ML24141A281
| ML24141A281 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/27/2024 |
| From: | Public Commenter Public Commenter |
| To: | NRC/NMSS/DREFS |
| NRC/NMSS/DREFS | |
| References | |
| 89FR960 | |
| Download: ML24141A281 (19) | |
Text
From: Paul Gunter <paul@beyondnuclear.org>
Sent: Tuesday, February 27, 2024 5:03 PM To: NorthAnnaEnvironmental Resource
Subject:
[External_Sender] CORRECTED comments of Beyond Nuclear and Sierra Club on North Anna Draft SEIS for SLR Attachments: FINAL_2024.02.27 Beyond Nuclear-Sierra Club corrected-updated comments_NAPS Draft SEIS-20240222.pdf
- Hello, Attached please find the corrected and updated Beyond Nuclear and Sierra Club comments on the North Anna Draft Site-Specific Environmental Impact Statement for Subsequent License Renewal.
Thank you, Paul
Paul Gunter, Director Reactor Oversight Project Beyond Nuclear 7304 Carroll Avenue #182 Takoma Park, MD 20912 Tel. 301 523-0201 (cell) www.beyondnuclear.org Federal Register Notice: 89FR960 Comment Number: 5
Mail Envelope Properties (CALTCGdnTM_PUetOraDZz-UUoY-M_8weYePUzNX1kwjScbijNZA)
Subject:
[External_Sender] CORRECTED comments of Beyond Nuclear and Sierra Club on North Anna Draft SEIS for SLR Sent Date: 2/27/2024 5:02:43 PM Received Date: 2/27/2024 5:03:11 PM From: Paul Gunter
Created By: paul@beyondnuclear.org
Recipients:
"NorthAnnaEnvironmental Resource" <NorthAnnaEnvironmental.Resource@nrc.gov>
Tracking Status: None
Post Office: mail.gmail.com
Files Size Date & Time MESSAGE 425 2/27/2024 5:03:11 PM FINAL_2024.02.27 Beyond Nuclear-Sierra Club corrected-updated comments_NAPS Draft SEIS-20240222.pdf 2390538
Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
February 27, 2024
United States Nuclear Regulatory Commission Office of Administration Washington, DC 20555-0001 ATTN: Program Management, Announcements and Editing Staff By Email: NorthAnnaEnvironmental@nrc.gov
SUBJECT:
Corrected and updated comments on Draft Environmental Impact Statement for North Anna Units 1 and 2 Subsequent License Renewal
To whom it may concern,
On February 22, 2024, as provided in 89 Fed. Reg. pp 960-963 (Jan. 8, 2024), Beyond Nuclear submitted comments in response to the US Nuclear Regulatory Commissions NUREG-1437, Supplement 7a, Second Renewal, Site-Specific Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Subsequent License Renewal for North Anna Power Station Units 1 and 2, [Draft Report for Comment]
(DSEIS). The comments included an expert technical report by Mr. Jeff Mitman.
Beyond Nuclear now re-submits those comments with the addition of the Sierra Club as an additional sponsor. The late addition of the Sierra Club was necessitated by an approval process that took longer than expected. No other changes to the comments have been made, other than to correct a minor clerical error in the summary of our comments that was provided in Beyond Nuclears February 22, 2024 letter.
This updated should be considered because it is being filed less than a week after the comment deadline and because the substance of the comments has not changed.
To repeat the content of our February 22, 2024 comment letter: NRC Staff has not justified an overall site-specific finding of no significant impact and that the extended operation (60- to 80-years) of the North Anna nuclear power stations on the environment is SMALL with any reasonable assurance.
Mr. Mitmans extensive comments focus on a number of impacts including postulated accidents and the potential environmental impacts of postulated accidents. His comments also cover the DSEIS review of the August 23, 2011 Mineral, Virginia earthquake which exceeded the North Anna Power Stations design basis for safety critical systems, structures and components but does not provide any discussion on or an analysis of the impacts of the exceedance on these safety systems.
Other areas covered by Mr. Mitmans comments scrutinize the adverse impacts of climate change on nuclear power operations noting that the Draft SEIS does not address climate changes impacts on accident risks in Section 3.11.6.9 or Appendix F.
Mr. Mitman notes, This omission constitutes a significant deficiency in the Draft EIS because climate change demonstrably affects the frequency and intensity of some external events and therefore has the potential to significantly increase accident risks.
Mitmans comments further recognize that t he DSEIS does not acknowledge that the National Academies under sponsorship of the National Oceanic and Atmospheric Administration (NOAA) has started a project to modernize the probable maximum precipitation (PMP) methodology important to address a changing climate. He states, This project will consider approaches for estimating PMP in a changing climate, with the goal of recommending an updated approach, appropriate for decision -maker needs.
PMP is a significant input into the design of critical infrastructure such as dam and reactor safety analysis directly and indirectly through its impact on probable maximum flood (PMF). Mr. Mitman points out, The NRC is well aware of this effort as they have already participated in at least one of the initial project workshops and that [t]he Draft SEIS is silent on this and is thus deficient.
Importantly, Mr. Mitman further identifies that the Draft SEIS does not address the environmental impacts of concurrent multi-unit accidents which represents a significant omission that despite the fact that the questions of multi-unit accident is not one of possibility, but of probability.
Thank you,
Signed by Paul Gunter Signed by Connor Kish
/pg/-------------- -------------/ck/-------------
Paul Gunter Connor Kish, Director Beyond Nuclear Sierra Club, VA Chapter 7304 Carroll Avenue #182 100 West Franklin Street Mezzanine Takoma Park, MD 20912 Richmond, VA 23220
2
Technical Review of U.S. Nuclear Regulatory Commissions Draft Site-Specific Environmental Impact Statement for Subsequent License Renewal of North Anna Power Station Units 1 and 2 With Respect to Accident Analysis
Submitted by Jeffrey T. Mitman to the U.S. Nuclear Regulatory Commission on behalf of Beyond Nuclear, Inc.
February 22, 2024
Introduction
This report presents my technical review of the accident risk analyses presented by the U.S. Nuclear Regulatory Commission (NRC) in the Draft Site-Specific Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 7a, Second Renewal Regarding Subsequent License Renewal for North Anna Power Station Units 1 and 2 (Dec. 2023) (ML2339A047) (Draft SEIS). My review focuses on Section 3.11.6.9 (Postulated Accidents), and Appendix F (Environmental Impact of Postulated Accidents).
I am submitting this technical review on behalf of Beyond Nuclear, Inc. (Beyond Nuclear) and the Sierra Club, who have participated in multiple environmental and safety proceedings regarding Virginia Electric Power Companys/Dominions (VEPCOs or Dominions) 2020 application for subsequent license renewal (SLR) for the North Anna reactors (NAPS). In 2023, I also prepared a technical review for Beyond Nuclear and the Sierra Club regarding the Draft License Renewal GEIS. Beyond Nuclear and the Sierra Club submitted my expert declaration and technical review in support of their comments on the Draft GEIS, Comments by Beyond Nuclear and the Sierra Club on Proposed Rule and Draft Generic Environmental Impact Statement for Renewing Nuclear Power Plant Licenses (May 2, 2023, corrected May 19, 2023)
(ML23123A411).
As set forth in the expert declaration submitted with Beyond Nuclears and the Sierra Clubs comments, I am a nuclear engineer with a significant level of expertise in risk analysis. I have more than 40 years of experience in the nuclear industry and 16 years of experience with the NRC. While at the NRC, I served as Senior Reliability and Risk Analyst, with significant responsibility for managing a number of risk analysis projects and teams. A copy of my curriculum vitae is attached to my comments here.
Comments on Section 3.11.6.9 Postulated Accidents
According to the Draft SEIS at Page 3-169 (lines 35-36): For design-basis accidents, site-specific analysis of design-bass accidents is in the North Anna Updated Final Safety Analysis Report (UFSAR). Because the UFSAR is part of the current licensing basis and also the subject of the NRC oversight program for operation during PEO [period of extended operation], the impacts of design-basis accidents are SMALL. But this claim is not consistent with the history of operation of NAPS, because the reactors experienced a beyond-design-basis event in 2011, the Mineral earthquake. While the Mineral Virginia
1
earthquake is discussed in the Draft SEIS, the NRC does not describe whether or how it brought the plant back into compliance with the design basis. My review of NRC-VEPCO correspondence indicates that nothing has been done in this regard, i.e., the earthquake has led to no changes in the NAPS design-basis accident (DBA) analysis, the design basis (DB) or its current licensing basis (CLB). Thus, the Draft SEIS lacks support for its claim that environmental impacts of operating NAPS are SMALL because the reactors are operating in compliance with their design basis. I would also note that the Mineral Earthquake is not discussed in any accident risk analysis in the Draft SEIS, either here or in Appendix F.
Comments on Appendix F Environmental Impacts of Postulated Accidents
Section F.1.1 Design-Basis Accidents (Page F-2 line 41) states: [T]he licensee is required to maintain the acceptable design and performance criteria (which includes withstanding design-basis accidents) throughout the operating life of the nuclear power plant, including any license-renewal periods of extended operation. On August 23, 2011 the NAPS experience the Mineral Virginia earthquake. This earthquake caused NAPS to exceed its licensed DBA. Thus, NAPS has not been maintained with its DB or its CLB.
At Page F-3 (lines 10 - 13) the Draft SEIS states: [T]he NRC completed its review of Fukushima-related information relevant to North Anna and concluded that no further regulatory actions were needed to ensure adequate protection or compliance with regulatory requirements, thereby reconfirming the acceptability of North Annas design basis. This assertion is incorrect. The correspondence documenting the NRCs review of Fukushima -related information relevant to North Anna did not reconfirm the acceptability of the entire NAPS design basis as claimed. At best, the correspondence confirmed the elements of the design of NAPS regarding seismic and flooding hazards. Nothing was said about the acceptability of North Annas design basis.
At Page F-3 (lines 42 - 46) the Draft SEIS states: Because the requirements of the existing design basis and any necessary aging management programs will be in effect for SLR, the environmental impacts of design-basis accidents as calculated for the original operating license application should not differ significantly from the environmental impacts of design-basis accidents during other periods of plant operations, including during the initial license renewal and SLR periods. This statement is incorrect. As stated in the previous comment, NAPS experienced a beyond design basis earthquake in August of 2011.
Thus, NAPS was not maintained within its design basis. Accordingly, the conclusion of the Draft SEIS regarding the impacts of design basis accidents is unsupported.
In Section F.3.1 New Internal Events Information (Section E.3.1 of the 2013 LR GEIS), starting at Page F-9 (line 28), the Draft SEIS presents a discussion of internal events risks. This discussion makes comparisons between the 2013 GEIS and the current Dominion PRA model for internal events. Typically, a reactor risk analysis of internal events considers internal flooding along with other internal events risks. In this case, however it is unclear from the text whether the Draft SEIS includes an analysis of internal flooding. If it has been excluded, the omission is significant and should be corrected.
At Page F-10 (lines 10 - 14) the Draft SEIS states: [T]he NRC staff concludes that no new and significant information exists for North Anna during the SLR term concerning the offsite consequences of severe accidents initiated by internal events that would alter the conclusion that the probability weighted consequences of severe accidents would be SMALL reached in the 1996 LR GEIS, the 2013 LR GEIS, and the North Anna initial LR SEIS. This conclusion is stated after reiterating the corresponding 2013 GEIS
2
CDF values for NAPS and the current NAPS internal events values as supplied by Dominion. However, the analysis underlying the conclusion does not comply with NRC guidance requiring that risk-informed decision-making must include consideration of uncertainties. See NUREG-1855 Rev. 1 "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2017.
At page F-10 (line 32) of Section F.3.2 External Events (Section E.3.2 of the 2013 LR GEIS), the Draft SEIS reports that the seismic CDF value for NAPS is 6E-5 per year. (This is consistent with the NRCs 2019 letter to Dominion giving a value of 6.3E-5 per year. See letter from NRC to Dominion re: North Anna Power Station, Units 1 And 2 - Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near -Term Task Force Recommendation 2.1:
Seismic, April 25, 2019 (ML19052A522 )). In contrast, Table E.3-12 in the Draft License Renewal GEIS (Page E-33) shows a mean all hazards CDF (i.e., including both internal events and external events) for PWRs of 6.1E-5 per year. Thus, the Draft SEIS value for seismic alone at NAPS is greater than the NRCs calculated average for all hazards combined in the Draft License Renewal GEIS. This discrepancy should be addressed.
In addition, Table E.3-10 of the Draft License Renewal GEIS (Page E-28) lists the mean fire CDF for all reactors as 4.5E-5 per year. For NAPS the sum of internal events (1.36E-6), seismic (6.3E-5) and fire risk (4.5E-5) totals 1.1E-4 per year. This value (1.1E-4) is significantly above both the Draft License Renewal GEIS all hazards value of 6.1E-5 and the 8.4E-5 internal events value used in the original 1996 License Renewal GEIS to make its decisions. In fact, per RG-1.174 Rev. 3 An Approach for Using PRA in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256), a value of 1.1E-4 per year places NAPS in a risk region: Applications that result in increases to CDF above 1E-5 per reactor year would not normally be considered (Page 28 of RG 1.174).
In this regard, it should be noted that Dominion has not completed a fire PRA on NAPS. Therefore, the above combined internal and external CDF sum of 1.1E-4 per year is based on an industry mean fire CDF of 4.5E-5 per year which corresponds to a median fire CDF of 4.6E-5 per year (from the Draft License Renewal GEIS, Table E.3-10 (Page E-28)). It should be remembered that half of all values are above the median value. Therefore, there is a 50% probability that the NAPS fire CDF is greater than the median value of 4.6E-5. Draft NUREG-1437 Rev. 2 Table E.3-10 reports fire CDFs for 68 reactors or about 75% of the current fleet. In this regard, I would also note that a quarter of the fleet has either not performed a fire PRA or has not reported their fire PRA results to the NRC. Because the NRC has only partial fire CDF data, it is very possible that a plant that has not performed a fire PRA has a fire CDF higher or even significantly higher than the highest results reported (i.e., Turkey Point Unit 3 with a reported fire CDF of 8.7E-5 per year - Draft License Renewal GEIS Appendix E Table E-3.10 at Page E-28). That is, NAPS could have a fire CDF significantly higher than the value used to make this important regulatory decision. With the tools available today, this is not acceptable.1
Page F-11 (lines 16 - 18) the Draft SEIS states: The staff also noted that the actions taken by Dominion and experience gained after the 2011 Mineral earthquake provide additional assurance regarding North Annas ability to handle a beyond-design-basis seismic event. The Draft SEIS gives no explanation for
1 At Page F-11 (lines 24 - 32) the Draft SEIS uses ratios and percentages to make arguments why margin exists between the SLR decision today and the original decisions for LR in 1996 GEIS. As documented by the illustration above, there is no need to use these ratio surrogates when actual CDF values are or should be available.
3
what actions taken or experienced gained has anything to do with the quantitative risk reduction or the resolving the design basis exceedance in 2011.
At Page F-11 (lines 28 - 29) the Draft SEIS states that the population dose risk was calculated in the initial SAMA analysis to be 50 person-rem per reactor year (RY). According to the NRC: This provides a ratio of the North Anna 1996 LR GEIS 95 percent upper confidence bound predicted population dose to North Anna initial license renewal total population dose risk of 30. This considerable margin offsets any increases in external events since the previous SAMA analysis. (lines 29 - 31). The NRC argues that a ratio of 30 ( 1,496 / 50 = 30 ) between the 1996 LR GEIS and the initial license renewal SAMA total population does risk supplies assurance that risks are low enough. But it appears that the initial LR 50 person-rem/RY calculation was at least in part based on an unrealistically low seismic CDF, given that the Dominion-reported seismic CDF is 6.3E-5 per year. The current base estimate analysis shows a ratio of external event risk to internal event risk of 75 [ = ( 3.9E-5 (fire CDF) + 6.3E-5 (seismic CDF) ) / 1.36E-6 (internal events CDF) ]. Thus, the external to internal events ratio of 75 increase swamps the 30 ratio reduction argued by the Draft SEIS.
In Section F.3.3 New Source Term Information (Section E.3.3 of the 2013 LR GEIS) Page F-12 (lines 17 -
18, the Draft SEIS) states: The NRC staff expects to incorporate the information gleaned from the SOARCA project in future revisions of the LR GEIS (NRC 2013-TN2654). NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report was published in November 2012. One would think that the subsequent decade would supply sufficient time to incorporate any insights gained from the SOARCA work into the Draft GEIS Rev. 2 or this Draft SEIS.
In Section F.3.5 Higher Fuel Burnup Information (Section E.3.5 of the 2013 LR GEIS) at Page F-14 (lines 1 -
- 4) the Draft SEIS reports the NRC Staffs conclusion that: [N]o new and significant information exists for North Anna SLR concerning offsite consequences due to higher fuel burnup that would alter the conclusions reached in the 1996 LR GEIS and 2013 LR GEIS or the North Anna initial LR SEIS. But on Page F-13 (lines 33 - 36) the Draft SEIS states: According to the 2013 LR GEIS, increased peak fuel burnup from 42 to 75 gigawatt days per metric ton uranium (GWd/MTU) for PWRs resul ts in small to moderate increases (up to 38 percent) in population dose in the event of a severe accident (emphasis added). It then goes on to say (Page F-13 lines 39 - 40): Dominion indicated that the average burnup level of the peak rod is not planned to exceed 60,000 MWd/MTU during the proposed SLR operating term. The Draft SEIS contradicts itself as it says that a 42,000 to 75,000 MWd/MTU can lead to a moderate increase and that Dominion may operate at the 60,000 MWd/MTU range. It appears that this should be a moderate increase and not a no new and significant information conclusion.
At Page F-14 (lines 26 - 32) Section F.3.6 Low Power and Reactor Shutdown Event Information (Section E.3.6 of the 2013 LR GEIS) references industry initiatives described in SECY 97-168 for the proposition that: [T]he offsite consequences of severe accidents, considering low power and reactor shutdown events, during the North Anna SLR term would not exceed the impacts predicted in either the 1996 LR GEIS or 2013 LR GEIS. SECY 97-168 discusses improvements in outage conduct achieved in the early 1990s. The benefits achieved by these industry changes were known before the 1996 initial LR GEIS was promulgated and were incorporated into the 1996 initial LR GEIS. Any residual improvements post the 1996 initial LR GEIS were obtained and certainly understood long prior to the 2013 GEIS update. Thus, the NRC should not double-count those industry initiatives as it does here. To the contrary, the NRC should consider expert industry evidence that the number of outage events increased during the decade between 2000 and 2010, see INPO SOER 10-2 September 7, 2010.
4
Section F.3.7 Spent Fuel Pool Accident Information (Section E.3.7 of the 2013 LR GEIS) (Pages F-14 to F-15) fails to capture new information documented in NUREG-2161 "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor,"
September 2014. While this is a BWR analysis it is relevant to PWRs also. The authors of the Draft SEIS appear to be unaware of the work. This section also fails to capture and address the fact that utilities are packing more spent fuel with higher burnup into spent fuel pools (SFPs), while at the same time not increasing the heat removal capacity of the SFP cooling systems. These facts are not considered in this analysis.
At page F-16 (lines 11 - 14) Section F.3.9 Uncertainties (Section E.3.9 of the 2013 LR GEIS) presents the NRC Staffs conclusion in the 2013 LR GEIS, the NRC staff concluded that the reduction in environmental impacts resulting from the use of new information (since the 1996 LR GEIS analysis) outweighs any increases in impact resulting from the new information. With a close reading of the sentence, it can be reduced to The staff concluded the reduction in impacts resulting from the use of new information outweighs any increases from the new information. This makes no sense.
This section is also seriously deficient because it purports to address risk uncertainties without complying with NUREG-1855 Rev. 1 "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2017 (ML17062A466). As stated in NUREG-1855, that document provides guidance on how to treat uncertainties associated with PRAs used by a licensee or applicant to support a risk-informed application to NRC. The NRC Commissioners themselves identified
[t]reatment of uncertainty as an important issue for regulatory decisions in NRC Final Policy Statement Use of PRA Methods in Nuclear Regulatory Activities, Federal Register, Vol. 51, p. 42622 (51 FR 42622), Washington, D.C., 1995.) As they explained: Uncertainties exist... from knowledge limitations... A probabilistic approach has exposed some of these limitations and provided a framework to assess their significance and assist in developing a strategy to accommodate them in the regulatory process. Id. The Draft SEIS should be revised consistent with the Commissions direction.
Section F.3.10 Summary and Conclusion (Section E.5 of the 2013 LR GEIS) at Pages F-17 to F-18 reiterates areas of risk reduction but fail to mention the increase in risk identified by new seismic analysis. See discussion above.
Page F-18 (lines 3 - 4) states: The LR GEIS estimated the net increase from the five areas listed above would be (in a simplistic sense) approximately an increase by a factor of 4.7. The Draft SEIS does not specify where the factor of 4.7 comes from in the LR GEIS. The Draft SEIS then goes on to combine this 4.7 factor risk increase with a factor of 25 risk reduction in the internal events CDF to compute a total risk reduction by a factor of 20.3 (25 minus 4.7) (at 4). Assuming that the increase factor of 4.7 and the decrease factor of 25 are correct, the math applied is wrong. Instead of subtraction, the two factors should be divided, i.e., 4.7 divided by 25 yielding a net reduction by a factor of 0.2 or 5. This reduction is significantly less than the factor of 20.3 claimed.
Section F.4.1 10 CFR 50.54(hh)(2) Requirements Regarding Loss of Large Areas of the Nuclear Power Plant Caused by Fire or Explosions starting at Page F-19 and continuing to Page F-20 discusses the risk improvements obtained from efforts by the NRC and licensees post September 2001. However, any risk reductions obtained by these efforts should have been incorporated into the NAPS initial license renewal in 2002 and certainly by the 2013 GEIS Revision 1 report. The NAPS PRA used as the basis of Dominion Environmental Report (ER) supplied as part of the SLR application in 2020 should have
5
captured any of these risk reductions. Assuming that the NAPS 2020 PRA does capture these risk reductions, the Draft SEIS should not be attempting to credit them again.
Starting at Page F-19 (at 46) and continuing on Page F-20 (lines 1 - 4) the Draft SEIS states: These enhancements included significant reinforcement of the defense capabilities for nuclear facilities, better control of sensitive information, enhancements in emergency preparedness, and implementation of mitigating strategies to deal with postulated events potentially causing loss of large areas of the nuclear power plant due to explosions or fires, including those that an aircraft impact might create. None of these enhancements can be quantified in the PRA as these risks were never incorporated in the base PRA to begin with. Thus, to imply that the quantified risk has been improved is misleading at best. A more accurate characterization is that the base PRA understated the risks. Take for example the stated risk reduction from aircraft impact improvements, the aircraft impact hazard was not incorporated into the base PRA. If a risk reduction is to be credited, then the risk should be acknowledged and added into the base risk analysis before any enhancement is credited.
At Page F-20 (lines 14 - 16) the Draft SEIS states: NRC requirements pertaining to nuclear power plant security are subject to NRC oversight on an ongoing basis under a nuclear power plants current operating license and are beyond the scope of license renewal. If security is indeed beyond the scope of LR and SLR, then the Draft SIES should not argue for risk reductions for enhance security.
At Page F-21 (line 15) Section F.4.3 Fukushima-Related Activities, the Draft SEIS states that [T]here was a partial meltdown of fuel in three of the reactors (emphasis added). Fukushima Dai-ichi Units 1, 2 and 3 melted enough fuel to generate at least three powerful explosions. Unit 1 has fuel debris below the reactor pressure vessel. These should not be characterized as partial meltdowns.
Starting at Page F-21 (lines 41 - 42) Section F.4.4 Operating Experience states: Section E.2 of the 2013 LR GEIS mentions the considerable operating experience that supports the safety of U.S. nuclear power plants. The 2013 GEIS in Section E.2 discusses operating experience that has led to improved performance. This discussion includes topics on: IPE/IPEEE, aging monitoring improvements, generic safety issue (GSI) 191 on sump performance, and the September 2011 terrorist attacks. All but GSI 191 have already been credited in Appendix F of this Draft SEIS. This Draft SEIS section is suggesting that there are some other risk improvements not previously captured in Appendix F. However, the 2013 GEIS examples are not new and have already been credited in this Draft SEIS. The NRC should not double-count factors that have already been considered.
At Page F-24 (lines 17 - 19) Section F.5.3 Dominions Evaluation of 1 Unimplemented North Anna Phase 2 SAMAs states: SAMAs related to creating a containment vent were screened out because this nuclear power plant modification has been evaluated industrywide and explicitly found to not be cost effective in Westinghouse large/dry containments. The Draft SEIS should supply a reference documenting this bald assertion.
Starting at Page F-24 (lines 45 - 46) and continuing onto Page F-25 (lines 1 - 2) Section F.5.4 Dominions Evaluation of SAMAs Identified as Potentially Cost-Beneficial at Other U.S. Nuclear Power Plants that Are Applicable to North Anna states: Of the results presented in Table E4.15-2 [of the NAPS ER], one case (labeled as emergency diesel generator (EDG)) yielded an internal events LLRF (Large Late Release Frequency) reduction of 57 percent. However, Dominion explained that the total change in the Maximum Benefit for the EDG case is well below 50 percent. The SAMA methodology has a threshold for continued evaluation of a 50% reduction. The case identified here has a risk reduction of 57% and
6
thus, exceeds the 50% threshold. The NRC should explain: Why wasnt this case explored in more detail as required by the SAMA methodology?
At Page F-25 (lines 7 - 8) the Draft SEIS states: The NRC staff reviewed North Annas onsite information and its SAMA Stage 1 process during an in-office audit at NRC headquarters (NRC 2020 -TN8100 see Appendix D). The supplied reference (NRC 2020 -TN8100) goes to ML20351A388. This reference document is a four-page letter from NRC to Dominion documenting the occurrence of the in -office audit. It does not document the audit findings. There is no Appendix D to this letter and thus, no documentation that: The staff found that Dominion had used a methodical and reasonable approach to identify any SAMAs that might reduce the maximum benefit by at least 50 percent and therefore could be considered potentially significant (Page F-25 lines 8 - 11). Thus, it is not possible to evaluate this claim. The NRC should supply the missing information.
Section F.5.6 Conclusion at Page F-26 (lines 29 - 33) the Draft SEIS states: The NRC staff reviewed Dominions new and significant information analysis for severe accidents and SAMAs at North Anna during the SLR period and finds Dominions analysis and methods to be reasonable. As described above, Dominion evaluated a total of 334 SAMAs for North Anna SLR and did not find any SAMAs that would reduce the maximum benefit by 50 percent or more. This conclusion is inaccurate. As discussed above Dominion found a EDG SAMA with a Phase 1 risk reduction of 57%.
At Page F-26 (lines 37 - 41) the Draft SEIS states: [T]he NRC staff did not otherwise identify any new and significant information that would alter the conclusions reached in the previous SAMA analysis for North Anna. Therefore, the NRC staff concludes that there is no new and significant information that would alter the conclusions of the SAMA analysis performed for North Annas initial license renewal. In the Draft SEIS, the NRC has not documented any effort to look for new and significant information beyond the work presented by Dominion or document in previous versions the GEIS or the NAPS supplemental EISs. This is not acceptable. The NRC should at a minimum review its own Generic Issues Program and the Office of Researchs ongoing research plan for relevant new information.
At Page F-26 (lines 42 - 43) the Draft SEIS states: In addition, given the low residual risk at North Anna, the substantial decrease in internal event CDF (emphasis added) Here the NRC fails to acknowledge the identified increase in seismic risk and the complete reliance on industry average fire risk evaluations as Dominion has not published any fire risk results. See discussion above.
Climate Change
Various sections of the Draft SEIS address climate change. However, the Draft SEIS does not address climate changes impacts on accident risks in Section 3.11.6.9 or Appendix F. This omission constitutes a significant deficiency in the Draft EIS because climate change demonstrably affects the frequency and intensity of some external events and therefore has the potential to significantly increase accident risks.
Moreover, the frequency and intensity of climate change effects are increasing over time. Given that the NRC is proposing to rely on the Draft SEIS for decisions that could affect reactor safety decades from now, the Draft SEIS must address these changing effects over the entire licensed lifetime of reactors, which may end 4 decades from now.
As discussed above, the Draft SEIS is already inadequate as a general matter for making broad generalizations about external event CDF based on extrapolations from internal event CDF values and
7
limited actual plant-specific values for external event CDF. Appendix F looks explicitly at external events focusing exclusively on seismic issues. It ignores other external events such as flooding, external fires (e.g., forest and wildfires), tornadoes, etc. Climate change has already started to increase the frequency and intensity of these events. See, for example, Climate change is probably increasing the intensity of tropical cyclones, March 31, 2021 NOAA, https://www.climate.gov/news-features/understanding-climate/climate-change-probably-increasing-intensity-tropical-cyclones; Climate Change Indicators:
Weather and Climate, EPA, https://www.epa.gov/climate-indicators/weather-climate; Global Warming and Hurricanes, NOAA Geophysical Fluid Dynamics Laboratory, April 11, 2023, https://www.gfdl.noaa.gov/global-warming-and-hurricanes/.
The NRC is well-aware of the issues of climate change and its impact on nuclear plant safety. After the Fukushima meltdowns, the NRC Office of Research initiated a research program to develop tools to assist in probabilistic and deterministic assessments of external hazards including seismic, high winds and flooding with a consideration of climate change. See NRC Probabilistic Flood Hazard Assessment Research Program Overview,, February 22 - 25, 2021 (ML21064A418) and Potential Impacts of Accelerated Climate Change, PNNL-24868, May 2016 (ML16208A282)). In addition, climate change has been a topic of discussion at the NRCs Regulatory Information Conference (RIC) in recent years. See Climate Change Impact on the Safety of Nuclear Installations, March 8-10, 2022 (ML22140A312)) &
Observations on Extreme Weather and Impacts on Nuclear Power Plants, EPRI ML22140A320, 2022).
Accident risk evaluations for climate change must be site-specific
The effects of climate change on accident risk are and will continue to be site-specific and not subject to generalization. Some reactors already have been identified as vulnerable to climate impacts and others
- like NAPS - have not been evaluated for their vulnerability. Given what we know about some U.S.
reactors, it is unacceptable not to provide a comparable analysis for NAPS.
Reactors for which climate vulnerability has been demonstrated or poses an unusual risk include Oconee, Turkey Point, and Duane Arnold. For example, the three reactors at the Oconee plant -- for which the NRC is now considering an application for subsequent license renewal -- lie downstream of two large dams. The design of the dams includes consideration of the maximum probable flood induc ed by the maximum probable precipitation (i.e., storm). Climate change has the potential to significantly increase the amount of precipitation falling on watersheds above the dams. Will the dams be able to pass these higher intensity storms and the resulting floods? See the attached declaration NRC Relicensing Crisis at Oconee Nuclear Station: Stop Duke from Sending Safety Over the Jocassee Dam for a thorough analysis.
Another example is the Turkey Point plant, located in a low-lying coastal area of South Florida. With climate change the already-occurring, sea level change will continue and possibly accelerate during the SLR period. Likewise, hurricane intensity, i.e., wind speed, rain fall and storm surge, will intensify.
As discussed below, the Duane Arnold plant in Iowa was prematurely and permanently shuttered after being hit with a Derecho with wind speeds exceeding 100mph. Climate change has been implicated in the severity of this extreme weather event (Hints of a derecho-climate change link, ten years after 2012 storm, Washington Post, June 29, 2022, https://www.washingtonpost.com/climate -
environment/2022/06/29/derecho -climate-change-severe-storm/)
8
Therefore, in order to provide a reasonably thorough and complete analysis of accident risks during the license renewal term, the Draft SEIS must address the continuing and growing contribution of climate change to accident risks at nuclear plants. And this evaluation must be conducted on a site-specific basis.
Effects of climate change considerations on Probabilistic Analysis
Climate change affects risk in two ways. First, it increases the likelihood or initiating event frequency of events. For example, increased storm frequency can lead to higher initiating event frequency for losses of offsite power (LOOPs). Second, climate change can increase the probability of failure of design features or mitigation equipment. A 2020 severe windstorm at the Duane Arnold plant (ML21139A091) illustrates this phenomenon. While the storm may or may not be directly attributable to climate change, it is a reasonable example of the type of severe weather effects that climate change can cause today and will cause in the future. In that case, a severe windstorm caused a loss of offsite power (LOOP). As a result of the LOOP, debris accumulated at the suction of the service water systems, which are necessary to cool the emergency diesel generators (EDGs) and the emergency core cooling system (ECCS) heat exchangers. The NRCs risk analysis of the event showed an increase in the failure probabilities of the service water system, the EDGs and the ECCS due to this climate -related external event. Consideration of these risks in an EIS would provide important information regarding climate -related accident risk as well as identification of mitigation measures to address those risks.
A third way that climate change affects risk analysis, which is unique to flooding risk, is the cliff edge effect. With most hazards if the severity is increased slightly, the stress on the system is increased somewhat proportionately. However, with many flood-related issues, a small increase in the hazard can cause a dramatic and often overwhelming impact on a structure. For example, a small increase in wave height could raise the flood height sufficiently to overtop a floodwall inundating the equipment the floodwall is designed to protect. Risk analyses for climate change-related flooding must look carefully at this cliff-edge phenomenon.
Finally, the National Academies under sponsorship of the National Oceanic and Atmospheric Administration (NOAA) has started a project to modernize the probable maximum precipitation (PMP) methodology (https://www.nationalacademies.org/our -work/modernizing-probable-maximum-precipitation-estimation#sectionSponsors). This project will consider approaches for estimating PMP in a changing climate, with the goal of recommending an updated approach, appropriate for decision -maker needs. PMP is a significant input into the design of critical infrastructure such as dam and reactor safety analysis directly and indirectly through its impact on probable maximum flood (PMF). The NRC is well aware of this effort as they have already participated in at least one of the initial project workshops.
PMP and PMF also impact reactor safety directly via their impact on local intense precipitation (LIP). The Draft SEIS is silent on this and is thus deficient. As this process is likely to take several years, if the SEIS cannot wait for resolution, then any plant issued a SLR prior to its resolution should be required to revisit the issue once the update is completed.
Multiunit Impacts
The Draft SEIS does not address the environmental impacts of concurrent multi -unit accidents. This is a significant omission, given the well-recognized independent contribution that multi -unit accidents make to accident risk. As discussed in a 2013 paper by NRC Staff member Suzanne Schroer and University of
9
Maryland Professor Dr. Mohammed Modarres, multi-unit site risk is neither formally nor adequately considered in either the regulatory or the commercial nuclear environment [citations omitted] despite the fact that the questions of multi-unit accident is not one of possibility, but of probability. Schroer and Modarres, An Event Classification Schema for Evaluating Site Risk in a Multi-Unit Nuclear Power Plant Probabilistic Risk Assessment, p. 1 (2013) (ML13217A335).
As recognized by Schroer and Modarres, the events at Fukushima Daiichi in 2011 underlined the significance and importance of accident events involving multiple units. Id, p. 1. And indeed, the NRC has been discussing how to address the issue of multi-unit nuclear power plant PRAs for many years, including a lessons learned report after the Chernobyl accident. Id. The Near Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (July 12, 2011) (ML111861807) specifically identified multi-unit accidents as an issue that should be investigated and addressed. As noted by the Task Force:
The accident at Fukushima has shown that prolonged SBO [station black out] and multiunit events are realities that must be addressed as part of EP [emergency planning]. While of low probability, these events have the potential for severe consequences that require an effective EP response. The Task Forces evaluation in this section focuses on a licensees capability to respond during these types of events. Currently, the United States has 29 single-unit sites, 33 dual-unit sites, and 3 triple-unit sites. The agency is currently reviewing new reactor applications that may add units to existing sites; however, no applicant has requested to bring the total number of units at a single site to more than four. In most cases, proposed quadruple-unit sites have physical separation between the two existing and the two proposed units.
Id., p. 51. While the NTTF focused its recommendations on safety improvements related to emergency planning, its conclusion that multi-unit accidents are realities with potentially severe consequences demonstrates their relevance to risk analysis for environmental impact studies.
The differences between single-unit PRAs and multi-unit PRAs are well-understood, as is the risk-significance of failing to address interdependent multi-unit events. As described in a recent paper by Taotao Zhou, Mohammad Modarres, Enrique López Droguett:
Conventional PRA studies have traditionally been restricted to single reactor units and are referred to as single-unit PRAs (SUPRAs). The SUPRAs include accident scenarios exclusive to one reactor unit, assuming the effects of other units are not critical. Hence, SUPRAs only consider the dependencies between the structures, systems, and components (SSCs) within a single reactor unit. These dependencies, referred to as intra-unit dependencies, are likely to induce multiple failure events that may overcome redundancies or diversities and ultimately lead to a class of SSC failures called dependent failures. Although these dependent events are usually much less frequent than the independent events, they have proven to be the most critical contributors to the likelihood of reactor core damage, environmental radioactive exposure, and overall plant risk. Typically, the influence of these dependencies is explicitly modeled in the PRA event tree and fault tree logics or implicitly treated as the type of dependencies commonly referred to as common cause failure events.
Multi-unit nuclear power plant probabilistic risk assessment: A comprehensive survey, Taotao Zhou, Mohammad Modarres, Enrique López Droguett, Reliability Engineering & System Safety, Volume 213, September 2021 (emphasis added).
10
(https://www.sciencedirect.com/science/article/abs/pii/S0951832021003070) As further explained by Profs. Zhou, Modarres, and Droguett:
These inter-unit dependencies can play critical roles in nuclear accident risks with the possibility of multiple core damages, including damages to the spent fuel pool and other radioactive waste storage facilities. Proper characterization of these site-level dependencies is thus critical to obtain an accurate risk profile of a nuclear power plant site. Examples of these inter-unit dependencies include certain initiating events simultaneously occurring in multiple units, a transient event in one unit affecting some or all of the other units, the proximity of the units to each other, shared structures, components (e.g., shared batteries and diesel generators),
common operation practices, and substantial procedural and other organizational similarities.
(emphasis added). Three important conclusions can be drawn from the study of multi -unit accidents by the NRC and independent researchers. First, multi-unit accident risks - including risks to reactors and fuel storage pools - are well-understood as reasonably foreseeable. Second, multi-unit accidents have unique characteristics that are not bounded by single -unit accident risk studies. Finally, the risks of multi-unit accidents are unique to reactor sites, and must consider the relative location of reactor units, fuel storage pools, and other onsite facilities. Therefore, multi-unit accident risks must be independently evaluated for each separate reactor site for which license renewal is considered.
11
CURRICULUM VITAE JEFFREY T. MITMAN Poolesville, MD February 22, 2024
QUALIFICATIONS Reliability and risk analyst with more than 40 years experience in the nuclear industry. Skills include evaluation and modeling of probabilistic risk analyses (PRA) and management of PRA projects and teams.
Highly experienced in low power and shutdown (LPSD) risk modeling issues. Solid record of bringing projects in on schedule and budget.
MAJOR ACCOMPLISHMENTS
- Transitioned NRC to detailed PRA models for LPSD significance determinations process evaluations.
- Guided development of and managed industrys first configuration risk management software tool.
- Obtained regulatory approval of EPRIs risk informed in-service inspection (RI-ISI) methodology.
- Managed first PRA of bolted spent fuel storage cask.
EXPERIENCE PRIVATE CONSULTANT (Poolesville, MD)
Nuclear risk analyst 2021-Present
- Reviewed Oconee Subsequent License Renewal application and prepared technical report on adequacy of environmental and safety analyses to address flooding risks.
- Reviewed and submitted comments on NRCs draft (2023) Generic Environmental Impact Statement (NUREG-1437 Revision 2).
US NUCLEAR REGULATORY COMMISSION (Rockville, MD) 2005 - 2021 Senior Reliability and Risk Analyst (NRC Office of Nuclear Reactor Regulation)
- Conducted Significance Determination Process (SDP) evaluations of reactor events including development and/or modification of required models.
- Served as lead analyst for LPSD event issues and concerns.
- Guided development of shutdown Standardized Plant Analysis Risk (SPAR) models.
- Conducted Human Reliability Analysis (HRA).
- Evaluated external event risk from dam failures.
- Served on NRCs Japan Team (part of USAID disaster assistance response team for Fukushima Daiichi accident), providing technical advice and support through the U.S. Ambassador to Japanese government.
- Participated in post NRCs Fukushima Near Term Task Force (NTTF) flooding guidance development.
- Developed NRCs guidance on crediting FLEX in risk-informed regulatory applications.
- Advised NRC National Fire Protection Association (NFPA) 805 team on issues related to shutdown fire risk.
- Performed evaluations of risk informed license applications.
Reliability and Risk Analyst (NRC Office of Nuclear Regulatory Research)
- Project Manager for the development of shutdown SPAR models ERIN ENGINEERING AND RESEARCH, INC. (Walnut Creek, CA) 2004 - 2005 Lead Senior Engineer
- Prepared configuration risk management evaluation of at-power fire risk.
- Prepared configuration risk management evaluation of loss of offsite power.
ABE STAFFING SERVICES (Palo Alto, CA) 2003 - 2005 Consultant to EPRI
- Brought project to closure involving Dry Cask Storage PRA project and team, involving Transnuclear bolted cask containing PWR fuel.
Jeffrey T. Mitman l Page 2
EPRI (Palo Alto, CA) 1998 - 2003 Project Manager
- Outage Risk Assessment and Management (ORAM-Sentinel)
- Grew first of a kind software application for performing configuration risk management in nuclear power plants.
- Conducted research in low power and shutdown risk; shutdown initiating event and event frequency derivation.
- Delivered multiple versions (including alpha, beta & production), testing and full documentation.
- Administered utility user group, marketing, contract preparation, technology transfer, technical report publication and training.
- Actively managed both development and application contracts with multiple suppliers and customers.
Managed annual $1M budget.
- Dry Cask Storage PRA: Initiated innovative analysis of Transnuclear cask containing PWR fuel.
- Managed unique team with diverse experience in both cask design and PRA backgrounds.
- Risk Informed In-service Inspections Project (RI-ISI): Lead team in obtaining regulatory approval of methodology to safely reduce piping weld inspection requirements using combination of probabilistic and degradation analysis.
- Responsible for methodology finalization and acceptance by industry and U.S. NRC.
- Conducted marketing, sales, contract preparation, technology transfer, training and technical report publication.
- Actively managed both development and application contracts with both suppliers and customers.
Managed annual $1M budget.
- Human Reliability Analysis Project: Managed project to bring consistency to on industry use of HRA methods.
- Responsible for EPRI HRA area, including development of HRA Calculator software and establishment of associated users group.
ERIN ENGINEERING AND RESEARCH, INC. (Palo Alto, CA) 1992 - 1998 Lead Senior Engineer Collaborated with EPRI ORAM-SENTINEL Project Manager in project development and administration, user group administration, contract preparation, technology transfer workshops, technical report generation and editing. Performed ORAM analysis of the Diablo Canyon plant. Performed ORAM Probabilistic Analysis of Perry spent fuel pool. Drafted and edited ORAM V2.0 Users Manual. Assisted in ORAM-SENTINEL software design, performed software debugging. Principle researcher and author of BWR outage contingency report. Prepared marketing and training, materials.
ABB IMPELL CORPORATION (King of Prussia, PA) 1990 - 1992 Lead Senior Engineer
- Design Basis Documentation: directed team of three engineers to review PECO Feedwater System Design. Wrote Design Basis Documentation reports for Limerick and Peach Bottom power plants, identifying licensing and design concerns by reviewing the system design as documented in drawings, calculations, vendor manuals, Technical Specifications, UFSAR, SER, SRP, 10CFR50.59 safety evaluations etc. and by interfacing with utility engineering personnel. Prepared Engineering Change Requests as necessary.
- Shift Outages: during Limerick Nuclear Power Plant refueling / maintenance outage. Coordinated all shift maintenance work and testing. Collaborated with all groups in power plant, allocating resources as needed to maintain schedule and reporting to senior plant outage management. Performed system reviews prior to placing them back in service. Conducted shift outage meetings. Tracked work group performance against schedule. Advised utility management on techniques for schedule and outage organizational improvements.
Jeffrey T. Mitman l Page 3
GENERAL ELECTRIC COMPANY (San Jose, CA) Experience Prior to 1990 Startup-Test Engineer
- Shift Startup Engineer: During power ascension phase coordinated all system testing on shift and startup interface with operations. During preoperational phase, acted as operations shift supervisor responsible for coordinating all system testing and flushing on shift from main control room. Updated senior utility management daily on testing status.
EDUCATION / PROFESSIONAL DEVELOPMENT
- BSE, Nuclear Engineering, University of Michigan, Ann Arbor, MI
- Introductory VBA class, University of California, Berkeley, CA
- Misc. business courses at various colleges and universities
- Senior Reactor Operator Certified
- GE Station Nuclear Engineering
- Effective Utilization of PSA, ERIN Engineering & Research, Walnut Creek, CA.
PROFESSIONAL ASSOCIATIONS
- American Nuclear Society (ANS) member since 1978.
- ANS Nuclear elected member of Installation Safety Division Executive Committee 2015 to 2021.
- ANS/ASME Risk Informed Standards Writing Group on Shutdown PRA Standard.
- ASME Section XI, Working Group on Implementation of Risk Based Examination.
- MIT Professional Summer Programs Guest Lecturer at Risk-Informed Operational Decision Management Course.
PAPERS
- 1. Technical Challenges Associated with Shutdown Risk when Licensing Advanced Light Water Reactors, PSAM12 2014. Co-author.
- 2. Comparing Various HRA Methods to Evaluate Their Impact on the results of a Shutdown Risk Analysis during PWR Reduced Inventory, PSAM11 2012. Co-author.
- 3. Uncertainty Analysis for Large Dam Failure Frequencies Based on Historical Data, PSAM11 2012. Co-author.
- 4. An Assessment of Large Dam Failure Frequencies Based on US Historical Data, PSA 2011. Co-author.
- 5. Generic Failure Rate Evaluation for Jocassee Dam, US NRC (ML13039A084), 2010. Co-author.
- 6. Development of PRA Model for BWR Shutdown Modes 4 and 5 Integrated in SPAR Model, to be presented at PSAM10 2010. Co-author.
- 7. Development of Standardized Probabilistic Risk Assessment Models for Shutdown Operations Integrated in SPAR Level 1 Model, PSAM9 2008. Co-author.
- 8. PRA of Bolted Dry Spent Fuel Storage Cask, Presented at ICONE12. 2004. Co-author.
- 9. Low Power and Shutdown Risk Assessment Benchmarking, Presented at PSA 02 2002. Co-author.
- 11. Derivation of Shutdown Initiating Event Frequencies, Presented at PSAM5 2000. Co-author.
- 12. Quantitative Assessment of a Risk Informed Inspection Strategy for BWR Weld Overlays, Presented at ICONE 8 2000. Co-author.
- 13. EPRI RI-ISI Methodology and the Risk Impacts of Implementation, Presented at SMiRT 11 1999. Co-author.
- 14. Application of Markov Models and Service Data to Evaluate the Influence of Inspection on Pipe Rupture Frequencies published. PVP 1999. Co-author.
- 15. Progress in Risk Evaluation of Outages, International Conference on the Commercial and Operational Benefits of PSA. 1997. Co-author.
- 16. Control of Reactor Vessel Temperature/Pressure during Shutdown, GE SIL 357. June 1981. Co-author.
Jeffrey T. Mitman l Page 4
SOFTWARE
REPORTS / STANDARDS
- 1. Requirements for Low Power and Shutdown PRA - ANS/ASME-58.22-2014 (Trial Use Standard).
- 2. Probabilistic Risk Assessment (PRA) of Bolted Storage Casks: Quantification and Analysis Report, EPRI 2003. 1002877. PM.
2002. 1003465. PM and principal investigator.
- 5. Guidance for Incorporating Organizational Factors into Nuclear Power Plant Risk Assessments: Phase 1 Workshop. EPRI and U.S. DOE 2002. 1003322. PM.
- 6. An Analysis of Loss of Decay Heat Removal Trends and Initiating event Frequencies (1989-2000):
- 7. Piping System Failure Rates and Rupture Frequencies for Use in Risk Informed In-service Inspection Applications: TR-111880-NP, EPRI 2000. 1001044. PM.
- 8. Application of Risk-Informed Inservice Inspection Alternative Element Selection Criteria. EPRI, Charlotte NC: 2000. TE-11482. PM.
- 9. Revised Risk-Informed Inservice Inspection Evaluation Procedure, EPRI 1999. TR-112657 Revision B-A. PM & co-author.
- 10. Piping System Failure Rates and Rupture Frequencies for Use in Risk Informed In-service Inspection Applications, EPRI 1999. TR-111880. PM.
- 11. Comparison between EDF and EPRI of Pipe Inspection Optimization Methods, EPRI Palo Alto, CA; Electricite de France, Paris, France: 1999. TR-113315. PM.
- 13. Evaluation of Pipe Failure Potential via Degradation Mechanism Assessment, EPRI Palo Alto, CA:
1998. TR-110157. PM.
- 15. Piping System Reliability Models and Database for used in Risk Informed Inservice Inspection Applications, EPRI 1998. TR-110161. PM.
TR-102975. PM.
- 25. Outage Risk Assessment and Management Implementation at Fermi 2, EPRI 1997. TR-109013. Co-author.
TR-102973. Principal investigator.