ML111570508

Declaration of Dr. Thomas L. Sowdon and Dr. Kevin R. O'Kula in Support of Entergy'S Answer Opposing Pilgrim Watch Request for Hearing on Post-Fukushima SAMA Contention

ML111570508
Person / Time
Site:	Pilgrim
Issue date:	06/06/2011
From:	O'Kula K, Sowdon T Entergy Nuclear Operations, Entergy Nuclear Generation Co, URS Corp
To:	Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML111570507	List: ML111570508 ML111570509
References
RAS 20429, 50-293-LR, ASLBP-06-848-02-LR
Download: ML111570508 (38)

Text

June 6, 2011 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Generation Company and ) Docket No. 50-293-LR Entergy Nuclear Operations, Inc. ) ASLBP No. 06-848-02-LR

)

(Pilgrim Nuclear Power Station) )

DECLARATION OF DR. THOMAS L. SOWDON AND DR. KEVIN R. OKULA IN SUP-PORT OF ENTERGYS ANSWER OPPOSING PILGRIM WATCH REQUEST FOR HEARING ON POST-FUKUSHIMA SAMA CONTENTION Dr. Thomas L. Sowdon (TLS) and Dr. Kevin R. OKula (KRO) state as follows un-der penalties of perjury:

I. INTRODUCTION A. Entergy Declarants

1. Dr. Thomas L. Sowdon

1. (TLS) I am Senior Project Manager with Entergy Nuclear Generation Company and Entergy Nuclear Operations (hereinafter and collectively, Entergy). In that capacity, I am responsible for managing a variety of projects related to both onsite and offsite emergency re-sponse planning

2. (TLS) My professional and educational experience is summarized in the Curricu-lum Vitae attached as Exhibit 1 to this Declaration. I have 35 years of experience in the nuclear industry in various positions related primarily to Radiation Protection, Health Physics and Emer-gency Planning responsibilities. I hold a Bachelors Degree in Nuclear Engineering from the University of Lowell, a Masters Degree in Radiation Health Physics from the Massachusetts

Institute of Technology, and a Doctor of Science Degree in Occupation and Environmental Epi-demiology from the Harvard School of Public Health. I am also a Certified Health Physicist in the power reactor specialty.

3. (TLS) In my capacity as an Entergy Project Manager, I am knowledgeable of the Severe Accident Mitigation Alternatives (SAMA) analyses prepared by Entergy for the Envi-ronmental Report that is part of the Pilgrim Nuclear Power Station (PNPS) license renewal ap-plication, as well as Entergys responses to the Nuclear Regulatory Commission (NRC) Staff Requests for Additional Information related to the SAMA analyses. I have reviewed the SAMA analysis and provided advice and input on its preparation.

4. (TLS) In my professional career, I have performed and have been responsible for the analyses of the radiological consequences of reactor accidents for both boiling and pressur-ized water reactors. This includes experience with Stone and Webster Radiation Protection Group, with Boston Edison Company as a Senior Radiological Engineer, and also with Entergy as Chief Radiological Scientist. Such analyses included the assessment of accident source terms, in-plant transport and behavior of fission and activation products, and environmental transport and biological effects of radioactive materials released to the environment.

2. Dr. Kevin R. OKula

5. (KRO) I am an Advisory Engineer with URS Safety Management Solutions (URS) LLC. My professional and educational experience is summarized in the Curriculum Vitae attached as Exhibit 2 to this Declaration.

6. (KRO) I have over 28 years of experience as a technical professional and manager in the areas of safety analysis methods and guidance development, computer code evaluation and 2

verification, probabilistic safety assessment, deterministic and probabilistic accident and conse-quence analysis applications for reactor and non-reactor nuclear facilities, source term evalua-tion, risk management, reactor materials dosimetry, and shielding. I obtained a Bachelor of Sci-ence degree in Applied and Engineering Physics from Cornell University in 1975, a Master of Science degree in Nuclear Engineering from the University of Wisconsin in 1977, and a Ph.D. in Nuclear Engineering from the University of Wisconsin in 1984.

7. (KRO) My education and training in Nuclear Engineering includes understanding the conditions under which uranium fuel materials are able to sustain a nuclear chain reaction. I have previous Probabilistic Safety Assessment (PSA) and severe accident analysis experience in analyzing reactor core phenomena under accident conditions, including scenarios where core degradation has occurred and the potential for recriticality exists. The severe accident analysis work in these efforts has included evaluating the fission products behavior and estimating the subsequent release of radionuclides into the environment.

8. (KRO) I have over 22 years of experience in using the MELCOR Accident Conse-quence Code System (MACCS) and the MACCS2 Computer Codes, and have taught MACCS2 training courses for the Department of Energy (DOE) at Lawrence Livermore Na-tional Laboratory, Los Alamos National Laboratory, Idaho National Laboratory and at DOE Safety Analysis Workshops. I was the lead author of a DOE guidance document on the use of MACCS2.1 Also, I am a member of the State-of-the-Art Reactor Consequence Analysis (SOARCA) Project Peer Review Committee that provides recommendations on applying 1

MACCS2 Computer Code Application Guidance for Documented Safety Analysis, DOE-EH-4.2.1.3-Final MACCS2 Code Guidance, Final Report, U.S. Department of Energy, Washington, DC (June 2004).

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MACCS2 in the context of accident phenomena and subsequent off-site consequences in the context of severe reactor accidents, to Sandia National Laboratories (SNL) and the NRC.

B. Pilgrim Watchs Proposed Late-Filed Contention on Fukushima

9. (TLS, KRO) We have reviewed and are familiar with Pilgrim Watchs late-filed contention concerning alleged lessons learned from the March 11, 2011 accident at Japans Fu-kushima Daiichi reactor complex, which was filed on May 12, 2011 in the NRC licensing pro-ceeding for the PNPS license renewal.2

10. (TLS, KRO) Pilgrim Watchs late-filed contention states:

The Environmental Report is inadequate post Fukushima Daiichi because En-tergys SAMA analysis ignores new and significant lessons learned regarding the possible off-site radiological and economic consequences in a severe accident.

PW Request at 1. In particular, Pilgrim Watch argues that the SAMA analysis code (MACCS2)

. . . underestimates consequences for a number of reasons two of which are based on lessons learned on new and significant information from Fukushima. PW Request at 3.

11. (TLS, KRO) The two bases that Pilgrim Watch asserts for its late-filed contention based on allegedly new and significant information from Fukushima are as follows:

1. The code limits the total duration of a radioactive release to no more than four (4) days, if the Applicant chooses to use four plumes occurring sequentially over a four day period. Entergy chose not to take that option and limited its analysis to a single plume having a total duration of the maximum-allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In any case either a 24-hour plume or a four-day plume is insufficient duration in light of lessons learned from Fukushima. The Fukushima crisis now stretches into its sec-ond month and shows that releases can extend into many days, weeks, and months; a longer release can cause offsite consequences that will affect cost-benefit analyses.

2. Computer codes in use are totally incapable of modeling an 8-week chain reac-tion that continues after a scram. MACCS2 is no exception. Like all the computer 2

Pilgrim Watch Request for Hearing on Post Fukushima SAMA Contention (May 12, 2011) (PW Request).

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codes, it is incapable of modeling a severe accident release that lasts 8 weeks or longer. The MACCS2 code used by Entergy, and all other codes, assumes that the reactor is scrammed when the accident begins, the reactor is scrammed, and that the production of all fission products ceases at that time.

PW Request at 3 (footnotes omitted).

12. (TLS, KRO) For support, Pilgrim Watch relies on an April 28, 2011 article au-thored by an unidentified individual posted to the Gerson Lehrman Group website. PW Request at 8-9. This article purports to analyze data, presented in graphs, from Tokyo Electric Power Company (TEPCO) concerning the levels of radioactive isotope iodine-131 (I-131) and two isotopes of cesium (Cs), Cs-134 and Cs-137, detected in water from sub-drains under each of the six Fukushima Daiichi reactors. Id. The graphs are reproduced at pages 10-13 of the PW Request. Among other things, the article asserts that the data from Unit 2 demonstrates un-equivocal[ly] that recriticalities are occurring because instead of seeing [an] expected decrease in I-131 levels relative to Cs-134 and Cs-137 . . . I-131 was seen to be increasing, instead of de-creasing as the physics said it should because of I-131s much shorter half-life. PW Request at 8, 9. The article goes on to assert that, [t]he only possible source of I-131 would be pockets of molten core in the Unit 2 RPV settled in such a way that the boron injected into the water is in-sufficient to stop the localized criticalities. Id. at 9 (footnote omitted).

13. (TLS, KRO) Our Declaration addresses the claims raised by Pilgrim Watch con-cerning the adequacy of the Pilgrim SAMA analysis in light of Fukushima. In summary, the evi-dence cited by Pilgrim Watch is not unequivocal that post-scram criticalities have occurred, or are occurring, in any of the damaged Fukushima Daiichi reactors. Many phenomena other than post-scram criticalities could give rise to the observed levels of I-131, Cs-134, and Cs-137 in plant systems and/or effluents. In any event, the potential for post-scram criticalities to occur in a damaged reactor core is not new information. Moreover, any radioactive releases arising from 5

the post-scram criticality events alleged by Pilgrim Watch would have comparatively small im-pacts that would not alter the Pilgrim SAMA analysis. Pilgrims SAMA analysis bounds the low-level, longer duration radioactive releases to the environment at Fukushima, whether or not caused by low energy, intermittent recriticalities in a damaged reactor core.

II. ALLEGED POST-ACCIDENT RECRITICALITIES A. Pilgrim Watch Alleges No Conclusive Evidence that Post-Scram Recriticali-ties Have Occurred

14. (TLS) Pilgrim Watch contends that high levels of I-131 have continued to be pro-duced after the Fukushima reactors were shutdown following the March 11, 2011 earthquake be-cause, [i]f criticality had stopped after the reactors scrammed, the I-131 would have largely de-cayed and would not [] be at the levels we have seen reported [] that exceed the Cesium read-ings. PW Request at 4. However, the evidence cited by Pilgrim Watch is far from conclusive that any post-scram criticalities have occurred at Fukushima. While it is possible that a recritical configuration developed periodically or intermittently in small, localized portions of the reactor core debris, many other phenomena could give rise to the relatively higher levels of I-131 re-ported in some locations at Fukushima. For example, the melting and boiling point differences and other chemical property differences between iodine and cesium, the timing of fuel becoming uncovered and percentage of fuel becoming damaged, thermal conditions, the geometry of the fuel, and other factors can all play a role.

15. (TLS) In particular, it is well known that iodine and cesium behave very differently in both wet and dry environments. First, iodine is far more volatile than cesium. Iodine boils at 365 degrees Fahrenheit (oF), while cesium boils at 1240 oF. This means that, as the reactor units cooled below the boiling point of cesium, the temperature would still be sufficient for the iodine to vaporize and be carried away from the reactor core and released into in-plant systems 6

and the environment. In addition, iodine and cesium have different levels of solubility under dif-ferent chemical conditions. These differences in the chemical properties of iodine and cesium could easily explain the comparatively higher levels of I-131 detected in the water in the sub-drains of the damaged Fukushima reactors.

16. (TLS) Pilgrim Watch attempts to argue that there should have been essentially no I-131 remaining when TEPCO reported the relative I-131 and Cs-137 data on April 27, 2011, because a period of about 40 days had elapsed since any of the reactors were critical. See PW Request at 9 (Because I-131 has no long-lived parent to feed it by parent decay, the levels of I-131 in scrammed reactors with intact geometry will decrease exponentially with an 8-day half-life; after 5 half-lives (40 days) the I-131 levels are only 3% of what they were at scram) (quot-ing Gerson Lehrman Article). This argument, although arithmetically correct, is misleading be-cause accounting for 5 or 6 half-lives of radioactive decay in this context reveals that substantial quantities of I-131 were still available for release into the environment even in the absence of recriticality after the initial event. At the time of the event on March 11, 2011, the total I-131 inventory in Units 1, 2 and 3 was on the order of about 1.7 x108 curies.3 After 5 or 6 half-lives, there would still be over 2.6 x 106 curies of I-131 remaining. This is equivalent to about 1017 Bq of activity. In short, the amounts of I-131 and Cs-137 observed during the period in question can be explained by the very substantial surviving inventory of I-131 combined with widely and rap-idly varying conditions (temperature, pressure, presence of salt water) in the reactor and support-ing systems, when considered in light of the substantially different chemical and physical charac-teristics of iodine versus cesium.

3 See NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents (Oct. 1988) at 2-3 (Table 2.2).

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17. (TLS) Thus, contrary to the claims made in the Gerson Lehrman Group article par-roted by Pilgrim Watch, the alleged detection of I-131 in elevated concentrations relative to the cesium isotopes is far from a conclusive or unequivocal indication that post-scram criticalities have occurred, or perhaps continue to occur, since the reactors shut down on March 11, 2011 af-ter the earthquake struck. The analysis provided in the article ignores the well-known differ-ences in chemical behavior between iodine and cesium, and the data simply do not support the articles assertion that post-scram criticalities are the only possible source for the elevated presence of I-131. Although it is possible that localized recriticalities could have occurred within portions of the molten cores that presumably have settled at the bases of the reactor pres-sure vessels, the data discussed in the Gerson Lehrman Group article provides no direct evidence of continued recriticalities in the Fukushima reactors, and the observed levels of I-131, Cs-134 and Cs-137 are readily explainable by phenomena other than recriticality.

18. (TLS) Indeed, if continued recriticalities were occurring, one would expect that two shorter-lived isotopes of iodine (I-132, I-134) that would have also been produced in any post-scram criticalities would also have been detected. The half-lives of these isotopes are 2.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 52.6 minutes, respectively. It is simply not possible to determine whether the detected I-131 resulted from a recent fission without comparing it to other isotopes of the same chemical which would all have the same boiling temperature, solubility in water, and other chemical prop-erties.

19. (TLS) In sum, the information on which Pilgrim Watch relies is by no means conclusive that ongoing criticalities are occurring at Fukushima Daiichi. There are many more factors that would affect the presence and relative levels of I-131 and Cs-134/137 than continued criticality. Pilgrim Watch has put forward no direct evidence (such as the presence of I-132 and 8

I-134) indicating that the I-131 detected in TEPCOs measurements is the result of recent or on-going recriticalities. Given the well-understood different phenomena that could give rise to the observed levels of I-131 and Cs-134/137 in the reactor sub-drains, the mere assertion that the stated evidence unequivocally indicates the existence of ongoing criticalities demonstrates that the unidentified author of the Gerson Lehrman Group article is fundamentally unqualified to opine on the subject.

B. The Potential for a Post-Scram Recriticality Is Not New Information

20. (KRO) It should also be noted that the potential for a recriticality to occur in a reac-tor core under severe accident conditions is not new information. Recriticality has been assessed in multiple severe accident safety studies, beginning as early as the WASH-1400 study published in 1975.4 Later severe accident analysis, such as the NRCs NUREG-1150 study,5 have consid-ered the potential for recriticality in the context of: (1) leading to a release of energy sufficient to fail the reactor vessel, or other reactor structure, system, or component; and (2) the subsequent release of radioactivity. For example, the NUREG-1150 study noted that in some BWR accident sequences, a period of time exists when the control blades may have melted and relocated while the fuel pellets are essentially in their normal configuration.6 Under these specific circum-stances, adding water to the core could potentially result in a critical condition; however, the possibility of an energetic excursion with the potential to fail the reactor vessel was assessed to be small.

4 WASH-1400 (NUREG-75/014), Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants (Oct. 1975).

5 NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, Final Summary Re-port (Dec. 1990) (NUREG-1150).

6 NUREG-1150, Vol. 3, App. D at D-27.

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21. (KRO) Supporting NUREG-1150 were a number of topical reports including one from Pacific Northwest Laboratory (NUREG/CR-5653, Recriticality in a BWR Following a Core Damage Event (PNL-7476) (Dec. 1990). This document analyzed the potential for recriticality occurring in BWRs upon the melting of the control blades while the fuel remained in its normal configuration and water was still present or added. This configuration could be supported for only a limited time (on the order of hours) after shutdown. Another aspect of the study was to describe the core debris that would form in later stages of the accident sequence. While debris beds could be formed, insufficient water would likely be available to support recriticality in the debris, and thus recriticality would be much less likely. Low levels of recriticality might be pos-sible intermittently, but over time, further spreading of the fuel debris and its dilution would make the material less likely to become recritical.

22. (KRO) Thus, the potential for recriticality occurring in severe accidents is an issue that has been looked at and assessed for decades. Even assuming that recriticality has occurred at Fukushima, it is not new information arising from the Fukushima events as claimed by Pilgrim Watch.

III. RADIOACTIVE RELEASE CHARACTERISTICS ASSUMED IN PILGRIMS SAMA ANALYSIS

23. (KRO) I agree with Dr. Sowdon that the evidence cited by Pilgrim Watch is not unequivocal that post-scram recriticalities have either occurred, or are occurring, in any of the damaged Fukushima Daiichi reactors in that many other phenomena other than post-scram re-criticalities could give rise to the observed levels of I-131. In any event, however, even assum-ing that post-scram recriticalities have occurred, or are occurring, any radioactive releases asso-ciated with these post-scram recriticality events, as well as any other ongoing low-level releases of radioactivity from Fukushima (e.g., due to evaporation and resuspension of radioactivity from 10

the Fukushima facilities and surrounding site), would be orders of magnitude lower than the large airborne releases resulting from different energetic severe accident release events postu-lated for the Pilgrim SAMA analysis. The Pilgrim SAMA analysis postulates a range of ener-getic severe accident release events with sufficient energy to breach the reactor engineering safety systems or barriers (e.g., breach the reactor vessel or fail the containment structure) or lead to their bypass. These include events that result in large fractions of the radioactive inventory of the core being released into the atmosphere and dispersed throughout the 50-mile SAMA region that would far exceed any of the ongoing low-level releases of radioactivity from Fukushima.

24. (KRO) Therefore, the type of releases that Pilgrim Watch claims need to be evalu-ated - i.e., low-magnitude releases that extend into many days, weeks, and months (PW Re-quest at 3-4, 7) - would have no measureable impacts on the results of the SAMA analysis com-pared to the large airborne releases due to energetic events assumed in the Pilgrim SAMA analy-sis, particularly in the 10 to 50 mile range from the Pilgrim plant where most of the population and economic consequences that drive the Pilgrim cost-benefit analysis occur. The large air-borne radioactive source terms resulting from the energetic severe accident events assumed for the Pilgrim SAMA analysis more than bound the radioactive releases from Fukushima, including any ongoing low-level releases to environment and any intermittent, localized recriticality events occurring within the damaged Fukushima reactor facilities. Therefore, Pilgrim Watchs claims would have no material effect on the Pilgrim SAMA analysis.

A. Pilgrim Watchs Claims Will Not Result in Consequences Sufficient to Alter the SAMA Cost-Benefit Analysis

25. (KRO) None of Pilgrim Watchs claims of ongoing releases or post-scram critical-ities concern large releases of radioactivity into the atmosphere caused by energetic events such as those postulated in the Pilgrim SAMA analysis. Such large releases would be required to re-11

sult in significant changes to the consequences important to the Pilgrim SAMA analysis, espe-cially in the range of ten to fifty miles. It is in this range where most of the population and eco-nomic consequences occur that drive the Pilgrim SAMA cost-benefit analysis. Because of the relatively minimal impact from the continued long-term low level releases such as those specu-lated by Pilgrim Watch to occur due to post-scram criticalities, such releases would not result in the additional costs necessary to alter the Pilgrim SAMA cost-benefit analysis and cause any ad-ditional SAMAs to become cost-beneficial. As discussed in the Entergy Contention 3 testimony, for the next SAMA to become potentially cost-beneficial, the benefit (risk averted) would need to increase by more than a factor of two, i.e., more than 100%.7

26. (KRO) As discussed in the Entergy Contention 3 Testimony at A15, a severe acci-dent is a beyond design basis accident that could result in substantial damage to the reactor core.

The SAMA analysis is a probabilistic analysis focused on long-term and spatially averaged im-pacts from severe accident events for the purpose of making cost-benefit evaluations. The analy-sis simulates the atmospheric transport, dispersion and deposition from a set of postulated radio-logical releases to predict the probabilistic consequences over the 50-mile region around the site.

In other words, a SAMA analysis is interested in average, long-term impacts such as population dose and economic cost consequences in a fifty-mile region from highly unlikely, severe acci-dent events.

27. (KRO) As also discussed in the Entergy Contention 3 Testimony at A18, the first step in the SAMA analysis is to determine the total severe accident risks, consisting of the off-site dose and economic impacts. Severe accident risk is determined using plant specific prob-7 Testimony of Dr. Kevin R. OKula and Dr. Steven R. Hanna on Meteorological Matters Pertaining to Pilgrim Watch Contention 3 (Jan. 3, 2011) (Exhibit No. ENT000001) (Entergy Contention 3 Testimony) at A47.

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abilistic safety assessment (PSA) models, also referred to as probabilistic risk assessment (PRA) models, to assess what can go wrong, how likely it is, and what are the resulting conse-quences. Other relevant background facts from the Entergy Contention 3 Testimony for pur-poses of the discussion here include:

• Entergy used the MACCS2 code to calculate consequence and risk values necessary for a SAMA analysis. The key consequence values of interest computed by MACCS2 are: (1) total off-site population dose (in units of person-sievert); and the (2) total off-site economic cost (in units of dollars). Entergy Contention 3 Testimony at A23.

• In order to obtain corresponding risk values for population dose and off-site economic costs, the off-site population dose and off-site economic cost consequence values are multiplied (outside of the MACCS2 code) by the calculated severe accident fre-quency results obtained from the plant-specific PSA and related information. This re-sults in the key risk values of interest for determining potentially cost-beneficial SAMAs, i.e., (1) population dose risk (PDR) in units of person-rem/year; and (2) the off-site economic cost risk (OECR) in units of dollars/year. Entergy Contention 3 Testimony at A23.

• The Pilgrim SAMA analysis considered 19 collapsed accident progression bins (CAPBs) or accident scenario source terms analyzed in the Pilgrim Plant specific PSA. As discussed further below in this Declaration, these were developed for the various postulated Pilgrim plant damage and by-pass configurations and represent a range of severe accident releases from small to very large. For each of the 19 CAPBs, a series of simulations were run using the MACCS2 code to evaluate postu-lated consequences under different meteorological conditions using Pilgrim site-specific meteorological data. The mean or average consequence results obtained for each of the 19 CAPBs were multiplied by their frequency of occurrence, and then summed to yield the overall PDR and OECR for the Pilgrim SAMA analysis. En-tergy Contention 3 Testimony at A37.

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28. (KRO) As discussed in the Entergy Contention 3 Testimony at A43, the results of the Pilgrim SAMA analysis shows that over 95% of the population dose risk and about 94% of the off-site economic cost risk occurs in the 10 to 50 mile range, and that 83% of the population dose risk and 79% of the off-site economic cost risk occurs in the 20 to 50 mile range. Table 1 below (Table 3 reproduced from the Entergy Contention 3 Testimony) shows the percentage of contribution to the Pilgrim SAMA PDR and OECR by distance interval from the Pilgrim plant.

Table 1. Contribution to Pilgrim SAMA PDR and OECR by Distance Interval8 Ring Distance In- PDR OECR terval (0-10 miles) 4.22% 6.18%

(10-20 miles) 12.42% 14.84%

(20-30 miles) 27.40% 26.09%

(30-40 miles) 37.83% 37.39%

(40-50 miles) 18.12% 15.50%

TOTAL 100.00% 100.00%

In short, population dose and off-site economic cost consequences for the Pilgrim SAMA analy-sis are dominated by exposure and contamination incurred in the 10-50 mile region due to the intersection of high population levels with the high exposure and contamination conditions, driven by the large airborne releases assumed in the Pilgrim SAMA analysis. Entergy Conten-tion 3 Testimony at A43.

29. (KRO) Any post-scram recriticality events that may have occurred or are still oc-curring at Fukushima would add little to the overall releases caused by the energetic events, such as the hydrogen explosions that occurred at the Fukushima facilities in the first week after the earthquake and tsunami. Similarly, any post-scram recriticality events would add little to the 8

Table 3 in the Entergy Contention 3 Testimony.

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overall releases due to the energetic events assumed in the Pilgrim SAMA analysis. This arises from a host of technical reasons not recognized or addressed in Pilgrim Watchs proposed new contention. These include the following:

• The low-enriched uranium9 fuel assemblies used in light water reactors, such as Pil-grim, require precise spacing and geometry, the absence of control materials or poi-sons and an appropriate ratio of water to fuel to sustain criticality and generate steady-state power during normal operation. Water is a necessary moderator for criti-cality to proceed. If changing conditions that occur in the reactor core as the fuel un-dergoes fission and is consumed are not managed during reactor operation, the nu-clear chain reaction will terminate because all of these requirements will not be met.

• Core degradation under severe accident conditions destroys the carefully designed geometry of the fuel assemblies and changes the water to fuel ratio needed to main-tain the chain reaction. The melting and mixing of the fuel with the fuel cladding, control material, and other reactor components in the core will act to stop the chain reaction as the core becomes molten, loses it shape, and becomes more diluted.

• The molten core, now better described as core debris, flows into the lower parts of the reactor vessel. As the molten core debris cools into irregular shape(s) and porosity it is difficult to sustain fission through the overwhelming majority of the core debris.

• The addition of water onto the core debris may infrequently lead to conditions favor-ing recriticality, but these will tend to be near the surface of the core debris, irregu-larly occurring and localized in pockets. At best, these portions of the core would be very small fractions of the fully functional core. Accordingly, the levels of I-131, Cs-137, and other radionuclides generated from potential intermittent recriticality in the core debris would at best be many orders of magnitude below the levels of radionu-clides produced in a fully functional reactor where all requirements are met over the full core volume. This situation sharply contrasts with the fully functional reactor 9

Often referred to as LEU, approximately 3-5 weight percent U-235.

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core inventory assumed under the severe accident conditions for the Pilgrim SAMA analysis.

• In addition, if a chain reaction does occur, it will not be sustainable for very long.

The water-to-fuel atom ratio will be favorable only momentarily and other geometry factors such as lack of efficient transfer of the energy from the reaction will tend to stop the nuclear chain reaction. In this respect, recritical events tend to be self-dispersive in nature such that once recritical, the energetics of the criticality are suffi-cient to break apart the critical combination of materials, thereby ceasing the chain reaction.

• Moreover, aside from the evolution of noble gases from the limited recriticality events, most of the fission products will be contained by the overlying water layer over the core debris that is necessary for recriticality. In other words, the fission products produced by the recriticality will be largely removed or scrubbed by the same water that gives rise to the recriticality.

• Finally, the energetics of this event in the core debris is significantly less than those accompanying the severe accidents considered in the Pilgrim SAMA analysis. The analysis in NUREG/CR-5635 suggested that favorable conditions might exist for a more energetic recriticality in the first day following an initiating event. Given the length of time that has passed since the Fukushima initiating event took place, the level of energy release from potential recriticality events will be very small at best, short term, and negligible compared to the large, elevated release source terms due to the energetic events that are the basis for the Pilgrim SAMA analysis.

30. (KRO) Thus, any occurring post-scram recriticalities would be small and localized and greatly less than the criticality that occurs at full power. Accordingly, it is technically incor-rect to imply that post-scram recriticality in a damaged reactor is equivalent to the core chain re-action at full-power in an undamaged core as Pilgrim Watch has done. The unsupported claim made by Pilgrim Watch at page 2 and elsewhere in its proposed contention that the fission chain reaction continues apace despite reactor scram is misleading and technically flawed. Further-16

more, any recriticalities would generally be low energy in nature and occur under some depth of water, which would further limit any airborne radioactive releases. The resulting incremental consequences, especially to high population areas within the 10-50 mile range that dominates the Pilgrim SAMA analysis, would be correspondingly very small.

31. (KRO) Any ongoing radioactive releases from Fukushima due to post-scram re-criticalities, as well as any other ongoing low-level releases of radioactivity from Fukushima, would therefore contrast sharply with the releases assumed in the Pilgrim SAMA analysis.

These ongoing releases would include any airborne releases due to evaporation and resuspension of radioactivity from the facilities at Fukushima, releases resulting from the occurrence of any alleged recriticalities, and any aquatic releases from leaks stemming from the recovery efforts.

These radioactive releases and any other ongoing releases would be very small compared to the large airborne releases due to energetic phenomena considered in the Pilgrim SAMA analysis.

For example, a long, but relatively small, release of radioactive iodine could occur over weeks and be observed as elevated concentrations in sub-drains and other liquid pathways or elevated airborne levels, but the actual dose received by the public from this type and level of release would be greatly exceeded by the much larger, short-term release of longer-lived, more dose-dominant radionuclides (e.g., Cs-137, Sr-90, and Pu-238 among others) that are associated with elevated, severe accident releases to the atmosphere assumed in the Pilgrim SAMA analysis.

32. (KRO) Lastly in this respect, Pilgrim Watchs assertion that a large problem cre-ated by the ongoing chain reaction is the calculation of food doses because of the inability of the MACCS2 code to model the continual production of I-131 and I-134 which can get to peo-ple both by milk and from fresh leafy-vegetable consumption, PW Request at 4, is immaterial to the results of the Pilgrim SAMA analysis. As explained above, the source term from a potential 17

localized, intermittent recriticality is many orders of magnitude below the source term consid-ered in the SAMA analysis. In addition, more than 80% of the population dose in the current Pilgrim SAMA analysis is incurred in the long-term phase after the accident. Although the dose pathways included over this period are groundshine, inhalation of resuspended radionuclides, and ingestion of contaminated food and contaminated water, contaminated food and water can be interdicted by authorities until levels are sufficiently safe. Moreover, for I-131 and I-134 with an 8-day and a 53-minute half-life, respectively, there is no real impact to the long-term food inges-tion dose compared to longer-lived radioisotopes. Thus, any change to the I-131 and I-134 ra-dionuclides do not impact food and water ingestion pathway contribution to long-term dose be-cause this iodine would either decay essentially to zero before consumption, or realistically, be interdicted in food to prevent consumption. Thus, contributions from these radioisotopes from potential, but not confirmed, recriticality events would not contribute to change the existing SAMA analysis outcome.

33. (KRO) In summary, the longer-term radioactive releases that Pilgrim Watch con-tends are ongoing at Fukushima, but are not accounted for by the MACCS2 code would not change the result in the SAMA analysis cost benefit calculation. This is because they are small relative to the releases postulated in severe accident analyses, and thus fail to cause sufficient impacts to the population dose and off-site economic cost risks which are dominated by expo-sures and contamination in the 10-50 mile region resulting from the large airborne releases due to the energetic severe accident events assumed in the Pilgrim SAMA analysis.

B. The Radioactive Releases Assumed in the Pilgrim SAMA Analysis Bounds the Fukushima Daiichi Releases

34. (KRO) The large airborne radioactive source terms resulting from the energetic se-vere accident events assumed for the Pilgrim SAMA analysis more than bound the radioactive 18

releases from Fukushima, including any ongoing low-level releases to the environment and any intermittent, localized recriticality events occurring within the damaged Fukushima reactor fa-cilities. The bounding nature of the Pilgrim SAMA releases refutes Pilgrim Watchs claim. As a result of the bounding nature of the Pilgrim SAMA releases, Pilgrim Watchs claims that MACCS2 considers radioactive releases of insufficient duration in light of lessons learned from Fukushima and is incapable of modeling an 8-week chain reaction that continues after a scram because it assumes that the production of all fission products ceases at that time, (PW Request at 3), or that the Fukushima situation shows that releases can extend into many days, weeks, and months and that a longer release can cause offsite consequences that will affect cost-benefit analyses (id.) are immaterial.

35. (KRO) Contrary to Pilgrim Watchs claim, the duration of an accident release is not the controlling factor for a SAMA analysis. As discussed above, the doses that the public would receive from a low-level release occurring over an extended period of time is greatly ex-ceeded by the larger, elevated releases due to the energetic events analyzed in the Pilgrim SAMA analysis. In this respect, the source terms assumed for the radioactive releases in the Pilgrim SAMA analysis have significant margin in severity over that represented by the events at Fuku-shima, even assuming the longer term, but low-magnitude, radioactive releases, including those from potential intermittent recriticality events.

36. (KRO) The overall source term in the case of a severe accident includes the type and amount of radionuclides, the heat energy in the plume associated with the release (which will cause the plume to rise), the height of the release, the timing of release, and the maximum plume duration considered. A separate source term is developed for each of the 19 postulated accident scenarios from the Pilgrim PSA or CAPBs previously mentioned. The 19 CAPBs are based on 19

the plant-specific Pilgrim PSA and account for postulated system, structure, and component fail-ures, the status of the reactor pressure vessel, the status of the containment, and accident se-quence timing. Each CAPB represents a different combination of plant feature status and release mechanism and have a characteristic frequency and source term release based on attributes of the accident. The CAPBs represent a range of plant radioactivity releases from small to very large and have different characteristics to describe the occurrence of core damage, the occurrence of vessel breach, primary system pressure at vessel breach, the location of containment failure, the timing of containment failure, and the occurrence of core-concrete interactions. The CAPBs used for the Pilgrim SAMA analysis include accident releases that are far more severe in magni-tude and are immediately airborne compared to those from any intermittent recriticality releases from Fukushima.10

37. (KRO) The full list of radionuclide groups and their constituent radionuclides con-sidered in the Pilgrim SAMA analysis are shown in Table 2 in Attachment 1. These include the longer-lived cesium (Cs), tellurium (Te), strontium (Sr), ruthenium (Ru), lanthanum (La), cerium (Ce) and barium (Ba) radionuclide groups in addition to shorter-lived noble gases (NG) and io-dine (I). The reactor inventory for each of the radionuclides created by ongoing fission in the core that are considered in the MACCS2 Pilgrim SAMA analysis is shown in Table 3 in Attach-ment 1. The atmospheric release fraction assumed for each of the radionuclide groups in Table 2 for each of the 19 CAPBs is shown in Table 4 in Attachment 1. The release fractions identify for 10 The 19 CAPBs evaluated in the Pilgrim SAMA analysis are qualitatively described in Table E.1-9 of Exhibit ENT000006, submitted to the Board in relation to Contention 3. In this Table, the barriers to release of radioac-tivity, primarily the fuel cladding, the reactor core water and reactor pressure vessel (RPV) and reactor con-tainment that are assumed to be breached or bypassed are described, along with the phenomenon that causes the failure or bypass.

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each CAPB the percentage of the cores radionuclide inventory for each radionuclide that is as-sumed to be released in the accident scenario represented by the CAPB.

38. (KRO) The radioactive releases for each of the radionuclides for each CAPB can be computed from the information in Tables 2-4 in Attachment 1. The source term for CAPB-15 (which contributes over 80% of the PDR and OECR to the Pilgrim SAMA analysis)11 is illustra-tive of the bounding nature of the CAPBs compared to the identified releases for Fukushima, even including postulated recriticality events. For example, the radioactive release for I-131 as-sumed for CAPB-15 is 5.35E+17 Bq,12 and the radioactive release for Cs-137 assumed for CAPB-15 is 6.38E+16 Bq.13 As discussed below, these are much larger than the radiological releases reported for Fukushima.

39. (KRO) An analysis of the radiological releases from the Fukushima reactors com-pared to the radiological releases assumed in the single-unit Pilgrim SAMA analysis shows that the Pilgrim SAMA analysis accounts for severe accident conditions that more than bound the reported releases from the Fukushima units. Cumulative radioactive releases from Fukushima have been estimated by the Japanese regulatory agencies.14 For I-131 and Cs-137, the estimates 11 CAPB-15 assumes core damage occurs followed by vessel breach, with late containment failure due to core-concrete interactions after the vessel breach, and ensuing containment failure occurring in the drywell or below the torus water level (ENT000006 at Table E.1-9).

12 0.276 (the radionuclide release fraction for iodine for CAPB-15 from Table 4) multiplied by 1.94E+18 Bq, the core inventory for I-131 from Table 3.

13 0.268 (the radionuclide release fraction for cesium for CAPB-15 from Table 4) multiplied by 2.38E+17 Bq, the core inventory for Cs-137 from Table 3.

14 Four units suffered damage, but only three units suffered core damage because only three units were operating at the time of the earthquake and ensuing tsunami. It is not clear what, if any, portion of the radioactive release es-timates result from possible releases from the Unit 4 spent fuel pool. Whatever the case, the Pilgrim SAMA analysis bounds the total release estimates reported to date.

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from the Nuclear Safety Commission of Japan are 1.5 x 1017 Bq and 1.2 x 1016 Bq, respec-tively.15

40. (KRO) As can be seen from comparing these releases to those for CAPB-15 dis-cussed in paragraph 39 of this Declaration above, the assumed releases of I-131 and Cs-137 from CAPB-15 are much larger than the releases from all of the Fukushima reactor facilities com-bined. As shown on Table 5 discussed below, the assumed release of I-131 from CAPB-15 is a factor of about 3.6 larger, and the assumed release of Cs-137 from CAPB-15 is a factor of about 5.3 larger. If the assumed releases for CAPB-15 are multiplied by a factor of three (to better compare, on a per reactor basis, the assumed Pilgrim SAMA release to the releases from the three Fukushima reactor units that suffered core damage in Japan), the release of I-131 from CAPB-15 is a factor of about 10.7 larger, and the release of Cs-137 from CAPB-15 is a factor of nearly 16 larger. Or, viewed differently, the release of I-131 from one of the Fukushima three reactors suffering core damage on average is only 9.3% of the I-131 release assumed in CAPB-15 and the release of Cs-137 on average for the three Fukushima reactors is only 6.2% of the re-lease of Cs-137 assumed in CAPB-15.

41. (KRO) Table 5 shows the releases of I-131 and Cs-137 for the 19 CAPBs com-pared to the releases reported by the Japanese regulatory agencies for Fukushima. Columns 6 and 7 of the Table compare the release of I-131 and Cs-137 for each CAPB to the reported re-lease for all of the Fukushima reactors. Columns 8 and 9 compare the release of I-131 and Cs-137 for each CAPB to the average release from each of the three Fukushima reactors (i.e., the total Fukushima releases are divided by three to provide a per reactor comparison to the assumed 15 TEPCO status briefing, The Great East Japan Earthquake and Current Status of Nuclear Power Stations (May 31, 2011).

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Pilgrim SAMA release). Table 5 shows that, for 14 of the 19 CAPBs evaluated in the Pilgrim SAMA analysis, the assumed releases for both I-131 and Cs-137 are larger than the average re-lease of I-131 and Cs-137 from each of the three Fukushima reactors. These 14 CAPBs consti-tute over 99% of the Pilgrim PDR and OECR. For five of the assumed releases (CAPB-8, -10, -

15, -17, and -19), the releases of both I-131 and Cs-137 are more than an order of magnitude lar-ger than the Fukushima average.

Table 5. Radionuclide Release Fractions of the PNPS Collapsed Accident Progression Bins (CAPBs) compared to Total Estimates from Fukushima Daiichi Ratio of Ratio of 137 Ratio of 131 Pilgrim Cs Ratio of 137 Pilgrim I Pilgrim Cs 131 137 Cs 131 to Total Pilgrim 131I Iodine Cesium I Released to Total I to One Release Released in 137 Cs to One fraction fraction in CAPB estimated 131 Reactor Mode CAPB estimated Reactor I - 137 from Cs -

from Fukushima Fukushima Fukushima Fukushima (Bq) (Bq)

CAPB-1 1.85E-07 1.85E-07 3.58E+11 4.40E+10 0.000002 0.000004 0.000007 0.000011 CAPB-2 4.81E-05 4.66E-05 9.32E+13 1.11E+13 0.0006 0.0009 0.0019 0.0028 CAPB-3 5.37E-05 4.97E-05 1.04E+14 1.18E+13 0.0007 0.0010 0.0021 0.0030 CAPB-4 4.90E-02 2.62E-02 9.49E+16 6.24E+15 0.63 0.52 1.90 1.56 CAPB-5 7.86E-02 3.68E-02 1.52E+17 8.76E+15 1.01 0.73 3.04 2.19 CAPB-6 4.02E-02 2.32E-02 7.79E+16 5.52E+15 0.52 0.46 1.56 1.38 CAPB-7 6.11E-02 2.94E-02 1.18E+17 7.00E+15 0.79 0.58 2.37 1.75 CAPB-8 2.98E-01 2.72E-01 5.77E+17 6.48E+16 3.85 5.40 11.54 16.19 CAPB-9 7.61E-02 7.07E-02 1.47E+17 1.68E+16 0.98 1.40 2.95 4.21 CAPB-10 2.80E-01 2.49E-01 5.42E+17 5.93E+16 3.62 4.94 10.85 14.82 CAPB-11 1.73E-01 1.41E-01 3.35E+17 3.36E+16 2.23 2.80 6.70 8.39 CAPB-12 5.84E-05 4.37E-05 1.13E+14 1.04E+13 0.0008 0.0009 0.0023 0.0026 CAPB-13 7.99E-03 5.99E-03 1.55E+16 1.43E+15 0.10 0.12 0.31 0.36 CAPB-14 2.88E-02 2.67E-02 5.58E+16 6.36E+15 0.37 0.53 1.12 1.59 CAPB-15 2.76E-01 2.68E-01 5.35E+17 6.38E+16 3.56 5.32 10.69 15.95 CAPB-16 6.71E-02 3.26E-02 1.30E+17 7.76E+15 0.87 0.65 2.60 1.94 CAPB-17 3.62E-01 3.37E-01 7.01E+17 8.02E+16 4.67 6.69 14.02 20.06 CAPB-18 9.76E-02 6.25E-02 1.89E+17 1.49E+16 1.26 1.24 3.78 3.72 CAPB-19 4.03E-01 3.77E-01 7.81E+17 8.97E+16 5.20 7.48 15.61 22.44 23

42. (KRO) Thus, the Pilgrim SAMA analysis accounts for severe accident releases that are many times greater than the releases that occurred at the Fukushima reactors.16 The severe accident releases assumed for the Pilgrim SAMA analysis more than bound the releases from Fukushima many times over, and would more than bound any continuing low-level releases such as those from postulated intermittent recriticality. Moreover, as discussed, any ongoing, gener-ally localized releases, would not significantly impact the PDR and OECR so as to alter the out-come of the current Pilgrim SAMA analysis.

43. (KRO) In summary, the Pilgrim SAMA analysis source term is quantitatively lar-ger than, and bounds the combined releases from, all of the Fukushima damaged reactor facilities and would more than bound any continuing low-level releases from Fukushima. When account-ing for the fact that Fukushima involves more than one damaged reactor, the large margin in the Pilgrim SAMA analysis is even more pronounced when considered on a per reactor basis. Be-cause of the large margins in the Pilgrim SAMA analysis, Pilgrim Watchs claims are immaterial to, and would have no impact on, the results of the Pilgrim SAMA analysis.

IV. CONCLUSION AND

SUMMARY

44. (TLS, KRO) Our testimony in this Declaration demonstrates that Pilgrim Watch raises no material challenge to the PNPS SAMA analysis performed in support of license re-newal. Contrary to Pilgrim Watchs assertions, the potential for post-scram recriticality events to occur in a damaged reactor core is not new information, and the evidence cited by Pilgrim Watch is not unequivocal that post-scram recriticalities have occurred, or are occurring, in any of the 16 The radionuclide release data from Fukushima will likely continue to be updated and may possibly be increased.

However, even if the data is increased, the releases assumed in the Pilgrim SAMA analysis have large margins over those reported from Fukushima. For example, even if Fukushima radionuclide release estimates were to double, CAPB-15 (which contributes over 80% of the PDR and OECR to the Pilgrim SAMA analysis) would still bound the estimated I-131 releases from all of the Fukushima facilities by about a factor of two (1.78) and the es-timated Cs-137 releases by about a factor of three (2.66).

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damaged Fukushima Daiichi reactors. This is because many other phenomena could give rise to the observed levels of I-131, Cs-134, and Cs-137. Moreover, Pilgrims SAMA analysis bounds the low-level, longer duration radioactive releases to the environment at Fukushima, whether or not caused by low energy, intermittent recriticalities in a damaged reactor core. Thus, any ongo-ing radioactive releases arising from the alleged post-scram recriticalities or from other causes would have comparatively small impacts that would not alter the Pilgrim SAMA analysis.

We declare under penalty of perjury that the foregoing is true and correct.

Executed in Accord with 10 C.F.R. 2.304(d) Executed in Accord with 10 C.F.R. 2.304(d)

Dr. Thomas L. Sowdon Kevin OKula, Senior Project Manager Advisory Engineer Entergy Nuclear Operations, Inc. URS Safety Management Solutions LLC 600 Rocky Hill Rd 2131 South Centennial Avenue Plymouth, MA 02360 Aiken, South Carolina 29803-7680 Phone: 508-830-8834 Phone: 803.502.9620; E-mail: tsowdon@entergy.com Email: kevin.okula@wsms.com 25

ATTACHMENT 1 Table 2. Radionuclides Considered in the MACCS2 Inventory Input17 Radionuclide Number of Constituent Nuclides Group label nuclides Noble Gas 6 Kr-85, Kr-85m, Kr-87, Kr-88, Xe-133, Xe-135 I 5 I-131, I-132, I-133, I-134, I-135 Cs 4 Rb-86, Cs-134, Cs-136, Cs-137 Te 8 Sb-127, Sb-129, Te-127, Te-127m, Te-129, Te-129m, Te-131m, Te-132 Sr 4 Sr-89, Sr-90, Sr-91, Sr-92 Ru 8 Co-58, Co-60, Mo-99, Tc-99m, Ru-103, Ru-105, Ru-106, Rh-105 La 15 Y-90, Y-91, Y-92, Y-93, Zr-95, Zr-97, Nb-95, La-140, La-141, La-142, Pr-143, Nd-147, Am-241, Cm-242, Cm-244 Ce 8 Ce-141, Ce-143, Ce-144, Np-239, Pu-238, Pu-239, Pu-240, Pu-241 Ba 2 Ba-139, Ba-140 Total 60 17 Table 5 from Testimony of Dr. Kevin R. OKula on Source Term Used in the Pilgrim Nuclear Power Station Severe Accident Mitigation Alternatives (SAMA) Analysis (Jan. 3, 2011) (ENT000012), Radionuclide Group Composition Used in MACCS2, based on Code Manual for MACCS2: Volume 1, Users Guide, NUREG/CR-6613 (SAND97-0594) (1998).

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Table 3. PNPS Core Inventory Used for the MACCS2 Input File18 Radionuclide Inventory (Bq) Radionuclide Inventory (Bq)

Co-58 1.15E+16 Te-131m 2.87E+17 Co-60 1.37E+16 Te-132 2.80E+18 Kr-85 1.88E+16 I-131 1.94E+18 Kr-85m 6.84E+17 I-132 2.85E+18 Kr-87 1.24E+18 I-133 4.07E+18 Kr-88 1.68E+18 I-134 4.45E+18 Rb-86 1.05E+15 I-135 3.83E+18 Sr-89 2.08E+18 Xe-133 4.07E+18 Sr-90 1.84E+17 Xe-135 9.68E+17 Sr-91 2.70E+18 Cs-134 3.96E+17 Sr-92 2.82E+18 Cs-136 8.51E+16 Y-90 1.58E+17 Cs-137 2.38E+17 Y-91 2.54E+18 Ba-139 3.75E+18 Y-92 2.84E+18 Ba-140 3.70E+18 Y-93 3.23E+18 La-140 3.77E+18 Zr-95 3.34E+18 La-141 3.48E+18 Zr-97 3.44E+18 La-142 3.35E+18 Nb-95 3.16E+18 Ce-141 3.36E+18 Mo-99 3.65E+18 Ce-143 3.27E+18 Tc-99m 3.15E+18 Ce-144 2.18E+18 Ru-103 2.76E+18 Pr-143 3.20E+18 Ru-105 1.84E+18 Nd-147 1.43E+18 Ru-106 7.52E+17 Np-239 4.26E+19 Rh-105 1.38E+18 Pu-238 2.96E+15 Sb-127 1.74E+17 Pu-239 7.51E+14 Sb-129 6.05E+17 Pu-240 9.40E+14 Te-127 1.69E+17 Pu-241 1.62E+17 Te-127m 2.27E+16 Am-241 1.65E+14 Te-129 5.68E+17 Cm-242 4.35E+16 Te-129m 1.49E+17 Cm-244 2.34E+15

• The unit of activity is the Becquerel (Bq) and is equal to 1 disintegration per second.

18 Table E-14 Updated PNPS Core Inventory, LRA Amendment 10, (Dec. 12, 2006), Attachment C, Response to Request for Additional Information on LRA Appendix E Concerning Severe Accident Mitigation Alternatives (ENT000010), Response to RAI 2.

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Table 4. Radionuclide Release Fractions in the PNPS SAMA Analysis CAPBs19 CAPB 19 Release Fractions from Table E.1-11, Collapsed Accident Progression Bin (CAPB) Source Terms, Pilgrim Li-cense Renewal Application Environmental Report, Attachment E, Severe Accident Mitigation Alternatives Analysis (2006) (ENT000006).

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EXHIBIT 1 RESUME THOMAS L. SOWDON, ScD 15 Alison Circle, Plymouth, MA 02360 508-747-3024 (Home); 508-830-8834 (Work)

EDUCATION Lowell Technological Institute Bachelor of Science Degree in Nuclear Engineering, June 1972 Massachusetts Institute of Technology Master of Science Degree in Nuclear Engineering with a Specialty in Radiological Health Physics, February 1988 Harvard School of Public Health Doctoral of Science in Epidemiology with minors in Biostatistics and Radiation Biology, June 1999 Doctoral Thesis title - Analysis of Cancer Incidence and Mortality in the Hi-roshima/Nagasaki Cohort using Generalized Additive Models New England Epidemiology Summer Program at Tufts University - 1989, 1991, 1994 HONORS AND AWARDS Appointed Adjunct Professor at Worcester Polytechnic Institute, 1991 through 1994 Certified Radiation Health Physicist - October 1983 Member Massachusetts Department of Public Health, Nuclear Incident Advisory Team (NIAT)

EXPERIENCE February 2009 to Present - Senior Project Manager, Emergency Preparedness Responsibilities/Experience:

- Responsible for Corporate and Fleet-wide projects supporting Emergency Preparedness Ac-tivities at all Entergy Sites

- Projects include WebEOC, Everbridge (Emergency Response Organization Recall System) and InForm (Offsite Agency Notification System), Ingestion Planning Exercises January 2000 to February 2009 - Manager of Emergency Preparedness - Entergy - Pil-grim Station

Responsibilities/Experience:

- Responsible for the Emergency Preparedness Program and Emergency Response Organiza-tion at Pilgrim Station including procedures, equipment, staffing, drills and exercises.

- Responsible for interface with Federal, State and local government agencies including MEMA, FEMA, RIEMA, MDPH and local CD Directors.

- Responsible for budget administration and control and budget contingencies for special pro-jects.

September 1977 to January 2000 - Boston Edison Company/Entergy - Pilgrim Station June 1987 to Present - Chief Radiological Scientist Pilgrim Nuclear Power Station Plymouth, Massachusetts Responsibilities/Experience:

- Manager for environmental epidemiology studies including review and evaluation of state and private studies and direction and oversight of consultant efforts.

- Technical representative for issues involving the biological effects of ionizing radiation as they apply to the occupational radiation worker and members of the general public includ-ing occupational and environmental radiation litigation.

- Primary interface with the Commonwealth of Massachusetts and the general public for is-sues involving the health concerns associated with ionizing radiation resulting from both normal plant operation and accidents.

August 1985 to June 1987 - Radiological Section Manager Pilgrim Nuclear Power Station Plymouth, Massachusetts Responsibilities/Experience:

- Fulfill the function of Radiation Protection Manager as specified in USNRC Regulatory Guide 1.8.

- Overall responsibility for administrative and technical performance of the Radiation Protec-tion Program including:

- Procedure development and modification

- Internal and external exposure control

- Internal and external dosimetry

- Respiratory protection equipment

- In-vivo and In-vitro bioassay 2

- Radiation detection instruments

- ALARA requirements

- Plant design change review and approval

- Radiological occurrences

- Routine surveillance

- High radiation area control

- Air sampling

- Radioactive source inventory and control

- Radiation work permits

- Shielding analyses

- Accident analyses

- Radiological environmental monitoring program

- Offsite dose calculations

- Meteorological monitoring program January 1981 to August 1985 - Environmental and Radiological Health and Safety Group Leader Nuclear Operations Support Department Braintree, Massachusetts Responsibilities/Experience:

- Overall responsibility for administrative and technical performance of the corporate Health Physics staff including:

- Procedure development and modification

- ALARA requirements for plant modifications

- Shielding specification and design

- Accident analyses

- Radiological environmental monitoring program

- Offsite dose calculations

- Health Physics support for the Emergency Plan September 1977 to January 1981 - Senior Radiological Engineer Braintree, Massachusetts Responsibilities/Experience:

- Management of the Radiological Environmental Monitoring Program including all re-quired reports

- Evaluation of sample analysis results

- Offsite dose assessment

- Issue and maintenance of the Offsite Dose Calculation Manual (ODCM)

- Meteorological monitoring program management

- Supervision of equipment maintenance, calibration and data reduction

- Calculation of atmospheric dispersion factors for offsite dose assessment May 1972 to September 1977 - Stone and Webster Engineering Corporation Responsibilities/Experience:

- Radiation shield design including:

3

- Determination of radiation source strengths and distributions

- Establishment of anticipated radiation zones throughout the plant

- Specification of the thickness and material for shielding

- Specification preparation for process and effluent radiation monitoring systems for both analog and digital systems

- Calculation of offsite doses due to routine and accidental releases

- Establishment of restricted area, exclusion area and low population zone boundary dis-tances

- Evaluation of population doses due to effluents, direct and scattered radiation from the plant 4

EXHIBIT 2 KEVIN R. OKULA Advisory Engineer URS Safety Management Solutions LLC 2131 South Centennial Avenue Aiken, South Carolina 29803-7680 Telephone: 803.502.9620 - Email: kevin.okula@wsms.com KEY AREAS:

• Computer Model Verification and Validation

• Severe Accident and Quantitative Risk Analysis

• Accident and Consequence Analysis for Design Ba-

• Level 2/3 Probabilistic Risk Assessment sis Accident Support

• Regulatory Standard & Guidance Development

• MACCS2 Code Applications

• New Reactor Design Accident Analysis and PRA

• Level 3 PRA Standard Development Support PROFESSIONAL

SUMMARY

Dr. OKula has over 28 years experience as a manager and technical professional in the areas of accident and consequence analysis, source term evaluation, commercial and production reactor probabilistic risk assessment (PRA) and severe accident analysis, safety software quality assurance (SQA), safety analysis standard and guidance development, computer code evaluation and verification, risk management, hydro-gen safety, reactor materials dosimetry, shielding, and tritium safety applications. He is a member of the American Nuclear Society (ANS) Standard working group ANS 58.25 on Level 3 Probabilistic Safety Assessment, and is a member of the Peer Review Committee for the Nuclear Regulatory Commissions (NRCs) State-of-the-Art Reactor Consequence Analysis (SOARCA) Program. Kevin was part of the Department of Energy (DOE) team writing DOE G 414.1-4, Safety Software Guide. He coordinated technical support for the DOE Office of Environment, Safety, and Health (EH) in addressing Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 on Software Quality Assurance (SQA), and was a consultant to DOE/EH-31 Office of Quality Assurance for disposition of SQA issues.

Dr. OKula was a member of the Partner, Assess, Innovate, and Sustain (PAIS) Safety Case team for the Sellafield Site in the United Kingdom in the early 2009 period. The PAIS team identified and began im-plementation of improvement opportunities in nuclear safety and related areas. Recommendations were documented in comprehensive reports to the Sites Nuclear Management Partners consortium in March 2009.

He is, or has supported, Atomic Safety Licensing Board (ASLB) relicensing issue resolution for several commercial plants including Indian Point, Prairie Island, and Pilgrim Nuclear Power Station, on severe accident mitigation alternatives (SAMA) analysis. He was also part of the accident analysis and PRA/severe accident teams supporting the Design Certification Document for the U.S. Advanced Pres-sure Water Reactor (US-APWR) a joint effort with URS Washington Division and Mitsubishi Heavy In-dustries (MHI). He has provided similar support for an alternative reactor technology, the Pebble Bed Modular Reactor (PBMR).

Kevin is coordinating WSMS support to the Quantitative Risk Analysis (QRA) for evaluation of hydro-gen events in a waste vitrification plant design, including fault tree and human factors areas. He is

also a contributor to the DOE response on the use of risk assessment methodologies as part of the DNFSB Recommendation 2009-1 implementation action for Risk Assessment. He led work in reviewing EIS food pathway consequence analysis performed on assumed accident conditions from the Mixed Oxide Fuel Fabrication Facility (MFFF), sited at the Savannah River Site. This project compared and evaluated the impacts calculated from three computer models, including MACCS2, GENII, and UFOTRI.

He is past chair of the American Nuclear Society (ANS) Nuclear Installations Safety Division (NISD),

and the Energy Facility Contractors Group (EFCOG) Accident Analysis Subgroup. He is a member of the Nuclear Hydrogen Production Technical Group under the ANSs Environmental Sciences Division, and is chair for the EFOCG Hydrogen Safety Interest Group. He was the Technical Program Chair for two ANS embedded topical meetings on Operating Nuclear Facility Safety (Washington, D.C., 2004) and the Safety and Technology of Nuclear Hydrogen Production, Control and Management (Boston, MA, 2007).

Dr. OKula was PRA group manager for K Reactor at the time of restart in the early 1990s. He led a suc-cessful effort demonstrating Savannah River Site (SRS) K-Reactor siting compliance to 10 CFR 100, and tritium facility compliance with SEN-35-91.

He was the project leader for independent Verification and Validation (V&V) of urban dispersion soft-ware for the Defense Threat Reduction Agency (DTRA) and is the current V&V project manager for the evaluation of several chemical/biological software tools for the U.S. Army Test and Evaluation Command (ATEC) and Chemical-Biological Program (Dugway Proving Ground (Utah) and Edgewood Chemi-cal/Biological Center in Maryland.

EDUCATION:

Ph.D., Nuclear Engineering, University of Wisconsin, 1984 M.S., Nuclear Engineering, University of Wisconsin, 1977 B.S., Applied and Engineering Physics, Cornell University, 1975 TRAINING:

Conduct of Operations (CONOPS), 1994 Harvard School of Public Health, Atmospheric Science and Radioactivity Releases, 1995 Consequence Assessment, (Savannah River Site, 1995)

U.S. DOE Risk Assessment Workshop (Augusta, GA, 1996)

MELCOR Accident Computer Code System (MACCS) 2 Computer Code, 1997, 2005 MCNPX Training Class (ANS Meeting, 1999)

CLEARANCE:

Active DOE Q PROFESSIONAL EXPERIENCE:

Washington Safety Management Solutions 1997 to Present Advisory Engineer and Senior Fellow Advisor Dr. OKula is a member of the State-of-the-Art Reactor Consequence Analysis (SOARCA) Project Peer Review Committee that provides recommendations on applying MACCS2 in the context of accident phe-nomena and subsequent off-site consequences in the context of severe reactor accidents. This activity 2

supports the efforts of Sandia National Laboratories (SNL) and the Nuclear Regulatory Commission (NRC) to provide more realistic assessment of severe accidents.

Dr. OKula is also part of the Level 3 PRA Standard working group charged with developing an ANSI/ANS standard for Level 3 PRA analysis. He participated in a team that conducted an SQA gap analysis on the bioassay code [Integrated Modules for Bioassay Analysis (IMBA)] based on DOE G 414.1-4 requirements. He identified safety analysis codes that were designated as DOE toolbox codes, and oversaw production of the first documents (QA criteria and application plan, code guidance reports, and gap analysis) for six accident analysis codes designated for the DOE Safety Software Tool-box. He provided support to DOE/EH-31 (now DOE/HSS) for addressing SQA issues for safety analysis software. He was a contributor to DOE G 414.1-4, Safety Software Guide on SQA practices, procedures, and programs.

Kevin has provided technical input for work packages on several recent commercial projects. In the first, he teamed with Entergy on MACCS2 code applications issues in the Severe Accident Mitigation Alterna-tives (SAMA) analysis area for the Pilgrim Nuclear Power Station. In the second, he was part of tritium environmental release analysis team that supported evaluation of tritium control and management areas for the Braidwood plant. A third effort developed an initial SAMDA document for the Mitsubishi Heavy Industries (MHI) US-APWR (1610 MWe evolutionary PWR), as well as complete a control room habita-bility study for postulated toxic chemical gas releases.

Kevin was part of a Washington Group team that developed a Design Control Document (DCD) for the MHI US-APWR using input information from MHI. He was Chapter lead on Chapter 15 (Transient and Accident Analysis), and later transitioned to severe accident evaluation and documentation support to Chapter 19 (PRA and Severe Accidents). He currently is the Chapter 19 lead for PRA and Severe Acci-dent for COLA development for the Pebble Bed Modular Reactor (PBMR).

Dr. OKula developed the outline, coordinated contributors, and assembled the first draft of the DOE Ac-cident Analysis Guidebook, a reference guide for hazard, accident, and risk analysis of nuclear and chemi-cal facilities operated in the DOE Complex. He is also the primary author and coordinator for the Acci-dent Analysis Application Guide for the Oak Ridge contractor. Dr. OKula also developed a one-day course and exam for the guide, which he later presented to the Oak Ridge, Paducah, and Portsmouth staff.

Dr. OKula also led an independent V&V review for the DTRA of the U.K.-developed Urban Dispersion Model (UDM) software for predicting chemical and biological plume dispersion in city environments, and is leading projects to verify and validate chemical/biological simulation suite software applications for the Dugway Proving Ground (Utah), and the Edgewood Chemical Biological Center (ECBC) in Mary-land.

Managing Member, Consequence Analysis Dr. OKula was responsible for the consequence analysis associated with accident analysis sections of Documented Safety Analysis (DSA) reports and other safety basis documents for SRS, Oak Ridge, and other DOE nuclear facilities. He also developed the methodology and identified appropriate computer models for this purpose. Additionally, Dr. OKula developed training to enhance consistency and stan-dardize analyses in the consequence analysis area. He was project manager for environmental assessment support to SRS on a transportation safety analysis using the RADTRAN code.

Dr. OKula coordinated development of a DOE Accident Analysis Guidebook involving over 10 sites and organizations. He also led the effort to produce Computer Model Recommendations for source term (fire, spill, and explosion), in-facility transport, and dispersion/consequence (radiological and chemical) areas.

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Westinghouse Savannah River Company 1989 to 1997 Group Manager Dr. OKula managed consequence analyses associated with accident analysis sections of DSA reports and other safety basis documents. He also developed the associated methodologies and identified appropriate computer models. He was a member of the management team supporting Criticality Safety Evaluation preparation assisting Safe Sites of Colorado and the dispositioning of final criticality safety issues for the decommissioning and decontamination of nuclear facilities at the Rocky Flats Environmental Technology Site.

In a teaming arrangement with Science Applications International Corporation, Kevin initiated discus-sions that led to development of an emergency management enhancement tool to risk inform likely source terms. Applied this approach to a Savannah River nuclear facility (K Reactor), and was part of the team to provide this methodology for use on the British Advanced Gas-Cooled Reactors (AGRs) (for the United Kingdoms Nuclear Installation Inspectorate). Model was knowledge-based, and required devel-opment of an Accident Progression Event Tree (APET) for the facility in question.

Dr. OKula managed the completion of the SRS K Reactor PRA program. He was the lead for develop-ment of the K Reactor Source Term Predictor Model and assisted with the core technology lay-up pro-gram to preserve competencies in reactor safety. He coordinated a 25-person group responsible for K Reactor probabilistic and deterministic dose analyses, and led the examination of reduced power cases at project termination. He developed risk and dose management applications to cost-effectively prioritize facility modifications.

Kevin interfaced with DOE Independent and Senior Review teams to finalize study acceptance, and tran-sitioned the risk assessment team to risk management functions for nuclear and waste processing facili-ties. In addition, he successfully prepared a 10 CFR 100 Siting white paper to resolve issues raised by the DNFSB, and teamed with DOE/HQ legal support to document resolutions. He led the development of a position paper demonstrating SRS Replacement Tritium Facility compliance with DOE Safety Policy (SEN-35-91).

Staff Engineer Dr. OKula led an analytical team quantifying the tritium source term during a Loss of River Water de-sign basis accident. He evaluated airborne tritium levels with multi-cell CONTAIN model, interfaced with a multidisciplinary team to resolve Operational Readiness Review concerns, developed an SRS-specific methodology for applying MACCS as a tool for Level 3 PRA Applications, and applied CON-TAIN code for K Reactor source term analysis.

E.I. du Pont de Nemours & Company 1982 to 1989 Principal Engineer, Research Engineer Dr. OKula performed risk analysis duties for the Savannah River Laboratory (SRL) Risk Analysis Group, after earlier conducting research activities for the Reactor Materials and Reactor Physics Groups.

He performed initial planning for offsite irradiation of test specimens to evaluate remaining reactor life-time for Savannah River reactor components.

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Westinghouse Electric Corporation 1975 Summer Student, Reactor Licensing Monroeville, PA American Electric Power Corporation 1973 to 1974 Co-op Student, Reactor Physics and Reactor Licensing New York, NY Long Island Lighting Company 1972 Summer Intern Riverhead, NY PARTIAL LIST OF PUBLICATIONS (2000-2010):

K. R. OKula, D. C. Thoman, J. Lowrie, and A. Keller, Perspectives on DOE Consequence Inputs for Ac-cident Analysis Applications, American Nuclear Society 2008 Winter Meeting and Nuclear Technol-ogy Expo, November 9-13, 2008 (Reno, NV).

K. R. OKula, F. J. Mogolesko, K-J Hong, and P. A. Gaukler, Severe Accident Mitigation Alternative Analysis Insights Using the MACCS2 Code, American Nuclear Society 2008 Probabilistic Safety As-sessment (PSA) Topical Meeting, September 7-11, 2008 (Knoxville, TN).

K. R. OKula and D. C. Thoman, Modeling Atmospheric Releases of Tritium from Nuclear Installations, American Nuclear Society Embedded Topical Meeting on the Safety and Technology of Nuclear Hy-drogen Production, Control and Management, June 24-28, 2007 (Boston, MA).

K. R. OKula and D. C. Thoman, Analytical Evaluation of Surface Roughness Length at a Large DOE Site (U), American Nuclear Society Winter Meeting, November 12-16, 2006 (Albuquerque, NM).

K. R. OKula and D. Sparkman, Safety Software Guide Perspectives for the Design of New Nuclear Fa-cilities (U), Winter Meeting of the American Nuclear Society, November 13 - 17, 2005 (Washington, D.C.).

K. R. OKula and R. Lagdon, Progress in Addressing DNFSB Recommendation 2002-1 Issues: Improving Accident Analysis Software Applications, Fifteenth Annual Energy Facility Contractors Group Safety Analysis Workshop, April 30 - May 5, 2005, Los Alamos, NM (2005).

K. R. OKula and Tony Eng, A Toolbox Equivalent Process for Safety Analysis Software, Fourteenth Annual Energy Facility Contractors Group Safety Analysis Workshop, May 1-6, 2004, Pleasanton, CA (2004).

K. R. OKula, D. C. Thoman, J. A. Spear, R. L. Geddes, Assessing Consequences Due to Hypothetical Accident Releases from New Plutonium Facilities (U), American Nuclear Society Embedded Topical Meeting on Operating Nuclear Facility Safety, November 14 - 18, 2004 (Washington, D.C.).

K. OKula and J. Hansen, Implementation of Methodology for Final Hazard Categorization of a DOE Nuclear Facility (U), Annual Meeting of the American Nuclear Society, June 13-17, 2004, (Pitts-burgh, PA).

K. R. OKula and Tony Eng, A Toolbox Equivalent Process for Safety Analysis Software, Fourteenth Annual Energy Facility Contractors Group Safety Analysis Workshop, May 1-6, 2004, Pleasanton, CA (2004).

K. R. OKula, et al., Evaluation of Current Computer Models Applied in the DOE Complex for SAR Analysis of Radiological Dispersion & Consequences, WSRC-TR-96-0126, Westinghouse Savannah 5

River Company (2003).

K. R. OKula, et al., Evaluation of Current Computer Models Applied in the DOE Complex for SAR Analysis of Radiological Dispersion & Consequences, WSRC-TR-96-0126, Rev. 3, Westinghouse Savannah River Company (2002).

K. R. OKula, A DOE Computer Code Toolbox: Issues and Opportunities, Eleventh Annual EFCOG Workshop, also 2001 Annual Meeting of the American Nuclear Society, Milwaukee, WI (2001).

PUBLICATIONS (1988-1999):

Dr. OKula authored or co-authored more than 20 publications between 1988 and 1999. Details are avail-able upon request.

PROFESSIONAL SOCIETIES AND STANDARDS COMMITTEES

• American Nuclear Society

• Health Physics Society

• Level 3 ANS PRA Standard Committee 58.2 6