Resolution of Generic Safety Issues: Task II.B: Consideration of Degraded or Melted Cores in Safety Review (Rev. 5) ( NUREG-0933, Main Report with Supplements 1–35 )
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The objective of this task was to enhance public safety and reduce individual and societal risk by developing and implementing a phased program to include, in safety reviews, consideration of core degradation and melting beyond the design basis.
ITEM II.B.1: REACTOR COOLANT SYSTEM VENTS
This item was clarified in NUREG-0737,98 requirements were issued, and MPA F-10 was established by DL/NRR for implementation purposes.
ITEM II.B.2: PLANT SHIELDING TO PROVIDE ACCESS TO VITAL AREAS AND PROTECT SAFETY EQUIPMENT FOR POST-ACCIDENT OPERATION
This item was clarified in NUREG-0737,98 requirements were issued, and MPA F-11 was established by DL/NRR for implementation purposes.
ITEM II.B.3: POST-ACCIDENT SAMPLING
This item was clarified in NUREG-0737,98 requirements were issued, and MPA F-12 was established by DL/NRR for implementation purposes.
ITEM II.B.4: TRAINING FOR MITIGATING CORE DAMAGE
This item was clarified in NUREG-0737,98 requirements were issued, and MPA F-13 was established by DL/NRR for implementation purposes.
ITEM II.B.5: RESEARCH ON PHENOMENA ASSOCIATED WITH CORE DEGRADATION AND FUEL MELTING
The three parts of this item are evaluated below.
ITEM II.B.5(1): BEHAVIOR OF SEVERELY DAMAGED FUEL
Items II.B.5(1) and II.B.5(2) were combined and evaluated together under Item II.B.5(1).
DESCRIPTION
Historical Background
For a number of key severe accident sequences, there are critical phenomenological unknowns or uncertainties that impact containment integrity assessments and judgments regarding the desirability of certain mitigating features. The phenomena fall into three broad categories: (1) the behavior of severely damaged fuel, including oxidation and H2 generation; (2) the behavior of the core-melt in its interaction with water, concrete, and core-retention materials; and (3) the effect of potential H2 burning and/or explosions on containment integrity. Steam explosions were also to be considered in this category. Previous work in these several areas received less attention since they related to accidents beyond the design basis. At the time this TMI Action Plan48 item was raised, RES was conducting major programs to support the basis for rulemaking and to confirm certain licensing decisions. Complementary efforts conducted within NRR were to address specific licensing issues related to the subject research.
(1) Behavior of Severely Damaged Fuel
(a) In-Pile Studies: Fuel behavior research was to include in-pile testing to help evaluate the effects of conditions leading to severe fuel damage. Such tests were being performed in the INEL Power Burst Facility (PBF) in FY 1983 and later in the Annular Core Research Reactor (ACRR) at SNL and in the NRU reactor at Chalk River National Lab, Canada. In the PBF, RES was to perform a series of in-reactor fuel experiments to determine the effect of heating and cooling rates on damage to the bundle, rod fragmentation, distortion, and debris formation. Fission product release and H2 generation were also to be measured during the testing. Separate effects studies were to be conducted on rubble beds in the ACRR at SNL.
(b) Hydrogen Studies: The objective of this work was to increase the understanding of the formation of H2 in a reactor from metal-water reactions, radiolytic decomposition of coolant, and corrosion of metals, and to determine its consequences in terms of pressure-time histories and H2 deflagration or detonation. This work was also to include: (1) the preparation of a compendium of information related to H2 as it affects reactor safety; (2) analysis of radiolysis under accident conditions; (3) a review of H2 sampling and analysis methods; (4) a study of the effects of H2 embrittlement on reactor vessel materials; and (5) a review of means of handling accident-generated H2 with recommendations on improving existing methods. Results of these studies were considered in the resolution of Issue A-48, "Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment," and were not considered further in this issue.
(c) Studies of Post-Accident Coolant Chemistry: The RES objective in this area was the development of a relationship between fission product release and fuel failure and the improvement of post-accident sampling and analysis techniques. This was to be accomplished by the investigation of fission product release in a variety of fuel failure experiments.
(d) Modeling of Severe Fuel Damage: The effort in this area was the development of models for fuel rods operating beyond 2200°F that suffer a loss in geometry in order to compute extensive damage phenomena (such as eutectic liquid formation, fuel slumping, oxidation, and H2 generation, fission product release and interaction with the coolant, rubble-bed particle size, extent of fuel and clad melting, and flow blockage).
(2) Behavior of Core-Melt
The RES fuel-melt research program was to develop a base and verified methodology for assessing the consequences and mitigation of fuel-melt accidents. The program addressed the range of severe reactor accident phenomena from the time when extensive fuel damage and major core geometry changes occur until the containment has failed and/or the molten core materials have attained a semi-permanent configuration and further movement is terminated. Studies of improvements in containment design to reduce the risk of core-melt accidents were also included.
The program was composed of integrated tasks that included scoping, phenomenological and separate effects tests, and demonstration experiments that provided results for the development and verification of analytical models and codes. These codes and supporting data were then used for the analysis of thermal, mechanical, and radiological consequences of accidents and for decisions related to requirements of design features for mitigation and performance confirmation. The technical scope of the program included work in the following areas: fuel debris behavior; fuel interactions with structure and soil; radiological source term; fuelcoolant interactions; systems analysis codes; and mitigation features.
Safety Significance
The results of the research programs described above were expected to find broad application in areas such as PRA, accident analysis, siting, evacuation planning, emergency procedures, code development, etc. Thus, these programs would have considerable value just as licensing improvement efforts. However, the programs had sufficiently well-defined scopes to permit some estimates of direct safety significance. These programs were directed at a better understanding of severely damaged and molten cores. Once a core is in this state, any safety significance has to be in the area of minimizing radioactive releases and consequent dose to the public.
Possible Solutions
It was assumed that means would be devised to reduce the probability of containment failure and release of activity to the environment. Completely different approaches could be suggested after the results of the research programs were known.
The "classical" engineering approaches to handling degraded or melted cores are filtered vents to prevent containment overpressure and core-retention devices (core catchers) to prevent containment basemat melt-through. These approaches were used for cost estimates, but the other priority parameters were not specific to these approaches.
PRIORITY DETERMINATION
Studies64 of this issue by PNL considered only containment basemat melt-through. The approach presented here was expanded to include other aspects. The effect on a PWR with a dry containment was considered, based partly on the availability of information. It was not expected that the results for other containments or for BWRs would be greatly different, at least in the context of the uncertainty of such an analysis.
Frequency Estimate
Essentially, all core-melts are assumed to result in containment failure in WASH-1400.16 To estimate the effect of being able to deal with a severely damaged core, this assumption was relaxed. The modes of containment failure for PWRs were defined as follows:
α - containment rupture due to a reactor vessel steam explosion
β - containment failure due to inadequate isolation of openings and penetrations
γ - containment failure due to H2 burning
δ - containment failure due to overpressure
ε - containment vessel melt-through
Assuming that the research programs were successful in leading to engineering solutions, reductions in the frequency of the various failure modes were estimated as follows:
α - 10% (Little can be done about steam explosions.)
β - 0% (This does not affect isolation failure.)
γ,δ - 90% (Venting containment should be quite effective if methods are available for sizing the vent and determining what filtration is needed.)
ε - 90% (Should be achievable if a core catcher can be designed.)
Consequence Estimate
The consequences were straightforward in the sense that the consequences of each release category have been studied. However, the reduction in consequences was more difficult to assess since the release from a molten core in a tight containment is still not zero. Instead, it depends on the containment design leak rate, the efficiency of filtration of a containment relief vent, etc. To allow for this, it was assumed instead that the prevented releases corresponding to the α, γ, δ, and ε failure modes were similar to a PWR-9 release. The results of this calculation are summarized in Table II.B-1. For a new (forward-fit) plant (which was the most likely candidate for implementation), the public risk reduction was estimated to be 1,600 man-rem.
Cost Estimate
Industry Cost: PNL estimated64 the cost of a core retention device to be $1.4M for a forward-fit. SNL estimated312 the cost of a filtered containment vent to be on the order of a few million dollars. Thus, the industry cost was projected to be $10M/reactor.
NRC Cost: PNL estimated64 the NRC cost to be $2.3M, assuming implementation at 134 reactors. In reality, implementation might take place at a far smaller number of plants due to considerations of containment type, backfit vs. forward-fit, etc. However, even if only 10 plants were affected, the NRC cost would be insignificant compared to licensee costs. Therefore, NRC costs were neglected.
Total Cost: The total industry and NRC cost associated with the possible solution was estimated to be $10M/reactor.
| Release Category | Frequency* (RY)-1 | % Reduction**in Frequency | ΔF (RY)-1 | R (man-rem) | FR |
|---|---|---|---|---|---|
| Table II.B-1 | |||||
| PWR-1 | 5.3 x 10-8 | 10% | 5.3 x 10-5 | 4.9 x 106 | 2.6 x 10-2 |
| PWR-2 | 6.7 x 10-6 | 90% | 6.0 x 10-6 | 4.8 x 106 | 2.9 x 101 |
| PWR-3 | 2.6 x 10-6 | 81% | 2.1 x 10-6 | 5.4 x 106 | 1.1 x 101 |
| PWR-4 | 2.1 x l0-5 | - | - | 2.7 x 106 | - |
| PWR-5 | 4.9 x 10-5 | - | - | 1.0 x 106 | - |
| PWR-6 | 6.3 x 10-5 | 90% | 5.7 x 10-7 | 1.4 x 105 | 8.0 x 10-2 |
| PWR-7 | 3.4 x 10-5 | 90% | 3.1 x 10-5 | 2.3 x 103 | 7.1 x 10-2 |
| PWR-8 | 8.0 x 10-5 | - | - | 7.5 x 104 | - |
| PWR-9 | 4.0 x 10-5 | - | -3.9 x 10-5 | 1.2 x 102 | -4.7 x 10-3 |
| TOTAL: | 4.0 x 101 | ||||
* Because the specific containment failure mode was of interest here, the frequencies above were "unsmoothed." This is in contrast to the calculations in WASH-140016 which assumed a 10% contribution in frequency from adjacent release categories.
** Release Category PWR-1 is made up entirely of α failures and thus was assigned a 10% reduction in frequency. Categories PWR-2, PWR-6, and PWR-7 are made up of γ, δ, and ε failures and were thus assigned 90%. Category PWR-3 contains both α and δ failures which results in a net assignment of 81%.
Value/Impact Assessment
Based on a potential public risk reduction of 1,600 man-rem/reactor and a cost of $10M/reactor for a possible solution, the value/impact score was given by:
CONCLUSION
Based on the factors considered above, this issue was given a high priority ranking (see Appendix C). However, after further evaluation by the staff, the issue was determined to be clearly within the realm of severe accident research and was reclassified as a Licensing Issue.1102 The issue was pursued138l as part of SARP Issue L2, "In-Vessel Core Melt Progression and Hydrogen Generation," documented in NUREG-1365.1382
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 issue. Because licensing and regulatory impact issue 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.
ITEM II.B.5(2): BEHAVIOR OF CORE-MELT
This item was evaluated in Item II.B.5(1) above and determined to be a high priority (see Appendix C). However, after further evaluation by the staff, the issue was determined to be clearly within the realm of severe accident research and was reclassified as a Licensing Issue.1102 The issue was pursued138l as part of SARP Issue L2, "In-Vessel Core Melt Progression and Hydrogen Generation," documented in NUREG-1365.1382
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.
ITEM II.B.5(3): EFFECT OF HYDROGEN BURNING AND EXPLOSIONS ON CONTAINMENT STRUCTURE
DESCRIPTION
Historical Background
TMI Action Plan48 Item II.B.5 called for research into the phenomena associated with severe core damage and core melting. Item II.B.5(3) addressed the effect of H2 burns and/or explosions on containment integrity.
Safety Significance
Whereas Items II.B.5(1) and II.B.5(2) dealt with (among other things) the generation of H2 via radiolysis, metal-water interaction, interaction of a molten core with concrete, etc., Item II.B.5(3) was concerned with the effects on the containment of the burning and/or detonation of this H2. If the containment retains its integrity, even a severe accident resulting in a damaged or molten core produces relatively low offsite consequences. Item II.B.5(3) also included the effect of steam explosions. Again, the emphasis here was not in preventing the explosion but, instead, in maintaining containment integrity.
Possible Solution
Most of the work on Item II.B.5(3) was couched in terms of a stronger containment.
PRIORITY DETERMINATION
Item II B.5(3) was, to a large extent, similar to Issue A-48, "Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment." Issue A-48 was somewhat more general in that it included the effects of a H2 burn or detonation on containment penetrations and on safety systems located within the containment, not just the structural response of the containment. In addition, Issue A-48 included measures for control of the H2 burn and thus had preventive as well as mitigative aspects. However, even though Issue A-48 was expected to use the results of Item II.B.5(3), Item II.B.5(3) was not integrated into Issue A-48 because: (1) the scope of Issue A-48 was still under discussion; and (2) Item II.B.5(3) included steam explosions as well as H2 burns.
Frequency/Consequence Estimate
In WASH-1400,16 the PWR sequences refer to steam explosion-induced containment failures as "α" failures; containment failures induced by an H2 burn are called "γ" failures. Sequences including these two failure modes can be found in Release Categories PWR-1, PWR-2, and PWR-3. It was assumed that the possible solution would result in a 90% reduction in the probabilities of the sequences involving these two failure modes. The results are tabulated in Table II.B-2 below.
| Release Category (F) | α Frequency (per RY) | γ Frequency(F) (per RY) | Consequences(R)(man-rem) | 0.9FR (man-rem/RY) |
|---|---|---|---|---|
| TABLE II.B-2 | ||||
| PWR-1 | 5.3 x 10-8 | - | 4.9 x 106 | 0.23 |
| PWR-2 | - | 7.0 x 10-7 | 4.8 x 106 | 3.00 |
| PWR-3 | 3.4 x 10-7 | - | 5.4 x 106 | 1.70 |
| PWR-7 | -3.9 x 10-7 | -7.0 x 10-7 | 2.3 x 103 | - 0.002 |
| TOTAL: | 4.9 | |||
The PWR-7 category has a negative contribution because a molten core still gives some release, even if containment failure is prevented. Thus, it was assumed that the events which would have been α or γ failures instead lead to PWR-7 releases.
Over a 40-year plant life, the risk reduction above corresponded to about 200 man-rem/reactor. This was calculated using WASH-140016 data for a PWR with a large, dry containment. BWR pressure-suppression containments and PWR ice-condenser containments have a much smaller free volume and thus are more susceptible to α and γ failures. Therefore, the risk for these plants could well be considerably higher.
Cost Estimate
Industry Cost: Without the results of research at the time of this evaluation, it was difficult to assess costs. A stronger containment could cost $15M, based on doubling the 3.5 foot wall thickness of a (150 ft x 200 ft) structure. (Such structures cost roughly $1,000/cubic yard of concrete.)
NRC Cost: NRC costs were considered to be negligible.
Total Cost: The total industry and NRC cost associated with the possible solution was $15M/reactor.
Value/Impact Assessment
Based on an estimated public risk reduction of 200 man-rem/reactor and a cost of $15M/reactor for a possible solution, the value/impact score was given by:
CONCLUSION
The public risk estimate for this issue was significant even for dry containments. Because of the difficulty in determining a cost-effective solution, the issue was given a medium priority ranking (see Appendix C). However, after further evaluation by the staff, the issue was determined to be clearly within the realm of severe accident research and was reclassified as a Licensing Issue.1102 The issue was pursued138l as part of SARP Issue L3, "Hydrogen Transport and Combustion," documented in NUREG-1365.1382
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.
ITEM II.B.6: RISK REDUCTION FOR OPERATING REACTORS AT SITES WITH HIGH POPULATION DENSITIES
DESCRIPTION
Historical Background
This TMI Action Plan48 item involved the review of operating reactors in areas of high population density to determine what additional measures and/or design changes could be implemented that would further reduce the probability of a severe reactor accident, and would reduce the consequences of such an accident by reducing the amount of radioactive releases and/or by delaying any radioactive releases, and thereby provide additional time for evacuation near the sites.
Risk studies were completed in 1981 for the Zion and Limerick sites and in 1982 for Indian Point. Although risk assessments of other sites were conducted by other NRC programs, e.g., National Reliability Evaluation Program (NREP), no further risk studies were envisioned as part of this issue. Further efforts directed towards this issue were review of the analyses and the possible implementation of site-specific fixes to reduce the risk at these sites. Special hearings were scheduled in FY 1982 to review possible design changes for Indian Point and follow-up work in connection with the accepted fixes was anticipated following these hearings.
Safety Significance
Concern existed over the potential for above-average societal risk due to accidents at reactor sites located near regions of high population densities.
Possible Solutions
As mentioned above, hearings were scheduled on possible fixes at the Indian Point site to reduce risk. The actual fixes that resulted from these hearings were unknown at the time of this evaluation. Nevertheless, it was assumed that fixes would be made to reduce the likelihood of the most dominant accident sequences contributing to the frequency of core-melt accidents.
PRIORITY DETERMINATION
Assumptions
Based on a review of similar Reactor Safety Study Methodology Application Program (RSSMAP) and Interim Reliability Evaluation Program (IREP) analyses, it was assumed that two sequences contributed to a large portion (50%) of the likelihood of a core-melt. It was further assumed that it was possible to reduce the frequency of each sequence by a factor of 10.
Frequency Estimate
Based on age and other related factors, it was believed that reactors in this category had an increased frequency of core-melt over the baseline plant (Oconee-3) by a factor of 5.5. Thus, the revised baseline core-melt frequency (F) was given by:
F = (5.5)(8.2 x 10-5/RY) = 4.5 x 10-4/RY
Assuming that the dominant sequences (50% of the frequency) could be reduced by a factor of 10, the revised core-melt frequency was (0.55)(4.5 x 10-4)/RY = 2.5 x 10-4/RY.
Consequence Estimate
Considering the same factors used above to estimate the core-melt frequency, the affected plants would have an exposure increase over the mean population density (340 persons/square-mile) and release fractions by a factor of 3. Thus, this exposure increase (R) was given by:
R = (3)(2.5 x 106 man-rem) = (7.5 x 106) man-rem
The baseline public risk was (4.5 x 10-4/RY)(7.5 x 106 man-rem) or 3,380 manrem/RY. The revised public risk was (2.5 x 10-4/RY)(7.5 x 106 man-rem) or 1,880 man-rem/RY. The resulting change in public risk was then 1,500 man-rem/RY resulting from the reduction in core-melt frequency of 2 x 10-4/RY. Over the estimated 27 years of remaining plant life, this would result in a total risk reduction of 40,500 man-rem/reactor.
Cost Estimate
Industry Cost: Licensee costs were estimated to be $4M/reactor to implement the changes required to reduce the two dominant sequences.
NRC Cost: NRC costs were estimated to be $22,000.
Total Cost: The total industry and NRC cost associated with the possible solution was $(4 + 0.02)M/reactor or $4.02M/reactor.
Value/Impact Assessment
Based on an estimated public risk reduction of 40,500 man-rem/reactor and a cost of $4.02M/reactor for a possible solution, the value/impact score was given by:
Other Considerations
The accident avoidance cost was estimated to be approximately $11M which would result in a potential cost saving of $7M, considering the $4M implementation costs.
CONCLUSION
Based on the above value/impact score, this issue was given a high priority ranking (see Appendix C). A staff review of PRAs submitted by the affected licensees was used to identify the strengths and weaknesses of the various plants and to assess the risk associated with their operation. A special adjudicatory proceeding was held from 1982 to 1983 during which time the issues regarding continued operation and risk of the Indian Point plants were heard. Following these hearings, the Commission concluded that neither shutdown of Indian Point Units 2 or 3 nor imposition of additional remedial actions beyond those already implemented by the licensees were warranted.806
The staff also reviewed the Zion PRA and concluded that the risk posed by the Zion plants was small. The dominant contributors to severe accidents at the Zion plants were examined and the staff recommended that: (1) the integrity of the two motor-operated gate valves in the RHR suction line from the RCS be checked each refueling outage; and (2) the diesel-driven containment spray pump be modified so that it could be capable of operating without AC power.806 Thus, this item was RESOLVED and new requirements were established. DL/NRR was responsible for managing the implementation of the above recommendations.806
ITEM II.B.7: ANALYSIS OF HYDROGEN CONTROL
DESCRIPTION
The TMI-2 accident resulted in a metal-water reaction which involved H2 generation in excess of the amounts specified in 10 CFR 50.44. As a result, it became apparent to the NRC that additional H2 control and mitigation measures would have to be considered for all nuclear power plants. The purpose of this TMI Action Plan48 item was to establish the technical basis for the interim H2 control measures on small containment structures and to establish the basis for continued operation and licensing of plants, pending long-term resolution of the H2 control issue
CONCLUSION
The long-term resolution of this issue was accomplished by rulemaking as part of Item II.B.8. A final rule was published on December 2, 1981 requiring inerting of the small BWR MARK I and II containments. In addition, based on Commission guidance, interim H2 control systems were required as a licensing condition for the intermediate volume Ice Condenser and MARK III containments. A proposed rule was published on December 23, 1981 (Federal Register 46 FR 62281) which required these systems for the intermediate volume containments. Except for pending construction permit (CP) and manufacturing license (ML) applications, no additional requirements for H2 control or H2 analyses were imposed at that time for large, dry containments. However, the proposed rule required that dry containments be analyzed to determine their ability to accommodate the release of large quantities of H2 (75% metal-water reaction). Also, H2 control requirements were established as part of the final Near-Term CP and ML Rule published on January 15, 1982.
Based on the accomplishments above, the basis for continued operation and licensing of plants with respect to the H2 control issue was established. Future work related to finalizing the proposed rule dealing with intermediate volume containments (Ice Condenser and MARK III) and large, dry containments continued as part of Item II.B.8.
ITEM II.B.8: RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS
DESCRIPTION
Historical Background
In the past, safety reviews concentrated on how to prevent a core from being damaged. Consequently, little attention was given to how a severely damaged core could be dealt with after damage occurred. Other subtasks within Task II.B were concerned with the study of the characteristics of degraded and melted cores (research programs) plus some immediate actions to be taken at plants in operation. Item II.B.8 envisioned both a short-term and a long-term rulemaking to establish policy, goals, and requirements to address accidents resulting in core damage greater than the existing design basis.
Item II.B.8 included an Advance Notice of Proposed Rulemaking (ANPRM) and an Interim Rule. The ANPRM was issued on December 2, 1980 (45 FR 65474) and the Interim Rule was issued in two parts: the first was issued in effective form in October 1981 (46 FR 58484) and the second was issued as a proposed rule on December 23, 1981 (46 FR 62281).
On January 4, 1982, SECY-82-1309 was forwarded to the Commission requesting reconsideration of the approach to long-term rulemaking. The events which prompted this request were as follows:
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The substance of SECY-82-1309 was that the uncertainty associated with long-term rulemaking was an inhibiting force on the industry. The paper then recommended that, since new applications were to be standardized, licensing could proceed on these standardized designs using the information available. PRAs and the safety goal would be used to assess plant safety. If plants needed safety features beyond the existing requirements to meet the safety goal, they could be included. This approach would not need rulemaking specifically directed at severe accident mitigation.
The Commission directed310 the staff to make several changes recommended in SECY-82-1.309 The staff then submitted revised papers SECY-82-1A311 and SECY-82-1B1405 that incorporated the changes directed by the Commission, including ACRS input. The evaluation of this item included consideration of Item II.B.7.
Safety Significance
Most of the engineered safety features at nuclear power plants of the existing generation were intended to prevent severe core damage. Relatively little attention was given in the past to dealing with a severely damaged or melted core. Once a core is damaged, the containment will still prevent the release of large amounts of radioactive material. However, once the core melts, the containment is likely to fail (although the hazard to the public varies widely, depending on the way in which the containment fails).
The degraded-core accident rulemaking was intended to require means for dealing with a damaged core. This translated into preventing the release of radioactivity and providing means for recovering from the accident. Specific items to be considered included the following: use of filtered, vented containment; H2 control measures; core retention devices ("core catchers"); reexamination of design criteria for decay heat removal and other systems; post-accident recovery plans; criteria for locating highly radioactive systems; effects or accidents at multi-unit sites; and comprehensive review and evaluation of related guides and regulations.
PRIORITY DETERMINATION
The safety significance of this issue was essentially the same as that of the research programs described in the analyses of Items II.B.5(1) and II.B.5(2) above. Examination of the estimated frequency of core damage and/or core-melt, coupled with estimates of the potential effectiveness of engineering solutions (and their cost) led to the recommended high priority for Items II.B.5(1) and II.B.5(2). In the same manner, Item II.B.8 had the potential for a significant (and cost-effective) reduction in public risk. In addition, it should be noted that some of the plant modifications contemplated were far more expensive to backfit than to forward-fit. Unnecessary delay could have reduced the costeffectiveness of the resolution to this issue.
CONCLUSION
Based on the above evaluation, this item was given a high priority ranking (see Appendix C). Work performed by RES on the H2 control aspect of the issue resulted in a Hydrogen Control Rule that was approved by the Commission and published in the Federal Register on January 25, 1985.807 The severe accident portion of the issue was addressed in April 1983 by a Policy Statement that set forth the Commission's intentions for rulemakings and other regulatory actions for resolving safety issues related to reactor accidents more severe than design basis accidents (48 FR 16014). Certain severe accident technical issues identified under the discussion of long-term rulemaking were to be dealt with for future and existing plants through procedures and ongoing severe accident programs identified in the Policy Statement and described more fully in Chapter IV of NUREG-1070.809 Thus, with the issuance of the rule on H2 control, this item was RESOLVED and new requirements were established.808
REFERENCES
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