Resolution of Generic Safety Issues: Issue 155: Generic Concerns Arising from TMI-2 Cleanup (Rev. 3) ( NUREG-0933, Main Report with Supplements 1–34 )
The TMI-2 Safety Advisory Board was established to provide the licensee, General Public Utilities Nuclear Corporation, with a qualified, independent appraisal of the cleanup of TMI-2, with particular emphasis on the assurance of public and worker health and safety. As a result of this appraisal, seven recommendations1362 were forwarded to the NRC for evaluation. These recommendations were treated as separate generic issues as outlined below.
ISSUE 155.1: MORE REALISTIC SOURCE TERM ASSUMPTIONS
During the TMI-2 accident, fission products did not behave as predicted with the analytical methods and assumptions used in the licensing process at that time and delineated in Regulatory Guides 1.3213 and 1.4214 and TID-14844.73 The earliest expert predictions were that major core damage had occurred. However, the NRC and the licensee believed that core damage was minimal and calculations were redone to confirm this view. Approximately 50% of the core was in a molten state, but there is evidence that only about 55% of the highly volatile fission products and noble gases were released from the reactor vessel with a major portion retained in the reactor building. There is also evidence that less than 5% of the medium and low volatile fission products were released from the reactor vessel.1362 These observations were based on research conducted since the TMI-2 accident.
It is now generally accepted that the chemical conditions in the reactor vessel were "reducing" in nature as opposed to "oxidizing." The elemental iodine was driven (or converted) to the iodide ion which very readily combined with available metallic ions. The water-soluble character of these chemical forms prevented a major release of iodine to the atmosphere of the containment or auxiliary buildings and only a few Curies were released to the environment. Throughout the TMI-2 accident sequence, the chemical state was maintained such that the water-soluble character was preserved.
With the completion of a large number of PRAs since the TMI-2 event, the Advisory Board believed that it should be possible to list accident sequences with chemical conditions similar to TMI-2. Such a listing could provide a guide as to which accidents might be regarded as hazardous, or less hazardous, relative to the possible escape of iodine and could be useful in the future design of safety features. Since some of the assumptions used for source term considerations at TMI-2 were flawed in this respect, the Board recommended that the source term be restated using current scientific knowledge.1362
At the time this issue was evaluated in February 1992, comprehensive revisions to 10 CFR Parts 50 and 100 were being pursued by the staff to reflect a better understanding of accident source terms and severe accident insights, as well as evaluate the impact of these phenomena on plant engineered safety features. A replacement for TID-1484473 was being formulated, based on previous severe accident research findings, to reflect the existing understanding of fission product release timing, iodine chemistry, and source term magnitude and composition. Thus, a solution to this issue had been identified and the issue was considered nearly-resolved.
In resolving the issue, the staff issued NUREG-14651465 which provided more realistic estimates of the fission product source term release into containment, in terms of timing, nuclide types, quantities, and chemical form, given a severe core-melt accident. Thus, the issue was RESOLVED with new requirements for future plants.1530 In an RES evaluation,1564 it was concluded that consideration of a 20-year license renewal period would not affect the resolution.
ISSUE 155.2: ESTABLISH LICENSING REQUIREMENTS FOR NON-OPERATING FACILITIES
At the time the TMI-2 event occurred, 10 CFR 50 contained regulations primarily for the design, construction, and operation of nuclear facilities but did not provide adequate guidance for the post-accident condition. Much was learned while the unit was being defueled and prepared for the post-defueling, monitored storage phase. The decommissioning rule1364 issued in 1988 addressed the safe removal of nuclear facilities from service and the reduction of residual radioactivity to a level that permits release of the property for unrestricted use and termination of the operating license. The options for compliance with this rule are described in NUREG-0586173 and include DECON, SAFSTOR, and ENTOMB. Decommissioning activities do not include the removal and disposal of spent fuel; these are considered to be operational activities.
Once a reactor is permanently shut down and defueled, it enters a storage phase until the licensee begins implementation of a decommissioning plan approved by the NRC. During the storage phase, requirements for security plans, operator licensing, emergency planning, etc., that were in effect while the plant was operational, may become unnecessary and burdensome to the licensee. Once all nuclear fuel is removed from the reactor site, the risk of an extraordinary accident, as defined in 10 CFR 50.54(w) and 10 CFR 140.11, is essentially eliminated. The Board recommended that regulatory guidance be developed for use by non-operating and defueled facilities during the storage phase prior to decommissioning.1362
This issue addressed changes in existing regulatory guidance that could significantly reduce licensee costs without any substantial change in public risk. Thus, it was classified as a Regulatory Impact issue. Staff stated in the Supplement to NUREG-0933 published in 1995 that revisions to 10 CFR 50.54(w) and 10 CFR 140.11(a)(4) might be necessary to address insurance coverage for non-operating and defueled facilities during the storage phase prior to decommissioning.1363
As a part of the improvements to NUREG-0933, the NRC staff clarified in SECY-11-0101, “Summary of Activities Related to Generic Issues Program,” dated July 26, 2011,1967 that the Generic Issues Program will not pursue any further actions toward resolution of licensing and regulatory impact issues. Because licensing and regulatory impact issues are not safety issues by the classification guidance in the legacy Generic Issues Program, these issues do not meet at least one of the Generic Issues Program screening criteria and do not warrant further processing in accordance with Management Directive 6.4, “Generic Issues Program,” dated November 17, 2009.1858 Therefore, this issue will not be pursued any further in the Generic Issues Program.
ISSUE 155.3: IMPROVE DESIGN REQUIREMENTS FOR NUCLEAR FACILITIES
The Board recommended1362 that the NRC undertake an effort to evaluate lessons learned at TMI-2 and incorporate them into the design of future nuclear plants. The recommendations suggested by the Board focused on recovery from a severe accident and were as follows:
|(1)||Prohibit the use of cinder blocks inside the reactor building (because they absorb so much contamination and become a radiological hazard) or designing the facility to be "robot friendly."|
|(2)||Utilize higher range radiation instrumentation in order to monitor the environment inside the reactor building during a severe reactor accident.|
|(3)||Based on design criteria and clear evidence that the TMI-2 containment building was not challenged, a reduction in criteria might be prudent based upon actual accident conditions. The NRC had reviewed in some detail the capability of reactor containment structures to withstand accident environments, including significant pressure increases; a review of these studies might be helpful and may lead to a reduction in design criteria. A similar effort for reactor vessels has not been undertaken and should be, considering the condition of the lower head of the TMI-2 reactor vessel with the severity of the accident.|
|(4)||TMI-2 has also demonstrated the need to provide access to the underside of a reactor vessel for remote inspections to determine the extent of possible damage in the aftermath of a severe reactor accident. The 52 instrument penetrations in the lower head of the TMI-2 reactor vessel have been a concern since the discovery of once-molten material on the lower head of the reactor vessel and thus lower head integrity has been a major concern during the recovery efforts. For future reactor vessel design, it was recommended that in-core instrumentation penetrate the head instead of the bottom.|
The four concerns outlined in this issue were evaluated separately below:
In accordance with 10 CFR 50, Appendix I, nuclear power plants are required to keep occupational risk exposure (ORE) as low as is reasonably achievable (ALARA). Cinder blocks constitute one of the materials that are used inside the reactor building of some operating plants as local shielding to meet this ALARA criterion. Prohibiting the use of cinder blocks inside the reactor building would have no impact on public risk in the event of a severe accident. The use of other shielding materials that do not absorb as much contamination has the potential for decreasing the decontamination time (and ORE) following a severe accident.
Designing future nuclear plants to be robot-friendly will require spatial considerations for the mobility of robots that could drastically increase design, engineering, and construction costs. However, as is the case above, the use of robots would have no impact on public risk in the event of a severe accident; only occupational risk would be affected.
From NUREG/CR-2800,64 the occupational dose from cleanup, repair, and refurbishment following a severe accident was estimated to be 19,860 man-rem. Even assuming that 50% of this dose can be reduced with either the elimination of cinder blocks or the use of a robot for cleanup and assuming a core-melt frequency of 10-5/RY and an average remaining reactor life of 28 years, the potential dose reduction is approximately 3 man-rem/reactor. Thus, this concern had negligible risk reduction potential and consideration of costs would only lower its priority ranking.
|(2)||The recommendation to utilize higher range radiation instrumentation in order to monitor the environment inside the reactor building during a severe accident was addressed by TMI Action Plan Item II.F.1. This item was clarified in NUREG-073798 and required implementation at all plants. Thus, this concern was previously addressed by the staff.|
For future plants, the Commission's Severe Accident Policy Statement established the criteria and procedural steps under which new designs for nuclear power plants could be acceptable for meeting severe accident concerns. Rather than a reduction of criteria, it is expected that future plants would have to achieve a higher standard of severe accident safety performance, including clarification of containment performance. The staff's plan of action in this area was presented to the Commission in SECY-92-292.1427 Operating plants were assessed under the Containment Performance Improvement Program (see Issue 157).
The mode of vessel failure, including investigation of the TMI-2 vessel, was being pursued by the staff as part of its severe accident research program.1382 The results of this research was expected to determine whether changes to future vessel design would be warranted. Thus, this concern was being addressed by the staff.
|(4)||The relocation of in-core instrumentation was expected to be addressed by NSSS vendors in the design of future plants which was subject to review and approval by the staff. For example, the bottom-mounted instrumentation penetrations were eliminated in the Westinghouse AP600 design to reduce building volume and costs significantly. Thus, this concern was being addressed by the staff.|
Of the four recommendations contained in this issue, two were addressed in other ongoing programs and one had been previously addressed by the staff. The remaining recommendation had negligible risk reduction potential and, therefore, was not considered to be safety-significant. Thus, this issue was DROPPED from further consideration as a new and separate issue. In an RES evaluation,1564 it was concluded that consideration of a 20-year license renewal period did not change the priority of the issue.
ISSUE 155.4: IMPROVE CRITICALITY CALCULATIONS
The Board believed that doubts still remained as to whether the TMI-2 core became critical, or was very close to critical, during the TMI-2 accident and recommended that the NRC establish guidelines that deal with criticality following a severe reactor accident.1362 These guidelines should take into account abnormal geometries and possible core conditions that could result from the accident. The Board believed that the accident scenario developed by the TMI-2 licensee was sufficiently detailed that a series of geometric configurations could be simulated for criticality calculations. Variables that could be estimated reasonably well included the presence of water, oxidation of cladding, melting and movement of fuel, melting of poison rods, and movement of poison.
The safety concern was addressed by DSR/RES in SARP Task 4.3: Investigate the Possibility and Consequences of Recriticality in Degraded BWR Cores.1382 The staff's study was documented in NUREG/CR-56531379 in which it was concluded that there was the potential for recriticality in BWRs, if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event would most likely not generate a pressure pulse significant enough to fail the vessel. Two strategies were identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached: (1) initiation of boron injection at or before the time of core reflood, if the potential for control blade melting exists; and (2) initiation of RHR suppression pool cooling to remove the heat load generated by the recriticality event and extend the time available for boration.
The issue was not considered to be a major concern for PWRs because of their design that includes a safety injection system for supplying borated water to the core. Furthermore, it was concluded in NUREG/CR-58561417 that, during a severe accident, an unmoderated recriticality of the molten, consolidated portion of a degrading core cannot occur at U235 enrichments characteristic of a PWR. Based on the staff's efforts in addressing the safety concerns in the SARP, this issue was DROPPED from further pursuit as a new and separate issue. In an RES evaluation,1564 it was concluded that consideration of a 20-year license renewal period did not change the priority of the issue.
ISSUE 155.5: MORE REALISTIC SEVERE REACTOR ACCIDENT SCENARIO
The TMI-2 event was a severe accident in which approximately 50% of the core was in a molten state at some point during the accident. Approximately 20 tons of the once-molten debris poured through the core support structure into the water-filled lower plenum and onto the lower head of the reactor vessel. Most codes in use at that time would have predicted a failure of the lower head under these conditions. The severity of the accident showed that the reactor vessel was more difficult to fail than was anticipated.
The Board recommended that in-vessel core-melt progression for severe accidents be studied further by the NRC and that the results be incorporated into existing codes and standards. The Board believed that codes should have the capability to reproduce the TMI-2 accident with reasonable accuracy before they can be accepted as predictive tools.1362
At the time this issue was evaluated in June 1992, the safety concern was being addressed by DSR/RES in SARP Issue L2: In-Vessel Core Melt Progression and Hydrogen Generation.1382 In considering core-melt progression, the staff was expected to treat BWRs and PWRs separately because of their different fuel assembly, control element, and lower plenum structures. Concerns common to both BWRs and PWRs are: (1) the integrity of core structures; (2) the mode of core material relocation; (3) hydrogen generation; (4) the mode of bottom head failure; and (5) the effects of water injection. The answers to the above concerns will be different because of the physical differences of BWRs and PWRs. TMI-2 data and the results of new experiments and model development were to be examined by the staff in its research. Based on the staff's efforts on SARP Issue L2, Issue 155.5 was DROPPED from further pursuit as a new and separate issue. In an RES evaluation,1564 it was concluded that consideration of a 20-year license renewal period did not change the priority of the issue.
ISSUE 155.6: IMPROVE DECONTAMINATION REGULATIONS
The Board believed that the decontamination techniques used throughout the nuclear industry for small activities were not applicable to large-scale activities and recommended that the NRC use the experience gained from the TMI-2 accident to prepare guidelines for decontamination and decommissioning of nuclear plants.1362
Traditionally, the NRC has not developed or approved decontamination techniques. Due to the many ways in which decontamination can be accomplished and the rapidly evolving technology in this area, it is not practical or beneficial for the NRC to establish guidelines for decontamination techniques. Rather, the NRC has focused on the development of criteria which set standards for exposure of workers and the public (e.g., 10 CFR 20), the levels of allowable residual contamination, and the handling and disposal of the radioactive waste generated. Efforts at establishing residual contamination criteria applicable to decommissioning were in progress as described below.
In June 1991, the Commission deferred1412 implementation of the Below Regulatory Concern (BRC) policy but reaffirmed its intentions to carry out its responsibilities to address issues related to waste disposal, consumer products, recycling of materials, and decontamination and decommissioning, as necessary, on a case--by-case basis in the manner in which these issues were considered, prior to the development of the BRC policy statement. In this regard, the staff was directed to continue its accelerated efforts in completing the technical basis for rulemaking on residual contamination criteria.
In accordance with SECY-92-045,1413 the staff proceeded with an enhanced participative rulemaking process to develop radiological criteria for decommissioning; this effort was tracked in the NRC Regulatory Agenda (NUREG-0936). Based on the above considerations, Issue 155.6 was DROPPED from further pursuit as a new and separate issue. In an RES evaluation,1564 it was concluded that consideration of a 20-year license renewal period did not change the priority of the issue.
ISSUE 155.7: IMPROVE DECOMMISSIONING REGULATIONS
The Board raised concerns over the requirements for the disposal of highly contaminated components from a nuclear plant during decommissioning and recommended that regulations be developed.1362
The TMI-2 experience was considered by the staff in the development of the decommissioning rule1364 in 1988. Industry options for complying with this rule are described in NUREG-0586173 and include DECON, SAFSTOR, and ENTOMB. As part of its resolution of Issue B-64, "Decommissioning of Reactors," the staff is currently developing an SRP11 Section for use in its review of licensee decommissioning plans. Concurrent with this effort is the development of two Regulatory Guides: DG-1005, "Standard Format and Content for Decommissioning Plans for Nuclear Reactors"; and DG-1006, "Records Important for Decommissioning of Nuclear Reactors." Thus, Issue 155.7 was DROPPED from further consideration as a new and separate issue. The related concern of decommissioning prematurely shutdown plants was addressed in Issue 155.2. In an RES evaluation,1564 it was concluded that consideration of a 20-year license renewal period did not change the priority of the issue.
- NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, (1st Ed.) November 1975, (2nd Ed.) March 1980, (3rd Ed.) July 1981.
- NUREG/CR-2800, “Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development,” U.S. Nuclear Regulatory Commission, February 1983, (Supplement 1) May 1983, (Supplement 2) December 1983, (Supplement 3) September 1985, (Supplement 4) July 1986, (Supplement 5) July 1996.
- TID-14844, “Calculation of Distance Factors for Power and Test Reactor Sites,” U.S. Atomic Energy Commission, March 23, 1962. 
- NUREG-0737, “Clarification of TMI Action Plan Requirements,” U.S. Nuclear Regulatory Commission, November 1980, (Supplement 1) January 1983.
- NUREG-0586, “Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities,” U.S. Nuclear Regulatory Commission, August 1988.
- Regulatory Guide 1.3, “Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors,” U.S. Nuclear Regulatory Commission, November 1970, (Rev. 1) June 1973, (Rev. 2) June 1974. 
- Regulatory Guide 1.4, “Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors,” U.S. Nuclear Regulatory Commission, November 1970, (Rev. 1) June 1973, (Rev. 2) June 1974. 
Memorandum for E. Beckjord from F. Gillespie, “Generic Concerns Arising from TMI-2 Cleanup,” February 21, 1991. 
- Memorandum for E. Beckjord from F. Gillespie, “Request for Generic Rulemaking Concerning Decommissioning Issues,” January 7, 1992. 
- Federal Register Notice 53 FR 24018, “10 CFR Parts 30, 40, 50, 51, 70, and 72, General Requirements for Decommissioning Nuclear Facilities,” June 27, 1988.
- NUREG/CR-5653, “Recriticality in a BWR Following a Core Damage Event,” U.S. Nuclear Regulatory Commission, November 1990.
- NUREG-1365, “Revised Severe Accident Research Program Plan,” U.S. Nuclear Regulatory Commission, August 1989, (Rev. 1) December 1992.
- Memorandum for J. Taylor et al. from S. Chilk “SECY-91-132—Evaluation of the Feasibility of Initiating a Consensus Process to Address Issues Related to the Below Regulatory Concern Policy,” June 28, 1991. 
- SECY-92-045, “Enhanced Participatory Rulemaking Process,” U.S. Nuclear Regulatory Commission, February 7, 1992. 
- NUREG/CR-5856 “Identification and Evaluation of PWR In-Vessel Severe Accident Management Strategies,” U.S. Nuclear Regulatory Commission, March 1992.
- SECY-92-292, “Advance Notice of Proposed Rulemaking on Severe Accident Plant Performance Criteria for Future LWRs,” U.S. Nuclear Regulatory Commission, August 21, 1992. 
- NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, February 1995.
- Memorandum for J. Taylor from E. Beckjord, “Resolution of Generic Issue 155.1, ‘More Realistic Source Term Assumptions,’” March 13, 1995. 
- Memorandum for W. Russell from E. Beckjord, “License Renewal Implications of Generic Safety Issues (GSIs) Prioritized and/or Resolved Between October 1990 and March 1994,” May 5, 1994. 
- Management Directive 6.4, “Generic Issues Program,” U.S. Nuclear Regulatory Commission, November 17, 2009.
- SECY-11-0101, “Summary of Activities Related to Generic Issues Program,” July 26, 2011. [ML111590814]