an existing structure suffering from ASR the same way we designed it?
In evaluating the degree to which ASR threatens compliance with NRC safety standards, it is important to bear in mind that analytical considerations related to the design of new structures are very different from the ones relating to the safety of existing structures. Analysis for design of new structures starts by amplifying the load by say 40 or 50%, and the response up to failure is assumed to be linear (this is indeed code-driven). In analyzing the safety of existing safety structures, one has to determine the exact nonlinear response beyond the elastic limit and --
most importantly -- determine the corresponding deformation that would be under-estimated in the former case. Only the latter approach can truly capture the impact of damage caused by ASR without over-simplification. See Figure 1. A simplified linear elastic analysis (used in design of new structures) will under-estimate the displacements and cannot capture either the failure load or the deformation. Safety assessment can only be performed through a nonlinear analysis.
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Figure 1 Design vs Analysis C Discussion of Expert Opinion C.1 Do you continue to hold the same opinions you expressed in your February 2019 expert report?
Yes. My conclusions about the adequacy of NextEras investigations are reflected in my report entitled Concerns Regarding the Structural Evaluation of Seabrook Nuclear Power Plant (Feb.
12, 2019) (Exh. INT007). As I stated there, both the FSEL testing program and the finite element analysis used by NextEras consultant SG&H (with important input from the laboratory tests) are substandard and inadequate to support any conclusion that the ability of the Seabrook containment to withstand a design basis earthquake has not been unduly compromised by the presence of ASR. Id. page 3. To summarize, insufficient attention has been given to the unique and complex nature of ASR. Therefore, based on my expertise and the published state of the art of ASR and basic principles of structural engineering (design vs safety assessment; linear vs nonlinear analysis), I concluded that the quality of the presented results is not sufficiently reliable to support their stated purpose of confirming regulatory compliance for the next 30 years.
As discussed in my expert report, the manner in which NextEras consultants have analyzed the impact of ASR on Seabrook is seriously deficient in five major respects.
- First, the concrete used in the FSEL testing program was not representative of the concrete at Seabrook.
- Second, the specimen (scaled) dimensions, loads and boundary conditions are not representative of Seabrook.
- Third, NextEras failure to address reported load displacement and cracking patterns; and Page l 8
- Fourth, NextEra relies on incorrect assumptions about ASR, including confusing material strength with structural strength and assuming that adding a design basis load to the Seabrook safety analysis can account for ASR.
By themselves, these errors, which were incorporated into NextEras finite element analysis, had the effect of rendering that analysis completely unreliable to support any conclusions about the safety of the Seabrook plant under earthquake conditions.
And the errors adversely affected the adequacy of parameters used in the monitoring program.
Both the monitoring program for ASR progression and the monitoring program for structural deformation depend on FSEL test results. And both programs are seriously deficient because of that dependence.
In addition, the problems with NextEras safety assessment and monitoring programs were compounded by the fact that NextEra and its consultants applied an analytical method to the FSEL data that was extremely simplistic and contains numerous significant flaws (ASR modeling and seismic analysis, among others.) By feeding erroneous and unreliable data into an analytical model that was already inadequate to address the complexity of ASR at Seabrook, NextEra compounded the problem and made it even worse.
In considering this issue, it is important to recognize that testing and analysis (any analysis) is a very tightly coupled process where the latter depends greatly on the reliability of the former.
Hence, the results of any finite element analysis whose cracking/failure/safety criteria depend on erroneous experimental data will consequentially be flawed.
This is what happened. First, NextEra relied on test results to conclude that Seabrook is currently safe to operate. Second, NextEra gauged the nature and degree of monitoring required based on the level of safety assurance it had obtained from the results of the flawed testing and analysis of the data. Third, NextEra based its acceptance criteria for determining the safety of Seabrooks operation and its parameters for monitoring ASR on the FSEL test results and SG&H analysis. These criteria were approved by the NRC Staff.
I also have an overarching concern about the absence of any credible peer review of NextEras work. NextEra relied on consultants with standard engineering experience, and did not seek review by independent ASR experts. The NRC Staff ultimately accepted scientifically unproven assertions. Given that this is the very first occurrence of ASR in a nuclear containment vessel, that the NRC has funded at least two major projects on ASR, both NextEra and the NRC should have ensured that their work would receive independent review by qualified experts in the field.
It is important to note that my conclusions are relevant not just to the continued operation of Seabrook out to the end of its current license term in March 2030, but also for Seabrooks renewed operating license term - which does not expire until 2050. This is because the license Page l 9
amendment will become part of NextEras Aging Management Plan for the license renewal term. Under the renewed license that has adopted the terms of the LAR, Seabrook may operate for a very long time based on (a) the LARs unjustified determinations of safe operation and (b) a monitoring plan whose parameters were devised to confirm those faulty findings and thus are inadequate to assess the true extent and progression of ASR at Seabrook. Finally, should there be other instances of containment structures suffering from ASR, its operator will then perpetuate the flawed process endorsed by the NRC.
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