LR-N11-0173, Comments Opposing NRC Acceptance of Petition for Emergency Enforcement Action for GE Boiling Water Reactors Utilizing Mark I Primary Containment Systems and Without Dedicated Class 1E Power for Cooling Elevated Irradiated Used Fuel...
| ML111570041 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/03/2011 |
| From: | Keenan J Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| LR-N11-0173, 2.206, G20110262 | |
| Download: ML111570041 (10) | |
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PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 JUN 03 2011 OPSEG LR-N11-0173 Nuclear LLC U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Hope Creek Docket 50-354: Comments Opposing NRC Acceptance of Petition for Emergency Enforcement Action for GE Boiling Water Reactors Utilizing Mark I Primary Containment Systems and Without Dedicated Class 1E Power for Cooling Elevated Irradiated Used Fuel Storage Pools Outside a Rated Containment Structure; EDO # G2011 0262
Dear Sir or Madam:
PSEG Nuclear LLC ("PSEG") is providing the following comments opposing NRC acceptance of the "Petition for Emergency Enforcement Action for GE Boiling Water Reactors Utilizing Mark I Primary Containment Systems and Without Dedicated Class 1E Power for Cooling Elevated Irradiated Used Fuel Storage Pools Outside a Rated Containment Structure" ("Petition"), dated April 13, 2011, submitted by Beyond Nuclear ("Petitioner"). As explained below, the Petition does not meet the acceptance criteria for further review under NRC Management Directive
("MD") 8.11, "Review Process for 10 C.F.R. 2.206 Petitions."
INTRODUCTION On April 13, 2011, Petitioner filed a petition under 10 C.F.R. § 2.206 requesting that the NRC take emergency enforcement action at certain General Electric ("GE") Boiling Water Reactor
("BWR") Mark I units to suspend their operating licenses in response to the accident at the Fukushima Daiichi nuclear power plant in Japan. PSEG is the licensee for Hope Creek, which is among the plants that are the subject of the Petition. PSEG hereby responds to the Petition and addresses the issues relevant to the acceptance review phase of the 10 C.F.R. § 2.206 process.
As described further below, the Petition does not meet the criteria for acceptance as a 10 C.F.R.
§ 2.206 petition, in that it fails to provide sufficient facts to warrant further inquiry by the Petition Review Board ("PRB"), and adequate alternative means for addressing Petitioner's issues exist.
Specifically, the historical issues raised in the Petition are not new and have been thoroughly evaluated over the last several decades. Also, other than the occurrence of the earthquake and the tsunami, the actual sequence of events, including the operation or failure of particular systems or components that resulted in the Fukushima Daiichi accident are not yet known, so any connection to the Mark I design is speculative at best. Further, none of the information in the Petition provides a basis for concluding that Mark I containments are unsafe to warrant the extraordinary action of suspending their current operating licenses. Therefore, PSEG requests that the PRB reject the Petition, finding that it does not meet the acceptance criteria for consideration under 10 C.F.R. § 2.206.
Document Control Desk Page 2 LR-N11-0173 DISCUSSION - PETITIONER'S REQUEST DOES NOT MEET CRITERIA FOR REVIEW AS A 10 C.F.R. § 2.206 PETITION The regulations in 10 C.F.R. § 2.206 require that a Petition under 2.206 "set forth the facts that constitute the basis for the request." NRC's guidance for its 2.206 petition acceptance process states that the staff will review a petition if it meets all three of the following criteria:
- The Petition contains a request for enforcement-related action such as issuing an order modifying, suspending, or revoking a license; issuing a notice of violation with or without a proposed civil penalty; etc.;
- The facts that constitute the bases for taking the particular action are specified. The petitioner must provide some element of support beyond bare assertion. The supporting facts must be credible and sufficient to warrant further inquiry; and
- There is no NRC proceeding available in which the petitioner is or could be a party and through which the petitioner's concerns could be addressed. If there is a proceeding available, for example, if a petitioner raises an issue that he or she has raised or could raise in an ongoing licensing proceeding, the staff will inform the petitioner of the ongoing proceeding and will not treat the request under 10 C.F.R. 2.206. 1 The Petition fails to meet two of the standards for acceptance as a 2.206 petition. First, it lacks sufficient supporting facts to warrant further inquiry. The historical record shows that the asserted facts in the Petition are incorrect and incomplete with respect to the historical Mark I containment issues, the Direct Torus Vent System ("DTVS"), and spent fuel pools.
Furthermore, there is no factual nexus established between the cited elements of the Mark I design and the causes of the accident at the Fukushima Daiichi plants. Second, NRC's ongoing Fukushima Task Force activities provide the opportunity for public participation with respect to the Fukushima Daiichi issues that are the subject of the Petition. For both of these reasons, the PRB should not accept the Petition for review under 10 C.F.R. § 2.206.
I. There is an Insufficient Factual Basis for Petition The NRC has repeatedly found that petitions which lack a credible and sufficient factual basis do not meet the criteria for review pursuant to 10 C.F.R. § 2.206. 2 In particular, mere allegations of safety concerns provide an insufficient factual basis for a 2.206 petition. Also, the absence of facts establishing that the safety concerns represent a threat to public health and safety at the cited nuclear plants is fatal to a petition that requests enforcement action at those plants. This Petition fails for both of these reasons.
The Petition cites certain design elements of BWR Mark I designs and broadly asserts that the reactor design has failed at Fukushima Daiichi "to reliably and adequately mitigate and contain significant arid mounting radiological releases to the atmosphere, groundwater and the ocean Management Directive 8.11, Part III, Section C(l)(a)(i)-(iii) (October 25,2000).
2 See Letter from Thomas Blount, Deputy Director Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation, to Bill Linton (April 28, 2011), available at ADAMS Accession No. ML111160151; Letter from Thomas Blount, Deputy Director of Policy and Rulemaking, Office of Nuclear Reactor Regulation, to Thomas Gurdziel (February 16,2010), available at ADAMS Accession No. MLl00420040.
Document Control Desk Page 3 LR-N11-0173 from multiple severe accidents in multiple GE BWR Mark I units.,,3 In making this assertion, Petitioner cites to historical information that allegedly asserts the inadequacy of the BWR Mark I design. Beyond Nuclear's Petition, however, contains incorrect and incomplete factual assertions with respect to historical Mark I containment issues identified in the 1970s and 1980s, installation of DTVS for combustible gas control, and spent fuel pool designs. The correct and complete factual record on each of these issues, discussed further below, demonstrates that these historical safety issues were appropriately evaluated and addressed at that time, and the necessary plant modifications were implemented.
Furthermore, the Petition fails to establish a nexus between the alleged Mark I design problems and the accident at Fukushima Daiichi, particularly given the preliminary nature of the information currently known about the accident. Petitioner's failure to provide any information that is not already known to the NRC and failure to raise any immediate safety concerns, coupled with the Petition's incorrect and incomplete factual information, means that the factual basis for the Petition is insufficient to warrant its review under 10 C.F.R. § 2.206.
A. Review of Historical Mark I Containment Issues Petitioner alleges that BWR Mark I pressure suppression containments are "undersized and vulnerable,,4 to "early failure under severe accident conditions including over-pressurization.,,5 Citing the "unreliability of ... [Mark I] containment design and construction,,,6 Petitioner asserts that the BWR Mark I containments "can malfunction as previously warned in 1972 and again in 1985 to be unreliable in severe accident conditions.,,7 In an attempt to support such arguments, Petitioner quotes at length from a 1972 memo by Dr.
Stephen Hanauer on the vulnerability of Mark I containments to early failure under severe accident conditions including over-pressurization, and alleges "US nuclear safety officials ... subsequently ignored over the decades [the safety concerns of Dr. Hanauer]."s Petitioner, however, fails to mention that Dr. Hanauer's concerns were resolved by 1978. In fact, the Atomic Energy Commission's ("AEC's") evaluation of the BWR pressure-suppression design began in the late 1950s and continued for years after Dr. Hanauer's 1972 memo. 9 In the Petition at 6.
4 Id. at 16.
Id. at 5.
Id. at 16.
7 Id at 19.
Id. at 6. Petitioner also refers to an AEC policy statement in 1972 to not accept further new construction applications and implies some connection to the Mark I issues identified by Dr. Hanauer. However, the policy statement bore no relationship to the Mark I issues, but, rather, it provided for the implementation of the National Environmental Policy Act of 1969 ("NEPA"). Part-50 Licensing of Production and Utilization Facilities Implementation of National Environmental Policy Act of 1969,36 Fed. Reg. 18071 (Sept. 9, 1971). While the AEC was revising its NEPA policies and procedures and license applicants were developing the necessary environmental submittals, the AEC did not issue a power plant construction permit or operating license for more than 17 months.
9 See, e.g., NUREG-0474, A Technical Update on Pressure Suppression Type Containments in Use in U.S.
Light Water Reactor Nuclear Power Plants, at Abstract (.Tuly 1978) ("The NRC staff, the [Advisory
Document Control Desk Page 4 LR-N11-0173 mid-1970s, the NRC (AEC's successor as the nuclear regulatory agency), with input from the nuclear industry, implemented a short-term program to test the hydrodynamic loads and dynamic structural responses. The long-term program which followed involved the Mark I Owners' Group - an industry group of those utilities with Mark I plants - conducting a full-scale testing program to further investigate these issues, while the NRC conducted research to confirm the test results. 10 In July 1978, the NRC published the results of the short-term testing program and some of the results of the long-term testing program in NUREG-0474. The NRC stated that "[b]ased upon the reviews that have been performed, the staff concludes that the pressure suppression concept for containment design is acceptable and safe."11 Importantly, in a separate enclosure to the NUREG, the NRC specifically addressed each of the issues raised by Dr. Hanauer in 1972 and concluded that they had been resolved by either: (a) subsequent or existing regulations or (b) subsequent testing establishing that they did not constitute a safety issue.
Petitioner also fails to discuss Dr. Hanauer's memo in 1978 which documented Dr. Hanauer's support for the licensing of the new BWR plants using the Mark I containment system.
Following the testing, evaluations and plant modifications documented in NUREG-0474, Dr.
Hanauer concluded in a June 20, 1978, memorandum to NRC Chairman Joseph Hendrie, "Thus, while we may yearn for the greater simplicity of 'dry' containments, the problems of both
'dry' and pressure containments are solvable, in my opinion, and the design safe, therefore licensable.,,12 Also, since 1978, there have been several other legal challenges to the operation of Mark I deSigned-reactors in the United States, similarly referencing Dr. Hanauer's 1972 concerns. The NRC has repeatedly confirmed that Dr. Hanauer's concerns were adequately addressed and, therefore, cannot form the basis for an acceptable challenge to these reactors.13 Therefore, contrary to Petitioner's assertions, not only were Dr. Hanauer's 1972 concerns not ignored for decades, but those concerns were addressed and resolved within six years.
Committee of Reactor Safeguards], and their respective consultants have been reviewing the various analytical and experimental programs over the past 20 years. The NRC staff has concluded, and the ACRS concurs, that there exists an adequate technical basis for justifying continued operation of licensed plants &
continued licensing of new plants that utilize pressure suppression containments.").
10 Id. at v.
11 Id. at 13.
12 Memorandum from Stephen Hanauer, Technical Advisor to Executive Director for Operations, Nuclear Regulatory Commission, to Joseph Hendrie, Chairman, Nuclear Regulatory Commission (June 20, 1978),
available at ADAMS Accession No. 7810060203 13 See, e.g., Iowa Elec. Light & Power Co. (Duane Arnold Energy Center), DD-80-7, 11 NRC 383, 385-86 (Feb. 13, 1980) (citing NUREG-0474) ("The NRC has given careful consideration to the concerns identified in Dr. Hanauer's memorandum of September 20, 1972, as well as new safety concerns associated with pressure suppression containments that have been identified [the hydrodynamic load issues]. Based upon the reviews that have been performed, the staff, including Dr. Hanauer, has concluded that the pressure suppression concept for containment design is safe.").
Document Control Desk Page 5 LR-N11-0173 Petitioner's citation to comments by Dr. Harold Denton, then Director of Nuclear Reactor Regulation (NRR) regarding the probability of failure of the Mark I containment in a severe accident,14 also is incomplete in that Petitioner omits the factual context of the comments, as well as the actions taken by the NRC and the nuclear industry to address those issues. Dr.
Denton's comments were based on a 1975 study of nuclear power plant risks.15 The study found that the Mark I risk was dominated by two scenarios: station blackout and anticipated transient without scram. The NRC later promulgated regulations for both of these scenarios and took other actions to reduce the probability of containment failure. 16 In addition, Dr. Denton also suggested in the same discussion cited by Petitioner that protection of Mark I containments from over-pressurization "would require positive ways to vent and filter before you get a high pressure buildup.,,17 Subsequently, all licensees conducted plant-specific analyses of severe accidents, and Mark I licensees evaluated the relative safety benefit of installation of DTVS in order to reduce containment pressure in the event of a severe accident, as discussed below. Therefore, the issues raised by Dr. Denton in 1986 were specifically addressed, resulting in plant-specific analyses and any necessary design modifications.
The Petition also mentions a 1989 10 CFR 2.206 petition filed by Anne Harlow, and states that Beyond Nuclear's Petition provides "new information and analysis [that] "supersede[s] any prior NRC Final Director's Decision" related to that 1989 petition. 18 In acknowledging the existence of an earlier petition, Petitioner is conceding that it is basing this Petition on the same historical factual arguments that were previously rejected by the NRC. The 1989 petition by Anne Harlow requested the NRC to take emergency action to "fix or close" all GE BWRs in the U.S, based on the statements of Dr. Hanauer and Dr. Denton. 19 The NRC denied the portions of that petition that related to Mark I reactors because the petition did not provide any information of which the NRC staff was unaware, and the petition did not provide sufficient evidence of either design flaws in Mark I containment or high risk to warrant suspending the operating licenses for those reactors. 20 Here, Petitioner asserts that the new information and analyses generated by the Fukushima Daiichi accident supersede the prior final Director's Decision on the Anne Harlow petition. However, Petitioner has failed to provide either a sufficient factual basis to warrant further inquiry or a factual nexus between the causes of the Fukushima Daiichi accident and the allegedly faulty design features of the Mark I reactor. Therefore, the Petition should be rejected, as it does not meet the acceptance criteria under 10 C.F.R. § 2.206.
14 Petition at 6.
15 Although Dr. Denton referred to NUREG-7S/014, Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants (Oct. 1975), available at http://www.nrc.gov/reading-rmidoc-collections/nuregs/staff/sr7S-014/, actual conditional containment failure probability is more directly addressed in NUREG-llS0, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants (Oct. 1990), available at http://nrc.govlreading-rmldoc-collections/nuregs/staff/sr11 SOl, the draft of which had been issued shortly before Dr. Denton made his comments.
16 See 10 C.F.R. § SO.63 (station blackout); 10 C.F.R. § SO.62 (anticipated transient without scram).
17 Brian Jordan, Denton Urges Industry to Settle Doubts about Mark I Containment, Inside NR.C., June 9, 1986.
18 Petition at 4.
19 Id.
20 Boston Edison Co. (Pilgrim Nuclear Power Station), DD-89-9, 30 NRC 791, 806 (Dec. 4,1989).
Document Control Desk Page 6 LR-N11-0173 B. Review of Direct Torus Vent Systems Petitioner cites as a basis for the requested emergency action the deliberate venting of "the recognized undersized and vulnerable Mark I pressure suppression containment temporarily under severe accident conditions,,21 at Fukushima Daiichi. Based on the venting of these reactors, Petitioner requests that a review of decisions about and implementation of hardened vent systems in U. S. Mark I containments be conducted by the NRC. 22 Again, Petitioner fails to identify that the plant-specific analyses were conducted by all Mark I plant licensees to evaluate plant-specific vulnerabilities to severe accidents, including the probability of core damage and the performance of the containment,23 and that the NRC reviewed and accepted those analyses. 24 In the 1980s, the NRC staff began reviewing the potential for accidents more severe than those the plants were licensed for and designed to mitigate. According to the NRC, while BWR Mark I containments had provisions for venting, it "typically consist[ed] of a ductwork system which has a low design pressure of only a few pounds per square inch" and "[v]enting under high pressure severe accident conditions would fail the ductwork" and "release the containment atmosphere into the reactor building.,,25 In order to enhance the ability of Mark I containments to prevent and mitigate the consequences of accidents beyond the design basis accidents, the NRC requested in Generic Letter 89-16 that all BWR Mark I plants install a "hardened vent," also known as a DTVS, and stated that for those Mark I plants that did not do so voluntarily, it would conduct a backfit analysis to determine whether to impose it as a requirement. 26 In response to Generic Letter 89-16 and the subsequent results of their Individual Plant Examinations ("IPEs"), all of the Mark I nuclear power plants demonstrated compliance to the NRC's satisfaction.27 However, because installation of DTVS was not ordered for all Mark I plants but was subject to a site-specific backfit analysis, each decision to install such a system was based on plant-specific analyses, subject to the NRC's review. Therefore, the NRC has already reviewed the issue of installation and operation of DTVS on a plant-specific basis, and Petitioner's assertions that such a review is now necessary ignores the historical record of the actions already taken by the NRC.
- c. Review Spent Fuel Pool Design Issues 21 Petition at 8.
22 Petition at 9.
23 See Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities - 10 C.F.R. 50.54(f) (November 23, 1988).
24 See NUREG-1560, Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance (Dec. 1997).
25 NUREG-0933, Resolution of Generic Safety Issues: Issue 157: Containment Performance (Rev. 1, June 1995).
26 Generic Letter 89-16, Installation of a Hardened Wetwell Vent (September 1, 1989).
27 NUREG-1560.
Document Control Desk Page 7 LR-N11-0173 Petitioner's allegations regarding the vulnerability of densely-packed, elevated spent fuel storage pools outside a primary containment structure and without Class 1E safety-related power systems for backup pool cooling are similarly incomplete and unsupported. Specifically, Petitioner merely refers to the characteristics of certain spent fuel pools and concludes that the Fukushima Daiichi accident demonstrates the vulnerability of such pools. This constitutes a bare assertion, lacking credible and sufficient supporting facts, and is therefore insufficient as a basis for acceptance of the 2.206 petition.
Beginning in 1975, the NRC examined the risks of beyond design basis accidents in spent fuel storage pools.28 The NRC concluded that the risks were orders of magnitude below those involving the reactor core. Among the reasons cited for this conclusion was the Simplicity of the spent fuel storage pool design: the coolant is at atmospheric pressure, the spent fuel is always subcritical and the heat source is low, and there is no piping which can drain the pool.29 Also, in the early 1980s, NRC began to reexamine spent fuel storage pool accidents due to the expansion of onsite fuel storage by means of high density storage racks, resulting in a greater heat load. In addition, some laboratory studies cited the possibility of fire propagation between assemblies in an air-cooled environment. 3o In 1983, NRC established Generic Safety Issue
("GSI") 82 related to the risk of beyond design basis accidents in spent fuel pools. Based on the results of further analyses, the NRC concluded that the risk and consequences of a spent fuel pool accident meet the NRC's Safety Goal Policy Statement for both BWRs with elevated spent fuel pools and for PWRs with pools at ground level. 31 In reaching those conclusions, the NRC explicitly evaluated the seismic capacity of elevated BWR spent fuel pools, concluding that any reduction in the structural capacity of the elevated BWR pool versus the ground-level PWR pool is offset by the lower conditional probability of a fire in the BWR fuel cladding versus in the PWR fuel cladding. 32 Therefore, the probability of a fire in the fuel cladding resulting from a loss of water from the spent fuel pool is approximately the same for both types of reactors. With respect to the capability of the elevated BWR pools to withstand earthquakes that are substantially more severe than the design basis earthquake, the NRC found that BWR pools have a highly conservative design margin of about a factor of eight, notwithstanding the higher elevation of the spent fuel pool.33 In further support for its Petition, Petitioner also cites to the lack of backup Alternating Current
("AC") electrical generators and reliable emergency backup Direct Current battery systems to provide emergency cooling to spent fuel pools. Again, however, Petitioner fails to address or even acknowledge existing systems and measures already in place. For example, because the 28 NUREG-75/0 14.
29 NUREG-0933, Resolution of Generic Safety Issues: Issue 82: Beyond Design Basis Accidents in Spent Fuel Pools (Rev. 3, June 2004).
30 Id.
31 NUREG-1353, Regulatory Analysis for the Resolution of Generic Issue 82, "Beyond Design Basis Accidents in Spent Fuel Pools," at 6-3 to 6-4 (April 1989).
32 Id. at ES-3.
33 Id. at ES-4.
Document Control Desk Page 8 LR-N11-0173 emergency diesel generators ("EDGs") supply power only to safety-related loads, they would not power the non-safety-related spent fuel pool cooling system. However, the EDGs would power both the safety-related Residual Heat Removal ("RHR") system, which can provide supplemental pool cooling, and the spent fuel makeup system, which can replace pool water.34 Thus, even in the event of a loss of offsite power, sufficient makeup water would be provided to the spent fuel pool to ensure adequate cooling of the spent fuel.
In the event of station blackout, when both offsite and onsite AC power is lost, water would need to be supplied to the pool to replace any water lost to boiling and evaporation. Petitioner cites the lack of safety-related electrical backup systems for pool cooling ability, but the fundamental issue is the cooling capacity, rather than whether electrical power can power specific cooling systems. To prevent exposure of the spent fuel, water could be supplied to the spent fuel pool through means that do not rely upon electrical power, such as diesel-driven fire water pumps or portable pumps.35 Further, due to the relatively large volume of spent fuel pools, the pools can survive for a relatively long period of time (e.g., at least one day and usually for many days or longer) without the need for makeup water. 36 This provides time to put the non-electrically-driven pumps into operation for spent fuel pool cooling water makeup.
In summary, in omitting significant facts about the risks of spent fuel pools in severe accident conditions and relying on a bare assertion of spent fuel pool vulnerability, Petitioner has failed to provide a sufficient factual basis for the Petition. This clearly does not meet the criteria for review under Section 2.206.
D. There is No Factual Nexus to Causes of Fukushima Daiichi Accident Petitioner alleges that the Mark I design elements mentioned in the Petition are the cause of the accident consequences at Fukushima Daiichi. However, what is now known about the causes of the accident is that the reactors experienced an earthquake and beyond-design basis tsunami in rapid succession on March 11, 2011. The relationship between any specific design features of the Mark I reactors in the U.S. and the accident in Japan is only speculative at this time, given the early stage of knowledge about all of the causes of the accident and its consequences. Without a connection between the Petitioner's concerns about the causes of the accident and the design features of the Mark I reactors that are the subject of the Petition, the Petition lacks the necessary factual basis for acceptance under Section 2.206.
II. Petitioner's Issues Are Already Being Addressed by the NRC Petitions to the NRC whose subjects are already under consideration by the NRC in another proceeding are consistently rejected as not meeting the acceptance criteria for a petition under 34 See, e.g., Final Safety Analysis Report [Plant name], Section 8.3.1.
35 Individual plants have put in place emergency procedures that provide such alternative sources of cooling water to the spent fuel pool, particularly in the aftermath of September 11, 2001.
36 NUREG-1353, at 4-23. This report states that boil-off in one day would only occur in a very limited scenario; i. e., full core discharge five days following reactor shutdown followed by a loss of spent fuel pool cooling and makeup water.
Document Control Desk Page 9 LR-N11-0173 Section 2.206. 37 If there is another NRC proceeding available in which Petitioner could participate and have its issues addressed, the NRC will not accept Petitioner's attempt to open an additional proceeding on the same factual basis. Here, the Petition fails because another avenue for Petitioner to raise its concerns has already been established in the NRC's Fukushima Task Force.
The general concerns about the Fukushima Daiichi accident articulated in the Petition are some of the same issues that the NRC is already addressing through a task force that was announced on March 21, 2011,38 which was over three weeks before Petitioner filed this Petition. The purpose of the NRC senior level task force is to identify any potential areas for the NRC to make improvements to its regulatory system based on the Fukushima Daiichi accident.
The short-term, 90-day phase of the task force review will focus on near-term operational or regulatory issues, while the longer-term review will "evaluate all technical and policy issues related to the event to identify additional research, generic issues, changes to the reactor oversight process, rulemakings, and adjustments to the regulatory framework."39 The NRC plans to begin the longer-term review after it has sufficient technical information from the accident, with a goal of beginning the longer-term review no later than when the short-term report is completed.
Petitioner, along with other members of the public, will have an opportunity to participate during the six months of the conduct of the longer-term review. The NRC's direction to the task force is that it "should receive input from and interact with all key stakeholders" during the longer-term review.40 In addition, the NRC plans to have the task force provide updates on the beginning of the longer-term review at the 30 and 60 day status updates on the shorter-term review. Reports on both the short-term and the longer-term reviews will be made public. Thus, the NRC already has in place a process that is addressing the subject of the Petition, and Petitioner will have the opportunity to participate in the process. Therefore, the Petition should be rejected because it fails to meet the third acceptance criterion for a 2.206 petition.
CONCLUSION The facts about the causes of the Fukushima Daiichi accident are still being investigated, so any conclusions based on the accident are speculative. All of the historical facts and concerns in the Petition have been previously considered in detail, and actions have been taken to address them. The NRC has underway a pair of task force reviews that will address the issues of the Fukushima Daiichi accident and provide for public participation in those issues. Therefore, the Petition should be rejected because it fails to meet the acceptance criteria for review by the PRB as a 10 C.F.R. § 2.206 petition.
37 See Letter from Thomas Blount, Deputy Director Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation, to Mark Leyse (April 26, 2011), available at ADAMS Accession No. ML111020301.
38 Memorandum from Gregory J aczko, Chairman, Nuclear Regulatory Commission, on NRC Actions Following the Events in Japan (March 21,2011), available at ADAMS Accession No. ML110800456.
39 Id. at 2.
40 Id. at 2.
Document Control Desk Page 10 LR-N11-0173 Sincerely, Associate General Nuclear Counsel C William M. Dean, Regional Administrator - USNRC Region I R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector - Hope Creek P. Mulligan, Manager IV, NJBNE L. Marabella, Corporate Commitment Tracking Coordinator Ken Yearwood, Hope Creek Commitment Tracking Coordinator