ML11346A301

Declaration of Joseph R. Lynch, Lori Ann Potts, and Dr. Kevin R. O'Kula in Support of Entergy'S Answer Opposing Pilgrim Watch Request for Hearing on a New Contention Regarding Inadequacy of Environmental Report, Post-Fukushima

ML11346A301
Person / Time
Site:	Pilgrim
Issue date:	06/27/2011
From:	Jeffery Lynch, O'Kula K, Potts L Entergy Nuclear Generation Co, Entergy Nuclear Operations, URS Safety Management Solutions
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
50-293-LR, ASLBP 06-848-02-LR
Download: ML11346A301 (49)

Text

June 27, 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 JOSEPH R. LYNCH, LORI ANN POTTS, AND DR. KEVIN R.

O'KULA IN SUPPORT OF ENTERGY'S ANSWER OPPOSING PILGRIM WATCH REQUEST FOR HEARING ON A NEW CONTENTION REGARDING INADEQUACY OF ENVIRONMENTAL REPORT, POST-FUKUSHIMA Mr. Joseph R. Lynch ("JRL"), Ms. Lori Ann Potts ("LAP") and Dr. Kevin R. O'Kula

("KRO") state as follows under penalties of perjury:

I. INTRODUCTION A. Entergy Declarants

1. Joseph R. Lynch

1. (JRL) I am the Manager of Licensing for the Pilgrim Nuclear Power Station ("Pil-grim"). My professional and educational experience is summarized in the Curriculum Vitae at-tached as Exhibit 1 to this Declaration.

2. (JRL) I have over 29 years of nuclear power experience and background in engi-neering, licensing/regulatory affairs, environmental compliance, complex problem solving, stakeholder communications, project management, cost control, budgeting and employee man-agement. I obtained my Bachelor of Science in Mechanical Engineering (BSME) from the Worcester Polytechnic Institute in Worcester, Massachusetts in 1982, specializing in thermody-

namics and fluid dynamics. I have undertaken graduate-level studies in Business Management, Communications, and Regulatory Compliance. I have also taken numerous internal and external management courses while previously working with the Yankee Atomic Electric Company and the Vermont Yankee Nuclear Power Station.

3. (JRL) In my current position as Pilgrim Licensing Manager, I am responsible for managing the Pilgrim Licensing Group, which supports the operation and regulatory compliance of the Station in accordance with NRC, State and Federal regulations, permits and statutes. I am responsible for the development of all necessary letters, licensing correspondence and regulatory approvals from the NRC, local, state and federal agencies required in support of plant operations.

I am familiar with all Pilgrim operational systems, including the operation, and maintenance of the Direct Torus Vent ("DTV"), and the Extensive Damage Mitigation Guidelines ("EDMGs"),

which are a series of requirements imposed by the NRC following the attacks on the World Trade Center on September 11, 2001. 1am familiar with the Pilgrim License Renewal Applica-tion, and the aging management measures that will be required of Pilgrim during its period of extended operation.

2. Lori Ann Potts

4. (LAP) I am a senior consulting engineer to Entergy Nuclear in the areas of Severe Accident Mitigation Alternative Analysis and Fire Probabilistic Risk Assessment. My profes-sional and educational experience is summarized in the Curriculum Vitae attached as Exhibit 2 to this Declaration.

5. (LAP) I have over 30 years of experience as a technical professional in the nuclear industry in the areas of safety analysis, probabilistic safety assessment, deterministic and prob-2

abilistic accident and consequence analysis, materials aging management, reactor engineering, and systems engineering. I obtained a Bachelor of Science degree in Nuclear Engineering from The Pennsylvania State University in 1981.

6. (LAP) I have previous Probabilistic Safety Assessment ("PSA") and severe acci-dent analysis experience in analyzing reactor, emergency system, and containment phenomena under accident conditions. I was the primary author of the industry SAMA guideline (NEI 05-

01) which was endorsed by the NRC. I have participated directly in the SAMA analyses for 8 nuclear plants, including the SAMA analysis for the Pilgrim plant, and have peer reviewed the SAMA analyses for 3 additional nuclear plants.

3. Dr. Kevin R. O'Kula

7. (KRO) I am an Advisory Engineerwith URS Safety Management Solutions

("URS") LLC. My professional and educational experience is summarized in the Curriculum Vitae attached as Exhibit 3 to this Declaration.

8. (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 verification, probabilistic safety assessment, deterministic and probabilistic accident and conse-quence analysis applications for reactor and non-reactor nuclear facilities, source term evalua-tions in both design basis and severe accident assessments, risk management, reactor materials dosimetry, and shielding. I obtained a Bachelor of Science 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 Univer-sity of Wisconsin in 1984.

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9. (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.

10. (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 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 Watch's Proposed Late-Filed Contention on Fukushima

11. (JRL, LAP, KRO) We have reviewed and are familiar with Pilgrim Watch's late-filed contention concerning alleged new and significant information resulting from the March 11, MA CCS2 Computer Code Application Guidancefor 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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2011 accident at Japan's Fukushima Daiichi reactor complex, which was filed on June 1, 2011 in the NRC licensing proceeding for the PNPS license renewal.2

12. (JRL, LAP, KRO) Pilgrim Watch's late-filed contention states:

Based on new and significant information from Fukushima, the Environmental Report is inadequate post Fukushima Daiichi. Entergy's SAMA analysis ignores new and significant issues raised by Fukushima regarding the probability of both containment failure, and subsequent larger off-site consequences due to failure of the direct torus vent (DTV) to operate.

PW Request at 1. In particular, Pilgrim Watch argues that, when preparing the Pilgrim SAMA analysis, Entergy "followed conventional NRC practice and assumed very low probabilities" that an accident could occur. Id. at 1-2, 6, 29. In addition, Pilgrim Watch contends that, in the event of an accident, Entergy erroneously assumed that there would be no containment pressure build-up, no significant delay in attempting to vent the containment because of operator error, no fail-ure or inoperability of the DTV, and no "catastrophic failure of the containment." Id. at 1-2, 6,

29. As a result, Pilgrim Watch contends that Entergy must "conduct a new analysis - based on what Fukushima has taught about reality" considering its purported new and significant informa-tion. Id. at 2.

13. (JRL, LAP, KRO) Pilgrim Watch identifies two pieces of new and significant in-formation that it claims must be accounted for in a new SAMA analysis. Fjirt Pilgrim Watch asserts that there is an increased probability of a containment failure based on the alleged (1) hesitancy of plant operators to use the DTV because it is unfiltered; and (2) likely failure of the DTV. PW Request at 9, 21-24. Second, Pilgrim Watch asserts that there is an increased prob-2 Pilgrim Watch Request for Hearing on New Contention Regarding Inadequacy of Environmental Report, Post Fukushima (June 1, 2011) ("PW Request").

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ability of a significantly larger volume of off-site radiological releases due to failure of the DTV.

Id. at 2, 5, 26, 28.

14. (JRL, LAP, KRO) Our Declaration addresses the claims raised by Pilgrim Watch concerning the adequacy of the Pilgrim SAMA analysis in light of Fukushima. In summary, the alleged failure of the Fukushima Units 1, 2, and 3 DTVs and subsequent alleged "catastrophic" failure of the primary containments for all three units is an incorrect portrayal of events. Al-though preliminary, the Government of Japan has prepared a comprehensive Report3 that sum-marizes known facts concerning the accident. That summary states that the DTVs were success-fully operated for at least Units 1 and 3. Furthermore, while it is clear that the reactor building structures, or secondary containments, of Units 1 and 3 were damaged by explosions likely caused by hydrogen accumulation and ignition within those structures, there is absolutely no evi-dence suggesting "catastrophic" failures of those units' primary containments, which house the reactor vessels. These units' primary containments continue to contain the overwhelming major-ity of their respective core inventories. Indeed, it is estimated that for Fukushima Units I and 3, approximately 99% of the radionuclide content remains contained. Report at IV IV-43, IV-75 (estimating core inventory release fractions for Fukushima Units 1 and 3). Although the known facts are less clear with respect to whether the Unit 2 DTV was operated and the status of its primary containment, the Report estimates that 93%-99% of the radionuclide inventory re-mains contained. See Report at IV IV-43, IV-59, IV-75 (estimating core inventory release fractions for Fukushima Units 1-3). Comparatively, the releases assumed in the Pilgrim SAMA 3 Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEP-CO's Fukushima Nuclear Power Stations, Nuclear Emergency Response Headquarters, Government of Japan (June 2011) ("Report"), which is attached to this Declaration as Exhibit 4.

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analysis for containment failure are much larger than the apparent releases from all three Fuku-shima units combined. 4

15. (JRL, LAP, KRO) In any event, Pilgrim Watch's allegations concerning the DTV failure, followed by primary containment failure, would have no significant impact on the Pil-grim SAMA analysis. The potential for the DTV to fail is not new information, and its failure is explicitly postulated in the Pilgrim SAMA analysis. A failure of the primary containment fol-lowed by substantial radiological release is also included in the analysis. Accordingly, Pilgrim Watch's claims are neither new nor significant, as they are already factored into the Pilgrim SAMA analysis, which Pilgrim Watch fails to materially dispute.

II. ALLEGED FAILURES OF THE FUKUSHIMA UNITS 1, 2, AND 3 DIRECT TO-RUS VENTS AND PRIMARY CONTAINMENTS

16. (JRL, LAP, KRO) The following paragraphs provide an overview of the design and operation of the Pilgrim DTV, summarize statements from the Report (Exhibit 1) indicating that, contrary to Pilgrim Watch's assertions, all three Fukushima DTVs did not fail, and all three Fu-kushima primary containments did not suffer catastrophic failures.

A. Pilgrim DTV Overview

17. (JRL) The focus of Pilgrim Watch's new contention is on the direct torus vent

("DTV"). Under some specific accident conditions, the DTV is intended to relieve pressure buildup in the primary containment. The pressure buildup can occur as a result of steam and hy-drogen production as a result of a core melt accident. The Pilgrim DTV provides a welded pip-4 With respect to what is known about the operation of the DTV and the hydrogen explosion events at Fukushima Units 1, 2, and 3, the Report prepared by the Government of Japan is consistent with a separate report prepared by the International Atomic Energy Agency entitled Mission Report: The Great East Japan Earthquake Expert Mis-sion, IAEA International Fact Finding Expert Mission of the Fukushima Daiichi NPP Accident Following the Great East Japan Earthquake Tsunami (May 24 -June 2, 2011) ("IAEA Report"). The relevant portions of the IAEA Report summarizing the sequence of events leading to the accident at Fukushima Daiichi are attached to this Declaration at Exhibit 5.

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ing system between the air space within the suppression chamber (which is also referred to as the torus or suppression pool), which is part of the primary containment where steam and other gases would collect, and the main stack, which is an elevated release point for the plant. Venting through the DTV requires no external power as the primary containment pressure provides the motive force. The system lineup is achieved by opening two separate valves, whose normal electrical and pneumatic power are "essential" (i.e., supported by multiple, redundant, dedicated electrical and pneumatic supplies), and the system is also designed to be operated manually. A 30 pound-force per square inch gauge ("psig") rupture disk is in the flowpath to preclude inad-vertent releases from the system. The Control Room Shift Manager has the authority to direct operation of the system in accordance with Pilgrim specific procedures. The system was de-signed, installed, and approved between 1986 and 1989 and has been subject to routine and regu-lar maintenance. Training on the operation of the system is part of the licensed operator training program.

18. (JRL) The NRC's Extensive Damage Mitigation Guidelines ("EDMGs"), which are a series of requirements implemented by the NRC following the events of September 11, 2001, further enhance operators' ability to utilize the DTV and address other severe accident mitigation parameters in circumstances where no external power sources may be available. Pro-cedural guidance, trained and licensed personnel, and pre-staged equipment are available for manual, local operation of both DTV valves, should the diverse and redundantly powered valves of the normal system, the containment atmospheric control system, and the DTV system not be available because of loss of power.

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B. Pilgrim Watch Erroneously Asserts that the Fukushima Daiichi Units 1-3 DTVs Failed to Operate

19. (JRL, LAP, KRO) Pilgrim Watch asserts that the DTVs at Fukushima Daiichi Units 1, 2, and 3 failed to operate. PW Request at 2 ("[a]ll three failed" to open); 10 ("[t]he DTV design failed three times at Fukushima"); 22 ("[a]t Fukushima, DTV systems failed three times -three out of three"). As a result, Pilgrim Watch contends that "as expected, all three con-tainments failed" in each Fukushima Units 1,2, and 3, resulting in significant offsite conse-quences. Id. at 2, 6.

20. (JRL, LAP, KRO) However, Pilgrim Watch's assertions stand in stark contrast with the preliminary summary of known facts of the Fukushima accident compiled by the Gov-ernment of Japan and presented in its preliminary Report on the Fukushima accident (Exhibit 1).

Furthermore, Pilgrim Watch appears to misunderstand that, whereas the reactor building struc-tures (or secondary containments) for Fukushima Units 1 and 3 were damaged, these events do not evidence failure of the primary containment structures which house the reactor vessel.

21. (JRL, LAP, KRO) The Report.was prepared by the Government of Japan and pro-vided to the International Atomic Energy Agency. Although neither Entergy nor its personnel assisted in the preparation of the Report and do not have any first-hand knowledge of the facts and circumstances surrounding the accident at Fukushima Daiichi, we are technically qualified to understand the Report's contents, particularly with regard to DTV operations.

22. (JRL, LAP, KRO) The Report describes itself as a "preliminary accident report" that "represents a summary of the evaluation of the accident and the lessons learned to date based on the facts gleaned about the situation so far" concerning the March 2011 accident at Fuku-shima Daiichi. Report at 2. The intent of the Report is "to provide as accurately as possible an 9

exact description of the facts of the situation," while "providing a clear distinction between known and unknown matters" as of May 31, 2011. Id. at 3. Despite its preliminary nature, the Report is one of the most comprehensive attempts to summarize the facts and circumstances of the Fukushima Daiichi accident of which we are aware to date.

23. (JRL, LAP, KRO) The Report does not support Pilgrim Watch's multiple asser-tions that each of the Fukushima Units' DTVs failed or was inoperable and thus did not perform their intended function. The Report details when DTV venting operations were undertaken for each Unit, referring to "wet vent," "wet well vent," and primary containment vessel ("PCV")

venting multiple times. For Unit 1, the Report states that "[w]et well venting of the PCV was carried out at 14:30 on March 12." Report at 10. Although work to open the vent "proceeded with difficulty" due to high radiation, TEPCO judged that venting had been achieved since the PCV pressure had been reduced by 14:30 on March 12. Id. at IV-34, IV IV-41. See also id.

at VI-2; IAEA Report (Exhibit 5) at 32 ("operators confinned a decrease in the dry well pressure, providing some indication that venting had been successful"). For Unit 3, "TEPCO carried out wet venting'"at 08:41 on March 12, and then again at 05:20 on March 14. Report at IV-34, IV-

73. See also id. at 13 ("A wet well vent operation of the PCV was carried out at 05:20 on March 14"); VI-2 (describing the increase in radiation levels detected after Unit 3 was vented); IAEA Report at 33 ("successful venting was confirmed by the decrease in dry well pressure").

24. (JRL, LAP, KRO) For Unit 2, the Report indicates that a venting operation was ini-tiated, but is not clear whether the DTV was successfully operated. In one location, the Report states that TEPCO "carr[ied] out vent operations at 11:00 on March 13 and 00:00 on March 15, but a decrease in [drywell] pressure could not be verified" as instrumentation was apparently not available. Report at IV-57; see also IAEA Report (Exhibit 5) at 33 ("successful venting of the 10

Unit 2 containment could not be verified"). Elsewhere, the Report provides that "[iun Unit 2, a decrease in [drywell] pressure was observed due to a wet venting at 21:00 on March 14." Report at VI-3. In another location, the Report states that "[a] PCV wet vent operation including that of small valves was carried out from around 11:00 on March 13". Id. at 11. Use of "small valves" indicates steps being performed prior to DTV operation.

25. (JRL, LAP, KRO) Thus, according to the Report (and contrary to Pilgrim Watch's assertions) the DTV was operated successfully for Units 1 and 3. At this time, the Report indi-cates that a DTV operation was initiated for Unit 2, but the Report is not clear as to whether the venting operation resulted in a decrease in primary containment vessel pressure. Although the Report questions the effectiveness of the venting system as implemented at Fukushima as an ac-cident mitigation measure, it goes on to state that the containment venting system will be en-hanced 5 Report at 33. Nonetheless, at least with respect to Units 1 and 3, the DTV appears to have performed its primary containment depressurization function. Therefore, Pilgrim Watch's various assertions that the Fukushima DTVs experienced a 100% failure rate are wrong.

C. The Design and Operations of Pilgrim's DTV Differ from the Fukushima DTVs

26. (JRL) Further undercutting the relevance of Pilgrim Watch's assertions is the fact that the design and operations of Pilgrim's DTV differ from the Fukushima DTVs. Thus, even if the Fukushima DTVs failed to operate as intended, such failure would have little bearing on the likelihood that Pilgrim's DTV would fail.

27. (JRL) Based on the information available from the Fukushima accident thus far, it appears that the design of Pilgrim's DTV is different from those installed at the Fukushima units As discussed in subsequent sections, the enhanced design of, and clear operational procedures for, Pilgrim's DTV already ensure its effectiveness.

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in several key respects. First, Pilgrim is a single unitiplant that does not share vent lines with other units. Second, the Pilgrim DTV was constructed of welded piping over its entire length, and designed, built, and qualified to the same criteria as the Pilgrim primary containment, and this level of quality is maintained until the piping exits the secondary containment. This means that the DTV pipe does not connect with any other systems until exiting the secondary contain-ment, thus minimizing the potential for any leakage of gases into the secondary containment (i.e.,

reactor building) such as which occurred at Fukushima. Report at 9. Additionally and impor-tantly, the Pilgrim DTV has by design multiple means of operation and can be operated with no electrical power sources. In addition, Pilgrim has diverse, redundant sources of offsite power.

The Japanese Government's Report on the accident states that, although TEPCO had multiple power lines entering the site, they were supported by'the same transmission towers which are subject to common mode failure. Report at IV-127. Pilgrim's design is different. The external electrical sources utilize different physical routing, and are spatially isolated from each other, with overhead and underground routes precluding failures of one source adversely affecting the other source.

28. (JRL) My review of the Report's discussion of the procedures for operating the Fukushima DTVs indicates that Pilgrim's procedures for DTV operation differ significantly from those governing the Fukushima DTVs. The Report's description of how those procedures were carried out at Fukushima also varies significantly from how Pilgrim's procedures would be car-ried out under similar circumstances. With respect to operation of the Fukushima DTVs, the Re-port states:

PCV vent from the [suppression chamber] (hereinafter referred to as "wet vent")

shall be given priority operation; and when the PCV pressure reaches the maxi-mum operating pressure before core damage, when the pressure is expected to reach about twice as high as the maximum operating pressure after core damage 12

and if [residual heat removal ("RHR")] is not expected to be recovered, wet vent shall be conducted if the total coolant injection from the external water source is equal to or less than the submergence level of the vent line in the [suppression chamber] or PCV vent from the [dry vent] shall be conducted if the vent line of the [suppression chamber] is submerged. The procedures for operation in severe accidents specify that the chief of emergency response headquarters shall deter-mine whether PCV vent operation should be conducted after core damage.

Report at IV-13. In other words, the Fukushima procedures call for DTV operation before maximum operating pressure is reached when RHR is available, or, if RHR is unavailable, the procedures call for DTV use before twice the maximum operating pressure is reached. In both cases, the DTV can be used only with authorization from the chief of emergency response head-quarters.

29. (JRL) The Report describes how these procedures were implemented during the accident. For example, for Unit 1, the Report states that the primary containment vessel pressure reached 0.7 MPag (-101.5 psig; 1 MPag = 145 psig), or approaching twice the maximum work-ing pressure of 0.427 MPags (-61.9 psig), before venting occurred. Report at IV-46. The Re-port suggests that the Minister of Economy, Trade, and Industry gave the final authorization for TEPCO to vent the containments of Units 1 and 2. Report at IV-34.

30. (JRL) Based on the stated summary of DTV procedures applicable at Fukushima and the events that actually occurred, the procedures governing operation of the Fukushima DTVs differ significantly from Pilgrim's DTV operating procedures in key respects.

31. (JRL) First, Entergy's operational and severe accident procedures clearly identify the actions that are to be undertaken by plant personnel under different plant circumstances.

These procedures require Entergy to vent the primary containment using the DTV long before the Fukushima operators attempted that same operation. Pilgrim Emergency Operating Proce-dures EOP-03 and 5.4.6 detail the steps that operators are to follow, starting at a containment 13

pressures of 2.2 psig, for venting using the standby gas treatment system ("SGTS") to restore containment pressure to less than 2.2 psig. Multiple piping pathways are available to reduce containment pressure below 2.2 psig. First is a two-inch (2") low capacity path; second is a twenty-inch (20") high capacity path; and third is the one-inch (1 ") Containment Atmospheric Dilution ("CAD") system path. These three paths are directed through the SGTS. Should use of these three paths fail to restore pressure to less than 2.2 psig, the procedures direct that the eight-inch (8") DTV is to be used before reaching a primary containment pressure of 56 psig. The DTV bypasses the bulk of the SGTS, but connects to the SGTS discharge piping that leads to the main stack, an elevated release point. Based on the data provided in the Report, it appears that pressure inside Fukushima Unit l's primary containment reached 101.5 psig before venting was performed, approaching twice the level (56 psig) at which Pilgrim's procedures would require opening the DTV.

32. (JRL) Second, Pilgrim's procedures provide the Control Room Shift Manager with the authority and direction to utilize the DTV long before reaching a level that could challenge the primary containment, so that authorization from someone outside the plant is not needed.

Based on multiple references in the Japanese Government Report, the level of authority required to allow use of the DTVs at Fukushima was a "Minister" level in the government. With multiple

• nuclear units involved, and infrastructure unavailable because of the earthquake, tsunami, and nuclear emergency, the delays in operating the DTV are therefore explainable, but would not be analogous to Pilgrim, where the decision and authority to operate the DTV rest with the control room Shift Manager.

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33. (JRL) It is also worth noting that Pilgrim operators receive classroom and simula-tor training on venting the containment, which is repeated approximately every two years. Also, the SGTS discharge piping to the main stack is operated routinely to manage drywell pressure.

D. Pilgrim Watch Erroneously Asserts that the Fukushima Daiichi Units 1-3 Primary Containments Catastrophically Failed

34. (JRL, LAP, KRO) Also incorrect are Pilgrim Watch's claims that the primary con-tainments for all of the Fukushima Units experienced catastrophic failure, purportedly as a result of DTV-related failures. According to our review of the Report, the primary containments for Units 1 and 3 suffered no catastrophic failure and continue to contain the overwhelming majority of the radioactive inventory.

35. (JRL, LAP, KRO) Pilgrim Watch overstates the significance of the hydrogen ex-plosions that occurred at Fukushima Units 1 and 3, which did not occur in the primary contain-ment. For Units 1 and 3, the Report provides ample evidence that the hydrogen explosions at Units I and 3 occurred in the reactor building structure, sometimes referred to as the "secondary" containment, and not in the primary containment. Report at 9, 32-33, IV-6. See also IAEA Re-port at 32 ("the first hydrogen explosion occurred at the site in the Unit I reactor building"), 33

("a hydrogen explosion occurred in the Unit 3 reactor building"). Further evidence that the pri-mary containments for Units I and 3 did not "catastrophically" fail is that data summarized from the Report show no indication of rapid pressure increase, followed by rapid pressure decrease, for these Units. Figure 1 below depicts ourunderstanding of where the explosions occurred, out-side the primary containment, for Units 1 and 3.

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Figure 1. Cross-sectional view of Fukushima Primary and Secondary Containments At FuklShinia D~illhI Unlt 1~'nd 3S[pent Fuel Pool~

Rector Vrý'c ;e

• Unit 2

~A':, p*~it ; Fri ut n ?i~ IT,,r nr rvCcn.l), fl,:

36. (JRL, LAP, KRO) The distinction between the primary containment and reactor building is important because the primary containment is the structure designed to contain radio-active releases from the reactor coolant pressure boundary. The reactor building houses the pri-mary containment, which is the robust steel structure that houses the reactor vessel containing the nuclear fuel. Although the leakage pathways have not been clearly identified, hydrogen (and radioactive material) was leaked to the Fukushima Units 1 and 3 reactor buildings, ultimately resulting in hydrogen explosions. The result is that some gases that were intended to vent di-rectly to the atmosphere first collected in the reactor buildings and then were emitted into the at-mosphere after the explosions.

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37. (JRL, LAP, KRO) The information on the status of Unit 2 is not as clear as it is for Units 1 and 3. With respect to Unit 2, it is suspected that a hydrogen explosion occurred in the torus room, i.e., the room housing the torus at the lower elevations of the reactor building. Re-port at 11, IV IV-35, IV-63. The location of the suspected explosion is depicted in Figure I above. Although the torus is part of the primary containment, the torus room is not. It is possi-ble, however, that the torus itself, which is also known as the suppression chamber, was dam-aged. Id. at V-33.

38. (JRL, LAP, KRO) Further demonstrating that the primary containments of Units I and 3 have not "catastrophically failed" are the estimates provided in the Report of how much of the radioactive inventories from each unit was released. For Unit 1, the Report states that TEPCO estimates that approximately 1% of total iodine and less than 1% of other radionuclides were released. Report at IV IV-43. A "cross-check" analytical analysis performed by Ja-pan's Nuclear and Industrial Safety Agency ("NISA") estimates the releases at 1% tellurium, 0.7% iodine, and 0.3% cesium. Id. at IV-43. For Unit 3, TEPCO estimates that 0.5% of iodine was released, and NISA estimates that 0.4% - 0.8% of iodine, and 0.3% to 0.6% of other nu-clides, were released. Id. at IV-75. These estimates are consistent with field measurements and computer code simulations of the accident progression in each of the affected units. Thus, esti-mates are that the overwhelming majority of the radioactive inventory for Units 1 and 3 remains contained.

39. (JRL, LAP, KRO) Even for Unit .2, whose status is not as clearly understood as those of Units 1 and 3, TEPCO estimates a release fraction of "less than about 1%" for iodine and other radionuclides, whereas NISA estimates a wider range of releases, i.e., 0.4% to 7% for iodine nuclides, 0.4% to 3% for tellurium nuclides, and about 0.3% to 6% for cesium nuclides.

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Id. at IV-59. Thus, even if the Unit 2 suppression chamber has been compromised, it appears that 93%-99% of the radionuclide inventory remains contained.

40. (JRL, LAP, KRO) The potential for hydrogen explosions at nuclear reactors is not new information. This potential has been recognized by the industry at least since the accident at Three Mile Island, and regulations are in place to ensure that combustible gases are controlled to minimize this potential (see 10 C.F.R. § 50.44). These regulations focus on preventing hydrogen explosions that would damage the primary containment because it is the structure that is de-signed to contain radioactive releases. Mark I primary containments, such as at Pilgrim, have an inert atmosphere (consisting of non-combustible nitrogen gas) to preclude the possibility of a hydrogen combustion event within the containment in the event of a design basis accident. In-deed, the Japanese Government Report notes that, whereas measures were in place to prevent hydrogen explosion within the Fukushima primary containments, such measures were not in place to prevent hydrogen explosion in the reactor building. Report at IV-6.

41. (JRL, LAP, KRO) In summary, the known facts as summarized in the Report con-tradict Pilgrim Watch's assertions that all three Fukushima DTVs failed, and that all three Fuku-shima primary containments suffered catastrophic failure.

11. PILGRIM'S SAMA ANALYSIS CONSIDERS ACCIDENTS INVOLVING FAIL-URE TO VENT THE TORUS, HYDROGEN EXPLOSION, PRIMARY CON-TAINMENT BREACH, AND LARGE RADIOACTIVE RELEASES

42. (KRO, LAP) Contrary to Pilgrim Watch's assertions, PW Request at 1-2, 6, 29, the Pilgrim SAMA analysis assumed realistic probabilities that an accident would occur, and consid-ered pressure buildup within the primary containment, operator error regarding failure to vent, failure of the DTV to operate as intended, primary containment breach, and large radioactive re-leases, as discussed in the following paragraphs.

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A. SAMA Analysis Background Relevant To Pilgrim Watch's Claims

43. (KRO, LAP) As described in the license renewal application environmental report

("ER"), the SAMA analysis is based on a three-level, plant- and site-specific probabilistic safety analysis ("PSA") (also referred to as probabilistic risk analysis, or "PRA"). The first two levels of the PSA are used to determine the radioactive releases from the plant used to calculate off-site consequences. The Level 1 PSA assesses the accident progression scenarios that could lead to core damage. The Level 2 PSA examines the response of the containment and its protection sys-tems to the accident phenomena and related physical processes from core damage accident se-quences that can lead to a radiological release.

44. (KRO, LAP) Plant damage states ("PDS") provide the interface between the Level 1 and Level 2 analyses (i.e., between core damage accident sequences and fission product release categories). In the PDS analysis, Level I results are grouped ("binned") according to plant char-acteristics that define the status of the reactor, containment, and core cooling systems at the time of core damage. This ensures that systems important to core damage in the Level 1 event trees, and the dependencies between containment and other systems, are handled consistently in the Level 2 analysis. A PDS therefore represents a grouping of Level 1 sequences that defines a unique set of initial conditions that are likely to yield a similar accident progression through the Level 2 containment event trees ("CETs") and the attendant challenges to containment integrity.

Table E. 1-3 of the license renewal application environmental report ("ER") provides a list of the risk-significant basic events (component failures, operator actions, and initiating events) from the Level I PSA model and indicates the SAMA(s) evaluated in the cost-benefit analysis for each of them. Table E. 1-8 of the ER summarizes the core damage accident sequence PDSs.

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45. (KRO, LAP) The Pilgrim Level 2 model includes two types of analyses: (1) a de-terministic analysis of the physical processes for a spectrum of severe accident progressions, and (2) a probabilistic analysis in which the likelihood of the various outcomes are assessed. The deterministic analysis examines the response of the containment to the physical processes during a severe accident and analyzes on a deterministic basis under what conditions the containment will fail. It includes consideration for several hydrodynamic and heat transfer phenomena that occur during the progression of severe accidents, including debris coolability, pressure spikes due to ex-vessel steam explosions, direct containment heating, molten debris filling the pedestal sump and flowing over the drywell floor, containment bypass, explosion of hydrogen, thrust forces at reactor vessel failure, and liner melt-through.

46. (KRO, LAP) The probabilistic element of the Level 2 analysis consists of a con-tainment event tree ("CET") with functional nodes that represent different phenomenological events and containment protection system status and determines probabilistically for each func-tional node the likelihood of containment failure or containment bypass. Table E. 1-5 of the ER lists the functional nodes considered in the Level 2 analysis and describes the failure mechanisms that contribute to each. As discussed in Subsections B-D below, these different nodes consider failure to vent, hydrogen explosion and containment breach, and large radioactive releases from the containment.

47. (KRO, LAP) Evaluation of the forty-eight PDSs through the CET resulted in hun-dreds of Level 2 accident progression sequences and thus hundreds of source terms for internal initiators, making calculation with MACCS2 cumbersome. Thus, following standard PRA prac-tice, the source terms were grouped into a smaller number of source term groups defined in terms of similar properties, with a frequency-weighted mean source term for each group. The conse-20

quence analysis source term groups are represented by 19 collapsed accident progression bins

("CAPBs"). The CAPBs 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 postulated consequences. The 19 CAPBs are presented in ER Table E. 1-9. The 19 CAPBs ac-count for postulated system, structure, and component failures, the status of the reactor pressure vessel, the status of the containment, and accident sequence timing. Each CAPB represents a different combination of plant feature status and release mechanism and has a characteristic fre-quency and source term release based on attributes of the accident. The CAPBs have different characteristics to describe the occurrence of core damage, the occurrence of vessel breach, pri-mary system pressure at vessel breach, the location of containment failure, the timing of con-tainment failure, and the occurrence of core-concrete interactions.

48. (KRO, LAP) The source term for each CAPB comes from the Level 2 deterministic analysis which considers the hydrodynamic and heat transfer phenomena that occur during the progression of each CAPB severe accident. To simplify, for each CAPB, the deterministic analysis considers what plant features have failed and what plant features have succeeded in that accident sequence. It then considers the hydrodynamic and heat transfer phenomena that would apply with the plant in that configuration and determines (1) the fraction of the core inventory of each of nine radionuclide groups present in the reactor core that would be released; (2) the heat energy in the plume associated with the release (which will cause the plume to rise); (3) the height of the release; (4) the timing of release; and (5) the release duration. These parameters are the source terms for each of the CAPBs reported in ER Table E. I1 -I1 and input to the MACCS2 analysis.

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B. Accident Probability

49. (KRO, LAP) Pilgrim Watch contends that the Pilgrim SAMA analysis assumed "very low probabilities.., that any accident would occur at all." PW Request at 1. To the con-trary, Pilgrim used a site-specific estimate of accident probability, which Pilgrim Watch does not otherwise challenge except for in this bare assertion. The Pilgrim PSA model used realistic es-timates of the probability that an accident will occur. Those initiating events that challenge nor-mal plant operation and that require successful mitigation to prevent core damage were identified using a structured, systematic process for identifying initiating events that accounts for plant-specific features. Generic analyses of similar plants were reviewed to ensure the list of chal-lenges included in the model accounts for industry operating experience. A systematic evalua-tion of each plant system was undertaken to assess the possibility of an initiating event occurring due to a failure of the system. Each initiating event frequency was calculated accounting for relevant generic and plant-specific data.

50. (KRO, LAP) For events initiated by natural phenomena, such as earthquakes, high winds, and floods, historical evidence of these types of events and the geological and topog-raphical configurations near the Pilgrim plant were considered in determining the initiating event frequency.

C. Failure to Vent

51. (KRO, LAP) Pilgrim Watch claims that the SAMA analysis is deficient because it does not consider, in the event of an accident, (i) pressure-build up within the containment; (ii) a significant delay in even attempting to vent the containment because of operator error; and (iii) failure/inoperability of the direct torus vent. Contrary to Pilgrim Watch's claims, the Pilgrim SAMA analysis considers each of these factors.

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52. (KRO, LAP) Following a core damage event with reactor pressure vessel failure, the Level 2 model assumes that failing to vent the containment could result in overpressure of the containment or in hydrogen explosion within the containment. As can be seen in ER Table E. 1-5, the node Early Containment Failure ("CFE") considers that the containment may fail soon af-ter failure of the reactor pressure vessel due to overpressure or hydrogen explosion. The CFE states:

This top event node considers the potential loss of containment integrity at, or be-fore, vessel failure. Several phenomena are considered credible mechanisms for early containment failure. They may occur alone or in combination. The phe-nomena are containment isolation failure; containment bypass; containment over-pressure failure at vessel breach; hydrogen deflagration or detonation; fuel-coolant interactions (steam explosions); high pressure melt ejection and subse-quent direct containment heating; and drywell steel shell melt-through.

ER Table E. 1-5 (emphasis added). Thus, the SAMA analysis does take into account the fact that pressure can build up in the containment following an accident.

53. (KRO, LAP) Furthermore, the probability of failure of each functional node is de-termined by a subordinate fault tree analysis. The subordinate fault tree for the Early Contain-ment Failure node includes operator failures to act as well as component failures that could result in failure to vent and an early failure of containment. CAPB 4 through CAPB 11 include acci-dent sequences in which early containment failure occurs (see ER Table E. 1-9). Thus, the buildup of containment pressure due to failure of venting - both from an operator's failure to vent and from physical failure of the DTV - contributes to, the accident sequences in CAPBs 4 through 11.

54. (KRO, LAP) Pilgrim Watch recognizes that failure to vent the primary containment is considered in the SAMA, although it points to only one basic event, CIV-XHE-FO-DTV, Op-erator fails to vent containment using direct torus vent (DTV). See PW Request at 23. Pilgrim 23

Watch claims, however, that the SAMA analysis fails to take into account physical failure of the venting system. PW Request at 23. That is not true. As demonstrated below in Table 1 (which summarizes information taken from ER Table E. 1-3), multiple failure modes account for physi-cal failure of the venting system.

Table 1. Containment Venting Failure Modes involving DTV (ER Table E.1-3)

Failure mode Description Failure of DTV valve AO 5042B to Represents random failure of DTV valve AO 5042B to open on open (CIV-AOV-CC-5042B) demand, resulting in loss of containment venting capability to control containment pressure Failure of DTV valve AO 5025 to Represents random failure of DTV valve AO 5025 to open on open (CIV-AOV-CC-A5025) demand, resulting in loss of containment venting capability to control containment pressure.

Failure of DC breaker to supply Represents random failure of DC circuit breaker 72-175 to pro-power to DTV valve AO 5042B vide power to DTV valve AO 5042B, causing failure of the (DC1-CBR-CO-72175) valve to open on demand, resulting in loss of containment vent-ing capability.

Failure of DC breaker to supply Represents random failure of DC circuit breaker 72-165 to pro-power to DTV valve AO 5025 vide power to DTV valve AO 5025, causing failure of the valve (DCI-CBR-CO-72165) to open on demand, resulting in loss of containment venting capability.

Failure of DTV valve AO 5042B Represents random failure of the control circuit of DTV valve control circuit (CIV-RCK-NO- AO 5042B, causing failure of the valve to open on demand, 5042B) resulting in loss of containment venting capability to control containment pressure.'

Failure of DTV valve AO 5025 con- Represents random failure of the control circuit of DTV valve trol circuit (CIV-RCK-NO-A5025) AO 5025, causing failure of the valve to open on demand, re-sulting in loss of containment venting capability to control con-tainment pressure.

Operator fails to vent containment Represents operator failure to recognize the need to using direct torus vent (DTV) (CIV- vent the torus for pressure reduction during loss of containment XHE-FO-DTV) heat removal accident sequences

55. (KRO, LAP) Although Pilgrim Watch recognizes that the Pilgrim SAMA analysis takes into account an operator's failure to vent, Pilgrim Watch nevertheless claims that Entergy did not "consider what actually happened at Fukushima - operators consciously deciding not to open the DTV for fear of serious contamination offsite." Request at 23. Pilgrim Watch's con-24

cern that the Pilgrim SAMA did not consider an operator's motive for failing to open a vent is incorrect and otherwise irrelevant. As previously discussed, the Report shows that the DTV was operated for Units 1 and 3, and at least attempted to be operated for Unit 2.

56. (KRO, LAP) Further, as shown in Table 1 above, the probability that the operators will fail to vent containment using the DTV is considered in basic event CIV-XHE-FO-DTV.

The failure probability for this event was calculated using PRA Human Reliability Analysis

("HRA") techniques. HRA evaluates the individual tasks necessary to perform an action, the time available to perform the action, the time it takes to perform the action, and factors which influence the ability of the operators to successfully perform the action. The factors influencing the ability of an operator to successfully perform an action are called performance shaping fac-tors. Consideration of the impact of each performance shaping factor is plant-specific and se-quence-specific. Also, the influences are confirmed by such techniques as talk-throughs, walk-downs, field observations, simulations, and examination of past events in order to be realistic.

Examples of performance shaping factors considered in the Pilgrim HRA include the following:

• Applicability and suitability of training and experience.

• Suitability of relevant procedures and administrative controls.

• Availability and clarity of instrumentation (cues to take actions as well as to con-firm expected plant response to the action).

• Time available and time required to complete the action, including the impact of concurrent and competing activities.

• Complexity of required diagnosis and response.

• Workload, time pressure and stress.

• Team/crew dynamics and crew characteristics.

• Available staffing and resources.

• Ergonomic quality of human-system interface.

• Environment in which the action needs to be performed.

• Accessibility and operability of equipment to be manipulated.

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" The need for special tools (keys, ladders, hoses, clothing such as to enter a radiation area).

• Communications (strategy and coordination) as well as whether one can be easily heard.

" Special fitness needs for situations expected to involve the use of heavy or awkward tools/equipment, carrying hoses, climbing, etc.

" Consideration of "realistic" accident sequence diversions and deviations (e.g., ex-traneous alarms, failed instruments, outside discussions, sequence evolution not ex-6 actly like that trained on).

57. (KRO, LAP) Thus, the Pilgrim SAMA analysis considers a wide range of factors affecting human performance. Pilgrim Watch nowhere explains how its concern that the SAMA analysis fails to consider an operator's motive for failing to open the vent - e.g*, fear of serious contamination offsite - would alter the Pilgrim SAMA analysis. In any event, failure to operate the vent, for whatever reason, would end in the same result - a loss of containment venting capabil-ity to control containment pressure. Thus, Pilgrim Watch has not shown any reason why the many factors considered in the Pilgrim SAMA analysis are inadequate in light of the events at Fuku-shima.

D. Hydrogen Explosion & Containment Breach

58. (KRO, LAP) Pilgrim Watch claims that the Pilgrim SAMA analysis fails to con-sider "containment failure/explosions that resulted in significant ongoing offsite consequences" and "catastrophic failure of the containment." PW Request at 2, 6. To the contrary, as de-scribed above for vent failure, early containment breach is considered in node Early Containment Failure ("CFE") of the Pilgrim PSA (Table E. 1-5 of the ER), which contributes to the accident sequences in CAPBs 4 through 11. The Level 2 model also considers that containment could be breached later in an accident via node Late Containment Failure ("CFL"). In these accident se-quences, venting is successful, but containment fails later due to long-term steam and non-6 See NUREG- 1792, Good Practices for Implementing Human Reliability Analysis (Apr. 2005) at Table 5-1.

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condensable gas generation from the attack of molten core debris on concrete. CAPBs 12 through 15 include accident sequences in which late containment failure occurs. Thus, contain-ment breach contributes to accident sequences in CAPBs 4.through 15.

59. (KRO, LAP) The potential for hydrogen explosion within the primary containment (which, as previously discussed, did not occur at Fukushima Units I and 3) is also considered in the Level 2 PSA model for Pilgrim. As described earlier, hydrogen explosion is considered a credible mechanism for early primary containment failure and therefore contributes to the CFE functional event node, which considers the potential loss on containment integrity at, or before, reactor vessel failure. The PSA fault tree combines initiating event probabilities, component failure probabilities, and human action failure probabilities. Thus, the subordinate fault tree for CFE includes component and operator action failures that could result in a buildup of hydrogen in the primary containment.

60. (KRO, LAP) Collapsed accident progression bins CAPB 4 through CAPB 11 in-clude accident sequences in which early containment failure occurs. Therefore, primary con-tainment hydrogen explosion has been appropriately considered in the Level 2 PSA model used in the Pilgrim SAMA analysis.

61. (KRO, LAP) The potential for hydrogen explosion within the reactor building is also appropriately considered in the Level 2 PSA model for Pilgrim, which was used in the SAMA analysis. The adverse impact of an explosion of hydrogen that had accumulated in the reactor building instead of being emitted through the plant stack would be a ground level radio-logical release to the environment, and no more radioactive material than had the release oc-curred through the plant stack. The CET in the Pilgrim Lev el 2 PSA model includes functional node Reactor Building ("RB"), which is used to assess the ability of the reactor building to retain 27

fission products released from containment. Success of event RB is defined to be a reduction in the containment release magnitude. Failure of event RB means that containment failures or by-passes are released directly to the environment rather than being held up in the reactor building.

62. (KRO, LAP) In summary, the Pilgrim SAMA analysis does consider the informa-tion concerning pressure buildup within containment, failure of the DTV, hydrogen explosions, and failure of containment that Pilgrim Watch contends is missing. Pilgrim Watch offers no ba-sis to contend otherwise.

E. Large Radioactive Release

63. (KRO, LAP) Pilgrim Watch contends that the "offsite consequences of contain-ment failure would be huge" and asserts that such huge consequences were not "properly fac-tored into Entergy's SAMA." PW Request at 24. See also id. at 34, Gunderson Affidavit ("huge amounts of radiation will be released"). To the contrary, large radioactive releases are consid-ered in the MACCS2 model to determine the offsite consequences from postulated severe acci-dents in the Pilgrim SAMA analysis. Indeed, the large radioactive releases considered in the Pil-grim SAMA analysis bound several times over the releases that occurred from Fukushima.

64. (KRO, LAP) To the extent that Pilgrim Watch contends that the Pilgrim SAMA does not consider that "huge amounts of radiation will be released" in light of events at Fuku-shima, its contention is wrong. The Pilgrim SAMA analysis postulates a range of energetic se-vere accident release events with sufficient energy to breach the reactor engineering safety sys-tems or barriers (e.g*, breach the reactor vessel or fail the containment structure) or lead to their bypass. The potential for direct release of radioactive material from containment to the envi-ronment is appropriately considered in the PSA model used in the Pilgrim SAMA analysis.

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65. (KRO) As discussed in the previous Declaration supporting Entergy's Answer op-7 posing Pilgrim Watch's first request for a hearing on a post-Fukushima SAMA Contention, comparison of the radiological releases assumed in the single-unit Pilgrim SAMA analysis shows that the Pilgrim SAMA analysis accounts for severe accident releases that more than bound the reported releases from all of the Fukushima units combined. Sowdon/O'Kula Declaration at ¶

41. Thus, the Pilgrim SAMA analysis has considered "huge" radioactive releases.

66. (KRO) Subsequent to the development of the comparisons in the Sowdon/O'Kula Declaration, the Japanese authorities increasedtheir estimate of the radioactive release from Fu-kushima by about 22% above the estimates used in the Sowdon/O'Kula Declaration. 8 This in-crease has no effect on the conclusions drawn from the comparisons made in Table 5 of the Sowdon/O'Kula Declaration. As noted there, "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 1-131 releases from all of the Fukushima facili-ties by about a factor of two (1.78) and the estimated Cs-137 releases by about a factor of three (2.66)." Sowdon!O'Kula Declaration at 24 n. 16. Thus, the radionuclide releases assumed in the Pilgrim SAMA analysis far exceed actual releases at Fukushima.

Entergy's Answer Opposing Pilgrim Watch's Request for Hearing on Post-Fukushima SAMA Contention (June 6,2011); Declaration of Dr. Thomas L. Sowdon and Dr. Kevin R. O'Kula in Support of Entergy's Answer Op-posing Pilgrim Watch Request for Hearing on Post-Fukushima SAMA Contention (June 6, 2011) ("Sowdon/

O'Kula Declaration").

8 The comparisons made in Sowdon/O'Kula Declaration were based on release estimates for cesium and iodine made by the Japan Nuclear Safety Commission ("NSC"), which estimated that 630 petabecquerels ("PBq") or 6.3E+1 7 Bq had been released from Fukushima. In June, the Nuclear and Industrial Safety Agency ("NISA"),

another Japanese regulatory authority, increased its original estimate of 370 PBq (which had been much lower than the NSC's estimate used in the Sowdon/O'Kula Declaration) to 770 PBq, which is approximately 22%

higher than the NSC estimate. The iodine equivalent radiological hazard for both the NSC and the NISA are de-termined by multiplying the estimated Cs-137 release by a factor of 40 to account for its higher radiological haz-ard, and adding this to the estimated iodine activity.

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67. (KRO) In addition, the fraction of the Fukushima Units' core inventories released into the environment, based on measurements and computer model backed calculations reported to date by the Japanese government, Report at IV IV-43, IV-59, IV-75, is more than bounded by the Pilgrim SAMA-basis CAPBs. Table 2 below shows the fraction of core inven-tory release for iodine, cesium and tellurium groups for each of the nineteen CAPBs in the Pil-grim SAMA analysis. The same fractional releases are then listed for Fukushima reactor units (1, 2, and 3) as determined by the Tokyo Electric Power Company ("TEPCO") and the Japanese regulatory agency, Nuclear and Industrial Safety Agency ("NISA"). The Table shows that frac-tional release from the three units is limited to approximately one percent of the inventory at re-actor shutdown for iodine (I), cesium (Cs), and tellurium (Te) groups, and that most of the air-borne release is due to Unit 2. All noble gases are considered fully released (-100%) under these circumstances. Also, NISA tended to be more conservative in its estimates, and provided a range in the Unit 2 release of iodine, cesium, and tellurium due to uncertainty with that unit.

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Table 2. Pilgrim SAMA I, Cs, and Te CAPB Release Fractions Compared to Those Reported in Government of Japan Report Iodine fraction Cesium frac- Tellurium frac-Release Mode frioio todin tion tion CAPB-I 0.00002% 0.00002% 0.000%

CAPB-2 0.0048% 0.0047% 0.00002%

CAPB-3 0.01% 0.005% 0.0002%

CAPB-4 4.90% 2.62% 0.004%

CAPB-5 7.86% 3.68% 0.004%

CAPB-6 4.02% 2.32% 0.15%

CAPB-7 6.11% 2.94% 0.13%

CAPB-8 29.80% 27.20% 0.00%

CAPB-9 7.61% 7.07% 0.001%

CAPB-10 28.00% 24.90% 1.11%

CAPB-1 1 17.30% 14.10% 1.00%

CAPB-12 0.006% 0.004% 0.00001%

CAPB-13 0.80% 0.60% 0.02%

CAPB-14 2.88% 2.67% 0.00%

CAPB-15 27.60% 26.80% 0.13%

CAPB-16 6.71% 3.26% 0.04%

CAPB-17 36.20% 33.70% 0.13%

CAPB-18 9.76% 6.25% 2.09%

CAPB-19 40.30% 37.70% 6.87%

Fukushima Unit 1 1% "<1% other radionuclides" (TEPCO estimate)

Fukushima Unit 2 (EC estimaUite21% "<1% other radionuclides" (TEPCO estimate)

Fukushima Unit 3 0.7% 03 em (TEPCO estimate)

Fukushima Unit 1 0.7% 0.3% 0%

(NISA estimate)

Fukushima Unit 2 (IAetmt) 0.4% -7% 0.3 -6% 0.4% -3%

Fukushima Unit 3 0.4% - 0.8% "0.3% - 0.6% other nuclides" (NISA estimate)

68. (KRO) The Pilgrim SAMA CAPB fractional releases are numerically larger, or bound the release estimates from the Fukushima units. When compared to the higher release es-timates for just Unit 2 provided by NISA, the Pilgrim SAMA analysis release assumptions are 31

larger than the Fukushima Unit 2 NISA estimate in 9 of the CAPBs for iodine, and in 8 of the CAPBs for cesium. While only one of the Pilgrim SAMA CAPBs bounds the Unit 2 NISA esti-mate of tellurium, the two isotopes reported in the Report, Te-129m and Te-132, have relatively short half-lives of 34 days and 3 days, respectively, and are not contributors to long-term SAMA risks (PDR and OECR), which dominate the SAMA analysis. Isotopes of strontium and cesium are major contributors to SAMA risks (PDR and OECR). An additional major conservatism in the Pilgrim SAMA analysis is that lower volatility, but high-risk, radionuclide groups are as-sumed released as part of the airborne source terms, whereas these have not been released in de-tectable quantities from Fukushima. This includes strontium (Sr), ruthenium (Ru), lanthanum (La), cerium (Ce), and barium (Ba).

69. (KRO, LAP) In summary, contrary to Pilgrim Watch's claims, the Pilgrim SAMA analysis considers "huge" radioactive releases. Indeed, the Pilgrim SAMA analysis considers radioactive releases far larger than those which have occurred at Fukushima.

IV. PILGRIM WATCH'S REMAINING CLAIMS ARE IMMATERIAL

70. (JRL, KRO, LAP) Pilgrim Watch makes several other incorrect or otherwise im-material assertions concerning Pilgrim's components. In summary, what Pilgrim Watch claims ought to be considered has, in fact, been considered in the Pilgrim SAMA analysis.

71. (JRL) Pilgrim Watch contends that the DTV is buried pipe subject to corrosion and therefore could fail. PW Request at 20-21. Pilgrim Watch's assertion is inaccurate. The DTV itself- i.e., the portion of piping installed when the DTV modification was made pursuant to NRC requirements - is within plant buildings. It is therefore not subject to corrosion as a result of being buried. The DTV is connected to the standby gas treatment system ("SGTS") discharge piping, which runs to the main stack to provide an elevated release point for suppression chain-32

ber gases. A. portion of the SGTS discharge piping is buried and is subject to inspection under the buried piping and tanks inspection program, further assuring that it will not corrode. In addi-tion, a portion of the SGTS discharge piping was internally inspected in April 2011, which re-vealed no evidence of corrosion.

72. (KRO, LAP) In any event, Pilgrim Watch's concern about the alleged corrosion and subsequent failure of the DTV is immaterial with respect to the Pilgrim SAMA analysis. As previously discussed, failure of the DTV is considered in the Pilgrim SAMA analysis.

73. (JRL) Pilgrim Watch also complains that the DTV has no redundancy. See, e.g.,

PW Request at 20. Among other things, Pilgrim Watch asserts that, "[a]bsent redundancy," the failure of required electrical power means that the DTV system would be unavailable, allegedly because there "would be no way to open the normally-closed isolation valves." Id. However, as previously discussed, prior to using the DTV, there are multiple, diverse and redundant systems available for depressurizing the primary containment. For example, the Containment Atmos-pheric Dilution System, and the low and high capacity paths through the Standby Gas Treatment System are available for this purpose. Further, the DTV system can be manually operated in the absence of electrical power. Indeed, the events at Fukushima demonstrate this precise circum-stance. See, e.g., Report at IV-41.

74. (KRO) Pilgrim Watch also complains that the Pilgrim DTV is not filtered, and that one cannot assume that the suppression chamber will scrub radioactive contaminants before re-lease into the environment because "there will not be sufficient water available to trap the radio-active materials of concern." PW Request at 9, 17-19, 21-22, 23. As such, Pilgrim Watch claims that the SAMA analysis must be redone to take this new and significant information into account. Id. at 19. However, contrary to Pilgrim Watch's claims, were the DTV to be operated 33

during an accident, the DTV would draw from the air space within the suppression chamber or torus. All gases from the drywell would pass through several feet of torus water prior to venting.

This water would effectively filter particulate and volatile fission products, depending upon the accident sequence and the temperature of the water. Furthermore, the additional volume pro-vided by the torus flow path would provide additional travel time for gas flow, and thus lead to longer times prior to release. The additional time would reduce the source term of the high-dose consequence, but short-lived radionuclides, most notably, the noble gases and shorter-lived io-dine radionuclides (1-132, 1-133, 1-134, and 1-135).

75. (KRO) As noted in the NUREG-1 1509 analysis of a similar plant, Peach Bottom, the pressure-suppression chamber can provide significant reduction in overall release to the envi-ronment. Especially for in-vessel severe accident releases, the radionuclides must pass through the suppression chamber "where substantial decontamination is possible. In sequences where the drywell spray system is operable, the ex-vessel release will also be mitigated by the spray or an overlaying pool of water. Both the in-vessel and ex-vessel releases will receive further attenua-tion in the reactor building before release to the environment. Even if the decontamination factor of some of these stages is small, the overall effect is to make the likelihood of a very large re-lease quite small." NUREG-1 150 at 4-15. The NUREG-l 150 study concluded that, while vari-able depending on the severe accident conditions, the overall effect of pressure-suppression chamber could be particularly large in the reduction of in-vessel releases and was estimated to lead to decontamination factors (fraction of the radioactive material removed) from 1.2 to 4000 with a median of 80 for flow through the safety relief valvelines. Id.

9 NUREG-1 150, Severe Accident Risks: An Assessment for Five Nuclear Power Plants, Final Summary Report (Dec. 1990) ("NUREG- 1150").

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76. (KRO, LAP) And in any case, Pilgrim Watch's concern regarding the unavailabil-ity of the suppression chamber to scrub contaminants before their release is immaterial. The Pil-grim SAMA analysis considers early containment failure and resultant large radioactive releases without scrubbing by the suppression pool. As can be seen in ER Table E. 1-5, the node Early Release to Torus (EPOOL) considers that an early containment failure may occur in a part of the primary containment that enables the large radioactive release to occur without scrubbing by the suppression pool. The node Early Release to Torus (EPOOL) states, This top event node considers the importance of early torus pool scrubbing in mitigating the magnitude of fission products released from the damaged core.

Success implies that fission product transport path subsequent to early contain-ment failure is through the torus water and the torus airspace. Failure involves a release into the drywell.

ER Table E. 1-5. Thus, the SAMA analysis does take into 'account the fact that an early contain-ment failure could occur without scrubbing of the release by the suppression pool. CAPB 8 through CAPB 11 include accident sequences in which early containment failure occurs in a part of the primary containment that enables the large radioactive release to occur without scrubbing by the suppression pool (see ER Table E. 1-9).

77. (KRO, LAP) Pilgrim Watch also contends that a "rational SAMA" would require redesign of the DTV to make it passive. PW Request at 12-13. However, in response to an NRC request for additional information, Entergy updated the Pilgrim SAMA analysis in July 2006 to evaluate the cost benefit of a passive design direct torus vent instead of the existing direct torus vent to mitigate failure of the operators to perform direct torus venting. The proposed SAMA would modify the air operated valves and the associated solenoid valves in the DTV pathway so that the air operated valves fail open on loss of air and nitrogen or on loss of power, which would remove the need for the operators to take action to open the valves. Evaluation of this SAMA 35

determined that it was not cost effective. Pilgrim Watch nowhere materially disputes or chal-lenges this evaluation.

V. CONCLUSION

78. (JRL, LAP, KRO) Contrary to Pilgrim Watch's assertions, the known facts to date do not show that the DTV failed to operate at any of the Fukushima units. Nor do they show that the primary containments for all three units suffered catastrophic failures. But, even if these as-sertions were true, there would be no need to re-do the Pilgrim SAMA analysis because that analysis already considers failure of the DTV to operate and failure of the primary containment.

Pilgrim Watch's Request nowhere disputes those aspects of Pilgrim's SAMA analysis that con-sider DTV failure and primary containment failure. Moreover, the releases considered in the Pilgrim SAMA analysis are much larger than the apparent releases from Fukushima.

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)

Joseph R. Lynch Lori Ann Potts Manager, Licensing Entergy License Renewal Services &

Pilgrim Nuclear Power Station ANO NFPA-805 Transition Project 600 Rocky Hill Rd. 1448 SR 333 Plymouth, MA 02360 Russellville, AR 72802 Phone: 508-830-8403 ANO-GSB-45 E-mail: jlvnch4(@entergy.com Phone: 479-858-3529 Email: Ipott9O(entnerg'y.corn Executed in Accord with 10 C.F.R. § 2.304(d)

Kevin O'Kula, Advisory Engineer URS Safety Management Solutions LLC 2131 South Centennial Avenue Aiken, South Carolina 29803-7680 Phone: 803.502.9620; Email: kevin.okLula(?diwsrs.corn 36

EXHIBIT 1 600 Rocky Hill Road (508) 830-8403 (Work)

Plymouth, MA 02360 (508) 728-1421 (Cell)

E-mail: jlynch4@entergy.com Joseph R. Lynch Jr.

Objective Senior Manager/Regulatory Affairs with 29 years of nuclear power experience and background in engineering, licensing/regulatory affairs, environmental compliance, creative problem solving, stakeholder communications, complex project manage-ment, cost control, budgeting and employee management. Strong strategic thinker and team builder.

Areas of Expertise " Regulatory Affairs/Licensing

" Project Management

" Design and Systems Engineering

" Environmental Health & Safety Compliance

• Oral and Written Communications Education Worcester Polytechnic Institute - Worcester, Massachusetts Bachelor of Science in Mechanical Engineering (BSME)

Specialized in Thermo/Fluids/Nuclear Graduate Studies in Business Management, Communications and Regulatory Compliance Numerous Internal and External Management Courses - Yankee Atomic Electric Company; Vermont Yankee Nuclear Power Corporation Professional experi- 2007-Present Entergy Nuclear Operations ence Licensing Manager Manages the PilgrimNuclear Power Station Licensing Group supportingthe operationand regu-latogy compliance of the station in accordance with NRC, State and Federalregulations,permits and statutes.

Development of all necessary letters, licensing correspondence and regulatory ap-provals from NRC, local, state and federal agencies required in support of plant operations.

" Responsible for communicating with regulators, governmental representatives and media on plant status, regulatory issues and emergent events.

" Management of a $ 6-8 million dollar annual department budget.

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2006-2007 Environmental Resources Management (ERM)

Senior Nuclear Consultant Directs ERM's national nuclear team, and leads, coordinates, or supports nearly all of the nu-clear-relatedwork on behalfof ERM nationally.

Responsible for the oversight of all technical and policy-related activities of the nuclear staff experts within ERM. Maintains a close awareness of current and emergent regulatory issues within the nuclear sector, developing trends, and indus-try initiatives. Recent projects include;

• Support of Groundwater Program development at several nuclear power plants in accordance with EPR1/NEI Guidance Documents and plant specific attributes.

• Regulatory affairs, licensing and permitting responsibility for nuclear client.

• Environmental, Health & Safety (EHS)/Due Diligence Assessments for several nuclear clients supporting Merger & Acquisition (M&A) efforts.

2003-2006 Yankee Atomic Electric Company Director, Regulatory Affairs, Licensing and Site Closure Directed and managed the Site Closure Project activitiesfor the clean-up and decommissioning of the Yankee Nuclear Power Station in strict compliance with NRC, State and Federalregula-tions,permits and statutes.

Development of strategies to obtain all necessary permitting, licensing and regula-tory approvals from NRC, local, state and federal agencies required to remediate all environmental hazards from the site in support of unrestricted re-use of the prop-erty.

" Responsible for communicating with community members, regulators, govern-mental representatives and media on project status, key company decisions and emergent issues.

" Authored the Site Closure Project Plan (SCPP), an industry first, comprehensive plan that integrated stakeholder input, corporate goals and regulatory, compliance by working with local, state and federal stakeholders to solicit input and accep-tance.

" Extensive outreach to stakeholders via written and verbal communication includ-ing site personnel, executive management, community advisory boards, local/state government leaders, town meetings, community and media events.

" Control and management of a $ 6-8 million dollar annual project budget.

2000-2003 Connecticut Yankee Nuclear Power Company Decommissioning Oversight Manager Directed and coordinated oversight of the construction and plant demolition activities at the Connecticut Yankee Nuclear Power Station in accordance with applicable regulatoy standards.

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Position responsibilities included, but were not limited to the following;

" implemented department activities in accordance with established and newly de-veloped station programs, policies and procedures.

" Assured a safety conscious work environment, including implementation of the Standards of Conduct and the Site Corrective Action Program.

" Communicated site performance to stakeholders including site personnel, execu-tive management, Independent Assessment groups, community members, regula-tors, governmental representatives and the media.

" Assured decommissioning activities did not affect safe storage of spent nuclear fuel.

" Assisted site executive management in establishing and implementing strategic plans.

1997-2000 Vermont Yankee Nuclear Power Corporation Department Manager - Design Engineering Responsiblefor management of plant design modifications, thermal/hydraulicdesign analyses and plant supportfunctional areasfor a twenty (20) engineer staff, including consultants, contractors and administrative support. Control and oversight of a $ 3-4 million dollar annual budget.

Position responsibilities included;

• Interface with U.S. Nuclear Regulatory Commission (USNRC) through Senior Resident and /or Region I Project Manager.

• Reviewed and approved all departmental design work products.

• Preparation, review and approval of Bases for Maintaining Operation (BMO).

• Preparation and maintenance of plant Design Basis Documentation (DBD).

• Preparation, review and approval of 10CFR50.59 Safety Evaluations.

• Employee performance appraisals, bonus/compensation determination and goal setting.

• Qualified adjunct instructor for providing training to engineering staff.

• Maintained the Department budget by tracking expenditures on capital and O&M Projects, contractor costs and employee salaries/benefits.

• Identified and developed Continuous Process Improvements (CPD) initiatives with plant management, supervision, and staff to improve overall performance of engineering work products.

1982-1997 Yankee Atomic Electric Company Manager- Design Engineering Fluid Systems (1995-1997)

Supervised the Design Engineering Fluid Systems stqf/ supporting the Vermont Yankee Nuclear Power Station.

Responsible for the review and approval of design change packages, calculations and analyses, 10CFR50.59 Safety Evaluations, Operability Determinations, Bases for Maintaining Operability, Design Basis Documents (DBD) and Department assigned commitments/corrective actions.

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" Responsible for oversight of the VY MOV, and Safety Classification Programs.

" Responsible for planning and scheduling of all assigned work, outage prepara-tion/implementation activities, project budget accountability and direction of con-tractor staff.

Motor-Operated Valve (MOV) Program Manager (1994-1995)

Directed Connecticut Yankee's efforts for planning and implementation of the Generic Letter 89-10 MO V Program testing and overhaul activities through the 1995 Refuel-ing Outage (RFO).

• Coordinated the completion of all documentation in support of NRC inspection and closure of the GL 89-10 imposed requirements for safety-related MOVs.

Lead Systems Engineer (1992-1994)

Project Managerfor the Millstone Unit 1 Hardened Wetwell Vent Syistem design and implementation.

• Instrumental in engineering and development of all design change documents.

• Responsible for the management of all work projects for Northeast Utilities, with-in the Systems Engineering discipline.

Senior Project Engineer (1990-1992)

Served as Project Managerfor the Yankee Nuclear Power Station (YNPS) high-pressure turbine retrofit and Main Condenser replacement projects.

• Provided project management, engineering and scheduling oversight.

• Worked extensively on condition assessment, performance monitoring and re-placement justification for.the YNPS Main Condenser.

• Negotiated the Contracts for acquisition of the HP Turbine and Main Condenser.

Systems/Senior Systems Engineer (1982-1990)

Deszgned, specified, and analyzed nuclearpowerplant fluid/air systems and equipment at the Yankee Nuclear Power Station.

" Provided technical assistance and installation supervision on primary and secon-dary plant equipment and systems.

" Designed/installed the Safe Shutdown System (Appendix R requirement for re-mote shutdown of YNPS), Emergency Diesel Generator (EDG) and Safety In-jection Building ventilation upgrades, EDG replacement and commercial grade dedication of the EDGs.

• Shift outage coordinator for the 1990 summer Refueling Outage at YNPS.

• Worked closely with the plant staff in planning, prioritizing and craft labor over-sight/support.

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EXHIBIT 2

• -Eritergy7 RESUME Lori Ann Potts EDUCATION B.S., Nuclear Engineering, The Pennsylvania State University, 1981 EXPERIENCE February 2008 - Consultant to Entergy Nuclear - NFPA-805 Transition Project Team Present Development of products and documents associated with the ANO-1 and ANO-2 Fire PRAs.

• Plant Boundary Definition, Plant Partitioning, and Fire Ignition Fre-quencies

• Fire PRA Component Selection and Fire-Induced PRA Model

• Post-Fire Human Reliability Analysis

• Fire Risk Quantification

• Fire PRA Peer Reviews Preparation, Execution, and Responses

• Risk assessments of Fire Protection non-compliances and issues February 2002 - Consultant to Entergy Nuclear - License Renewal Project Team Present

• Mechanical Aging Management Reviews and Development of Aging Management Programs for DC Cook, Pilgrim, Vermont Yankee, Pali-sades, ANO-2, J.A. Fitzpatrick, Indian Point, and Cooper License Re-newal Projects

• Responding to NRC questions on submitted applications

• Coordinating and reviewing evaluation of severe accident mitigation alternatives (SAMA) for ANO-1, ANO-2, Pilgrim, Vermont Yankee, J.A. Fitzpatrick, Indian Point, Cooper, and Grand Gulf Environmental Reports

• Peer reviewed SAMA analyses for Beaver Valley, Columbia, and Palo Verde

• Developed industry SAMA guideline (NEI 05-01) 01/1994 - 08/2001 Consultant to Entergy Operations - Arkansas Nuclear One (ANO) Nuclear Safety Analysis

• Project Manager on ANO-2 Probabilistic Safety Assessment (PSA)

Model Update

• Risk Analysis of the ANO-2 Power Uprate

• Risk Sensitivity Analysis of alternate repair criteria for ANO-1 Steam Generator tubes containing Intergranular Attack

• Power Uprate/Steam Generator Replacement modification of ANO-2

• Fortran code to calculate time to boil and time to core uncovery upon

loss of shutdown cooling

• Reviewed Safety Analyses for ANO-2 Steam Generator Replacement and Power Uprate

• Created ANO PSA Analysts' Deskguide

• Updated ANO-1 and ANO-2 PSA models and associated analyses 11/1989 - 06/1993 Arkansas Nuclear One (ANO), Entergy Operations, Inc.

Senior Engineer, Nuclear Engineering Design (05/91-06/93)

• Lead Engineer for Unit I PSA

• Responsible for documentation of Design Basis for reactivity related design basis accidents on Unit 2 Reactor Engineer I11, System Engineering (11/89 - 05/9 1)

• Responsible for Unit 2 Core Protection Calculators, Core Operating Limits Supervisory System and Excore Nuclear Instrumentation

• Defined core offload, shuffle and reload sequence for Unit 2 Cycle 9

• Startup Physics Testing and Core Monitoring Surveillances 02/1988 - 11/1989 Plant A. W. Vogtle, Georgia Power Company Senior Plant Engineer, Outage Management

• Project-2 scheduling for Refueling and mid-cycle Outages

• Performed Critical Path Analyses, Plots and Reports 01/1987 - 08/1987 Consultant to Pilgrim Power Station Senior Systems Specialist, I&C (03/87 - 08/87)

• Ensured Neutron Monitoring, Radiation Monitoring, Reactor Water Level, Turbine Generator Protection and Controls, Recirculation Sys-tem Controls and Communications systems were operational and ready for start-up of the plant from its extended outage

• Acted for Lead Systems Specialist, I&C and Electrical in his absence; supervising six engineers NPRDS Administrator (01/87 - 03/87)

• Reviewed plant Maintenance Requests, Malfunction Reports and De-sign Changes for reportable failures

" Generated failure and out-of-service reports 0 1/1985 - 12/1985 Consultant to Arkansas Nuclear One Principal Engineer

• Document Research and Engineering Evaluation for all components in both units to generate a computerized Component Database

" Supervised four technicians and five engineers 04/1984 - 12/1984 Clinton Power Station, Illinois Power Company 2

Plant Maintenance Engineer

• Resolved abnormal condition reports, commitments and audit findings

• Walk-downs of systems being turned over from Start-Up to Operations

• Developed maintenance program for Environmental Qualification of equipment

• Assisted in development of Computerized Maintenance Management System 05/1981 -03/1984 EG&G Idaho, Inc.

Three Mile Island, I&C and Electrical Program Engineer

" Performed failure and survivability testing of electrical components and instrumentation within the damaged Unit 2 reactor building

" Developed test procedures, instructed technicians in use of equipment and directed performance of tests

• Reviewed resultant data and test plans for off-site examinations

• Presented technical reports describing the program status to DOE, NRC and the industry 05/1980 - 08/1980 EG&G Idaho, Inc.

Three Mile Island, Intern

• Participated in and graphed data from survivability testing of electrical components and instrumentation within the damaged Unit 2 reactor building 3

EXHIBIT 3 KEVIN R. O'KULA Advisory Engineer URS Safety Management Solutions LLC 2131 South Centennial Avenue Aiken, South Carolina 29803-7680 Telephone: 803.502.9620 - Email: kevin.okula(iwsms.com KEY AREAS:

" Computer Model Verification and Validation 9 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 e MACCS2 Code Applications

• New Reactor Design Accident Analysis and PRA a Level 3 PRA Standard Development Support PROFESSIONAL

SUMMARY

Dr. O'Kula 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 Commission's (NRC's) 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. O'Kula 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 Site's 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 ANS's 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. O'Kula 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. O'Kula 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-2

nomena and subsequent off-site consequences in the context of severe reactor accidents. This activity supports the efforts of Sandia National Laboratories (SNL) and the Nuclear Regulatory Commission (NRC) to provide more realistic assessment of severe accidents.

Dr. O'Kula is also part of the Level 3 PRA Standard working group charged with developing an AN-SI/ANS standard for Level 3 PRA analysis. He participated in a team that conducted an SQA gap analy-sis 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 over-saw 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 Toolbox. 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 Altema-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. O'Kula 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. O'Kula also developed a one-day course and exam for the guide, which he later presented to the Oak Ridge, Paducah, and Portsmouth staff.

Dr. O'Kula 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. O'Kula 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. O'Kula 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. O'Kula 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. O'Kula 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 Unit-ed Kingdom's Nuclear Installation Inspectorate). Model was knowledge-based, and required develop-ment of an Accident Progression Event Tree (APET) for the facility in question.

Dr. O'Kula 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. O'Kula 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. O'Kula 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. O'Kula, D. C. Thoman, J. Lowrie, and A. Keller, Perspectives on DOE Consequence Inputsfor Ac-cident Analysis Applications, American Nuclear Society 2008 Winter Meeting and Nuclear Technol-ogy Expo, November 9-13, 2008 (Reno, NV).

K. R. O'Kula, F. J. Mogolesko, K-J Hong, and P. A. Gaukler, Severe Accident MitigationAlternative 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. O'Kula 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. O'Kula 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. O'Kula and D. Sparkman, Safety Software Guide Perspectivesfor the Design of New Nuclear Fa-cilities (U), Winter Meeting of the American Nuclear Society, November 13 - 17, 2005 (Washington, D.C.).

K. R. O'Kula 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. O'Kula and Tony Eng, A "Toolbox" Equivalent Processfor Safety Analysis Software, Fourteenth Annual Energy Facility Contractors Group Safety Analysis Workshop, May 1-6, 2004, Pleasanton, CA (2004).

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

K. O'Kula and J. Hansen, Implementation of Methodologyfor FinalHazardCategorizationof a DOE Nuclear Facility (U), Annual Meeting of the American Nuclear Society, June 13-17, 2004, (Pitts-burgh, PA).

K. R. O'Kula and Tony Eng, A "Toolbox" Equivalent Processfor Sqfety Analysis Software, Fourteenth Annual Energy Facility Contractors Group Safety Analysis Workshop, May 1-6, 2004, Pleasanton, CA (2004).

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K. R. O'Kula, et al., Evaluation of Current Computer Models Applied in the DOE Complex for SAR Analysis of RadiologicalDispersion & Consequences, WSRC-TR-96-0126, Westinghouse Savannah River Company (2003).

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

K. R. O'Kula, 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. O'Kula 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