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This petition calls for the NRC to issue an order to the Seabrook 

licensee requiring immediate implementation and enforcement of

ACI 349.3R and ASTM C 856-11 code standards requiring core 

sampling, and petrographic testing for the mechanical properties of 

tensile strength, Poisson’s ratio, modulus of elasticity, and 

compressive strength—specifically for walls of the Containment 

Building and Spent Fuel Pool at Seabrook Station.

I. INTRODUCTION

Provision “G” of Seabrook Station’s operating license (1990 as amended 

11-02-2015 per ML053130320) states “The issuance of this license will 

not be inimical to common defense and security or to the health and 

safety of the public.” However, this declaration of assurance by the 

Nuclear Regulatory Commission (NRC) has been undermined since 2009, 

when Alkaline Silica Reaction (ASR) was first detected during an on-site 

inspection. ASR continues to spread through the concrete structures of 

the reactor complex, and in so doing degrades their structural integrity. 

C-10 Research and Education Foundation has studied ASR and 

concludes that its presence at Seabrook is indeed inimical to public 

health and safety. To date, the NRC and NextEra have not determined 

the rate or extent of the ASR degradation, or confirmed the presence of 

ASR in the Containment Building. What is known is that ASR is a 

progressive and irreversible condition. Quoting NRC: “...affected SSC

[structures, systems, and components] should be considered operable but 

degraded, and below full qualification... Full qualification will be attained 

when the testing and analysis plans developed to address the ASR issues 

are completed and the long-term resolution is incorporated into the UFSAR

[Updated Final Safety Analysis Report].” (from revised NRC OP October, 

2011; excerpted from email to D. Grinnell, C-10 11/21/13.) Since the 

revelation of ASR in 2009, Seabrook has never achieved “full 

qualification” for operability – nor are they likely to do so, because 

Seabrook’s ASR has caused its concrete structures continually to 

degrade and fall further below full qualification. Unfortunately, to-date 

“research in this topic exists, but it has not reached a level that would 

allow the repair and prediction of service life.” (Page 50, NIST / ANSI 

Codes and Standards for the Repair of Nuclear Power Plant Structures”)

In 2014, C-10 filed a Petition for Rulemaking with NRC, to require that 

all licensees comply with the American Concrete Institute (ACI) standard 

349.3R “Evaluation of Existing Nuclear Safety-Related Concrete 

Structures,” and American Society for Testing and Materials (ASTM) 

standard C 856-11, “Standard Practice for Petrographic Examination of 

Hardened Concrete”. This would ensure that inspection frequency and 

methodology for ASR detection be standardized throughout the US 
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commercial reactor fleet. As of today, use of these industry standards is 

merely recommended, not mandated by NRC.

II. PURPOSE AND JUSTIFICATION

As of this filing, NRC has taken no action to require NextEra to follow ACI 

and ASTM codes. In the case of NextEra’s investigation into the extent of 

ASR at Seabrook, there are aspects of these standards that they appear 

to ignore. One consequence of the lack of adherence to these codes might 

be the adoption by NextEra of several assumptions that have proven to 

be scientifically inaccurate.

NRC staff made note of the lack of regulatory guidance, on September 19, 

2014, before NRC’s Advisory Committee on Reactor Safeguards (ACRS):

“...goals [under 10 CFR 50.65] be established commensurate with safety 

and, where practical, take into account industry-wide operating 

experience. In practice, for concrete structures, this usually translates into 

periodic visual inspection; however, specific inspection criteria related to 

ASR are generally not included.” Meanwhile, as far back as ten years, 

NRC made note of concrete degradation at the Seabrook reactor: 

“...visual examination of concrete containment surface (VT-3C) performed 

in October 2005.... identified numerous areas of spalled concrete that 

equal or exceed a depth of 1 inch. According to evaluation criteria in ACI 

349.3R-02, Sect.5.1 spalled areas that exceed a depth of 3/8in. and 4 in. 

in any direction must be evaluated.” (RAI B.2.1.28-2, p. 30) 

Furthermore, NUREG/CR -7171, “A Review of the Effects of Radiation on 

Microstructure and Properties of Concretes Used in Nuclear Power 

Plants” highlights the inadequacy of visual inspection: “The primary 

method utilized for detection of ASR is through visual examinations 

indicating evidence of expansion, relative movements between structural 

elements, and cracking. ASR capable of being detected visually, however, 

is probably in a fairly advanced stage of development.” 

Because the standards contained within ACI 349.3R and ASTM C 856-11 

represent the accumulated wisdom of “industry-wide operating 

experience” for concrete structures ostensibly sought after by NRC, NRC 

staff helps to make our argument for adoption and enforcement of these 

standards.

In the absence of regulatory enforcement for ASR testing, NextEra gives 

every appearance of avoiding the on-site (that is, at the reactor) testing 

that is an accepted and crucial part of proper petrographic analysis. In
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testimony before the ACRS / Plant License Review (PLR) Subcommittee, 

July 10, 2012, Office of Nuclear Reactor Regulation’s Senior Structural 

Engineer, Abdul Sheikh noted the 22 percent reduction in compressive

strength found in certain safety-critical buildings at Seabrook Station. 

He points out that this reduction in strength may in fact be considered 

much greater, when we consider that in the absence of ASR, the concrete 

should have strengthened by more than 20% since construction. (When 

compared to the expected higher strength value of 4800 psi, the tested 

strength reflected more than a 30% reduction.) Sheikh further states, 

“Since then the applicant has not performed any test to determine the rate 

of degradation of shear, tensile strength, bond strength on the concrete in 

the last 18 months. They haven’t, as I pointed out before, they haven’t 

extracted any cores from the containment... And it is a well-known fact 

that the visual examination cannot rule out the presence of ASR. You have 

to do some confirmatory tests.” (NRC ML 122070)

In the specific case of the Containment Building walls, C-10 is very 

concerned that the destructive power of the radiation from the reactor

core may have a “coupling” impact, along with ASR, in accelerating the 

deterioration of this most vital safety structure. NUREG/CR-6927, 

“Primer on Durability of Nuclear Power Plant Reinforced Concrete 

Structures – A Review of Pertinent Factors,” Oak Ridge National 

Laboratory, makes several references to this issue: (p.25) “Nuclear 

heating occurs as a result of energy introduced into the concrete as the 

neutrons or gamma radiation interact with the molecules within the 

concrete material. The heat generated may have detrimental effects on the 

physical, mechanical, and nuclear properties of the concrete.... 

Determination of whether any deterioration that may occur in concrete 

properties is due to radiation damage or thermal effects can be difficult.” 

(p.26) “...gamma radiation may cause a reduction in compressive 

strength... tensile strength of concrete is significantly reduced... with the 

decrease of tensile strength caused by neutron radiation more pronounced 

than the decrease in compressive strength...” Furthermore, there is an 

indication that nuclear radiation can significantly increase the reactivity of 

silica-rich aggregates to alkali (i.e. alkali-silica reaction).” 

Even with the abundance of evidence for the need of thorough strength 

testing for the Containment Building, Mel Gray, Branch Chief for 

Engineering, NRC Region 1, remarked to C-10’s Debbie Grinnell, in an 

email dated 8/5/14: “...My staff is not aware of any cores taken from the 

Seabrook... containment structure.”

C-10’s expert witness on this matter, Dr. Paul Brown, Professor of 

Ceramic Sciences and Engineering at Penn State University, underscored 

the vital importance of tensile strength testing for concrete in his 

9/19/14 testimony before the ACRS / PLR Subcommittee. He stated, in 
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part: “...we can’t simply rely on someone giving us some compressive 

strength data to really understand what’s going on in the structure...a 

deteriorated concrete is characterized both for its compressive and its 

tensile strength... And you can understand it fairly simplistically. If you 

take a solid metal object and you just squeeze on it, you take a roll of 

quarters and you squeeze on it, the mechanical properties in compression 

may not be significantly compromised. But if you try and do a tensile test, 

you’ll see a very, very different set of results.”

C-10’s Ms. Grinnell asked Dr. Brown a few simple questions in 

preparation for this petition—via email, 11/20-21/15. These exchanges 

follow, with Dr. Brown’s answers in italics:

(D.G.) “Is there a professional reason that cores done in a nuclear 

plant containment building would not show data that would reveal 

reliable results?” (P.B.) “There is no reason.” 

(D.G.) “Is there a reason cores cannot be done in a nuclear plant 

containment building as containment is too massive a building to draw a 

core safely?” (P.B.) “There is no reason.”

(D.G.) “Do you know if a core is removed from a containment wall, 

would it damage the building?” (P.B.) “No it will not have a meaningful 

effect on the building...No good technical reason not to do cores.”

After six years of study and analysis of ASR at Seabrook Station –

Seabrook being the first commercial atomic plant in United States known 

to be so affected – NRC has not yet required that standard code testing 

be performed, in order to discover the full extent of the concrete 

degradation in walls supporting the two most safety-critical areas of the 

facility: the atomic reactor containment structure and the highly 

radioactive spent fuel pool. Insistence on such testing is in fact NRC’s 

obligation. In part, this lack of oversight may be due to the inexperience 

of NRC with the phenomenon of ASR – and the absence of ASR-specific 

regulations within the larger framework of NRC codes and standards. In 

the view of the petitioner, the lack of action to discover and report the 

extent of ASR within Seabrook’s safety-critical structures is also the 

result of a seeming unwillingness on the part of NRC to mandate the 

steps needed to determine the actual current material properties of 

concrete (determined for tensile strength, Poisson’s ratio, modulus of 

elasticity, and compressive strength as stated above) in those locations 

within the facility whose structural integrity is most crucial to the public 

health and safety. 

Seabrook’s containment must remain an extremely robust concrete 

structure, in order to limit the radiation leaking out. Excessive leakage 

can expose plant workers to radiation and prevent or impede their efforts 

to mitigate an accident. Excessive leakage can expose members of the 

public to radiation and increase their chances of experiencing radiation-
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induced illnesses and fatalities. In light of the fact that both the 

aforementioned structures contain and must isolate from the biosphere

some of the deadliest and long-lived toxins on earth, any realistic “aging 

management program” must begin with an accurate assessment of the 

damage already done by ASR.

NextEra has elected to pursue a large scale, off-site research test 

program to resolve Seabrook’s containment non-conformance to design 

basis with reduced margins, rather than do standard code testing in 

Seabrook Station’s actual in-situ concrete containment building. The 

NRC has not required NextEra to take core samples from the plant’s 

structures or test specimens. (email to D Grinnell from NRC’s Seabrook 

Project Manager Glen Dental., October 12, 2015)

Several with the NRC Region 1 staff, NRR staff and experts have stated 

repeatedly that many of NextEra’s assumptions were scientifically 

unfounded, and that the NRC should declare its “...Action Plan—to 

provide a commitment and schedule to obtain core samples from all 

Category 1 concrete structures to confirm the presence or absence of 

ASR concrete degradation” (comments by WJR on Draft TIA for ASR at 

Seabrook, 2011.)

David Wright of the Union of Concerned Scientists carefully explained 

several of NextEra’s false assumptions in a letter to NRC dated 11/4/13, 

“Continuing Problems with Monitoring Concrete Damage at Seabrook,” 

co-signed with C-10 Executive Director Sandra Gavutis. For instance,

reliance on the “crack width index” as an “unjustified measure of ASR”; 

the pursuit of “replica tests” at the University of Texas in the place of 

tests which could be performed on the actual concrete structures from 

Seabrook Unit 2; the premature conclusion that the steel reinforcement 

is not corroding, given the presence of “potentially aggressive water 

migration through the concrete...;” these are just some of Dr. Wright’s 

points. Concerning NextEra’s avoidance of Seabrook in situ core 

extraction and testing, Dr. Wright quotes Dr. Paul Brown, C-10’s expert 

witness cited above: “It is well understood that drilled cores are extracted 

from an existing structure and have been subjected to the service 

environment associated with that structure. This in no way invalidates the 

result of the testing. The results of core testing are generally understood 

within the relevant engineering community. The NextEra preposition 

misuses the cautionary language of ASTM C42 and appears to be an 

attempt to avoid accumulating data which might be regarded as 

problematic.”

Although the NRC has stated that NextEra’s off-site research test 

program must represent the actual in-situ conditions of Seabrook’s 

containment, NextEra has refused the NRC Region I staff, NRR staff, and 
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experts’ recommendations that code testing should and must be done. 

NextEra continues to state assumptions that are not scientifically 

supported as their basis to refuse to do core sample code testing on 

Seabrook’s Containment Building, a seismic Category 1 structure. 

Without the actual in-situ core testing, no one can know the extent or 

rate of ASR within the reactor containment at Seabrook. 

Therefore, we request this Emergency Enforcement Action. As Emeritus 

Professor of Engineering from UMass Lowell, and member of the 

guidance system development team for the Apollo space program Dudley 

Shepard stated in his 3/25/15 comment in support of C-10’s Petition for 

Rulemaking, “Each of the players must be responsible for their role in 

the solution... Without NRC enforcement oversight, the solution process 

breaks down, and the problem remains. There is no other way.”

Sandra Gavutis, Executive Director

C-10 Research & Education Foundation

44 Merrimac Street, Newburyport, MA

01950