ML20236W599
| ML20236W599 | |
| Person / Time | |
|---|---|
| Issue date: | 07/06/1998 |
| From: | Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Diaz N, Dicus G, Shirley Ann Jackson, Mcgaffigan E, The Chairman NRC COMMISSION (OCM) |
| Shared Package | |
| ML20236W593 | List: |
| References | |
| NUDOCS 9808060096 | |
| Download: ML20236W599 (30) | |
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July 6, 1998 MEMORANDUM TO: Chairman Jackson Commissioner Dieus Coiamissioner Diaz Commissioner McGaffigan FROM:
L Joseph Callan Executive DirectorfMOperations
SUBJECT:
AGENCY PROGRAM PLAN FOR HIGH-BURNUP FUEL PURPOSE:
The purpose of this memorandum is to inform the Commission of an agency wide program plan to deal with issues related to utilization of fuel up to the current limit of 62 GWd/t bumup (average for the peak rod) and a strategy for assessing requests for bumups beyond that limit.
This program plan addresses a wide range of issues and was prepared during the past year by RES, NRR, NMSS, and AEOD, with RES taking the lead.
BACKGROUND:
Previous memoranda to the Commission have discussed bumup-related problems with control rod insertion in operating reactors and with regulatory criteria for reactivity accident analysis (September 13,1994, November 9,1994, March 7,1996, November 25,1996, July 15,1997, and December 18,1997). On March 25,1997, the Commission was briefed on a broader range of high-bumup fuelissues. Following that briefing, a Staff Requirements l
Memorandum directed the staff, among other things, to assign a primary point of contact with i
responsibility for integrating the related activities within the NRC. I assigned that responsibility to Ashok Thadant in his new role in RES. One of Mr. Thadani's first actions in that role was to i
direct the staff to prepare an agency wide program pian for high-bumup fuel. The plan would cover the broad range of issues discussed at the Commission briefing.
DISCUSSION OF THE PROGRAM PLAN:
The program plan uses several strategies under the Strategic Plan's goal of preventing radiation-related deaths and Enesses due to civilian nuclear reactors. In particular, it focuses efforts on activities that pose the greatest risk, it maintains a research capability that will provide technical independence, and it encourages the industry to propose regulatory criteria that can be endorsed by the NRC.
Predecisional budget infonnation has CONTACT:
Ralph O. Meyer, RES been removed from the Attachment 415 6789 9808060096 990714 PDR ORG NREB l
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The Commissioners 2
I The program plan addresses nine issues that were discussed with the Commission at the briefing of March 25,1997.
- 1. Cladding integrity and Fuel Design Limits
- 2. Control Rod insertion Problems
- 3. Criteria and Analysis for Reactivity Accidents
- 4. Criteria and Analysis for Loss-of-Coolant Accidents
- 6. Fuel Rod & Neutronic Computer Codes for Analysis
- 7. Source Term and Core Melt Progression
- 8. Transportation and Dry Storage
- 9. High Enrichments (>5%)
Each issue in the program plan is discussed in the following manner. (a) the issue is defined, stating the origin of the concem raised by high-bumup operation; (b) a risk perspective is presented on the issue; (c) a near-term assessment is summarized, explaining why it is satisfactory to wait 3-5 years, in some cases, for research results to achieve final resolution; (d) related NRC research is described; and (e) a description is given of what will constitute final resolution.
The first two issues, related to cladding integrity, fuel design limits, and control rod insertion, are being satisfactorily addressed by industry activities for current fuel designs and the current bumup limit of 62 GWd/t. The last two issues on transportation, storage, and high enrichments are related to expected future actions and indicate the need for new activities rather than the existence of current concems A users request for research assistance on ore of these issues (transportation and dry storage) was recently issued, and needs for additional research on the high-enrichment issue will be defined later this year. These research requests will have to be prioritized, however, as resources are not sufficient to conduct all requested programi..
The issue on source term and core melt progression is not being addressed actively. A brief consideration of bumup-related factors leads the staff to conclude that it is unlikely that high bumup will have a significant effect on source terms or core melt progression. The current source term is thus considered to be adequate for the foreseeable future. A more thorough assessment of these possible effects, utilizing recent French data, had originally been planned for FY98, but reductions in the severe accident area have eliminated funding for that work.
The remaining four issues are being actively addressed and some highlights are mentioned here. It should be noted that related NRC research described in this paper is intended to confirm safety at the current bumup limit While this research may also provide insights for bumup axtensions, it is the staff's intention to shift the responsibility to the industry for providing research to support extensions beyond the current bumup limit of 62 GWd/t. This point is discussed further in the section on licensing and research strategy.
Criteria and Analysis for Reactivity Accidents For all transients and accidents analyzed in a licensing safety analysis, one of the most important high-bumup effects is accelerated oxidation of the fuel cladding and the related loss of
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cladding ductility. The reactivity accidents are particularly sensitive to this effect. The specific accidents of concem here are the rod ejection accident in a PWR and the rod drop accident in a BWR, and these accidents are evaluated with regulatory criteria from Regulatory Guide 1.77 and
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Standard Review Plan 4.2. As discussed in the July 15,1997, memorandum from L. Callan to
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the Commission, these regulatory criteria are considered to be non-conservative in light of some currently available test data from foreign test reactors on reactivity-initiated accidents. Staff assessment of these data, however, has concluded that there is no reason to change currently approved bumup levels, unless the confirmatory research program demonstrates a need for change. This is because the probability of these accidents is low and generic plant transent calculations imply that energy inputs during such transients are low and will remain below the relevant test data failure levels. Nevertheless, there are large uncertainties associated with these limited test results because the database has substantial limitations (see Table 4 on p. 9 of the attachment). Thus confirmatory work is warranted. To provide a more definitive safety assessment for reactivity accidents, the staff will participate in new programs through intemational agreements and will reassess present conclusions in 3-5 years when significant new data become available. Two intemational programs will provide these data. One involsas a new water loop in the French Cabri reactor, NRC is discussing joint U.S. support for this program with the Electric Power Research Institute (EPRI). The other involves a new high-temperature capsule in the Japanese Nuclear Safety Research Reactor (NSRR), a program for which we have an information exchange.
Criteria and Analvsis for Lons-of-Coolant Accidents (LOCA)
NRC's regulatory criteria for LOCAs (10 CFR 50.46) will be affected by enhanced cladding oxidation and related effects that are experienced at high bumup. However, the current criteria are conservative for fresh fuel and may prove to be adequate at high bumup provided that the oxide accumulation prior to the accident is taken into account. Thus, there is no reason to change currently approved bumup limits unless the confirmatory research program demonstrates a need for change.
In FT 1997, a major experimental program was initiated by the staff to establish a database for confirming or revising LOCA criteria and models utilizing typical high-bumup fuel from U.S.
power reactors. Cooperation in this program is being obtained from EPRI, DOE, and saveral foreign agencies. The cooperation from EPRI is being implemented through the recent Memorandum of Understanding (MOU) with RES on Cooperative Nuclear Safety Research (SECY-97 239) and is substantial. EPRI has taken full responsibility for obtaining fuel rods from appropriate power reactors, for procharacterizing the rods, and for shipping them to our laboratory. The first shipment of fuel rods is expected later this year. EPRI has also taken an 3
active role in protest planning for the tests.
l Consistent with the MOU, cooperation with EPRI will be limited to obtaining the experimental data. Interpretation of the data will be done independently by NRC and EPRI to avoid conflicts of interest. Confirmation of existing criteria and models at current bumup levels, or an indication of need for revision, will be available from the new database starting around 2000.
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Criteria and Analvsis for BWR Power Oscillations (ATWS)
For BWR power oscillations that follow an ATWS event, fuel vendors currently use a criterion from Regulatory Guide 1.77 and Standard Review Plan 4.2 that was intended for the pulse-type reactivity accidents (e.g., the rod drop accident). This was believed to be a conservative application of that criterion. Based on the test results just discussed for the reactivity accidents, the conservatism in this application is also in question. However, the power oscillations are slower and probably less damaging than the sharp pulses used in the RfA tests and do not necessarily imply unacceptable fuel damage for the power oscillations. Thus, there is no need to change currently approved bumup levels, unless the confirmatory research program demonstrates a need for change.
9 The staff is doing detailed fuel rod calculations to understand differences in fuel behavior between the power oscillations and a rod drop pulse. Inquiries are also being made about the.
possibility of performing fuel tests with BWR-type oscillations in test reactors in several foreign programs. The final course of action will depend on the results of ongoing analyses and other factors such as interactions with the industry and intemational research organizations.
Fuel Rod & Neutronic Comouter Codes for Analysis Three types of analysis codes, whose results are affected by fuel bumup assumptions, are used by the NRC. One is a steady state fuel rod behavior code that is used to provide input for transient analysis. Another is a transient fuel rod code that can analyze fuel and cladding behavior during transients like LOCA and the BWR power oscillations. The third is a neutron kinetics code that is used to calculate local power for plant transients like a PWR rod ejection accident or BWR power oscillations.
NRC's steady-state fuel rod code had not been kept up to date and was not providing the staff with an adequate tool for reviewing industry submittals. This deficiency was recognized earlier and has been rectified with the issuance of the peer-reviewed FRAPCON 3 code in December 1997. Similar deficiencies were present in NRC's transient fuel rod code, and that code is now being improved. It is anticipated that 3-dimensional neutron kinetics analysis will be needed for some of the reactivity transients. That need is being addressed by adopting Purdue University's PARCS kinetics code and coupling it with NRC's thermal-hydraulic codes. Coupling of these codes should be completed later this year, and assessment activities are planned.
Computer code maintenance and improvement in these areas will have to be conducted on a l
continuing basis as new cladding types, core materials, and bumup ranges are introduced by j
the industry.
Licensing and Research Strategy As discussed earlier, NRC-funded research is directed toward confirming safety at the currently approved bumup levels. However, the program plan provides a licensing and research strategy for further bumup extensions beyond the current limit of 62 GWd/t. There are two unique aspects of this strategy.
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First, the industry will have to develop a data base for revised (or confirmed) regulatory criteria for the extended bumup range. In the past, the NRC has always performed its own research and developed its own data base from which it defined regulatory criteria like the LOCA limits in 10 CFR 50.46 and the criteria for reactivity accidents in Regulatory Guide 1.77. NRC research budgets have declined to a level where supporting such research is no longer possible. Thus the industry will have to perform this research and develop the database.
Second, the staff will encourage the industry to develop criteria and other guidelines that may be needed to obtain NRC approval for an extended bumup range. These would be submitted to NRC and, if endorsed, could replace current Regulatory Guide and Standard Review Plan criteria. This, too, is budget driven and is consistent with the role of industry as suggested in the agency's Strategic Plan.
SUMMARY
The staff has prepared a program plan for high-bumup fuel that: (a) addresses a range of issues that were previously discussed with the Commission, and (b) provides a licensing and research strategy for confirming the safety of currently approved bumup levels and for considering further bumup extensions that the industry is expected to request. For all issues, a basis is given for concluding that there is no need to change the current bumup limit of 62 GWd/t (average for the peak rod) unless confirmatory research work demonstrates a need for change.
Confirmatory work is under way for issues where the basis involves large data uncertainties and analyses. The licensing and research strategy for bumup extensions involves a shift in responsibility to the industry for work that the NRC traditionally does to establish regulatory criteria and guidelines.
1 This memorandum and the attached program plan have been reviewed by the Office of the General Counsel and the Chief Financial Officer, who have raised no objections.
Attachment:
Agency Program Plan for High-Bumup Fuel l
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' Attachment AGENCY PROGRAM PLAN FOR HIGH-BURNUP FUEL l
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l Introduction
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r A goal of the NRC's Strategic Plan is to prevent radiation-related deaths and illnesses from civilian nuclear reactors by avoiding reactor accidents in which substantial damage is done to the
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d reactor core. This is accomplished in part with a number of fuel damage criteria that serve as reactor " speed limits" to prevent postulated events from developing into severe accidents. Fuel damage criteria are the focos of this program plan.
In the 1970s when most of these criteria and related analytical methods (computer codes) were
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being established, high bumup was thought to occur around 40 GWd/t (average for the peak rod).
l Data out to that bumup had been included in data bases for criteria, codes, and regulatory j
decisions, ar.C it was believed that some extrapolation in bumup could be made. By the mid 1980s, however, unique changes in pellet microstructure had been observed from vendor and intemational data at higher bumups along with increases in the rate of cladding corrosion (breakaway oxidation). It thus became clear that something new was happening at high bumups and that continued extrapolation of transient data from the low bumup data base was not appropriate. Thus, on October 4,1993, NRR issued a fo. mal request to RES for assistance on high bumup fuels, and that request initiated the first NRC research in this area in more than a i
decade.
4 The NRC's research that has been completed since that time and is planned for the near future will be described in this plan in the context of confinning previous decisions to permit fuel bumups in licensed reactors up to 62 GWd/t (average for the peak rod). Future approvals for extensions in burnup above the present limit will require additional research and analysis by the industry, and a licensing and research strategy for such approvals is outlined in the final section of this plan. The agency's Strategic Plan encourages the industry to develop codes, standards, and guides that can be endorsed by the NRC and carried out by the industry. This method of addressing extended bumup limits is incorporated in the discussion of the licensing strategy.
. 'Ihe Strategic Plan also incorporates an approach to focus on regulated activities that pose the greatest risk to the public. Therefore, a risk perspective has been developed and is applied for
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each of the issues described below.
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Ria Ps;oective During the past year, a list of potential high-burnup issues has been identified L sed on observed l
operational problems, experimental results from test programs, and an understanding of basic phenomena. This list is given in Table 1. and was discussed with the Commission on March 25, 1997.
Table 1. List of Potential High-Burnup Issues
- 1. Cladding Integnty and Fuel Design Limits l
- 2. Control RodInsertion Problems l
' 3. Criteria and Analysis for Reactivity Accidents l
- 4. Criteria and Analysis for Loss-of-Coolant Accidents l
S. Criteria and Analysis for BWR Power Oscillations (ATWS)
- 6. Fuel Rod & Neutronic Computer Codes for Analysis l
- 7. Source Term and Core Melt Pmgression l
- 8. Transportation and Dry Storage l
- 9. High Enrichments (>5%)
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To help determine which issues warrant greater efforts for resolution, risk concepts have been employed. Of course, consideration of compliance and defense in depth also affect this determination, and a balance will be seen in later sections that discuss each issue.
In general, a reactor event sequence does not produce significant risk unless fuel melting and its resulting large fission product release are possible during the sequence. Therefore, the issue of cladding integrity and associated fuel design limits for normal operation (including anticipated
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operational occurrences), which relate to compliance with General Design Criterion 10 (GDC-10), are not significant from a risk point of view and are not included in probabilistic risk assessments (PRAs).
Control rod insertion (scram) must be capable of preventing fuel damage during normal operation including anticipated transients (GDC-26). Nevertheless, one class of events i
considered in regulation assumes that scram does not occur (anticipated transients without scram, l
ATWS), and that class of events is addressed separately below. Control rod insertion in i
combination with other reactivity control systems must also be sufficient to ensure coolable core geomeiry for postulated accidents (GDC-27). Thus, the risk associated with the failure of control rods to insent is significant, and control rod insertion also has strong compliance and defense-in-depth implications.
r The large reactivity-initiated accidents (RIAs) have the potential to produce unacceptable fuel damage. These are the rod drop accident in a BWR and the rod ejection accident in a PWR.
Probabilistic profiles have been developed at Brookhaven National Laboratory for these RIAs.
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Risk assessments are not available because RIAs have very low probabilities of occurrence and have been considered outside the range ofinterest for modem PRAs.
An early NRC study (1975) examined the probability of a rod drop accident in a BWR that l
would result in an energy deposition of more than 280 cal /g (unacceptable fuel damage).. That study was reexamined in 1986 and reaffirmed. If adjustments are made to that study to obtain the probability that a rod drop will result in an energy deposition of only 100 cal /g (a damage limit that might be more appropriate for high-bumup fuel), the resulting probability is ~107. For reference, any generic safety issue with an associated core damage frequency of 10-' or less l
would be dropped from further consideration by the NRC using a prioritization scheme based on l
principles of the Commission's safety goals.
Past studies are not availdle on failrx probabilities for PWR rod ejection accidents, and no failures have occurred in control rod drive mechanisms in over 2400 reactor years of operation world wide; therefore, an estimate has been made. From this observeion, it is estimated that the failure probability is no larger thar. -2x104 per reactor year, which is consistent with estimates of the frequency of pipe breaks based on mechanistic models. It is further assumed that only half of the rods that could be ejected would result in prompt criticality, which is then assumed to result l
in unacceptable fuel damage. Further, prompt criticality is expected to happen only when the -
l reactor is at hot zero power, which is less than 1% of the time. Combining these factors leads to an estimate of-104 per reactor year. This value isjust within the range ofinterest for generic issue consideration.
Loss-of-coolant accidents also have the potential to cause unacceptable fuel damage and they have been studied extensively in recent PRAs. From the PRA data base, which was constructed from licensees' individual plant examinations, core damage frequencies are seen to have the following ranges (Table 2).
Table 2. Core Damage Frequencies per Reactor Year Large Break LOCA Small & Medium Break LOCA PWR 3x10 8 to lx10-7 PWR 5x108 to lx107 BWR 2x104to Ix104 BWR 9x104 to 1x104 Loss-of-coolant accidents are significant risk contributors in PRAs, and these frequency ranges are seen to be within the range of consideration for generic safety issues.
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4 Finally, power oscillations in a BWR ATWS also have the potential to cause unacceptable fuel damage. Current licensing analyses show that fuel enthalpy remains below 280 cal /g and thus conclude that coolable geometry is maintained and there is no fuel dispersal. There is, however, concern that the 280 cal /g value may be too high, especially for high-burnup fuel. Probabilistic analyses do not exist with a lower value, but an upper bound on the probability of exceeding such a value would be the probability of occurrence of the ATWS in the first place. This probability is ~1x104. In general, ATWS events are found to comprise about 15% of the total risk for BWRs.
Plam to Resolve the Issues In the following sections, each of the issues identified in Table I will be discussed. Risk perspectives will be discussed, along with compliance and other considerations, to help determine appropriate regulatory actions and research efforts. A near-tenn assessment of each issue will be described to show why, in some cases, it is satisfactory to wait 3-5 years for research results in order to achieve a more final resolution. Where applicable, related NRC research programs will be described along with their schedules. And finally, the expected final resolution of each issue will be outlined. Schedules for research programs and overall resource requirements associated with this program plan are given in the Attachment.
While not explicitly discussed below, it should be noted that the NRC staffs activities to addrest these issues involve significant external interactions. Much of the research is now done in cooperative programs. Some of these, like the Haldo Project, are international projects in which we participate as a member. Others, like the collaboration with France, Japan, and Russia on reactivity accidents, are arranged with bilateral agreements. In other cases, EPRI and DOE participate in NRC research projects with memoranda of understanding. Technical discussions are maintained with the nuclear industry through daily regulatory activities, the Regulatory Information Conference, the Water Reactor Safety Information Meeting, ACRS meetings, and other special workshops.
- 1. Cladding Integrity and Fuel Design Limits (a) Description ofIssue Geacra! Design Critedon 10 states the principle that specified acceptable fuel design li"aits (SAFDLs) should be met to assure that integrity is maimabed in the first barrier few:ntion of fission products - the fuel rod cladding - durit:g normal operation and anticipated operational occurrences. That is, cladding defects (also simply called cladding failures) should not occur under those conditions. The following list identifies l
some of the SAFDLs that are dexribed in the Standard Review Plan (principally in ?RP 4.2).
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e Stress limits e Fatiguelifetime e Rod pressurelimits I
e Hold down spring capability e Pellet moisture limits-e Uniform claddir.g strain less than 1%
i e No pellet centerline melting I
e Mechanied loads less than 90% ofirradiated yield stress e Avoidance of critical heat flux (limits on Deperture from Nucleate Boiling Ratio, DNBR, and Minimum Critical Power Ratio, MCPR) l l
It is likely that some of these fuel design limits would be affected by pellet
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microstructural changes or the reduction in cladding ductility that accompany high burnup, thus affecting cladding integrity. Recent experience suggests that cladding failures, or cladding damage that might lead to failures, have occurred as the restdt of achieving higher bumups or attempting to reach higher burnups.
(b) Risk Perspective There is no significant contribution to plant risk from cladding failures during normal operation and anticipated operational occurrences because only small fission product releases are possible without core melting. however, it appears that high burnup fuel design has contributed to an increase in the severity of fuel cladding failures that have J
occurred with gross fuel release during normal operation and during fuel handling. The attendant fuel particle contamination within and outside the plant are a safety concem.
(c) Near-Term Assessment See Final Resolution.
(d) Related NRC Research None.
(e) Final Resolution Although bumup related problems have occurred, there has been, nevertheless, an overall j
trend during normal operation ofimproved fuel performance in the past fifteen years.
l Figure 1 shows this improving trend in the number of fuel assemblies containing fuel rods with defects, normalized per giga Watt of generated electricity, for the period from 1980 to 1996. Figures 2 and 3 show the actual number ofindividual fuel rod failures during the last few years, and it can be seen from the totals for the most recent years that this average is about 1-2 fuel rods per reactor per year. Compared with the number of I
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6 fuel rods in a typical core (~50,000), this defect rate is very low.
Reactor coolant cleanup systems are designed to accommodate on the order of 1% fuel failures (about 500 rods), and off-site dose limits can be met at that level. Nevertheless, the intent of GDC-10 is to maintain the first barrier to fission product release (defense in depth), which means near zero failures. From the discussion above, it is clear that the goal of near zero failures is being met. NRR will continue to monitor industry performance in this area and assess the significance of fuel failures, trends, and root causes.
- 2. Control Rod Insertion Problems (a) Dewription ofIssue in late 1995 and early 1996, several control rods failed to insen fully during scrams at two PWRs (South Texas and Wolf Creek). All of the affected control rods were positioned in high-bumup fuel assemblies. Upon inspection of the rods and fuel assemblies, the control rods were found to be in good condition, but the fuel assemblies were deformed. Related evidence was found in North Anna and at a number of plants of similar design in Europe.
(b) Risk Perspective While the incomplete rod insertion events at Wolf Creek and South Texas resulted in loss of only a small amount of shutdown margin, these events could be precursors of events with more serious consequences. Review of the Wolf Creek data indicated that thimble tube distortion'high in the core had the potential for control rods sticking at those locations. Continued operation under these conditions could have resulted in loss of significant shutdown margin.
(c) Near-Term Assessment Very shortly after these incidents occurred, the staffissued Bulletin 96-01 requesting special actions to ensure compliance with the current licensing basis for the facilities with respect to shutdown margin and control rod drop times. Those actions included additional training for operators, a review of control rod operability based on the recent events, testing of rods starting at the next appropriate shutdown, and review of scram data for anomalous indications.
(d) Related NRC Research None.
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(e) Final Resolution The root cause for the observed cases ofoontrol rod sticking was determined by Westinghouse to be the fuel assemblies' response at high burnup to several aspects of fuel design including creep, oxide thickam, operating temperature, holddown spring stiffness, thimble tube thickness, and dashpot dimensions. To verify the safety of operating plants, the staff considered a requirement for increased control rod testing frequency if control rods were located in high-bumup fuel assemblies. A supplement to Bulletin 96-01 was drafted to require such testing, and the draft was issued for public comment. While the draft was out for comment, additional technical information was provided and the industry indicated that it was redesigning the fuel assemblies and improving core management to eliminate the problem. Thus the staff has decided that the bulletin supplement is not needed. Utility awareness has been heightened, and it is believed that utilities have the means and ample motivation to avoid this problem in the future. The staff will continue to monitor industry performance in this area, and the issue needs to be addressed by the industry in any submittal for new fuel designs or bumup extensions.
- 3. Criteria and Analysis for Reactivity Accidents (a) Description ofIssue The specific accidents of concem are the rod drop accident in a BWR and the rod ejection accident in a PWR. For these postulated accidents, the NRC uses criteria to ensure that fuel rods remain coolable and that fuel particles are not dispersed into the coolant (280 cal /g peak fuel enthalpy) and to indicate the occurrence of cladding failure (DNB, MCPR, 170 cal /g peak fuel enthalpy) for the purpose of dose calculations. Tests in the French CABRI reactor in late 1993 with some highly degraded commercial fuel resulted in l
cladding failure at very low fuel enthalples (-30 cal /g average for a fuel rod) and l
substantial fuel dispersal. Analysis of these and similar tests showed that failures were occurring in a partially brittle manner, as a result of the mechanical expansion of the pellets, rather than by dryout and overheating of the cladding as addressed by the current criteria it thus appears that the current criteria may not achieve their purpose for high-bumup fuel.
(b) Risk Perspective From the general discussion of risk presented.above, it is seen that the frequency of l
occurrence of a BWR rod drop accident is below the range ofinterest for consideration as a generic issue whereas the frequency for a PWR rod ejection accident is just within that range. Nevertheless, their consideration is explicitly required by GDC-28. "Iherefore, it l
would seem appropriate to analyze these events more realistically (i.e., in a less conservative manner) than in previous analyses because of the low risk.
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8 (c) Near-Term Assessment Shortly after leaming of the CABRI data in 1994, an assessment was made based on probability and power level, and the staff concluded that there was no safety concem requiring immediate regulatory action (Taylor memo to the Commissioners,9/13/94).
NRC and U.S. industry calculations, which were presented at an OECD conference (Cadarache,1995), further suggested that there would be no cladding failure (and hence no fuel dispersal) during these accidents, provided that heavily oxidized fuel with spalling is avoided.
(d) Related NRC Research No test program of this kind has been in operation in the U.S. for over 15 years, so the NRC entered into formal agreements with France (CABRI test reactor), Japan (NSRR test reactor), and Russia (IGR test reactor) to obtain data from current prograns. The NRC also initiated generic plant calculations (mentioned above) and an assessment (largely in house) of the test data and plant calculations. Results of this assessment were documented in ajoumal article and in Research Information Letter (No.174, March 3, 1997). Based on those results, RES suggested tentative interim criteria shown in Table 3.
Table 3. Tentative Interim Criteria for RIAs Oxide spalling:
none allowed Cladding failure:
100 cal /g (enthalpy increase)
Coolability:N 280 cal /gN(enthalpy limit) <30 GWd/t No cladding failure >30 GWd/t l
WLoss of coolability is equated to fragmentation of the rod (several pieces) at low bumup and dispersal of fuel particles through cladding defects at high bumup.
Mhere is evidence that the 280 cal /g value should be reduced to 230 cal /g, but this is not a high-bumup issue per se.
A fixed bumup limit was not given for these criteria because burnup does not seem to be the most important variable (it is oxidation). The data base for these criteria includes bumups to 64 GWd/t and oxide thicknesses to 130 microns (one test), with most oxide thia-below 80 microns; however, no new phenomena, like pellet microstructural I
changes or breakaway oxidation of the cladding, are expected just above these values.~
'Ihe main limitation appears to be that oxidation should not be so severe that spallation occurs because that introduces known phenomena that can cause localized embrittlement.
Although the test programs just mentioned provide valuable data for an interim assessment, these programs have also provided enough understanding of the related
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%= to know that the current data base has substantial limitations (Table 4). As a result, there remains considerable disagreement among the intemational community as to what fuel enthalpy value constitutes an appropriate safety limit. In view of the data uncertainties, the staff does not believe that adopting revised criteria is appropriate at this time.
Table 4. Limitations of Current Data Base Pulse:
too narrow in NSRR and most of CABRI tests Cladding:
mostly opsolete varieties of Zircaloy Coolant:
sodium in CABRI and stagnant water in NSRR Temperature: too lowin NSRR Pressure:
too lowin CABRI and NSRR To address these uncertainties in a cost-effective manner, the staff will participate in new programs through intemational agreements. In France, a new water loop will be constructed to test more current PWR cladding types with prototypical pulse widths, water as the coolant, and appropriate coolant flow to investigate cladding failure and the effects of dispersed fuel particles. In Japan, a new high-temperature, high-pressure capsule will be constructed to test more current PWR and BWR cladding types, and pulse-width effects will be cross checked with the French program.
Test schedules for NRC's participation in these RIA test programs are shown in the Appendix, and significant new results are not expected for 3-5 years. The costs ofNRC's participation in these programs is highly leveraged.
1 (e) Final Resolution i
i Based on our current interpretation of the data, generic safety assessments, and the low probability of BWR rod drop and PWR rod ejection events, no reanalysis will be required f
for extant approvals. When significant results from the new test programs become available (3-5 years), this confirmatory assessment will be revisited and modified if necessary.
- 4. Criteria and Analysis for Loss-of-Coolant Accidents I
(a) Description ofIssue For these postulated accidents, the NRC uses cladding embrittlement criteria (2200*F peak cladding temperature,17% cladding oxidation) to ensure that coolable geometry is not lost; and related models must be used in safety analyses for oxidation kinetics, ballooning, rupture, and flow blockage to demonstrate that long-term cooling is
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maintained. Additional analyses are performed to show that seismic and blowdown loads do not fragment the bel or interfere with control rod insertion during and after a LOCA.
The criteria, models, and analyses being used today were based on data from unirradiated l
l cladding, yet the burnup process will likely have an effect. High burnup fuel rods can l
accumulate heavy oxide coatings during normal operation and experience some loss of ductility (embrittlement) from related hydrogen absorption. In a few cases, measured oxide levels have approached the 17% limit locally. Further, the enhanced " breakaway" oxidation that is observed in some cladding types suggests that the oxidation kinetics at high temperature would be increased. Thus it is likely that the criteria and models for LOCA analysis will be affected at high bumup, although it is not clear that high-burnup l
fuel will become limiting.
(b) Risk Perspective l
Core damage frequencies for LOCAs are as high as ~5x104 for PWRs and ~9x104 for l
BWRs, and while these numbers are quite small on an absolute scate they represent significant risk contributors. The compliance issue is also strong for LOCAs as the embrittlement criteria mentioned above are written directly into the regulations (10 CFR 50.46). Thus the question of high-bumup effects on LOCAs is being given a high priority.
(c) Near-Term Assessment Unlike the reactivity accidents, there is no evidence at this time that the criteria or analytical conclusions for LOCAs are deficient. Current embrittlement criteria, which are conservative for fresh fuel, may be adequate at high bumup provided that the initial oxide I
accumulation is taken into account. Preliminary tests in France indicate that this will be i
the case. The amount of oxidation that is predicted to occur during a LOCA transient is often small compared with the 17% limit, so the remaining margin may accommodate a l
large amount ofinitial oxidation from normal operation. Further, changes at high burnup in cladding ductility are likely to be in the favorable direction (less deformation and flow blockage with less ductility). Fuel vendor calculations also show that high-burnup fuel has lower peaking factors and is less limiting for LOCA analysis than fresh fuel, so in general there may be more margin all around to accommodate changes due to bumup.
However, in one recent case involving Westinghouse fuel rods with a bumable absorber, analysis showed that the 17% limit might be emdad for a LOCA at some time in the future if the ap-*d amount of oxidation at high burnup were included. Those cases are being addressed on a case-by-case basis to ensure continued compliance.with regulatory requirements. An information notice is being drafted to alert all licensees of these high-bumup effects and their potential impact on the requirements of 10 CFR 50.46.
L___________-_
l I1 Changes may also be needed in allowable structural loads for earthquakes during and after a LOCA because the strength and ductility of high-burnup cladding will not be the same as for fresh material. Elastic analyses are usually performed and resulting loads are compared with ASME allowable values. Analyses for fresh fuel show ample margins, and the increased strength at high bumup would seem to enlarge those margins. But this method of assessment presumes that the material being analyzed is ductile, whereas substantial losses in ductility occur at high burnup for some heavily oxidized fuel rods.
'Ihese effects are being addressed in the research program described below.
l (d) Related NRC Research Fuel behavior during loss-of-coolant accidents is assessed with embrittlement criteria and I
several types of analyses. (a) The initial stored energy in the fuel is calculated with the f
NRC's FRAPCON-3 code or similar vendor and licensee codes. (b) During the transient, the amount of oxidation and the peak cladding temperature are calculated for comparison with the embrittlement criteria, and the deformation of the rod (amount of ballooning) is 1
calculated to provide related flow blockage. NRC's FRAPTRAN code can calculate these quantities, and models of these phenomena are usually built into vendors' systems codes.
(c) Systems codes like the NRC's TRAC-P and TRAC-B codes calculate the entire plant i
transient, including the long-term cooling phase. (d) Finally, finite-eleraent structural mechanics codes are used to calculate the fuel assembly and core response to seismic and LOCA loads.
l The FRAPCON-3 and FRAPTRAN codes will be discussed in a following section (Issue l
6). Plant systems codes, which describe thermal, hydraulic, and neutronic behavior of the reactor, are not in the scope of this program plan. Nevertheless, fuel-related models developed in this program will be fed into the plant systems codes as appropriate.
l Embrittlement criteria and fuel behavior correlations for LOCA analysis are being investigated in a program at Argonne National Laboratory (below), and fuel structural response is also being addressed in that program.
l In FY97, a major program was initiated to establish a data base for LOCA criteria and models utilizing typical high-burnup fuel from U.S. power reactors. The program is being carried out in the hot cells at Argonne National Laboratory and will also provide fundamental mechanical properties, measured under temperature and rate conditions that are applicable to a wide range of postulated transients and accidents. Cooperation on obtaining and preparing fuel rods for the tests is being obtained from EPRI and DOE, and collaboration on technicel matters is also being obtained from France, Japan, and Russia.
A detailed test plan has been prepared for this program and is being followed.
Experimental techniques am being developed at this time, and fuel rod acquisition is expected in 1998. Test schedules are shown in the Appendix.
\\
I 12 (e) Final Resolution Confirmation of existing criteria and models at current burnup levels, or an indication of need for revision, will be available from the new database staning around 2000.
l The 280 cal /g criterion that is used as a limit for reactivity accidents is also used for.
BWRs to demonstrate the absence of cladding fragmentation, fuel dispersal, and related phenomena during an anticipated transient without scram (ATWS) with power oscillations. Since test results for the reactivity accidents show that fuel died can occur at much lower fuel enthalpies for high-bumup fuel, the 280 cal /g limh may not
]
ensure coolable geometry for high-bumup fuel subjected to the power oscillations.
(b) Risk Perspective l
i From the discussion above, the probability of a BWR ATWS with power oscillations is
~1x10-5. Although this would be a conservative estimate of the probability of causing unacceptable fuel damage (level unknown), it is high enough that this issue should be pursued.
(c) Near-Te:m Assessment The ATWS oscillation transient is much slower than the reactivity accident pulse. More heat transfer will occur such that expansion-driven stresses on the cladding are reduced j
and cladding temperatures are increased, thus reducing the likelihood of cladding failure.
Further, fuel fragmentation and dispersal may be reduced or eliminated without the I
prompt expansion of fission gases on grain boundaries. Therefore, questions about the adequacy of the 280 cal /g limit do not necessarily imply unacceptable fuel damage for such power oscillations. Because of the low probability of this event, there is no l
immediate safety concern, and research activities have been initiated to address this situation.
l (d) Related NRC Research l
BWR ATWS oscillations are being analyzed in house with the FRAPTRAN transient fuel rod behavior code. These calculations will attempt to estimate fuel enthalples, cladding stresses, and fission gas behavior so that the fuel duty can be compared to that during a reactivity pulse transient (e.g., rod drop accident). Only approximate results can be expected at this time because NRC's fuel rod codes have historically been designed for thermal calculations rather than mechanical calculations, and such improvements in the
E s
1 13 1
l codes will not be available for a couple of years.
l Inquiries are also being made about the possibility of performing BWR-type oscillations in test reactors in several foreign programs. There is no facility in the U.S. that could l
perform such tests without large startup costs. Specific research on cladding fragmentation, fuel dispersal, and fission product release is needed to support the assessment of this postulated event.
(e) Final Resolution The final course of action will depend on the results of ongoing analyses and other factors such as interactions with the industry and intemational research organizations.
- 6. Fuel Rod & Neutronic Computer Codes for Analysis l
(a) Description ofIssue NRC uses FRAPCON-3, a steady-state fuel behavior computer code, to audit similar vendor codes that are used to calculate LOCA stored energy, end-of life rod pressure, gap l
activity, and to perform other licensing analyses. FRAPTRAN, a transient code, is also used by NRC for special calculations and to interpret test results. Although the vendors l
were using fuel codes that had been updated for high bumup applications, at the time that reviews were being done of vendor requests tc go to 62 GWd/t, NRC's codes had not been updated for about 10 years and had been validated out to only about 40 GWd/t (rod average). Thus NRC's ability to deal with high bumup fuel issues has been hampered by outdated analytical tools.
(
For reactor power calculations, neither the industry nor the NRC is, as a rule, using 3-D neutronics codes. Postulated accidents like the rod ejection in a PWR, the rod drop in a
(
BWR, and the BWR ATWS power oscillations are very localized in nature and cannot be analyzed w:ll without 3-D kinetics codes. While some industry 3-D codes have been l
, abmitted for NRC review, most licensing codes do not have this capability or involve overly simplifying assumptions. NRC also occasionally uses its own 3-D neutronics codes for special analyr:s, but those codes are not coupled with the NRC's principal thermal-hydraulic codes. To accommodate the reduction in resistance to failure that fuel cladding experiences at high burnups,3-D licensing analyses may be needed to avoid penalties associated with the current conservative kinetics models. Such 3-D codes would require NRC review and appmval, i
in addition to the transition to a more dimensional ki setics analysis, there are several specific features of the kinetics codes that may need to be modified to address localized l
high bumup effects. One is the local power peaking during rapid power pulses (critical l
or prompt critical) that may not be treated conservatively by codes that use fuel rod l
14 bundles as the smallest calculational node rather than individual fuel rods. Another is the reduction in the delayed neutron fraction that results from the buildup of plutonium isotopes at very high bumups. These and other high-bumup code features need to be examined :arefully.
(b) RiskPeri,pective ne NRC's and the vendors' fuel codes can be used for a range of applications including safety analyses for LOCA. Because of the risk significance of LOCAs, these codes are therefore important from a risk perspective. The neutronics codes can also be used for a range of applications including the rod drop and rod ejection accidents, which are not particularly large risk contributors. However, this same capability is also needed to analyze the BWR ATWS, which has potentially greater risk significance, and other power transients.
(c) Near-Term Assessment The need for improved NRC fuel rod codes was identified early, and a major part of that work has now been completed. For the neutronic codes, a long lead time will be required to prepare codes for NRC review and to conduct a review if such codes are submitted by the industry.
(d) Related NRC Research The steady-state fuel code, which is used most frequently by NRC in licensing activities, has been updated as FRAPCON-3 and is applicable to about 65 GWd/t (rod average). A peer review was conducted, and the code and its documentation were issued in December 1997. Work is currently underway to upgrade the transient fuel code, FRAPTRAN, and this phase of the work will continue in FY98-99. To date, improvements in both' codes have focused on thermal analysis, and additional improvements are needed in the mechanical models. Further updates of the thermal models will also be made as additional data become available at higher burnups and for higher concentrations of burnable poisons. Detailed work plans have been developed for FRAPCON-3 and FRAPTRAN, and these plans are being followed. Schedules for the code-related research are shown in the Appendix.
The NRC's TRAC-P and TRAC-B plant systems codes do not currently have 3-D B fiics models, so the RAMONA-4B neutronics code has been used by NRC for iahp ~kt 3-D studies of the BWR rod drop accident with high-bumup fuel. To improve NRC's plant analysis capability, Purdue University's PARCS 3-D kinetics code is being coupled with TRAC-P and TRAC-B to provide the full three-dimensional capability. To estimate the effects oflocal power peaking within fuel bundles, the PARCS code is being compared with the Russian BARS. code, which has pin-ta-pin modeling. The effects of
e 15 local power peaking, delayed neutron fraction, and other high-bumup effects will be assessed in an ongoing program at Brookhaven National Laboratary.
(e) Final Resolution Maintaining steady-state and transient fuel codes and updating them for new bumup ranges and new fuel and cladding materials will be a continuing activity. 'Ihe fuel vendors will also continue to update their codes as new data become available for higher bumups and modified fuel designs. However, resolution of this issue will occur for the fuel codes when the current FRAPTRAN update has been completed to install the high-bumup thermal models developed for the recently updated FRAPCON-3 code.
For the 3-D neutronics codes, resolution will be largely achieved when NRC's new 3-D capability, with the coupled PARCS code, becomes available in 1998. There will be, nevertheless, continuing activities to maintain this capability and to keep the codes updated for new fuel materials and new operating conditions. Industry submittals of new 3-D codes for high-bumup applications would require a long lead time for staff review.
- 7. Source Term and Core Melt Progression (a) Description ofIssue During a severe accident, the progression of the accident sequence is strongly dependent on the way molten material develops in the core. Radiological releases, in tum, are detennined by the progression of the accident. Estimated releases for a spectrum of severe accidents have been used to develop the recent NUREG-1465 source term. This source term, however, may not be applicable for fuel irradiated to high bumup levels (in excess of about 40 GWd/t as noted in NUREG-1465). It is known that at higher bumups l
the gap inventory will increase, fuel particle behavior will be different, and the isotopics will shift. It is also known that cladding becomes more brittle at higher bumups, potentially resulting in earlier cladding failure and fuel relocation during a severe l
accident.
(b) Risk Perspective i
Since risk is the product of probability and consequence, understanding core melt progression and having a source term are &=y even to determine risk for various events. The severe accident source term is also used for analysis of consequences of a LOCA, and LOCA is a risk-significant design-basis accident. Thus, determining the effect of high bumup on source terms and core melt progression is itself very important from a risk perspective.
l
16 (c) Near-Term Assessment The main effects that might impact source terms and core melt progres.: ion at high burnup are (a) a reduction in the amount of unoxidized zirconium in the core, (b) embrittlement
'of the fuel cladding, (c) an increase in the release of fission gases from fuel pellets during normal operation, (d) fragmentation of fuel pellets, and (e) a shift in the spectrum of fission products produced as plutonium fission becomes more important.
A reduction in the amount of unoxidized zirconium metal in the core could diminish the severity of core melt and subsequent ex-vessel phenomena by lowering the reaction heat from metal oxidation. However, the amount ofpreoxidation of the cladding will be less than 17% of the wall thickness because of the restrictions of 10 CFR 50.46 and it likely to be much lower than that for newer cladding alloys, so this beneficial effect would be small. Non-molten fuel relocation may occur due to cladding embrittlement, particularly for scenarios involving delayed reflood or depressurization, but this is not expected to significantly afTect the overall outcome of uninterrupted core melt accidents. Gap activity comprises only a small part of the source term so that even large changes in gap activity would not have a big effect on the source term. (However, some licensing analyses use only gap activity, e.g., the fuel handling accident, and for those, consideration will need to be given to the increased gap activity resulting from the use of higher bumup fuel.) Fuel fragmentation has been observed at high bumup, but it appears that dispersal of fragments occurs by washout and there may be no means to get thai material into the atmosphere as aerosol particles. In contrast, particulate releases included in the source term are lifted from the core as high temperature gases that j
condense as aerosol particles. The source term itself consists of release fractions and i
therefore would not be affected by isotopic shifts. Those shifts would be accounted for in I
the generation analysis (e.g., with the ORIGEN code) and any changes are expected to be 1
small.
I Considering these factors, it is unlikely that high burnup will have a significant effect on source terms or core melt progression.
(d) Related NRC Research The applicability limitation of 40 GWd/t, mentioned in NUREG-1465, came from the data range of the HI and VI fission product tests at Oak Ridge National Laboratory.
Similar measurements on higher burnup fuel specimens are being made by CEA (France) at Grenoble, and results from those tests are available to the NRC through NRC's l
Cooperative Severe Accident Research Program. An assessment of the effects of high bumup on core melt progression and the source term, utilizing recent French data, was scheduled for FY98, but funding is no longer available for this work.
17
~
(e) Final Resolution The current source term is considered to be adequate for the foreseeable future.
- 8. Transportation and Dry Storage (a) Description ofIssue Two aspects of transportation and dry storage of spent fuel that might be affected by high burnups are the nuclide inventory and long-term cladding integrity. The nuclide inventory in tum affects shielding, heat sources, and potential releases of activity. As in reactors, the spent fuel cladding is the first barrier for retention of fission products. The cladding's integrity affects potential releases of fission products and the ability of licensees to safely retrieve the spent fuel for ultimate disposal.
(b) Risk Perspective A dry cask PRA has been initiated at Brookhaven National Laboratory and could provide a risk perspective for this issue, although the funding to complete this work is not available.
(c) Near-Term Assessment This issue addresses future actions that are now under consideration. Vendors of spent fuel casks have applied for storage of fuel with bumups up to 65 GWd/t (average for the peak rod), which is well above the bumup level for which current methods and assumptions have been approved.
(d) Related NRC Research Research will be defined in FY98 to address the two topics of nuclide inventory and long-term cladding integrity.
(c) Final Resolution Final resolution depends on the outcome of this future research and will occur when spent fuel casks are approved for fuel with high bumups.
l l
- 9. High Enrichments (>5' '
(a) Description of issue To date, the validation of criticality safety codes, and a.wiated cross section libraries,
l l
1 P
l l
l I8 for LWR fuels has concentrated on enrichments less than 5%. Neither benchmarks of code performance nor the bases for extrapolating code performance in the enrichment range of 5-10% have been well established. Moving into this range will require care r
l because the physics of criticality begins to change as enrichments reach 6% and beyond, l
l where single moderated assemblies can go critical and criticality of weakly moderated or l
unmoderated systems becomes possible. Enrichments above 5% will require redesign of some fuel fabrication and handling equipment and fuel transportation packages. The possibility of recriticality during severe accident core melt sequences should also be 1
l addressed as this could alter the pr'ogression of such accidents.
l l
(b) Risk Perspective i
Risk studies have not been performed.
(c) Near-Term Assessment This is an emerging issue. Some enrichment facilities and fuel fabricators have formally l
stated the intent to go to enrichments greater than 5%. (Aspects of the same criticality validation issue have already arisen in the ongoing downblending of surplus HEU to 5%
l l
enrichment. One downblending facility recently received an infraction for having failed to validate its criticality analysis methods in the 5-10% enrichment range.)
1 (d) Related NRC Research I
Ongoing research at ORNL on the ranges of applicability of criticality validation is aimed in part at helping address this issue. Needs for any additional research in this area, such as analytical benchmark studies, new experimental benchmark data, and severe accident considerations, will be defined in FY98.
(e) FinalResolution Final resolution depends on the outcome of this future research and will occur when higher enrichments are approved.
j Iicentina and Ree-ch Strateev The data and analyses described above (some of which will not be completed for 3-5 years) are intended to provide confirmation of acceptable fuel behavior for current fuel designs up to the present bumup limit. To obtain higher bumup limits, additional data and analyses of a similar nature will have to be provided. To avoid the need for extensive confmnatory work in the future, sufficient data and analyses will have to be provided prior to receiving NRC approvals.
19 In the past, the NRC has always performed the research needed to derme regulatory criteria, and the industry has performed research to develop methods of demonstrating compliance with those criteria. In recent years, NRC's research budget has declined to a level that the NRC can no longer support such research. nus, if the industry wants further bumup extensions, it will have to develop a data base for revised (or confirmed) regulatory criteria. The staff will make it clear to the industry that such research must be non-proprietary, to ensure that resulting criteria are fully scrutable, and the NRC staff must have full access to those research programs. If NRC resources are available, the NRC will actively participate in those research programs; however, the industry will be expected to take the lead in this work.
In accordance with the NRC's Strategic Plan, the staff will also encourage the industry to develop codes, standards, guides - and, by inference, fuel damage criteria - that can be endorsed by the NRC and carried out by the industry. Fuel behavior would have to be addressed during normal operation, transients, and postulated accidents, and, at a minimum, the high-bumup issues identified above would have to be covered. A program for monitoring fuel performance should be included in the industry proposal.
Also in line with the Strategic Plan, these codes, standards, guides, and criteria could be focused on events that pose the greatest risk to the public, based on probabilistic risk assessment concepts and other approaches for determining high-and low risk activities. If found acceptable, these codes, standards, guides, and criteria would be incorporated in a regulatory guide and endorsed by the staff. He review, public comment, and issuance process would likely take 12-18 months from receipt of a comprehensive industry proposal. Demonstration ofcompliance with the provisions of the guide would follow for a particular fuel design and bumup limit.
To develop a data base necessary to justify further bumup extensions, suitable fuel rod specimens would have to be available for testing under transient and accident conditions. For this purpose, the NRC would encourage the irradiation oflead test assemblies (LTAs) with typical bumup histories up to the proposed licensing limit and positioned in near limiting core locations. NRC would further encourage the irradiation of segmented test rods in the LTAs to facilitate subsequent testing. The NRC would also consider limited cooperation with the industry in the data phase of such test programs as that would make important data available to l
the NRC for its own independent assessment.
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3 Appendix HIGH BURNUP FUEL RESEARCH SCHEDULES Issue 1. Cladding Integrity and Fuel Design Limits None issue 2. Control Rod Insertion Problems None issua 3. Criteria and Analysis for Reactivity Accidents l
CABRITest Reactor (France) l l FY 98 l FY 99 i FY 00 l FY 01 l FY02 l
l sodium Loop Tests l
[ Water Loop Tests l l Water Loop Construction & Installation l
l
)
NSRR Test Reactor (Japan) l FY 98 l FY 99 l FY 00 l FY01 l FY02 l
1 l Ambient Pressure, Temperature Capsule l
l l High Pressure, Temperature Capsule l
l IGR Test Reactor and RIAR Hot Cells (Russia) 1FY98 l FY 99 l FY 00 l FY 01 l FY02 l
l l
l Document IGR l
l i
l Zr-Nb Mechanical Properties l
1 NeWonics and Fuel Codes l
in-House Assessment l FY'98 i FY 99 l FY 00 l FY 01 l FY02 l
l In-House Analysis and Assessment as Needed l
l l
l, 2
l Issue 4. Criteria and Analysis for Loss-of-Coolant Accidents ANL Hot Cells l FY M i FY 99 l FY 00 i FY 01 l FY 02 l
l OxidatenTests l
[
l Ring Stretch Tests l l Tube BurstTests l
%d3 Criteria Tests l
I l In-House Analysis and Assessment as Needed l
Issue 5. Criteria and Analysis for BWR Power Oscillations (ATWS) in-House Assessment l FY 98 l FY 99 l FY 00 l FY01 l F 02 l
l FRAPTRAN Analysis l l To Be Determined l
lasue 6. Fuel Rod and Neutronic Computer Codes for Analysis PNNL Code improvement and Assessment lFY98 l FY 99 l FY 00 l FY01 l FY 02 l
l FRAPCoN j l FRAPCON Mechanical Models
-l l FRAPTRAN improvement l l FRAPTRAN Assessmant l
[ Halden Test Reactor Data Assessment and Utilization l
I Issue 7. Source Term and Core Melt Progression To Be Determined E
.~
y 3
Issue 8. Transportation and Dry Storage I
h-House Assessment l W98 l W 99 l W OO l W O1 l W C2
.l Ioennevan l
1 l
)
l issue 9. High Enrichments (>5%)
h-House Assessment l W 98 l FY 99 i W OO l W 01 l FYO2 l
\\
IDennev**
I J
l b__
I
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