ML041960480
| ML041960480 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/01/2004 |
| From: | Curran D Blue Ridge Environmental Defense League, Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP |
| To: | Atomic Safety and Licensing Board Panel |
| Byrdsong A T | |
| References | |
| 50-413-OLA, 50-414-OLA, ASLBP 03-815-03-OLA, RAS 8089 | |
| Download: ML041960480 (197) | |
Text
- MATED CORRESPONDN July l, 2004 UNITED STATES OF AMERICA USNRC NUCLEAR REGULATORY COMMISSION July 6, 2004 (1
- 30PM)
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD OFFICE OF SECRETARY RULEMAKINGS AND In the Matter of: ) ADJUDICATIONS STAFF DUKE ENERGY CORPORATION
) Docket Nos. 50-413-OLA 50-414-OLA (Cata wba Nuclear Station, )
Units 1 and 2) )
PREFILED W'RITTEN TESTIMONY OF DR. EDWVIN S. LYMAN REGARDING CONTENTION I On behalf of Blue Ridge Environmental Defense League ("BREDL"), Dr. Edwin S.
Lyman hereby submits the following testimony regarding BREDL's Contention I.
Q.1. Please state your name and describe your professional qualifications to give this testimony.
A.I. My name is Dr. Edwin S. Lyman. I am a Senior Scientist with the Global Security Program at the Union of Concerned Scientists, 1707 H Street, NW, Suite 600, Washington, D.C.
20006. My education and experience are described in my curriculum vita, which is attached to my testimony as Exhibit A.
I am a qualified expert on nuclear safety and safeguards issues. I hold a Ph.D., a master of science degree, and a bachelor's degree in physics. For over eleven years, I have conducted research on security and environmental issues associated with the management of nuclear materials and the operation of nuclear power plants. My research has included the safety and security implications of using mixed oxide fuel as a substitute for uranium fuel in nuclear power plants. I have also published articles on this topic. A list of my publications is included in my C.V.
Q.2. What is the purpose of your testimony?
A.2. The purpose of my testimony is to discuss my views on BREDL Contention I, which was admitted for litigation by the Atomic Safety and Licensing Board ("ASLB") in LBP-04-04, Memorandum and Order (Ruling on Standing and Contentions) (March 5, 2004). BREDL Contention I asserts that Duke Energy Corporation's ("Duke's") license amendment request j-Onp Ic4te ':-- Y- 0 65 C V-1J
("LAR") to test plutonium mixed oxide ("MOX") fuel at the Catawba nuclear power plant is inadequate because Duke has failed to account for the differences between MOX and low enriched uranium ("LEU") fuel behavior; nor has Duke accounted for the impact of these differences on Duke's analysis of loss of coolant accidents ("LOCAs").
Q.3. What materials have you reviewed in preparation for your testimony?
A.3. I have reviewed Duke's LAR and related correspondence, including Duke's responses to Requests for Additional Information ("RAIs") by the NRC Staff. I have also reviewed the body of literature which has been developed regarding the behavior of MOX and other types of reactor fuel under LOCA conditions. I am also familiar with relevant NRC documents, including correspondence regarding this license amendment application, reports and correspondence concerning characteristics and behavior of MOX fuel, and correspondence and reports concerning the behavior of LEU fuel under LOCA conditions. In addition, I am familiar with regulations and guidance of the U.S. Nuclear Regulatory Commission ("NRC") and the U.S.
Department of Energy ("DOE") governing plutonium processing facilities. Finally, I am familiar with U.S. and foreign government reports regarding testing of LEU fuel under accident conditions.
Q.4. Please summarize the conclusions you have reached regarding the adequacy of Duke's LAR application to account for the differences between MOX and LEU fuel.
A.4. In my professional judgment, Duke's design-basis loss of coolant ("DB-LOCA") analysis is inadequate because it does not address the uncertainties associated with relocation effects that M5-clad MOX fuel may experience under LOCA conditions. These uncertainties relate to Duke's assertion that the action proposed in the MOX LTA LAR will not result in a violation of the emergency core cooling system (ECCS) acceptance criteria in 10 C.F.R. § 50.46: peak cladding temperature ("PCT"), maximum cladding oxidation, and the preservation of a coolable core geometry.
The phenomenon of fuel relocation has been observed in experiments with irradiated LEU fuel under LOCA conditions. While to my knowledge no similar experiments have been done on MOX fuel, there are technical reasons to believe that the impact of fuel relocation effects during a LOCA may be more severe for MOX fuel rods than for LEU fuel rods of the same burnup, due to differences in characteristics such as fuel fragment sizes and fuel-clad interactions. Moreover, calculations in Duke's LAR indicate that MOX fuel is generally more limiting than LEU fuel with respect to DB-LOCAs. Therefore, the consequences of fuel relocation, and the non-conservatism associated with neglecting them, may be of greater concern for MOX fuel rods than for LEU fuel rods with respect to compliance with LOCA regulatory criteria.
Duke has failed to address these uncertainties in MOX fuel behavior, and therefore its LTA application is unacceptable to satisfy the requirements of 10 C.F.R. § 50.46 with respect to PCT, maximum cladding oxidation, and coolable geometry of fuel. In addition, by failing to address the uncertainties in MOX fuel behavior, Duke has not demonstrated compliance with the general reasonable assurance standard in 10 C.F.R. § 50.40(a).
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I do not believe, however, that these uncertainties can be addressed with mere calculations or analyses based on LEU performance. In my professional opinion, the only satisfactory way to address these uncertainties would be to conduct integral tests of MOX fuel assemblies under LOCA conditions in such a manner that the impacts of the phenomena I have previously described can be measured with reasonable accuracy and precision.
Q 5: Please explain how the regulations in 10 C.F.R. § 50.46 apply to Contention I.
A.5: NRC regulations at 10 C.F.R. § 50.46 establish acceptance criteria for emergency core cooling systems for light-water nuclear reactors. Essentially, the regulation sets design limits for behavior of the reactor fuel under LOCA conditions. Appendix K to Part 50, whose requirements are referenced in 10 C.F.R. § 50.46(a)(1), sets forth ECCS "evaluation models," i.e., assumptions about the behavior of reactor fuel that are to be used in determining whether it complies with the criteria in 10 C.F.R. § 50.46.
10 C.F.R. § 50.46 and Appendix K apply only to uranium-based fuel, but Duke has requested an exemption from this limitation so that these requirements will apply to MOX fuel. I believe that it is generally appropriate to apply the requirements of 10 C.F.R. § 50.46 to MOX fuel, as long as Appendix K is not strictly applied to exclude consideration of relocation of the fuel during LOCAs.
The regulations in 10 C.F.R. § 50.46 sets forth fuel performance limits in three categories that have importance with respect to performance of MOX fuel: peak cladding temperature ("PCT"),
maximum cladding oxidation, and coolable geometry. Section 50.46(b)(1) requires that that PCT "shall not exceed 22000 F." Section 50.46(b)(2) provides that the "calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation." Section 50.46(b)(4) also requires that "[c]alculated changes in core geometry shall be such that the core remains amenable to cooling."
Q.6: Please explain why you think the Appendix K evaluation models for the MOX LTA core should include consideration of fuel relocation during LOCAs.
A.6: Appendix K does not include consideration of fuel relocation. The NRC did contemplate including fuel relocation as a criterion in Appendix K, but claimed to have resolved the question in Generic Issue 92. Memorandum from Ralph Meyer, NRC Office of Nuclear Regulatory Research, to John Flack, NRC Regulatory Effectiveness and Human Factors Branch, re: Update on Generic Issue 92, Fuel Crumbling During LOCA (February 8,2001) (NRC ACN #
ML010390163) (hereinafter "Meyer Memorandum"). A copy is attached to my testimony as Exhibit B. See also Memorandum from Ashok C. Thadani, Office of Nuclear Regulatory Research, to Samuel J. Collins, Office of Nuclear Reactor Regulation, re: Information Letter 0202, Revision of 10 CFR 50.46 and Appendix K (June 20, 2002) (hereinafter "Thadani Memorandum") (NRC ACN # ML021720690). A copy is attached to my testimony as Exhibit C. More recently, the NRC has acknowledged that omission of fuel relocation effects is a non-conservatism in Appendix K with a very large potential impact on PCT, and that an early "resolution" of this issue (i.e., Generic Issue 92) may have been in error or is no longer 3
applicable because of new information. See; Meyer Memorandum and Thadani Memorandum, Attachment 4 at 4-5.
As I will discuss in more detail later in my testimony, certain characteristics of MOX fuel appear to exacerbate the effects of fuel relocation, thus leading to higher PCTs and greater maximum cladding oxidation. While there are several other known non-conservatisms in Appendix K, this one in particular appears to be relevant to the MOX LTA LAR because of its disproportionately large impact on the MOX LTAs compared to the LEU assemblies that comprise the remainder of the core. Given the potential impact on PCT of relocation effects, it is not appropriate to omit consideration of this phenomenon from the Appendix K models that Duke uses to establish that loading of the MOX LTAs into Catawba will result in compliance with 10 CFR § 50.46 criteria.
Q.7: Please explain what you mean by fuel relocation during a LOCA.
A.7: According to NRC, "fuel relocation refers to the movement of fuel pellet fragments into regions of the fuel rod where the cladding has ballooned during a LOCA transient." Thadani Memorandum, Attachment 4 at 4-5.
Q.8: Please explain why fuel relocation could increase the severity of a LOCA.
A.8: Fuel relocation increases the local linear heat generation rate within the ballooned area. Thus it could increase the severity of a LOCA by resulting in a greater fuel rod peak cladding temperature (PCT) than in a situation in which fuel relocation did not occur. Because transient oxidation during a LOCA increases with an increase in PCT, fuel relocation could also result in a greater maximum cladding oxidation. Finally, the greater local linear heat generation rate requires a greater coolant flow around the ballooned area to ensure long-term core coolability. See slides presented by A.
Mailliat and J.C. M6lis, IRSN, at "PHEBUS STLOC Meeting" with NRC Staff (October 23, 2003) (NRC ACN # ML032970624) (hereinafter October 2003 IRSN Presentation"). A copy is attached to my testimony as Exhibit D.
Q.9: Please discuss the potential magnitude of the impact of fuel relocation on PCT and maximum cladding oxidation for uranium oxide (UO2) fuel.
A.9: The most recent calculations of the impact of fuel relocation on PCT of which I am aware were conducted by the Institut de Protection et de Sfiret6 Nucleaire (LPSN, now IRSN) and published in 2001. In that study, the authors used the CATHARE2 computer code to calculate the impact of fuel relocation on the large-break LOCA PCT for a high-bumup U0 2 fuel rod as a function of the "filling ratio," or the ratio of the volume of the relocated fuel material to the volume of the ballooned region. For the scenario evaluated, the authors found that the PCT in the absence of relocation effects was 970'C. For a filling ratio of 70%, the maximum considered, the PCT was 1144TC. For a filling ratio of 40%, the PCT was about 20'C greater than for the no-relocation case. Thus the maximum impact on PCT of relocation in this study was a APCT of +
174 0 C (313'F) for high-burnup U0 2 fuel. It is not clear from the study whether higher filling ratios, and hence larger impacts on PCT, are possible. C. Grandjean, G. Hache and C. Rongier, "High Burnup U0 2 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation After Burst," OECD/NEA Proceedings of the Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-4
Provence (March 22-23, 2001) (hereinafter "Grandjean, Hache, and Rongiere). A copy of this paper is attached to my testimony as Exhibit E. The NRC staff appears to be familiar with this result. See Thadani Memorandum, Attachment 5 at 4.
The study also evaluated the impact on the maximum cladding oxidation for the ruptured region (two-sided oxidation). The equivalent cladding reacted (ECR) calculated by the Cathcart-Pawel rate law (a surrogate for "maximum cladding oxidation") was 12.6% for the no-relocation case, and 19.7% for the 70% filling ratio case. Thus the maximum impact on ECR resulting from relocation was calculated as AECR = 7.1%.
Q.10: Please explain why the impact of fuel relocation on the severity of a LOCA could be greater for MOX fuel than for U0 2 fuel at the same burnup.
A. 10: Experts have concluded that MOX fuel may experience more severe relocation effects than U0 2 fuel at the same burnup. The IPSN study above did not explicitly consider MOX fuel, but stated that "it must be pointed out that that results of corresponding calculations with ... high burnup MOX fuels would have been more severe with regard to acceptance limits." Grandjean, Hache and Rongier at 7.
IRSN, the successor to IPSN, has reiterated these concerns, stating in a recent presentation that for MOX fuel, a "higher initial energy" and an "enhance [sic] of fuel relocation impact" results in greater increases in PCT and ECR associated with relocation. V Guillard, C. Grandjean, S.
Bourdon and P. Chatelard, "Use of CATHARE2 Reactor Calculations to Anticipate Research Needs," SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory, slides at 8-9 (May 25-26, 2004) (NRC ACN # ML041600261). A copy of this paper is attached to my testimony as Exhibit F. In the abstract for this presentation, the authors state that "a lack of knowledge on theses [sic] parameters [important for relocation] for irradiated U02 and particularlyMOXfuel [emphasis added] may lead to reduce [sic] safety margins."
MOX fuel may experience more severe relocation effects than U0 2 fuel at the same burnup because several characteristics that are important for relocation may be less favorable for MOX fuel. These include pellet fragment size and fuel-clad interaction.
Q.11: Please explain the basis for your concern regarding the pellet fragment size of MOX fuel and its impact on fuel relocation in a LOCA.
A. 11: The IPSN calculations cited above demonstrate the high sensitivity of fuel relocation-induced increases in PCT and ECR to the filling ratio. The filling ratio, in turn, is a function of the average particle size of the relocated fuel fragments, in that smaller particles will in general result in greater packing of the relocated area and hence higher filling ratios.
The fuel relocation phenomenon has been observed in LEU fuel for rod burnups exceeding around 48 GWD/t. See Grandjean, Hache and Rondier at 2 (2001). This suggests that vulnerability to fuel relocation is associated with the development of the high-burnup "rim" region known to emerge in LEU fuel for burnups exceeding about 40-45 GWD/t. IPSN states 5
that "fuel fragmentation is clearly associated to [sic] burnup, with finer fragments at higher BU."
See Grandjean, Hache and Rondier at 2 (2001).
For During manufacture of MOX fuel using the MIMAS process (which will be used for the Duke LTAs), plutonium agglomerates --- macroscopic clumps of plutonium-rich particles ---
occur in the fuel. Because the fissile material is concentrated in these clumps, very high local burnups result, due to the fact that the fission is occurring in a heterogeneous fashion. The ratio of local bumup within the agglomerates is on the order of 4-6 times the rod-average burnup, depending on the irradiation time. For instance, the agglomerate bum-up reaches about 60 GWD/t when the rod average is only around 18 GWD/t, and reaches 100 GWD/t when the rod average is only 28.4 GWD/t As a result, high-bumup rim-like regions emerge in the outer layers of the plutonium agglomerates for much lower rod-average burnups than 40-45 GWD/t, because the local burnups within the plutonium agglomerates increase much more rapidly than the rod-average burnups. Thus it is reasonable to expect that the onset of fuel relocation in MOX fuel may occur at lower rod-average burnups than in LEU fuel. This would imply that MOX fuel will be vulnerable earlier in its irradiation history (and consequently for a longer time) than LEU fuel.
Also, the particle size distribution in MOX fuel will be smaller than in LEU fuel at the same rod-average bumup, to the extent that fine fragments are generated in the ultra-high bumup plutonium agglomerate regions.
Fuel fragmentation can also be caused by the stress induced by the stored-energy redistribution during the blowdown phase of a LOCA. A. Mailliat and M. Schwarz, "Need for Experimental Programmes on LOCA Issues Using High Bum-Up and MOX Fuels," NUREG/CP-0176, Proceedings of the Nuclear Safety Research Conference at 436 (May 2002) (NRC ACN #
ML021710793) (hereinafter "Mailliat and Schwarz"). A copy of this paper is attached to my testimony as Exhibit G. Because MOX fuel has a lower thermal conductivity and a higher radial temperature gradient than LEU fuel, it could experience greater fuel fragmentation during the blowdown and more severe relocation effects as a result.
According to two out of four NRC experts who participated in the 2001 PIRT panel on LOCAs and high-bumup fuel, the composition of fuel (i.e. a specified MOX composition) is of "high importance" for consideration of fuel relocation effects because it "may affect the amount of fine grain material after relocation. Fuel structure and mechanical properties are influenced by fuel type." See NUREG/CR-6744, "Phenomenon Identification and Ranking Tables for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High-Burnup Fuel,"
Appendix D, Table D-1 at D-67 (December 2001) (NRC ACN # 013540623) (hereinafter "NUREG/CR-6477"). Relevent portions of this report are attached to my testimony as Exhibit H. One expert concluded that fuel composition was of moderate importance to relocation, stating that "the consequence of fuel fragments relocation (higher local decay heat and higher cladding temperature) could be more effective with MOX fuel than with U02 fuel" but that "the viscoelastic properties of the MOX should impair the fuel fragments relocation at high burnup."
Id. at D-67. A fourth expert concluded that fuel composition would be of only low importance to relocation. Id. at D-67. This difference of expert opinion highlights the inadequacies of the experimental database with regard to integral tests of MOX fuel under design-basis LOCA conditions, and underscores the significant uncertainties in Duke's design-basis LOCA analysis.
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Q.12: Please explain the basis for your concern regarding the effects of fuel-clad interaction within the MOX LTAs and their impact on fuel relocation in a LOCA.
A.12: I am concerned about differences between MOX and LEU fuel with respect to fuel-clad bonding and the impact of such differences on fuel relocation behavior during a design-basis LOCA. According to IPSN (now IRSN), tight fuel-clad bonding may delay the onset of fuel relocation. Mailliat and Schwarz at 433. Tight bonding has also been observed at the Halden reactor in Norway to retard the rate of balloon formation. Nuclear Energy Agency, NEA/CSNI/R(2003)9, Ongoing and PlannedFuel Safety Research in NEA Member States at 79 (March 5, 2003). Relevant excerpts of this report are attached to my testimony as Exhibit l. During NRC's recent expert elicitation (PIRT) process on LOCA issues for high-burnup fuel, all four participating experts agreed that "chemical and mechanical bonding between the fuel pellet and the cladding ... " was of high importance to the fuel relocation phenomenon, because "bonding could significantly affect the relocation characteristics by impeding pellet fragment movement."
NUREG/CR-6744, Table D-l at D-69. It has been confirmed that MOX fuel is more resistant to clad failures due to pellet-clad mechanical interaction (PCMI) than LEU fuel, even at high burnups. Nuclear Energy Agency, NEA/NSC/DOC(2004)8, International Seminaron Pellet-CladInteractionswith Water Reactor Fuels, at 20 (May 6, 2004).
Relevant excerpts of this report are attached to my testimony as Exhibit J. This phenomenon is not well-understood but may imply that the pellet-clad bond is weaker for MOX fuel, in which case MOX fuel may have a greater propensity to earlier and more extensive fuel relocation than LEU.
In Duke's April 14, 2004, Response to BREDL's first set of discovery requests, Duke stated that the Framatome design-basis LOCA analysis for the MOX LTAs did not assume any fuel-clad bonding and was therefore "conservative" with respect to the requirement that the degree of cladding swelling not be underestimated. Id. at 14. However, in the absence of an assessment of whether and to what extent the pellet-clad interaction is weaker in MOX fuel than in LEU fuel, there is no way of knowing the degree to which this assumption is conservative for MOX fuel. Therefore, Duke's failure to properly account for this phenomenon contributes another uncertainty to the safety margin associated with Duke's design basis LOCA calculation.
Moreover, there is evidence to contradict Duke's assertion that "deterministic LOCA evaluations typically based on data taken from unirradiated cladding" are conservative with respect to clad swelling. According to IPSN (now IRSN), results from the PBF-LOC experiments found that irradiated rods experienced greater clad deformation than unirradiated rods during design-basis LOCA conditions. See Mailliat and Schwartz at 432. There is simply no way to determine whether Duke's design-basis LOCA analysis underestimates or overestimates the degree of clad swelling (and hence the degree of fuel relocation) for MOX LTAs without additional experimental data from integral LOCA tests of high-burnup MOX fuel rods. Given the lack of data, BREDL finds unpersuasive the NRC's 1999 speculation, quoted by Duke in its April 14, 2004 set of responses to BREDL's discovery requests, that "a major effect is not expected" with regard to differences in pellet-clad bonding between MOX and LEU. Id. at 15.
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Q.13: Please explain the basis for your concern regarding clad balloon size and its impact on the severity of fuel relocation affecting the MOX LTAs in a LOCA.
A.13: The MOX LTAs will use M5 cladding, as compared to the Zircaloy-4 or ZIRLO cladding that is extensively used in US PWRs. According to IRSN, M5 will form larger balloons than Zircaloy-4 in a design-basis LOCA because it remains more ductile during irradiation. October 2003 IRSN presentation to NRC at 24. The greater retained ductility of M5 as a function of burnup compared to Zircaloy-4 can result in a greater M5 balloon size during a design-basis LOCA for fuel rods of the same burnup. Larger balloons increase the space available for fuel fragments to fall and hence result in a greater propensity for fuel relocation during a LOCA, with an associated increase in PCT and local clad oxidation.
Q.14: A group of experts from Electricite de France (EDIF), Framatome ANP and the French CEA recently challenged IRSN's assertion that M5 cladding would form bigger balloons during a LOCA than zircaloy-4 in a presentation at Argonne National Laboratory. Please explain your view of this position.
A. 14: I do not believe the EDF presentation responds adequately to the issue that IRSN has raised. Their claim is that the Edgar creep tests --- which indicated a greater ductility and a larger balloon size for M5 than for zircaloy-4 --- are not the appropriate tests to actually evaluate balloon size during LOCAs. Ramp tests utilizing pre-hydrided cladding samples, which EDF asserts are more representative of LOCA conditions, indicate that the balloon size for M5 is not actually greater than for zircaloy-4.
Obviously, a ramp test would be more similar to the conditions experienced during a LOCA than a steady-state creep test. However, neither creep tests nor ramp tests utilizing pre-hydrided but unirradiated cladding materials adequately simulate all the relevant phenomena that could affect balloon formation during a LOCA involving high-burnup fuel. For example, a well-known property of M5 cladding is that it generates a thinner oxidation layer during normal irradiation as a function of burnup than zircaloy-4. Zircaloy-4 at high burnups tends to generate a thick oxidation layer that's prone to spalling. Spalling wvill cause spatial inhomogeneities in the clad temperature that negatively affect ductility, leading to earlier cladding ruptures during a LOCA and hence smaller balloon sizes. I don't think that the ramp tests described by EDF take that effect into account. Therefore, I don't believe that the EDF presentation fully addresses the differences that would be observed in actual irradiated fuel with regard to the ductility and the balloon size of M5 compared to that of zircaloy-4.
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This question remains unresolved because there is an absence of experimental data on the performance of high-burnup, M5-clad fuel, under design-basis LOCA conditions. The Electric Power Research Institute (EPRI) and Areva (parent company of Framatome ANP) apparently continue to deny NRC access to samples of irradiated high-burnup M5-clad LEU fuel for testing at Argonne National Laboratory. Letter from Ashok C. Thadani, NRC, to David Modeen, EPRI (April 21, 2004) (ADAMS ACN # ML041130490). A copy of this letter is attached to my testimony as Exhibit K. This lack of cooperation can only cause further delays in the ability of NRC to obtain the experimental data it needs to confirm the safety of high-burnup M5-clad fuel (whether LEU or MOX).
I would underscore the admission of M. Blanpain of AREVA during the ACRS Reactor Fuels Subcommittee Meeting on April 21, 2004 that MOX fuel irradiated in France is predominantly clad in Zircaloy-4, and only "some M5 fuel rods with MOX for experimental purposes" have been used in France. See Transcript at 61-62. For some reason, France is reluctant to use M5-clad MOX fuel domestically and is primarily producing it for export to Germany (and now to the United States). However, even in Germany the use of M5-clad MOX has been extremely limited. And I am unaware of any integral LOCA tests performed with irradiated M5-clad MOX fuel.
Q.15: Please explain the basis for your concern regarding the impact of fuel relocation on the ability of the MOX LTA core to satisfy the regulatory requirement for coolable core geometry.
A. 15: As stated above, fuel relocation increases the local linear heat generation rate The maximum flow blockage that will preserve a coolable geometry depends on the assumed heat source and the heat transfer properties of the fuel bundle. As IRSN points out, acceptable bundle blockage ratios were derived based upon arrays of unirradiated fuel rods, and did not take into account fuel relocation and its associated impacts on the redistribution of the decay heat source within the fuel rods. IRSN presentation to NRC at 29 (October 23, 2003). IRSN restated its concern in a recent presentation:
"The impact of fuel relocation in fuel rod balloons, as was observed in all in-reactor tests with irradiated fuel, leading to an increase in local power (lineic and surfacic) ..., on the coolability of the blocked region, is still fully questionable and should be addressed by specific analytical tests with a simulation of fuel relocation."
C. Grandjean and G. Hache, "LOCA Issues Related to Ballooning, Fuel Relocation, Flow Blockage and Coolability," SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory at 23 (May 25-27, 2004) (emphasis in original). A copy of this paper is attached to my testimony as Exhibit L.
Thus, any analysis that does not take this into account is incomplete and is likely to be non-conservative. Lack of consideration of this phenomenon will be of greater concern for the MOX LTA core to the extent that the MOX LTAs have a smaller margin to regulatory limits than LEU fuel.
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Q. 16: Please explain the basis for your concern regarding the smaller safety margins for MOX fuel with respect to peak clad temperature in a LOCA.
A.16: As Duke's calculations have demonstrated, the PCT in a design-basis LOCA is higher for a MOX rod than for an LEU rod in the same position in the core. Duke MOX LTA LAR at 3-43 (February 27, 2003). The margin to the 10 CFR §50.46 PCT limit of 22000 F is therefore smaller for a MOX rod than for an LEU rod in the same position.
At high burnups, the linear heat generation rate for MOX fuel is generally higher than that for LEU fuel. This, in turn, results in increased centerline temperature and stored energy, therefore reducing the margin to design-basis LOCA regulatory limits. BREDL maintains that every reduction in margin associated with MOX fuel use, coupled with the non-conservatism of ignoring fuel relocation effects, reduces confidence in Duke's design-basis LOCA analysis of the MOX LTA core.
Because there is little or no experimental data to conclusively validate the impact of relocation on either LEU or MOX fuel, a design-basis MOX LTA LOCA analysis that takes relocation into account would be highly uncertain --- with a resulting large uncertainty in the calculation of the relocation-associated increase in PCT of a MOX LTA fuel rod compared to the relocation-associated increase in PCT of an LEU fuel rod. For instance, if the MOX filling ratio is 70% and the LEU filling ratio is only 40%, because of a greater quantity of fine fragments in the MOX fuel, the increase in PCT could be nearly three hundred degrees Fahrenheit greater for MOX than for LEU (assuming that no other MOX-related effect, such as a greater initial linear heat generation rate, results in an even more severe increase in PCT associated with relocation).
The PCT calculated by Duke for the MOX LTA is 201 8'F. Obviously, a relocation-associated increase in PCT of, say 31 3'F (associated with a 70% filling ratio for LEU fuel), would result in an exceedance of the 2200'F limit by 131 0F. On the other hand, if the LEU filling fraction is closer to 40%, the increase in PCT would only be about 40'F, and the LEU fuel would still be in compliance with the regulatory limit. Thus the MOX LTAs could well be limiting with respect to LOCA compliance if relocation is fully accounted for.
These significant uncertainties should be reflected in Duke's analysis, and NRC approval should be contingent upon a demonstration that uncertainties of this magnitude do not undermine reasonable assurance of adequate protection of the public health and safety. I do not believe that such a finding can be made, given the potential severity of the relocation phenomenon and its associated uncertainties.
Q.17: Please discuss how, in your opinion, the gaps in the experimental database for the behavior of high-burnup, MI5-clad MOX fuel during LOCAs can be reduced.
A. 17: The only way to fully address the uncertainties associated with the behavior of high-burnup, MS-clad MOX fuel during LOCAs is to conduct integral LOCA tests of such fuel, fabricated with the same specifications as the lead test assemblies that are under consideration here, and irradiated to a range of burnups, including the maximum of 60 GWD/t that Duke has requested 10
in its LAR. The proposed Phebus test series would likely make a substantial contribution to reducing the level of uncertainty associated with MOX fuel behavior during LOCAs.
These integral tests could be supplemented with separate-effects tests specifically designed to look at fuel relocation as a function of burnup for both MOX and LEU fuel, and to measure the relative susceptibility to relocation of the two types of fuels. The Halden IFA-650 test, which I understand is being designed to examine fuel relocation effects in LEU fuel, could help to resolve some of these questions. But similar tests on mixed oxide fuel will also be needed. And separate effects tests cannot reproduce the complex, interrelated set of thermal-hydraulic and mechanical phenomena that would occur during a LOCA and would affect fuel relocation.
Q.18: Does the Staffs Safety Evaluation Report (SER) provides any insight into the issues raised by Contention I?
A.18: The SER doesn't address the issues that we've raised concerning the impact of relocation.
So to that extent, it doesn't affect my conclusions at all. Members of the Staff admitted during the ACRS subcommittee meeting on the LTA LAR application that they have not done their own independent calculations to confirm Duke's LOCA analyses. The Staff has only checked Duke's results for internal consistency, rather than doing any of its own simulations. Therefore, to the extent that the Staff claims to have independently verified the adequacy of Duke's LOCA analysis, I do not believe that claim is correct.
Q.19: Does this conclude your testimony?
A.19. Yes.
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CERTIFICATE OF SERVICE I hereby certify that on July 1, 2004, copies of the Written Prefiled Testimony of Dr. Edwin S.
Lyman Regarding Contention I were served on the following by e-mail, as indicated below. The exhibits to Dr. Lyman's testimony, plus a hard copy of his testimony, were served by hand or by Federal Express on the following day, as indicated below:
Ann Marshall Young, Chair Susan L. Uttal, Esq.
Administrative Judge Antonio Fernandez, Esq.
Atomic Safety and Licensing Board Margaret J. Bupp, Esq.
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LIST OF EXHIBITS TO PREFILED WRITTEN TESTIMONY OF DR. EDWIN S. LYMAN REGARDING CONTENTION I EXHIBIT A: Curriculum Vita for Dr. Edwin S. Lyman EXHIBIT B: Memorandum from Ralph Meyer, NRC Office of Nuclear Regulatory Research, to John Flack, NRC Regulatory Effectiveness and Human Factors Branch, re:
Update on Generic Issue 92, Fuel Crumbling During LOCA (February 8, 2001) (NRC ACN# ML010390163)
EXHIBIT C: Memorandum from Ashok C. Thadani, NRC Office of Nuclear Regulatory Research, to Samuel J. Collins, NRC Office of Nuclear Reactor Regulation, re: Research Information Letter 0202, Revision of 10 CFR § 50.46 and Appendix K (June 20,2002) (NRC ACN # ML021720690).
EXHIBIT D: slides presented by A. Mailliat and J.C. M6lis, IRSN, at "PHEBUS STLOC Meeting" with NRC Staff (October 23, 2003) (NRC ACN # ML032970624)
EXHIBIT E: C. Grandjean, G. Hache and C. Rongier, "High Burnup U0 2 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation After Burst," OECD/NEA Proceedings of the Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-Provence (March 22-23, 2001).
EXHIBIT F: V. Guillard, C. Grandjean, S. Bourdon and P. Chatelard, "Use of CATHARE2 Reactor Calculations to Anticipate Research Needs," SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory (May 25-26, 2004) (NRC ACN
EXHIBIT G: A. Mailliat and M. Schwarz, "Need for Experimental Programmes on LOCA Issues Using High Bum-Up and MOX Fuels," NUREG/CP-0 176, Proceedings of the Nuclear Safety Research Conference (May 2002) (NRC ACN # ML021710793)
EXHIBIT H: NUREG/CR-6744, "Phenomenon Identification and Ranking Tables for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High-Burnup Fuel," Appendix D, Table D-1 at D-67 (December 2001) (NRC ACN #
013540623)
EXHIBIT I: Excerpts from NEA/CSN/R(2003)9, Ongoing and Planned Fuel Safety Research in NEA Member States, Compiled from SEGFS Members' Contributions (October 2002)
I.
EXHIBIT J: Excerpts, NEA/NSC/DOC (2004)8, International Seminar on Pellet-Clad Interaction in Water Reactor Fuels, organized by CEA Cadarache/DEN/DEC, in cooperation with OECD/NEA, IAEA, EDF, FRAMATOME, ANP, COGEMA (March 9-11, 2004)
EXHIBIT K: Letter from Ashok C. Thadani, NRC, to David Modeen, EPRI (April 21, 2004) (ADAMS ACN # ML041130490)
EXHIBIT L: C. Grandjean, Georges Hache, LOCA Issues Related to Ballooning, Fuel Relocation, Blockage, and Coolability: Main Findings from a Review of Past Experimental Programs, presented at SEGFSM Topic Meeting on Local Issues (May 25,-
27, 2004) (NRC ACN # ML041600354) ii
z %
EXHIBIT A
Edwin Stuart Lyman Curriculum Vitao Education Ph.D, Cornell University, Theoretical Physics, August 1992.
M.S., Cornell University, Physics, January 1990.
A.B., suitna ctun laude, New York University, Physics, June 1986; Phi Beta Kappa.
Professional Experience May 1.2003 - Present Senior Staff Scientist, Union of Concerned Scientists June 2002- April 2003: President, Nuclear Control Institute, Washington, D.C.
July 1995-May 2002: Scientific Director, Nuclear Control Institute, Washington, D.C.
August 1992-June 1995: Postdoctoral research associate, Center for Energy and Environmental Studies, Princeton University, Princeton, NJ.
Spring 1995: Preceptor for Environmental Studies 302, "Perspectives on Environmental Issues:
Values and Policies."
Spring 1994: Lecturer, Woodrow Wilson School. Preceptor for WWS 304, "Science, Technology and Public Policy."
July 1988-June 1992: Graduate research assistant, Newman Laboratory of Nuclear Studies, Cornell University, Ithaca, NY. Conducted thesis research on high-energy physics under the supervision of Prof. S.H.-H. Tye.
August 1986- June 1988: Andrew D. White Graduate Fellow, Physics, Cornell University.
Publications J. Beyea, E. Lyman and F. von Hippel, "Damages from a Major Release of 137 CS Into the Atmosphere of the United States," Science and Global Security 12 (2004) 125-136.
G. Bunn, C. Braun, A. Glaser, E. Lyman and F. Steinhausler, "Research Reactor Vulnerability to Sabotage by Terrorists," Science and GlobalSecurity 11 (2003)85-107.
R. Alvarez, J. Beyea, K. Janberg, J. Kang, E. Lyman, A. Macfarlane, G. Thompson and F.
von Hippel, "Reducing the Hazards from Stored Spent Power-Reactor Fuel in the United States,"
Science and Global Security 11(2003) 1-51.
E. Lyman, "Revisiting Nuclear Power Plant Safety" (letter), Science 299 (2003), 202.
E. Lyman, "The Limits of Technical Fixes," in NuclearPower and The Spread ofNuclear Weapons: Can We Have One W~ithout the Other?" (P. Leventhal, S. Tanzer and S. Dolley, eds.),
Brasseys, Washington, DC, 2002, 167-182.
E. Lyman, "The Pebble-Bed Modular Reactor: Safety Issues," Physics and Society, American Physical Society, October 2001.
E. Lyman, "Public Health Risks of Substituting Mixed-Oxide for Uranium Fuel in Pressurized Water Reactors," Science and GlobalSecurity 9 (2001), 1.
E. Lyman and S. Dolley, "Accident Prone," Bulletin of the Atomic Scientists, March/April 2000,42.
E. Lyman and H. Feiveson, "The Proliferation Risks of Plutonium Mines, " Science and Global Security 7 (1998), 119.
E. Lyman and P. Leventhal, "Bury the Stuff [Weapons Plutonium]," Bulletin of the Atomic Scientists, March/April 1997, 45.
E. Lyman, "Weapons Plutonium: Just Can It," Bulletin of the Atomic Scientists, November/December 1996, 48.
F. von Hippel and E. Lyman, "Appendix: Probabilities of Different Yields," addendum to J.
Mark, "Explosive Properties of Reactor-Grade Plutonium," Science and Global Security 4 (1993),
125.
F. Berkhout, A. Diakov, H. Feiveson, H. Hunt, E. Lyman, M. Miller, and F. von Hippel, "Disposition of Separated Plutonium," Science and GlobalSecurity 3 (1993), 161.
E. Lyman, F. Berkhout and H. Feiveson, "Disposing of Weapons-Grade Plutonium,"
Science 261 (1993) 813.
P. Argyres, E. Lyman and S.H.-H. Tye, "Low-Lying States of the Six-Dimensional Fractional Superstring," Phys. Rev. D46 (1992) 4533.
S.-w. Chung, E. Lyman and S.H.-H. Tye, "Fractional Supersymmetry and Minimal Coset Models in Conformal Field Theory," Int. J. Mod. Phys A7 (1992) 3337.
R. Alvarez, J. Beyea, K. Janberg, J. Kang, E. Lyman, A. Macfarlane, G. Thompson and F.
von Hippel, "Reducing the Hazards from Stored Spent Power-Reactor Fuel in the United States,"
Science and GlobalSecurity 11 # 1 (2003) pp. 1-60.
D. Hirsch, D. Lochbaum and E. Lyman, Bulletin of the Atomic Scientists (May/June 2003).
2
Articles to be Published, Submitted for Publication or In Preparation T. Taylor, E. Lyman, S. Erickson and J. Regester, "Criticality Weapons: A Fifth Class of WMD," in preparation.
Selected Reports E. Lyman, "Safety Issues in the Sea Shipment of Vitrified High-Level Radioactive Wastes to Japan," report sponsored by the Nuclear Control Institute, Greenpeace International and Citizens' Nuclear Information Center Tokyo, December 1994.
E. Lyman, "Interim Storage Matrices for Excess Plutonium: Approaching the 'Spent Fuel Standard' Without the Use of Reactors," PU/CEES Report No. 286, Center for Energy and Environmental Studies, Princeton University, August 1994.
E. Lyman, "The Solubility of Plutonium in Glass," PU/CEES Report No. 275, Center for Energy and Environmental Studies, Princeton University, April 1993.
Selected Invited Talks "U.S. Nonproliferation Policy, Plutonium Disposition and the Threat of Nuclear Terrorism,"
seminar on "Recycling Plutonium: Risks and Alternatives," sponsored by the Green Group, European Parliament, Brussels, Belgium, January 9, 2003.
"Current Status of the U.S. Plutonium Disposition Program," seminar, Princeton University Program on Science and Global Security, Princeton University, Princeton, NJ, June 12, 2002.
"Controlling Fissile and Radioactive Material," Public Health Summit on Weapons of Mass Destruction, sponsored by Physicians for Social Responsibility and the UCLA School of Public Health, Ackerman Hall, UCLA, Los Angeles, June 2, 2002.
"Assessing the U.S. Government Response to the Nuclear Terrorism Threat After 9/11,"
presentation to the Joint Atomic Energy Intelligence Committee, McLean, VA, May 9, 2002.
"Upgrading Physical Protection at Nuclear Facilities to Address New Threats," MIT Security Studies Seminar, MIT, Boston, MA, April 18,2002.
"Perspectives on New Plant Licensing," presentation at the U.S. Nuclear Regulatory Commission Briefing on Readiness for New Plant Applications and Construction, Washington, DC, July 19, 2001.
3
"Regulatory Challenges for Future Nuclear Plant Licensing: A Public Interest Perspective,"
U.S. NRC Advisory Committee on Reactor Safeguards (ACRS) Workshop on New Nuclear Plant Licensing, Washington, DC, June 5, 2001.
"The Future of Nuclear Power: A Public Interest Perspective," 2001 Symposium of the Northeast Chapter of Public Utility Commissioners, Mystic, CT, May 21, 2001.
Statement at the U.S. Nuclear Regulatory Commission Briefing on Office of Nuclear Regulatory Research Programs and Performance, May II, 2001.
"Barriers to Deployment of Micro-Nuclear Technology," presentation at the workshop on "New Energy Technologies: A Policy for Micro-Nuclear Technologies," James A. Baker III Institute for Public Policy, Rice University, Houston, TX, March 19-20, 2001.
"Aging Research and Public Confidence," presentation at the U.S. Nuclear Regulatory Commission 2001 Regulatory Informnation Conference (RIC), Washington, DC, March 14, 2001.
NRC Reactor Safeguards Activities," presentation at the U.S. Nuclear Regulatory Commission 2001 Regulatory Information Conference (RIC), Washington, DC, March 14, 2001.
"DOE's Nuclear Material Stabilization Approach: The Failure of Transparency," Embedded Topical Meeting on DOE Spent Nuclear Fuel and Fissile Material Management, American Nuclear Society Annual Meeting, San Diego, CA, June 2000.
"The Status of Reactor Safeguards Initiatives," presentation at the U.S. NRC 2000 Regulatory Information Conference, Washington, DC, March 29, 2000.
"Safety Questions Concerning MOX Fuel Use in Proposed U.S. Reactors," Sixth International Policy Forum on the Management and Disposition of Nuclear Weapons Materials, sponsored by Exchange/Monitor Publications, Washington, DC, June 1999.
"Transparency and Plutonium Disposition," ISIS Workshop on Comprehensive Controls on Plutonium and Highly Enriched Uranium: Long-Term Problems and Prospects for Solutions, sponsored by the Institute for Science and International Security, Washington, DC, June 1997. I "Ship Transportation of Radioactive Materials," presentation to the Marine Board of the National Research Council, U.S. National Academy of Sciences, Woods Hole, MA, June 20, 1996.
"The Importation and Storage of High-Level Radioactive Wastes at Rokkasho-Mura:
Safety Concerns," presentation at the Public Forum on High-Level Nuclear Waste and Reprocessing," Aomori, Japan, April 16, 1996.
"Perspectives on U.S. Options for Disposition of Excess Plutonium," Third International 4
Policy Forum on the Management and Disposition of Nuclear Weapons Materials, sponsored by Exchange/Monitor Publications, Landsdowne, VA, March 21, 1996.
"Addressing Safety Issues in the Sea Transport of Radioactive Materials," presentation to the Special Consultative Meeting of Entities Involved in the Marine Transport of Nuclear Materials Covered by the INF Code," International Maritime Organization, London, March 4-6, 1996.
"Prospects and Unsolved Issues for Plutonium Immobilization," INESAP/JANUS/UNIDIR Fissile Cutoff Workshop, Palais des Nations, Geneva, June 1995.
"An Intermediate Solution for Plutonium from Dismantled Nuclear Warheads," Annual Meeting of the German Physical Society, Berlin, Germany, March 1995.
"The Sea Transport of High-Level Radioactive Waste: Environmental and Health Concerns," Channel Islands International Conference on Nuclear Waste, St. Helier, Jersey, United Kingdom, January 1995.
Conference Papers E. Lyman, "The Congressional Attack on RERTR," 25"' International Meeting on Reduced Enrichment for Research and Test Reactors, RERTR-2003, Chicago, IL, October 2003.
E. Lyman, "Nuclear Plant Protection and the Homeland Security Mandate," 44h Annual Meeting of the Institute of Nuclear Materials Management, Phoenix, AZ, July 2003.
E. Lyman and A. Kuperman, "A Reevaluation of Physical Protection Standards for Irradiated HEU Fuel," 24h International Meeting on Reduced Enrichment for Research and Test Reactors, RERTR-2002, Bariloche, Argentina, November 2002.
E. Lyman, "Material Protection, Control and Accounting at the U.S. MOX Fuel Fabrication Plant: Merely and Afterthought?" 43 rd Annual Meeting of the Institute of Nuclear Materials Management (INMM), Orlando, FL, June 2002.
E. Lyman, "Terrorism Threat and Nuclear Power: Recent Developments and Lessons to be Learned," Symposium on Rethinking Nuclear Energy and Democracy after 9/11, sponsored by PSR/IPPNW Switzerland, Basel, Switzerland, April 2002.
E. Lyman, remarks for Expert Panel on Advanced Reactors, Nuclear Safety Research Conference, U.S. Nuclear Regulatory Commission, Washington, DC, October 2001.
E. Lyman, "The Future of Immobilization Under the U.S.-Russian Plutonium Disposition Agreement," 42 nd Annual Meeting of the Institute of Nuclear Materials Management (INMM),
Indian Wells, CA, July 18, 2001.
5
I E. Lyman, comments in the Report of the Expert Panel oil the Role and Direction of Nuclear RegulatoryResearch, U.S. Nuclear Regulatory Commission, May2001.
E. Lyman, "Can the Proliferation Risks of Nuclear Power be Made Acceptable?" Nuclear Control Institute 20'h Anniversary Conference, Washington, DC, April 9, 2001.
E. Lyman and P. Leventhal, "Radiological Sabotage at Nuclear Power Plants: A Moving Target Set," 41"t Annual Meeting of the INMM, New Orleans, LA, July 2000.
E. Lyman, "Comments on the Storage Criteria for the Storage and Disposal of Immobilized Plutonium," Proceedings of the Institute for Science and International Security Conference on "Civil Separated Plutonium Stocks--- Planning for the Future," March 14-15, 2000, Washington, DC, Isis Press, 135.
E. Lyman, "The Sea Shipment of Radioactive Materials: Safety and Environmental Concerns," Conference on Ultrahazardous Radioactive Cargo by Sea: Implications and Responses, sponsored by the Maritime Institute of Malaysia, Kuala Lumpur, Malaysia, October 1999.
E. Lyman, "A Critique of Physical Protection Standards for Irradiated Materials," 40'h Annual Meeting of the INMM, Phoenix, AZ, July 1999.
E. Lyman, "DOE Reprocessing Policy and the Irreversibility of Plutonium Disposition,"
Proceedings of the 3rd Topical Meeting on DOE Spent Nuclear Fuel and Fissile Materials Management, American Nuclear Society, Charleston, SC, September 8-11, 1998, 149.
E. Lyman, "Japan's Plutonium Fuel Production Facility (PFPF): A Case Study of the Challenges of Nuclear Materials Management," 39h Annual Meeting of the INMM, Naples, FL, July 1998.
E. Lyman, "Safety Aspects of Unirradiated MOX Fuel Transport," Annex 2b of the Comprehensive SocialImpact Assessment of MOX Use in Light Water Reactors, Citizens' Nuclear Information Center, Tokyo, November 1997.
E. Lyman, "Unresolved Safety Issues in the Storage and Transport of Vitrified High-Level Nuclear Waste," 38h Annual Meeting of the INMM, Phoenix, AZ, July 1997.
E. Lyman, "A Perspective on the Proliferation Risks of Plutonium Mines," proceedings of the Plutonium Stabilization and Immobilization Workshop, U.S. Department of Energy, Washington, DC, December 12-14, 1995, CONF-951259, p, 445.
E. Lyman, "Assessing the Proliferation and Environmental Risks of Partitioning-Transmutation," Fifth International Summer Symposium on Science and World Affairs, Cambridge, MA, USA, July 1993.
6
Op-Eds and Letters to the Editor L. Gronlund and E. Lyman, New York Times, 2004.
E. Lyman, "Troubles at Indian Point," New York Times, January 25, 2003.
E. Lyman and P. Leventhal, "Nonessential Nukes" (op-ed), Washington Post, November 26, 2002.
P. Leventhal and E. Lyman, "Shipping Plutonium," New York Times, July 12,2002.
E. Lyman, "Indian Point Reactor," New York Times January 27, 2002.
E. Lyman, "Spent Nuclear Fuel," New York Times, June 3, 2001.
E. Lyman and P. Leventhal, "Better Plutonium Plan," New York Times, February 5, 1998.
E. Lyman, "A Safer Plutonium Plan," Washington Post, August 24, 1997.
P. Leventhal and E. Lyman, "Who Says Iraq Isn't Making a Bomb?" International Herald Tribune, November 2, 1995.
H. Feiveson and E. Lyman, "No Solution to the Plutonium Problem," Washington Post, July 29, 1994.
E. Lyman, "Getting Rid of Weapon Plutonium," Bulletin of the Atomic Scientists, July/August 1994.
7
EXHIBIT B February 8, 2001 MEMORANDUM TO: John Flack, Assistant Branch Chief Regulatory Effectiveness and Human Factors Branch Division of Systems Analysis and Regulatory Effectiveness Office of Nuclear Regulatory Research FROM: Ralph Meyer, Senior Technical Advisor IRA!
Safety Margins and Systems Analysis Branch Division of Systems Analysis and Regulatory Effectiveness Office of Nuclear Regulatory Research
SUBJECT:
UPDATE ON GENERIC ISSUE 92: FUEL CRUMBLING DURING LOCA During the summer of 2000, RES held several meetings with experts to develop Phenomenon Identification and Ranking Tables (PIRTs) for a loss-of-coolant accident (LOCA) with high-bumup fuel (NUREG/CR to be published). During that meeting it became clear that fuel crumbling and relocation into the ballooned section of fuel during a LOCA remains an issue of interest in Europe. IPSN in France is still expressing concern about the absence of an accounting of this effect in LOCA analysis, and the Halden project in Norway is planning new in-reactor tests on this phenomenon during the next couple of years (HP-1 085, October 2000).
As a result of this recent expression of interest, I briefly reviewed GI-92 and its priority ranking of low that led to its being placed on the drop list. My view is that there has been an error in this ranking. It appears that estimated temperature increases resulting from this effect were applied only to increases in fission product release and were not accounted for in determining if the core had lost its coolable geometry.
At this time, we are pursuing resolution of the issue with results from the expected Halden tests in the 2003-2004 time frame. Because we are going to resolve this issue and because the issue is so old, we have decided not to attempt to re-prioritize the issue at this time. As soon as this issue can be resolved with the experimental results that are expected from Halden, we will document the resolution of GI-92.
Distribution SMSAB R/F DSARE R/F RMeyer R/F PNorian C:\GSI08.wpd OAR in ADAMS? (Y or N) Y ADAMS ACCESSION NO.: Ml n1n019n0is TEMPLATE NO. REs-nnDl PubliclyAvailable?(YorN) Y DATE OF RELEASE TO PUBLIC fl7I1al1 SENSITIVE? NO To receive a copy of this document, indicate Inthe box: C' =Copy without enclosures E' = Copy with enclosures N' = No copy OFFICE SMSAB l REAHFB I ABC:REAHFB l AD:DSARE L NAME RMeyer:adl REmrit JFlack FEltawila DATE 02/02/01
- 02/05/01
- 02/07/01
- 02/08/01*
V-W, Z L -Z, EXHIBIT C
4- i UNITED STATES NUCLEAR REGULATORY COMMISSION
-WASHINGTON, D.C. 20555-0001 June 20, 2002 MEMORANDUM TO: Samuel J. Collins, Director Office of Nuclear Reactor Regulation FROM: Ashok C. Thadani, Director Office of Nuclear Regulato r %;44C7 6,,
SUBJECT:
RESEARCH INFORMATION LETTER 0202, REVISION OF 10 CFR 50.46 AND APPENDIX K A significant amount of research has been performed since approval of the Emergency Core Cooling System (ECCS) rule in 1973. This research has enabled an improvement in the predictive capability and understanding of postulated accidents in nuclear power plants. In general, this research demonstrated the existence of a large safety margin between the regulatory acceptance limits and expected plant behavior during a loss-of-coolant accident (LOCA). As a result, 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," was amended in 1988 to allow the optional use of realistic physical models to analyze LOCAs. The basic 10 CFR 50.46 acceptance criteria were not revised in the 1988 rulemaking change and only minor revisions were made to Appendix K.
Because the 10 CFR 50.46 acceptance criteria are applicable to specific cladding alloys and the Appendix K models may contain some excessive conservatism, the staff in SECY-01-0133 [1]
recommended '(A) changes to the technical requirements of the current 10 CFR 50.46 related to acceptance criteria and evaluation model(s) and (B) development of a voluntary risk-informed alternative to reliability requirements in 10 CFR 50.46."
Reference 1 also identified several possible changes that might be made in Appendix K. These were:
- 1. Implementing the 1994 ANS Decay Heat Standard.
- 2. Replacing the Baker-Just model for metal-water reaction with the Cathcart-Pawel model.
- 3. Deleting the requirement for steam cooling only for reflood rates of less than 1-inch per second.
- 4. Deleting the prohibition on return to nucleate boiling during blowdown.
Efforts related to items 1 through 4 above that were discussed in Reference 1 have been completed, and the purpose of this memorandum is to inform you of the findings. Findings
Samuel J. Collins 2 can be grouped into three areas; those that pertain to rulemaking and the 10 CFR 50.46 acceptance criteria, those that are related to rulemaking and a revised version of Appendix K, and those that are related to non-conservatisms in the existing Appendix K. This third set of findings can be pursued outside the rulemaking process of the first two areas.
Findings on 10 CFR 50 Acceptance .Criteria for Rulemaking First, it remains technically acceptable to retain all of the existing requirements in 10 CFR 50.46 and Appendix K in their present form as an option such that no model changes or reanalysis would be required. Thus, the current requirements and methods of analysis would be grandfathered.
Second, the peak cladding temperature limit and the maximum cladding oxidation limit in 10 CFR 50.46 could be replaced by a performance-based requirement that would be independent of the particular zirconium-based cladding alloy being considered. In particular, this performance-based approach could allow the deduced peak cladding temperature limit to be above the current 2200 oF limit, provided that assurance of fuel rod integrity throughout a LOCA is demonstrated. The 1 percent total hydrogen generation requirement may be deleted, since the combustible gas requirements are considered in 10 CFR 50.44, "Standards for combustible gas control system in light water-cooled power reactors." The other two acceptance criteria, Coolable Geometry and Long Term Cooling, are already performance-based and therefore need not be revised.
These revisions should provide valuable increases in operating margins without reducing fuel rod integrity due to a LOCA. Further, these revisions do not specify a particular zirconium-based cladding material. They are generic, and apply to all zirconium-based alloys. Thus, it will no longer be necessary to obtain an exemption in order to allow a new zirconium-based cladding.
Findings on Appendix K to 10 CFR 50 for Rulemaking The technical efforts have shown that it is possible to replace correlations prescribed by Appendix K with correlations that are significantly more accurate. In doing so, some parameters are estimated more accurately, but this is accompanied by a reduction in overall analysis conservatism as might be expected.
As known conservatism is removed from Appendix K, there is the possibility that overall results produced by the revised Appendix K Evaluation Models might become non-conservative. Thus, there must be a process to ensure that calculations based on a revised version of Appendix K retain appropriate conservatism. This is consistent with the Commission Opinion for the original ECCS rule and with the conclusions rendered in SECY-86-318 [2], the Commission paper that ultimately resulted in the 1988 rule change. Provided that a process is established to ensure overall conservatism, it would be technically justifiable to make the following revisions:
(1) Replace the 1971 ANS decay heat standard with the 1994 ANS decay heat standard in a new optional Appendix K provided that there is an appropriate selection of user-specified values
Samuel J. Collins 3 and a determination of decay heat uncertainty. It is recommended that Regulatory Guide 1.157 [3) be updated to endorse the 1994 ANS Decay Heat Standard for use in a best estimate analysis. Details concerning application of the 1994 standard are described in Attachment 1.
(2)Replace the Baker-Just correlation with the Cathcart-Pawel correlation to account for heat release from the chemical reaction of steam with zirconium-based cladding materials in a new optional Appendix K. Regulatory Guide 1.157 could also be updated to endorse the.Cathcart-Pawel model for calculating metal-water reaction heat release for all zirconium-based cladding alloys in a best estimate analysis. In all cases, an adjustment should be made to account for the effect of enhanced oxidation at high pressures when analyzing small break LOCAs.
Calculation of cladding oxidation for comparison with the 17 percent limit should continue to be made with the Baker-Just correlation because that correlation was used in establishing the 17 percent limit. Attachment 2 contains a discussion of these changes.
(3)Delete the requirement for only reflood steam cooling for reflood rates less than one inch per second in a new, optional Appendix K. Attachment 3 discusses the basis for this change.
(4)Retain the prohibition on return to nucleate boiling during blowdown in a new, optional Appendix K. Attachment 3 discusses the basis for this recommendation.
Additional Findings Relevant to Appendix K Evaluation Models The staff, in SECY-01-0133 11] also recommended that an assessment of recognized non-conservatisms associated with Appendix K be performed. Non-conservatisms are un-modeled physical processes and simplifications now permitted under Appendix K that may result in an underprediction of the peak cladding temperature or the maximum cladding oxidation. Several non-conservatisms were Identified as part of this effort. These include (a) subcooled and saturated boiling in a downcomer annulus during the reflood phase of a LOCA and the resulting void generation and phase separation, (b)downcomer entrainment and inventory reduction due to steam bypass during reflood, and (c) fuel relocation following cladding swelling during a temperature transient. New Evaluation Models making use of a revised, optional Appendix K should conservatively account for these processes.
The need for retaining some conservatism Inthe Evaluation Model has become quite clear in this recent work. As documented in Attachments 4 and 5, the suggested revisions and non-conservatisms have a significant effect on analysis results. Hence, within the'regulatory framework, applicants making use of the new Appendix K should be required to ensure that the results are sufficiently and demonstratively conservative, and that the Evaluation Models appropriately account for non-conservatisms. Evaluation Models making use of the proposed revisions should be considered new Evaluation Models subject to a technical review.
Finally, it is recommended that a new Regulatory Guide be written to provide guidance to stakeholders on application of the 1994 ANS Decay Heat Standard and on a revised version of
Samuel J. Collins 4 Appendix K. The Regulatory Guide should discuss acceptable user selected inputs when using the 1994 ANS Standard and determination of decay heat uncertainty, and provide guidance on Evaluation Models using the new Appendix K. In addition, this Regulatory Guide should include sufficient guidance to ensure that appropriate levels of conservatism remain. This Regulatory Guide is expected to be useful to the industry as well as to the staff in making the review process efficient.
Coordination The findings discussed by this memorandum have been presented to the ACRS Subcommittees on Materials and Metallurgy, Thermal-Hydraulic Phenomena, and Reliability & Probabilistic Risk Assessment. Drafts of this memorandum and its attachments were reviewed by NRR and ACRS, and their comments have been considered. Many of the findings were also discussed with interested stakeholders at a public meeting earlier this year.
Attachments: As stated
REFERENCES:
[1( SECY-01-0133, "Status Report on Study of Risk-informed Changes to the Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations on Risk-informed Changes to 10 CFR 50.46 ECCS Acceptance Criteria," July 23, 2001.
[2] SECY-86-318, "Revision of the ECCS Rule Contained in Appendix K and Section 50.46 of 10' CFR Part 50," October 28, 1986
[3] USNRC Regulatory Guide 1.157, "Best Estimate Calculation of Emergency Core Cooling System Performance," May 1989.
.W I .
- Attachment 4
'IL Attachment 4 Appendix K Non-Conservatisms
2 Attachment 4 Introduction The Office of Nuclear Regulatory Research (RES) has Investigated several models and correla-tions required by Appendix K of 10CFR 50.46 in support of Risk Informed Regulation. As part of those efforts, the staff stated in SECY-01-0133 that potential non-conservatisms related to Appendix K would be considered. uNon-conservatismsu refer to those physical processes and modeling features that are not Appendix K requirements, and may result in lower peak cladding temperatures (PCT) or equivalent clad reacted (ECR) than would be realistically expected in a loss of coolant accident (LOCA). The staff isconsidering rulemaking revisions that would replace the 1971 Decay Heat Standard by a more realistic decay heat standard. Other models, including the Baker-Just correlation for metal-water heat release, steam cooling for reflood rate below 1-inch per second, and the prohibition on return to nucleate boiling during blowdown are also being considered for revision. Ineach of these cases, the revision will result in a reduction in the exist-ing conservatism in Appendix K. Thus, the non-conservatisms assume greater importance in Appendix K based Evaluation Models (EM) for LOCA analysis if not otherwise accounted for or if the existing conservatism associated with Appendix K is reduced.
RES has reviewed information made available to the staff by-vendors, produced as part of previ-ous rulemakings, and obtained through several experimental research programs. Non-conser-vatisms are discussed and recommendations and guidance are provided on how these non-conservatisms should be incorporated into existing and future regulatory decisions concerning revision of Appendix K and 10 CFR 50.46.
Sources of Non-Conservatisms Non-conservatisms in Appendix l' can be attributed to one of three different sources:
- 1. Thermal-hydraulic processes and fuel behavior that have been observed in experimental pro-grams since 1973: Since the original rulemaking, many experimental programs have been con-ducted to gain a better understanding of nuclear reactor thermal-hydraulics. With this improved understanding, physical processes not recognized, or considered important in 1973 are now found to play an important role in large and small break LOCAs.
- 2. Large code uncertainties: Uncertainties in predicting the PCT and ECR exist because of sim-plifications that are made in representing some physical processes, nodalization and numerical methods used by a computer code, and models and correlations that are applied outside of their original database. Additional uncertainties exist due to variations in plant operating conditions.
Because of these uncertainties, there isthe possibility that the 10CFR 50.46 limits can be exceeded. This was a topic of concern described by the staff in 1986 in SECY-86-318, ("Revision of the ECCS Rule Contained InAppendix K and Section 50.46 of 10 CFR Part 50) which recom-mended that the Appendix K decay heat guidelines not be relaxed unless model uncertainties were accounted for.
- 3. Specific models required by Appendix K may themselves be non-conservative. It is possible that models specified by Appendix K are non-conservative for some applications. An example is the Dougal-Rohsenow correlation for post critical heat flux heat transfer, which was found after the original rulemaking to be over-predict heat transfer. This was corrected as part of the 1988 rulemaking. There are numerous other models specified by Appendix K for use in LOCA analy-
4~ //
.1.
3 Attachment 4
' - sis. Few have been sufficiently assessed so that their assumed conservatism for LOCA thermal-hydraulic conditions has been quantified. Currently, none of the Appendix K specified models are suspected as being non-conservative. It should be noted however, that very few of the models and correlations specified by Appendix K have been rigorously assessed so as to demonstrate conservatism.
The recent RES review of 10 CFR50.46 and Appendix K concludes that there are three major issues that require careful consideration as Appendix K conservatisms are removed as part of new rulemaking. These are downcomer hydraulics, fuel relocation, and the overall uncertainty associated with LOCA Evaluation Models. A discussion on each issue follows.
A. Downcomer Hydraulics Downcomer hydraulics refers to two processes that were not anticipated In the original 1973 Rulemaking, nor recognized at the Ume of the 1988 Appendix K revision. The first process is downcomer boiling, which are the processes of subcooled and saturated boiling that may occur as fluid Inthe downcomer Is brought to saturation by heat released by the core barrel, reactor vessel walls, and lower plenum metal. The second process Is reflood downcomer bypass, which refers to the entrainment and carry-over of downcomer fluid to the break by steam that flows cir-cumferentially around the downcomer from the Intact cold legs. During the initial part of reflood, the downcomer water level is at an elevation near the bottom of the cold legs. Thus, high velocity steam entraining droplets from this stratified interface will occur; decreasing the downcomer level. Both of these processes are relatively *new". That is, the neither process was recognized as potential non-conservatisms until the early 1990's. Their effects can be observed in experi-mental data as well as in recent calculations with realistic thermal-hydraulic codes.
Downcomer Boil.;-,
The issue of downcomer boiling was first reported to the staff by Westinghouse through a series of meetings and exchange of information [1-3]. Large break LOCA calculations performed using a realistic thermal-hydraulics code showed that a second reflood clad temperature rise and the PCT frequently occurred after downcomer boiling took place. This secondary reflood tempera-ture rise was attributed to a loss in gravitational head In the downcomer due to the voids that were generated when boiling began. This loss in head significantly reduced the flooding rate, and allowed a prolonged secondary heatup to take place. Evaluation Models based on Appendix K do not necessarily capture this phenomenon, since modeling of the downcomer and subcooled boiling may be overly simplified in those types of codes.
It Is instructive to note the reasons why downcomer boiling has only recently been, observed and become a concern in large break LOCA analysis. Stored heat in thick metal structures is released slowly due to the thickness of the structures, and wall-to-fluid convective heat transfer coefficients. In a short reflood transient, one in which the core is quickly quenched, the down-comer fluid may not have sufficient time to reach its saturation temperature. As plants uprate in power however, large break transients necessarily become longer due to the increased decay heat that must be removed. This allows sufficient time for the fluid temperature to increase to sat-uration and boiling in the downcomer begins while the core still has considerable energy. Thus, the downcomer boiling process is dependent on the length of the transient For long transients, boiling and voiding in the lower plenum during reflood may also become important.
4 Attachment 4 Examples of downcomer boiling and their impact of large break LOCA calculations using Best Estimate thermal-hydraulic codes are available in the public domain. Reference (4] documents a calculation for a 4-loop Westinghouse PWR where downcomer boiling initiates a secondary reflood temperature excursion and an increase in the peak cladding temperature of roughly 222 K (400 F). Similar impacts can be seen for a CE/ABB System 80+ unit in References [5] and [6].
An important point, is that prediction of downcomer boiling is not restricted to one particular ther-mal-hydraulic code, nor any one particular type of PWR.
Experimental verification of downcomer boiling is limited. There are few tests that show an effect of downcomer boiling on reflooding rate or peak cladding temperature. This is because most reflood test facilities have been designed for low pressure operation and the Initial stored metal heat in the test facilities is much less than that in a full scale PWR. As a result, downcomer metal heat is non-conservatively scaled in most facilities, and the effect on reflood rate or PCT is not observed in the tests. In the few facilities with sufficient downcomer metal heat, an increase in cladding temperatures and reduction in reflood rate is apparent. Reference [71 provides a sum-Mary of downcomer boiling observed in experimental tests, and associated scaling issues.
Currently, there are no specific criteria in Appendix K that require downcomer boiling to be Included as part of an Evaluation Model for loss of coolant accident analysis. Section A, Item 6 of Appendix K requires only that metal heat be accounted for. It does not provide guidance on the level of detail necessary to model subcooled and saturated boiling in a downcomer, and thus potentially allows an inaccurate and non-conservative modeling of these complex'processes.
Reflood Downcomer Bvpass Emergency core cooling (ECC) bypass refers to process by which water in the downcomer is swept around the annulus to the broken loop. Typically, this process is of concern during the blowdown and refill periods of a la.ge break LOCA, when steam velocities are high and in coun-terflow to the ECC in the downcomer. During the reflood period ECC bypass has also been found to occur, although the'physical processes involved are different than those in the blowdown and refill periods. Reflood downcomer bypass refers to the entrainment and sweep out of water from the top of the downcomer. The water is entrained by steam flowing from the intact'loops across the top of the downcomer liquid. ECC liquid Injected to the intact cold legs may also become entrained in the steam, but can also condense part of the steam flow reducing its effectiveness to entrain flow in the downcomer.
Experimental verification of reflood bypass can be seen in the results of UPTF Tests 2 and 25, and CCTF Tests C2-4 and C2-9. These tests showed a strong relation between downcomer water level and ECC entrainment. High rates of entrainment and ECC bypass during reflood were observed when the water level in the downcomer approached the bottom of the cold'legs.
The tests also confirmed significant core - downcomer level oscillations, which helped contribute to ECC bypass. The entrainment of downcomer water reduces the driving head for core reflood, similar to the downcomer boiling effect. The effect of reflood downcomer bypass was concluded to be non-conservative In Reference 8, although the impact on PCT was not expected to be large. In a later study [9] however, it was concluded that the UPTF and CCTF experimental tests under predicted the effect in a PWR, and thus a larger increase in PCT due to reflood down-comer bypass was possible. Therefore, reflood downcomer bypass is considered a non-conser-vatism not appropriately accounted for in Appendix K.
- . 5 Attachment 4 B. Fuel Relocation Fuel relocation refers to the movement of fuel pellet fragments into regions of the fuel rod where the cladding has ballooned during a LOCA transient. This relocation of fuel causes a local increase in the linear power density (kW/ft) in the ballooned region.and higher cladding tempera-tures'compared to cases where the fuel does not relocate. The fuel relocation issue has been "previously considered by the staff as GenericIssue 92 (GI-92), 'Fuel Crumbling During LOCA."
Several experimental investigations using irradiated fuel rods have documented the existence of fuel relocation under LOCA conditions.'These include the PBF-LOC tests [10, 11] in the U.S., the FR2 tests [12] in Germany, and the FLASH5 [13] test in France. In each of these tests, fuel relo-cation occurred with pellet fragments from upper locations falling into the ballooned region of burst cladding. As reported in recent work' byIPSN [14], the fuel relocation phenomenon is not restricted to high burnup fuel, as some data Indicated fuel relocation could occur at bumups as low as 48 GWd/L Theoriginal resolution to GI-92[15] concluded that fuel relocation was a non-conservatismnot appropriately accounted for by Appendix K, but that the estimated effect on large break LOCA peak cladding temperature of +46 F was bounded by other analysis conservatisms[16]. The Issue of fuel relocation during a LOCA however, remains a topic of concern'in Europe, and test programs in both the U.S. and abroad are attempting to obtain new experimental data to quantify the effect More recent information however (14, 17] suggests that the fuel relocation effect on PCT may be significantly larger than that assumed in GI-92. .Fuel relocation during LOCA there-fore, should be considered an Appendix K non-conservatism with at least a +46 F impact on PCT..
until new data is available to help quantify the effect.
Currently, the Office of Nuclear Regulatory Research Is pursuing resolution of the issue through participation in an experimental program to be conducted in the 2003-2004 timneframe.Because the issue is old, and the experimental results sould be available in the near future, GI-92 has not been re-prioritized. With the new experimental information, it should be possible to better quan-tify the effect of fuel relocation.
C. Code and Evaluation Model Uncertaintyi The purpose of determining the'uncertaintles associated with a safety analysis is to provide assurance that for a postulated accident the applicable limits specified by50.46(b) are not exceeded. ndeveloping the original ECCS rule, it was clear that uncertainties and the retention of sufficient conservatism in the rule were considered important. As quoted from the Commis-sion Opinion [18] on the ECCS Rule (12/28/73),
'The Commission realizes that the knowledge in regard to a number of facets of the analysis of a loss of coolant accident is imprecise;isit partly for this reason that there is an on-going Water Reactor Research Program. The Commission Is confident, however, that the criteria and evalua-tion models set forth here are more than sufficient to compensate for remaining uncertainties in the models or in the data.
Continuing research and development will provide a more extensive data base for such items as heat transfer coefficients during blowdown and spray and reflood cooling, oxidation rates for zir-conium, fission product decay.he at, steam-coolant interaction, oscillatory reflood flows, fuel den-sification, pump modeling and flow blockage. With the additional data it may become practical to
/ 6 Attachment 4 assign a statistically meaningful measure of precision to the calculation. It is probable that, with a better data base, some relaxation can be made in some of the required features of the evaluation models. However, the Commission believes that any future relaxation of the regulations should retain a margin of safety above and beyond allowances for statistical error."
While the Commission Opinion is primarily concerned with sufficient conservatism, it is clear that the Commission's intent was to bound the statistical error" associated with the analysis meth-ods. This was also a main concern of the staff in SECY-86-318 (19], which recommended that the Appendix K decay heat guidelines not be revised unless model uncertainties were accounted for.
Appendix K to 10 CFR 50 currently includes requirements such as the use of the ANS 1971 Decay Heat Standard for decay heat plus 20 percent, use of the Moody break flow model, assumption of the worst single failure, etc. In addition, Appendix K identifies other analysis mod-els such as use of the modified Baroczy correlation for two-phase pressure friction multipliers, as "acceptable." There is no assurance that the models identified as acceptable InAppendix K are necessarily conservative for all of the plant designs or accident scenarios to which they may be applied. The selection and Implementation of these acceptable models and correlations, along with other unspecified models, are determined by the applicant. As noted in Reference 20, the models and correlations contained Inthermal-hydraulic codes for LOCA have numerous simplify-ing assumptions and questionable assumptions in their implementation. Thus, there is no guar-.
antee that the models identified as acceptable in Appendix K have sufficient conservatism to compensate for recognized or unanticipated non-conservatisms such as downcomer boiling or fuel relocation If the 1971 ANS Decay Heat Standard were replaced with a more realistic esti-mate of decay heat.
It is useful to make a distinction between "code uncertainty," and 'overall calculational uncer-tainty." The code uncertainty refers to the limit of accuracy that a thermal-hydraulic comouter code can calculate the value of a specific'parameter such as the peak cladding temperate i (PCT) or the equivalent cladding reacted (ECR) given a set of initial and boundary conditions.
The code uncertainty is due to performance of the models and correlations that are part of the thermal-hydraulic computer code, in addition to uncertainties associated with numerical methods.
The overall calculational uncertainty represents the sum total of the code uncertainty plus other sources of uncertainty that may affect the results. These include factors such as the fuel behav-ior, power distribution, break size and location, equipment availability, pump and valve perfor-mance, and plant initial temperature distribution. It also includes uncertainty associated with the experimental data used in the code assessment process. When possible, these uncertainty sources are often conservatively bounded in a safety analysis. The "statistical error" discussed in the Commission Opinion is interpreted to be the "overall calculational uncertainty" in this attach-ment.
Since the 1988 rulemaking change, new information has been made available to the staff con-ceming both code and overall calculational uncertainties. First, three different vendors have pre-sented to the staff statistically based methodologies using Best Estimate thermal-hydraulic codes. Ineach case, the uncertainties derived by comparing the predicted 95th percentile PCT to the 50th percentile PCT exceeds 300 F [21, 22, 23]. These relatively large values are due to the numerous models and correlations that are incorporated into a thermal-hydraulic code, and to the uncertainties associated with those individual models. It is important to realize that while Appendix K Is prescriptive, there are many models and correlations that are "ad hoc" and have relatively poor agreement with experimental data, or are overly simplified. As noted by the ACRS
7 Attachment 4
[20], The science of multiphase flow and heat transfer has not reached apoint where predictions can be made soley from a basis of secure fundamentals (as they can for many viscous single-phase flows, for example). Codes have evolved as an elaborate tapestry of interwoven working assumptions and approximate equations and correlations that have proved to be useful. Longev-ify of these engineering methods is no assurance of maturity, nor does it guarantee that the codes need no further development and improvement as new questions arise.x The staff has performed its own calculations usidig a recent version of RELAP to investigate code stability and uncertainty. Of particular interest is the effect of the exponentially increasing heat generation rate due the metal-water reaction no matter what model was used. For example, in one case using the Cathcart-Pawel correlation, a peak cladding temperature of 2550 0F at an ini-tial power of 1.0665 times the nominal power was obtained. When the power was Increased In the fourth decimal place to 1.0670, the cladding temperature increased to the melting tempera-ture. This Is very typical behavior for all codes and shows the Importance of predicting'a peak cladding temperature with a high degree of confidence.
An Appendix K based Evaluation Model can be expected to have an'overall calculational uncer-tainty at least as large as those reported to the staff using realistic codes, since the thermal-hydraulic codes used in such EMs are significantly less sophisticated. The magnitude of realistic '
code uncertainty is approximately equal to the reduction' il PCT expected if the decay heat were reduced in a large break LOCA calculation [24]. Thus, the 1971 ANS Decay Heat Standard cur-rently compensates for other models that have a high uncertainty. With a reduction in Appendix K conservatism by replacing conservative models with more realistic ones, there is the possibility that the revised Appendix K would produce lower peak cladding temperatures than a best esti-mate thermal-hydraulics code for the same set of boundary and initial conditions. Indeed, the staff has already been presented with information showing that an Appendix K Evaluation Model with a realistic estimate of decay heat predicts lower clad temperatures than a nominal best esti-mate calculation without unc.-rtainty 125]. This is a clear demonstration that Appendix K Evs:.a-tion Models have significant inaccuracies, and that conservatism in the 1971 decay heat standard compensates for other code shortcomings.
Second, the development of the Code, Scaling, Uncertainty, and Applicability (CSAU) [26] meth-'
odology concluded that both code and overall uncertainties vary in magnitude as a function of time. It was demonstrated that uncertainties propagate from initiation to the end of the transient.
Uncertainties that are small for one scenario, may become very largeIn a different scenario or If the transient length becomes significant. The implication of thisis Important, as the staff consid-ers power uprates and small break LOCA analysis. As plants upratein power, transients can become longer due to the higher rate'of decay heat. In a small LOCA, transients can be several thousands of seconds in duration even at current power levels. This leads to the possibility that' models and correlationswith small uncertainties can have a very large effect on PCT for a long transient when this uncertaintyis propagated. Thus, a simple estimate of code uncertainty is not appropriate, and can be non-conservative if applied without regard to transient length.
Therefore, as Appendix K based Evaluation Models are made more realistic by replacing selected models and correlations, itis important to determine the overall uncertainty and incorpo-rate the uncertainties so that they are functions of time.
8 Attachment 4 D. Conclusions and Recommendations A significant amount of research has been performed since approval of the original ECCS rule.
This research enables significant improvement in the predictive capability of hypothesized acci-dents In nuclear power plants. In several cases, it Is now possible to replace correlations pre-scribed by Appendix K with correlations that are significantly more accurate. In doing so, the estimate of critical parameters such as peak cladding temperature and equivalent clad reacted becomes more accurate, but this is accompanied by a loss of analysis conservatism.
This isparticularly important if the requirements to use the ANS 1971 Decay Heat Standard and the Baker-Just correlation for metal-water reaction are replaced with the ANS 1994 Decay Heat Standard and the Cathcart-Pawel correlations respectively. The 1971 Decay Heat Standard is:
generally considered to be sufficiently conservative such that it compensates for model inaccura-cies and known non-conservatisms. The proposed revision of other conservative model features, such as the steam cooling requirement for flooding rates below one inch per second and the pro-hibition on return to nucleate boiling during blowdown would further reduce conservatism in mod-els that apply Appendix K.
-However, as known conservatisms are removed from Appendix K, there is the possibility that results produced by Appendix K based Evaluation Models will become non-conservative. There must be clear assurance that calculations based on a revised version of Appendix K retain an appropriate level of conservatism. This is consistent with the Commission Opinion for the original ECCS rule and with the conclusions rendered in SECY-86-318. Evaluation Models using a revised version of Appendix K should:
(1)Account for the effects of downcomer boiling, and ECC bypass during the reflood phase. The Evaluation Models should be capable of calculating subcooled and saturated boiling In a down-comer annulus for conditions expected during reflood, ..Iu bhould be capable of calculating the resulting void generation and phase separation.
(2)Account for the reduction in downcomer inventory during reflood due to steam bypass. The Evaluation Model should be capable of determining the rate of entrainment as a function of downcomer level.
(3)Account for the effects of fuel relocation following cladding swell during an accident. The Evaluation Model should account for the local increase in power and increase In fuel'- clad con-ductance in the relocation zone.
(4)Require that when the calculated ECCS cooling performance is compared to the acceptance criteria set forth in either an existing or revised version of 10 CFR 50.46, there is'a high level of probability that the criteria would not be exceeded. This statement should require that any Evalu-ation Model making use of the new Appendix K provide reasonable assurance that the results produced by it are sufficiently and demonstratively conservative.
One option that the staff has used in the past to verify conservatism in an Evaluation Model is by quantifying the code and overall uncertainty. Uncertainties that have been Identified as important include those from the code constituent models and correlations, models describing fuel behav-ior, plant initial and boundary conditions, and component performance during a hypothesized accident. Text that would require this, and has been used by the staff to address the concerns related to item (4)above Is:
9 Attachment 4 "Comparisons to applicable experimental data must be made and uncertainties in the analysis method and Inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that,.When the calculated ECCS cooling performance is compared to the critera setforth in paragraph (b) of 10 CFR 50.46, there is a high level of probability that the criteria would not be exceeded."
'The staff has previously provided guidance that allows the use of realistic models in a LOCA analysis. An interim approach was discussed in SECY-83-472 [271, and Regulatory Guide 1.157
[28] provided guidance on full Best Estimate calculations of core cooling system performance.
Both documents discuss estimation of code uncertainties and other features required of codes using a realistic model for decay heat. The guidance on code and overall uncertainties contained in these documents could be applied in reviews of Evaluation Models and analyses that make use of any of the Appendix K revisions proposed by SECY-01-01 33.
E. References
[1] Letter from H. A. Sepp (Westinghouse Electric Company, NSBU-00-5970) to US NRC (J.
Wermiel), "1999 Annual Notification of Changes to the Westinghouse Small Break LOCA and Large Break LOCA ECCS Evaluation Models, Pursuant to 10 CFR 50.46(a)(3)(ii), May 12, 2000.
121 Letter from S. Dembek, NRC, to H. A. Sepp, Westinghouse Electric Company, "Potential Non-Conservative Modeling in Approved Evaluation Models," November 2, 2000.
[3] Memorandum from G. S. Shukla, Project Manager, Section 2, PDIV&D, to S. A. Richards, Director, PDIV&D, "Summary of March 27, 2001 Meeting with Westinghouse on Downcomer Boiling Modeling," April 16, 2001.
[4] Watts Bar Nuclear Plant Final Safety Analysis, UFSAR Amendment 2.
[5] Palazov, V. V., and Ward, L.W., OA System 80+ RELAP5IMOD3 Model for Downcomer Boil-ing Following a Large Break LOCA,' ISL-NSAD-NRC-01-009, Jan. 2002.
[6] Pottorf, J., and Bajorek, S. M., 'Large Break LOCA Safety injection Sensitivity for a CEIABB System 80+ PWR," Proc. 10th International Conf. on Nuclear Engineering, ICONE10-22519, April 2002.
(7] Memorandum from S. M. Bajorek, NRC to J. E. Rosenthal, NRC, "Downcomer Boiling Techni-cal Summary,' May 22, 2002.
[8] MPR Associates, Inc., Summary of Results From the UPTF Downcomer Separate Effects Tests, Comparison to Previous Scaled Tests, and Application to U. S. Pressurized Water Reac-tors,' MPR-1163, July 1990
[9] Memorandum from F. Eltawila, DSARE to G. Holohan, DSSA, 'Evaluation of Proposed Changes to 10 CFR 50 Appendix K," Sept. 11, 2000.
[101 Broughton, J. M. et al.,"PBF LOCA Test Series Tests LOC3 and LOC5 Fuel Behavior Report," NUREG/CR-2073, June 1981.
I 10 Attachment 4
[11] Broughton, J. M. et al.,"PBF LOCA Test LOC6 Fuel Behavior Report," NUREG/CR-3184, April 1983.
[12] Karb, E. H. et al., OLWR Fuel Rod Behavior in the FR2 In-Pile Tests Simulating the Heatup Phase of a LOCA," KfK 3346, March 1983.
[13] Bruet. M. et al., "High Burnup Fuel Behavior during a LOCA Type Accident: The FLASH5 Experiment," IAEA Technical Committee Meeting on Behavior of Core Material and Fission Prod-uct Release in Accident Conditions InLWRs, Cadarache, France, March 1992.
[14] Grandjean, C., et al. "High Bumup U02 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation After Burst," OECD Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-Provence, March 2001.
[153 Memorandum from R. Mattson to T. Speis, 'Fuel Crumbling During LOCA," February 1983.
[161 Emrit, R., Riggs, R., Milstead, W., Pittman, J., and Vandermolen, H., "A Prioritizafion of Generic Safety Issues," NUREG-0933, June 2000.
[173 Memorandum from R. Meyer, Senior Technical Advisor, to J. Flack, DEHFB, "Update on Generic Issue 92: Fuel Crumbling During LOCA,"
[181 USAEC,"Opinion of the Commission, from the Rulemaking Hearing on Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Docket No. RM-50-1, CLI-73-39, December 28,;1973
[191 SECY-86. ,I8, -sislon of the ECCS Rule Contained In Appendix K and Section 50.46 of 10 CFR Part 50", October 28, 1986.
[201 Memorandum from D. A. Powers, Chairman ACRS to R.A. Meserve, Chaiman U.S. NRC,
'Issues Associated with Industry Developed Thermal-Hydraulic Codes," January 11, 2001.
J21J Young, M. Y., Bajorek, S. M.,.Nissfey, M. E., and Nguyen, S. B., "Best Estimate Analysis of the Large Break Loss of Coolant Accident," Proc. of the 6th International Conference on Nuclear Engineering, ICONE-6265, May 1998.
[22] Framatome ANP, "Realistic Large Break LOCA Methodology for Pressurized Water Reac-tors,' EMF-2103(P), Rev. 0, August, 2001. (Proporietary).
[23] Hamon, D. A., ODuane Arnold Energy Center SAFERIGESTR - LOCA Loss of Coolant Acci-dent Analysis," NEDC-31310P, August 1986 (Proprietary).
[24] Ward, L. W., Palazov, V., Prelewicz, D., and Lauben, G. N., 'Assessment of Proposed Modifications to Appendix K of 10 CFR 50.46," ISL-NRC-403-00, Dec. 2000.
[253 Nissley, M. E., Westinghouse Electric Corp.j "Comparison of Best-Estimate Plus Uncertainty and Appendix K LBLOCA Analyses," Presentation to USNRC, June 28, 2001.
- 11 Attachment 4
[26] Technical Program Group, TQuantifying Reactor Safety Margins, Application of Code Scaling Applicability, and Uncertainty Methodology to a Large Break Loss of Coolant Accident,' NUREG/
CR-5249, December 1989.
(27] SECY-83-472, OEmergency Core Cooling System Analysis Methods,' November 17, 1983.
[28] USNRC Regulatory Guide 1.157, 'Best Estimate Calculation of Emergency Core Cooling System Performance," May 1989.
I I Attachment 5 Attachment 5 Effect of Proposed Revisions on Evaluation Model Results
2 Attachment 5 Introduction An important consideration with regards to the revisions proposed by Reference 1, is the effect they may have on Evaluation Model results. This section documents the results of several sensi-tivity studies and estimation of the impact of these model changes and non-conservativisms. The information was derived from several sources including staff assessments, contractor studies, and reports obtained from vendors and the public literature.
The effect of proposed changes can be significantly different depending on break size. There-fore, two tables are provided; one for large break LOCA and another for small break LOCA.
Because the large break LOCA analysis frequently limits core power, more work has been per-formed to investigate the effects of various parameters for that scenario.
Large Break LOCA Table 1 lists the sensitivity estimates for large break LOCA. For large break, the effect on PCT of decreasing decay heat in the analysis by replacing the 1971 ANS Decay Heat Standard with a more realistic estimate has been found to be worth several hundred degrees F. Existing studies have used the 1979 ANS Standard, which is more conservative than the proposed 1994 ANS Decay Heat Standard. Thus, it is reasonable to expect that analyses performed with the 1971 ANS Standard replaced with the 1994 Standard will result in a large break PCT reduction of nearly 500 "F. (While highly plant dependent, this amount of LOCA margin would allow plant power upratings of 10 to 20%.)
The effect of replacing the Baker-Just correlation with the Cathcart-Pawel correlation for heat release due to clad-steam chemical reaction has been es: mated to be less than 100 "F. The esti-mates assume that the clad temperatures are sufficiently high such that the metal-water reaction rate is important in the calculation (TIad > 1900 'F). Thus, for plants where the peak clad temper-atures remain low either due to decay heat model reduction or behavior of the plant itself, the effect of replacing Baker-Just with Cathcart-Pawel is negligible.
Three studies are cited in the table as providing a rough estimate of the non-conservatism asso-ciated with neglecting downcorner boiling. The Westinghouse estimate of +400 "F is the differ-ence between the first and second reflood PCTs for a typical 4-loop PWR. The second PCT is the direct result of the downcomer boiling and would not ha: ? occurred if the process had been neglected in the Evaluation Model. The +700 "F estimate < :m the RELAP simulation is consid-ered to be high due to excessively high interfacial drag calculated by that code. Together how-ever, the two estimates suggest that the downcomer boiling effect is significant and may offset any benefit obtained by a red;:c;on i.; decay heat.
The effect of fuel relocation is highly variable. In the earliest investigation available (by EG&G),
experimental results were used to determine the +46 "F increase in PCT. This was used in the initial resolution of Generic Safety Issue 92 (Fuel Crumbling During LOCA). More recent work performed in France [9] however suggests the effect to be significantly larger and a strong func-ti In of the packing fraction that cccu'S when the fuel relocates. It is likely that new test data will be needed to fully resclve the issue.
3 Attachment 5 There are several estimates of code uncertainty. In these estimates, realistic codes were used, which are considered to be significantly more accurate than Appendix K based Evaluation Mod-els. From Reference IO an estimate can be made of the code uncertainty by comparing the 95th and 50th percentile estimates of the PCT. For the case examined it was +340 "F. It is important to note however that the uncertainty Eased on a difference between a 99th (or 100th) and 50th percentile PCT is significantly larger. Since an Appendix K calculation is intended to bound all possible conditions, then the uncertainty based on the 99th and 50th percentile PCTs is more appropriate for an Appendix K code. In this case, the code uncertainty would be much larger than
+340 "F.
The other estimates of code uncertainty [11, 12, 13] are similar to that in Reference 10. In some cases, the exact value is not listed in Table I in order to insure that proprietary information is not disclosed in this document. The precise values can be obtained in the References cited. The important point is that the magnitude of the uncertainties are large, and are comparable to the reduction in PCT obtained by cecay heat relaxation. Thus, the "excess" conservatism in the 1971 ANS Decay Heat Standard is seen to compensate for code deficiencies.
- v "
4 Attachment 5 Table 1: Large Break LOCA APCT Estimates Process l APCTBasis/Comments Decay Heat -260 to Recent Westinghouse estimate based on App. K EM calcula-
-450 OF tions [2]. ANS 1971 + 1.20 replaced with ANS 1979 + 2a. Cal-culations performned using BASH-EM.
Decay Heat -372 OF NRC contractor RELAP calculations for CE 2700 MWt (Mill-stone 2) plant [3]. ANS 1971 + 1.20 replaced with ANS 1979 +
2a.
Decay Heat -460 0F 1984 Westinghouse study on Appendix K relaxation [4] .
Metal Water -45 to -55 Recent Westinghouse estimate assuming the Baker-Just corre-Reaction OF lation is replaced with Cathcart-Pawel for metal-Water reaction heat [23. Calculations performed using BASH-EM.
Metal Waler -75 OF NRC contractor RELAP calculations with Baker-Just replaced Reaction . by Cathcart-Pawel [3].
Metal Water -65 OF 1984 Westinghouse study on Appendix K relaxation [4] .
Reaction Downcomer +400 OF Westinghouse estimate from Best Estimate EM calculations for Boiling IN 4-loop PWR [5].
Downcomer +810 OF NRC contractor calculations using RELAP5 for a CE System Boiling 80+ (3600 MWt) nit f61.
Downcomer +63 0F Estimate based on WCOBRArTRAC calculations for an uprated Boiling + CE System 80+ unit [7]. Both downcomer boiling and ECC Reflood bypass during reflood were found to be important and contrib-Bypass ute to increases in PCT. '
Fuel Reloca- +46 OF EG&G estimate based on experimental tests in PBF (Power tion Eurst Fccility) to address Generic Safety Issue (GSI) 92 [81.
Fuel Reloca- +313 OFp csuUl reported in technical paper by IPSN [9] using tion CATH.ARE for a Framatome PWR (similar to a Westinghouse 3-lvop PV:R). A burst zone 70% filling fraction assumed.
Code Uncer- +340 0F ! APCT between 95th and 50th percentile uncertainty in a W tainty 4 !°op?F V!R for WCOBRAJTRAC calculation [10].
Code Uncer- +300 'F Difference between the 95th and 50th percentile PCTs for a taint- :-%ast'ncrouse RESAR-3S plant using TRAC-PF1(MOD1 111].
Code Uncer- > +275 OF :mmrt'vme ANP large break code uncertainty using realistic tainty 'rstooc' RELAP [121 Code Uncer- > +400 OF Go. code uncertainty using SAFERIGESTER [13) t3inty
A . Z-6 Attachment 5 References
[1] USNRC, "Status Report on Study of Risk-informed Changes to the Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed Changes to 10 CFR 50.46 (ECCS Acceptance Crite.,a)," SECY-01-0133, July 23, 2001.
[2] DiMuzio, C. A., Westinghouse Electric Corp., "BASH-EM Parametric Study Results," Presen-tation to USNRC, June 28, 2001.
[31 Ward, L. W., Palazov, V., Prelewicz, D., and Lauben, G. N., "Assessment of Proposed Mod-if-at ons to Appendix K of 10 CFR 50.46," ISL-NRC-403-00, Dec. 2000.
[41 Cadek, F. F., Gresham, J. A., Hochreiter, L. E., and McIntyre, B. A., 'Potential Thermal Margin Available From Changes in thi Appendix K Rule," Proc. of the International NuclearPowerPlant Topical Meeting, Taipei, Taiv; an, 1C34.
[5] Nissley, M. E., Westlnghou e Electric Corp., "Comparison of Best-Estimate Plus Uncertainty and Appendix K LBLCCA An*..yses," Presentation to USNRC, June 28, 2001.
[<l Palazov, V. V., arnd .'ard, '. ,., "A System 80+ RELAP5/MOD3 Model for Downcomer Boil-in3 Fcilowing a Large Break LCOA," ISL-NSAD-NRC-01-009, Jan. 2002.
[7] Pottorf, J., 'Rea!z;tic Simrn-ution of a Loss of Coolant Accident in a Combustion Engineering System 601130+ PV',\,> M.S. 7 -esin. Kansas State University, 2001.
(8] Emrit, R., Riggs, R., Milstead, W., Pittman, J., and Vandermoledn, H., 'A Prioritization of Ceneric Safety Issues," ?'UR7.'-0-C33, June 2000.
[9] Grandjean, C., et al. 'High Burnup U02 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation After Curst," OECD Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-Provence, March. 2031.
[a 0 Nissley. M. E., 'AWesting!;cuse Electric Corp., 'Comparison of Best-Estimate Plus Uncertainty 2'nd Appendix K LE;LOCA An:'>se-," Presentation to USNRC, June 28, 2001.
[11] Technical Program Grcup (TPG), "Quantifying Reactor Safety Margins," NUREGICR-5249, C-cc. 1989.
(12] Framatome ANR, 'Re _-tic La ge Break LOCA Methodology for Pressurized Water Reac-
's,' Et.IF-21C WF';, PeV Ov..' uzu-,' 2C;J1. (Proprietary).
(13] Hamon, D. A., Duane Arnold Energy Center SAFER1GESTR - LOCA Loss of Coolant Acci-
- ' nt Analy sis," ' JC -^ - . ' . a, A. gus't I906 (Proprietary).
(14] Bajorek, S. M., et al.,"Realisti: Small- and Intermediate-Break Loss-of-Coolant Accident Analysis Using WANCOBRA/7.FAC," Nuclcar Technology, Vol. 136, Oct. 2001.
EHB T EXHIBIT D
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-...:..After any LOCA transient a core geometry which preserve its coolability must be guaranteed ineans : e
. :-Shattering of the fuel rods has to be avoided
.pi---' Core coolability has to be maintained
'2'4
- Check that Criteria are not To derive Criteria V violated PE for reactor LOCAc PHEBWS STLOC MEETING',Ociobcr 2311*, 2003, Washinigtoi, D.C.USA A. AIA ILLIA T el al 17 N
n~~~~~~J .:, S y PFtI;te TO DERIVE CRITERIA 5 .
It mneans: to know the quantities which control the cladding residual ductility back to cold conditions and core coolability, the values not to exceed. j Suhinforma-tionr-~,e;~g-Jn;
- amous.p ft .'eij~~a i _( X me-2 ~
(P.CT)' adTEqui~ly n ldRatd~ R ~ ' 1 ~ ~ 4*:~~2 TO CHECK THAT CRITERIA ARE NOT VIOLA TEDFOR REACTOR LOCAs It ineans: To demonstrate, through calculations tools, that whatever is the kind ofS LOCA transient,, nowhere in the core, the criteria values are exceeded For a models PlE-B US STLOC MEETING, October 23th, 2003, Washington, D.C USA A. AfALLLIA Tel d 18
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A-a Hydrogen Uptake of MS'- Is eKxtremely low upper bound all 600ppm PHEBUS STLOC MJEETING, October23t11, 2003, Washington, D.C USA A. AIMILLIA Tet al 24
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6 PJIEBUSSTLOCAIEETING, October 23th, 2003, Washington, D.C USA A MAILLIA T el al 33
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Such a programme.hg thle PHEBUS Faci i will take advante accumulated when .the-programswith PIEIEBUSSTLOC MEETING, October2311s, 2003, Washington, D.C USA Am ALIILLM Tel al 35
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EXHIBIT E 01 097 2 MOXLTA-sta I
4
- I -
UGH BURNUP UO2 FUEL LOCA CALCULATIONS.
- - TO EALUATE EPOSSIBLE AC . **.
- OF UEL'RELOCATION AFR BUR C. GRANDJEANG. HACHE, C. RONGIER institut deProtection et de SUretc Nucl6aire CEN V'.adarache, FRANCE.
..
- Abstract -
A literatur revew. conducted at UPSN, of available resukti of past'LOCA 'in-pi eperinments with irradiated
- fuel 'has revealed that irr6a'tiedrods behavior-wassig zyu1itfferentfrom that of znirnri ted rods usner sInilarconditions.
- . *Partielarb, as suggestedfrom the resulsfrom PBF-LC FR2 and FLASH5 experiments indicating a geperal.
. occurrence of the relocation of fragmentedfude within the ballo~hed cladding, a main concern was raised regardingthe poisible imact offuel relocationion peak cladt:emperatureand.local oxidation rate.
In View of obtiining-some,insight ito thefuel rodperformancefollowingfuelrelocationin'aPWR high BU U0 2 rod, wder LOCi . calculations were being peformned. using the French CATHARE-.2 code with.specific rrnsi e u ffrotncs'-so as to dci a lacumlto oftcr bwt~ss i~n thc ruptured shotc rod Flddng-Main rltsi iht Jeti thc cladtemperatrcas signiintbutsh*U;cmainselow'e ECCS acceptanccelini on lCT. On the other hand. the maximum cladding oxidation rate imay exceed the 17%
However, alternativeembrinlement criteriabased on residualthilnwss ofductile metaL such as the Chung and.
-Kassner critcria, indicate a fair remaining margin to the therjmal shock embrittlenment limi4, whercas the
- andlng embrittlement limit may be exceeded * . ,.' . .
- ~*
1 INTRODUJIQN In the foilun o1 h b~is hat'w.er jointly conducted by lPSN and EDF in6rwdei ato investigate.
tlie-behavidrof high burnup fuelcladdingiuidi LOCA conditions,; lSN has b&en re-examining the
- problem or Liss-of-C(.olant-A~cidents with consideration of specific aspects related to fuel and-cliddingirradiation, soas to identify be remaining needs for friher studies and experinmentialdata.
These concerns have led IPSN to initiate new studies in order to provide the-answers to-pending.
- questions
- iegaiding the bohavior of irradiated rods and'as-efriblies underLOCAonditions.
In a preliminary. sep, in view of obtaining some insight into the fuel rod performance following fuel
.relocatioo in aPWR high BU U0 2.rod uhder LOCA, calculations were being perforned, using the
.French CATHARE-2 code w~ith specific modifications in. the fuel routines so as to describe a fuel accumulaiioni after burst in-the rupttiurd mesh of lhe sod cladding.
-. GROUN
. .1. Irradiatedfuelrod-behavior.-
- 21
- Uterature review there.exists a few number of available results of such experiments, with irradiated fuel rods under LO.CA conditions: main' issues were. found in results from the. PBF-LOC tests in the USA[.1,2],. the FR2 tests in Geniany[3];gnd the FLASH5[4] test in France.
011019 MOXLTA.s,( ,
A process of fuel relocation was clearly evidenced from the experimental observations 'made in these tests series : in all irradiated rods of the PBF-LOC, FR2 and FLASH5 tests, fuel relocation has
- occurred as a'result of slumping of pellets frariients from upper locations into the.swolleni regionof the burst'cladding. Fuel relocation phenomena. is not restricted to high burnup fuel since fuel fragmen'tation occurs as soon as low bumup levels (it was thus noticed on LOC5-7B rod, fresh rod preconditioned up to 48 MWd/t).
A main question concerns the instant of fuel relocation occurrence in these experiments. It is not easy to make it perfectly clear for most of tests but, in FM2 tests E3 and E4 that were specially instrumented for that purpose, it was demonstrated that the fuel movement initiation occurred at the time of cladding burst, possibly initiated by the pressure difference between rod plenum and coolant channel with assistance of gravity slumping.
The fuel movement was probably favored in PBF-LOC and FR2 experiments where the fuel-cladding
- gap. was not totally closed,'-due respectively to low burnup or to the inverted rod internal external
- pressure difference during initial irradiation at low temperatur. A tight bonding between fuel and' clad was supposed to counteract the fuel motion inception. However, in FLASH5 experiment with high burnup fuel (50 GWd/t), and in spite of a low 'clad ballo'oning (not, higher than 16%) post-tesf examinations hive shown that fuel fragments were no more stuck to the cladding: the transienit temperature rise combined to clad deformation may be sufficient to suppress fuel-cladding bonding.
.2.12 Main concern For irradiated fuel rods, as observed in the PBF-LOC results, the clad deformation is expected to be larger than for fresh rods, as a result of a more uniform temperature distribution associated to pellet- .
clad gap reduction following clad creepaown during rod irradiation. The increase in claddeformation will leave more space for fuel fragments to relocate. Since the fuel fragmentation is clearly associated to burnup, with finer fragments at higher BU, a pellet stack slumping is likely to -occur after burst.
resulting in more or less compact filling of clad balloons, A major question is then what could'be the impact on peak clad temperature and final oxidation ratio of the local increase in lineic and surfacic power and of the associated local decrease in fuel-clad gag ?*-
- It should be emphasized'that this question is particularly important for U0 2 fuel at beginning-of-life and for MOX fuel at end-of-life where power generation is not reduced unlike for U0 2 fueL 2.1.3 Early evaluations ' . .
The State-of-the-Att Review performed by PD. PARSONS et al. for CSNI/PWG-2 and'published in-1986[5], thus after PBF-LOC and FR2 tests completion, reports two calculation studies addressing the impact of fuel relocation on peak clad temperature.
2.1.3.1 CalculationsinSweden The first one was conducted~in 1978-79 in Sweden by Bergquist[6], within the frame of the ECCS evaluation for the Ringhals 3 power plant. It consisted of a series of parametric transient calculations, performed with the TOODEE-2 code, so as to evaluate the re;ponse on clad temperature with/without fuel relocation in the balloon after rod burst.
The main assumption for fuel relocation was a uniform redistribution of fuel in the deformed meshes of the balloon with a density taken to 50% of the theoretical density in the base case; a fuel' thermal conductivity of 0.6 W/mIK and heat exchange between fuel and cladding dealt with an' exchange coefficient of 5000 W/rIN2 K. In the reference case, the hoop strain of the -most deformed mesh of the' balloon was 42%, and the peak clad ternerature (PCI) without fuel relocation did not exceed 2000.0 F: :
(=10930 C).
Calculations with fuel relocation showed that the evolution of clad temperature in the ruqtured mesh.
is essentially dependent of the power rating in that mesh, in relation with fuel average density . *..
011020 4
barnL qwet
I
- at 43 kW/m linear power (Fq = 2.09),. the clad temperature evolution remains of classical shape, with aPCraround2o5 0°F(1121C);
- at 47.87kW/m linear powr (Fq = 232) the clad tenperature evolution.exlibits a significant rate.
increase-around 20000 F with a subsequent temperature esc latioi after45 s in'the transient;
-. at 43 kWMm linear power, but with a 60% fuel theoretical derisity in the balloon (instead of 50%),
the clad temperature evolution again exhibits a significant rate increase after 45 s, reach of the 2200"Flinit around 62s, followed by subsequent temperature escalaton.
Although these early calculations had to be i'6nsidered with large reservations, it may look surprising that they were not much discussed-nor compared to countdr-cilculations, in consideration of the possible importance of calculated trends with respect to safety analysis.
2.1.32 INEL Calcublatons The second evaluation was a steady state. thermal analysis of a ballooned fuel rod following a fuel redistribution, the amrount of which based on PBF-LOC tests results. This anialysis wai performed by T.R Yackle[7] as a response to a NRC request; it is also mentioned by Broughton in the PBF-LOC3/LOC5 test report (1].
Fuel redistribution in the ballooned cladding is modeled by a series -ofup to 7 concentric rings of
- different width to take account of large particles of original fuel and sirll particles of additional fuel, neighbor rings being separated by gas gaps: Only radial heat transfer is considered, with a rod power.
corresponding to ANS decay heat 100 s afte rscrarn (- 3% original power), and a flat radial power... .
profile. A cladding surface heat transfer coefficieht:of 60 Whiti?/-vas auixuedJ a fuel thermal.
- conductiviity constant -at 2.6 WlrmlK and no 'radiativetransfet bet veen fuel particles.
Ihe amount of fuel redistribution has been determined from the results of the PBF LOC-3 and LOC-5 tests. A line fit through the available, data of fuel relative increase as functioni of cladding relative volume increase indicates an average filling ratio closed to 0.65. Three calculations have then been considered, corresponding to clad strain of 0, 44 and 89%. The following table gives the temperatures at fuel centerline and clad outside surface obtained in these-three calculations.:
Clad strain (%) .' . d4()J*. Tct jj')* . .
0 1095 .1806.
.44 1120 -: .1620 89 1320 ' 2450 For the worst case, with 89% clad strain allowing the redistribution of 160% additional fuel, the.
outside clad temperature is 225 K larger than for the reference case (without deformation), while the corresponding increase on maximum fuel teniperatur6 is. 1270 K. Y * * . .. ..
The conclusion that was drawn at that time appears presently quite surprising, as it was stated that' i'uel relocation into a balloon (with conditions.such those calculated) will not pose a significant.
problem during a LOCA since both fuel and clad temperatures remain well below the corresponding melting points"...t may be thought that the relatively close occurrence of the ThI-2 accident is likely e to explain such shift of concerns from LOCA to Severe Accident-issues. ' ' .-. . .:
3 IPSN CALCULATIONS . :. : ;
3.1 Reference code andcalculationprocedute :
The French CATHARE-2 code hasibeen chosen as.abase tool due 'to its capability to provid a bt *.
estimate evaluation of the thermal-lydraulic evolution in hot assembly. as.well as: in inean core .
subchannels. The code organization allows to run. stand-alone: calculations of the fuel module (the 011021 MarMT. . S'j(-
I
II
- CATHACOMB module) in order to pravide rajoidly informaition ibout the behaviour of specific fue'l-
- rods subjected to a given set of hydraulic conditions ; these hydraulic conditions will thus not be
- influenced by the behaviour of such specific. rods: The hydraulic conditions may bi rectrieved from 'a
.previous A ARE whole calculation, or may be that of the cun -CATHARE comuaon
%~hichthe stand-alone fuel m6dule is carried out in parallel.'.
- For the purpose of this studythe calculations were performed as follows:
-in a first-step a wvhole tATHARE-2 computation was run, for a large. break LOCA- transient occurring on a typical French PWR,, witiran input deck corresponding to a fresh fuel "me-an core"
- rod and a high burnup U0 2 fuel "hb't assembly" rod ; this'comptitation provided the hydraulic fil used as input-in the~foUo~iing calculations; ..
- in a .secornd step :qseries of stad-aloie CATHACOMB calculations were run, *using -the
- .reiuycraehydraulic file fra specif hot'rod 6f thrhtaseblyand withthe inclusion of specific modifications in some fuel routines inordem t simulate fuel rebocationuafter burst in
-therpftured-nxtieshesenwifcatons will be briefly described in the following.
3.2 Code versiod 7he CAs tARE-2 ' 13L code iofsion was used as startinh version, acdording to the known iprovements implementedin this versionuto calsate thr refloodingtphase of the LOCA tradsievnt Slight modelag improveehnts have been,'adedr in bhe fuel routines so ae to compute at the end of each time step the oxidafuio weight gisn (uing both Cadwart-Pawel .aker-iust and rate lanrs) as well
- as-the thic ness.eof tbisZrO oxidatioh cllatyer w. thb p6mier:vriable ilows a direct ca:culatim of.thq.
- ~ uiaioxid-aiion raite'ECRI while the iatteik 'a' i~ble allows to 4erive the rmainngthe sso thecentral
- h-Zr layer. w .
.3.3 uBasic-nput optionsn crthecinitialCATHARE-Z whole calculatio
-3n3. Accident s transient condifonst , H .
' , .fi assumptions are:in agrenient ivitf rthse retaine indtheoStanilard Safety Report for 90Gi
- MV/c Fren'ch PWRs: ...
hubie eturded 1reak oncoldleg of thn.' lovp bean'g the pressurizer, i t f . , * ,.
r-o teat. I2%,nojuina powrtacjen iiton.
- ~~~residual p~imif. ANS714 20%. . .-
3:.2 Fuel iods description . ..........
., ', 'Basically, tHreefuel iodsiay bdescribedo fas CATdARs calculatinon:
.m-rthementicprerod,with a weight of (Nth-L)xNai Siwhere N. is the number of assemblies in the core and N. the numbes of active rods per assembly cache hot assembly th ean roa with a weight of Nat
- one(or several) hontrod(s) in the hotabssemblW'
'Each of these rods is descri'bed in terths of geomrtry, cladding oxide thickness pro-ile, power profile.
A typical axialameshing with 40 meshes was chosen.
- In agreement with theseptions retained in studies perfornyed-
.sensitivity.
at EDF somc years ago, the iecpwr~fte baicrrod - that of beginning of life (BOL) while onl th-hot
-shoe chcFnchosen to 5 GWj/tU 'radiai ;the ratio. Fofhotrod poweito mea rod
- ,e.;was dgube to 128 b(as compared to th6 135 value for BOL case), and the ratio F, of hot rod power to hot assembly mean rol poweryws kept to 1.05 identical to BOL case.
- 011 022 MOXI TA.Sqfy I
i I
I
I
- Prior to the LOCA initiation, the reactor core is.then supposed to be subjected to a transient evolution that brings the.three above mentioned -rodsrespectively to 68.20 kW, 83.25 kW and 87A1 kW, and with a truncated cosine axial power profile.
The irradiated rods bear an external oxide layer on the zircaloy clad, with a thickness profii typical of 4 cycles irradiation, the Maxirmrn thickness reaching 106 pm-at 2.79 m elevation. The pellet-cladding gap is supposed to be closed in the irradiated rods, at transient initiation. The internal pressure in hot conditions in irradiated rods is significantly higler (-15.6 Mpa) than in mean core fresh rods. . .-
Two hydraulic channels are associated to. the mean core and hot assembly rods. The thermo-mechanical behaviour of.the hot rod(s) is inilienced by the. hot assemblj chanW hydraulics during
.hee blowdown ahd refill phases and by the mean core chaneiel hydraulicsd the refloodin phase.
3.4 Reference case behaviour.(wlthout fuel rekicatlon)..
A-reference'calculation without fuel relocation was first performed for the hot rod of the hot
.asscrnbly. It must be pointed out that a best-estimate treatment of the clad ballooning and burst for the irradiated rod was not searched here: the standard clad deformation and burst models for fresh fuel Were kept unchanged in CATHARE.
However, in consideration .of the results of the TACIR experinents (oxidation and quenching tests) on irradiated cladding 18], having-clearly indicated that the initial oxide scale was no more protective for high temperature oxidation, it was chosen to suppress the-protective effect of the initial oxide
,'scale, Iwads.transieit oxidation'eof th d, thalis n l ctie in ihe,standard oxidation model
- of CATHARE.: . '- ' ' *'
. Z..
The rod cladding appeared to rupture at 30.2 seconds on mesh 24 (elevation 2.15 m) wiith a hoop strain of 563%. AU the foll6wing results, unless explicitly stated, will refer to.the ruptured mesh elevation. .. .
t
.Figure ] displays the evQlutlon of the fuel centeiline and clad outside temperatures: the clad outside temperature rises to a maximum of, 9700 C while -the fuel centerline temperature remains below 1100°C during the heatup phase. .
,. .'Figure 2 displays the equivaleit cladding reacted ECR-evolution; as. calculated with Cathcart-Pawel
.:>'rate'elaw, for the. ruptured mesh and -the -two -neighbor rneshes. Foir the non rupturid'meshes, due to unprotected oxidation on the external face only; the oxidation iate ECR is increased by about 1;7% in absolute value, while on the ruptured mesh, due to two-sided oxidation, the ECR rises-from an initial value at 9.2% to 12.6% at the end ofthe transient.' .
- ' ... c mp the equivalent-cladding reacted ECR eiolutions, as calculated with Cathcart-Pawel
- and Baker-Just rate laws; for the ruptured mesh. It can be noticed that both 'correlations give very CIose, results in the corresponding range of clad tempratre Acpac riterion on clad maximum oxidation rate (<17%) is clearly well satisfied.
Figure 4 displays the evolution of the remaiingithickness.of the clad P-Zr layer, showing the sharp drop in thickness (from -520 to -330 prm) corresponding to clad ballooning up to rupture, followed by a.slow decrease corresponding to high temperature oxidation. The final-thickness remains Just above 300pm,,indicating a fair remaining resistance to thermal shock embrittlemnent, with reference to. ebrittlernent criterion proposed' by Chung and Kassner[9], while the handling limit proposed by -
theseauthors is just reached.
3.5 Fuel relocation case.
3.5.1 Basic assumptions and. modeling options .
- ,'Withj-eference tb the FR2 experimental results discussed before in section 2;1.1, we assumed that fuel
-pellets crumbling and relocation occurred'immediately'after the cladding burst, leading to a partial 01 1 023 YOXU? .Sf',
I
I I
filling of the inside, volume of the bailooi dcl adding ruptured mesh. This volume was calculated as that of a cylindrical volume with clad inner radius at'burst and mesh height. It was then assumed that.-
this ruptured mesh volume was filled hornogeneously with fuel fragments up to an user's inpu~t filling rate (- ritio of dense fuel 'volume to new mesh volume). A base calculation was performed -with a filling ratio of 61.5% corresponding, to a value measured in the FR2 experiment E5. Two other calculations were conducted with values of the filling ratio of 40% and 70% in order to evaluate the sensitivity of the results to this main parameter.
The fuel fragments were assimilated to spherical particles with user's input diameter. These fragments are in contact with the cladding, leading thus to a closed fuel-cladding gap.
The effective thermal conductivity of the fuel fragments was derived from the Inurai[0] 6orrelation.
and taking into account the radiative 'transfers between'particles according. to the Yagi theoretical model. The resulting model, so-called Imun-i Yagi" mode, had been implemented in' the SFD code ICARE2 of IPSN after it had been validated against Sandia DCI experiment , ' :
According to the results of the.TAGCIR experiments mentioned.in-previous'section,ithe protective effect of initial.oxide scale towards LOCA.trarisient oxidation was again suppressed in all the following calculations involving fuel relocation.
3.5.2 Results of the base case (61.5% filling ratio)
A basic calculation was performed with a filling ratio of 615% corresponding. to a local value measured on a sample taken from the ballooned region (with.67.5% total circumferential elongation) of the the FR2 experiment ES. The particle diameter was taken as the average size detemined in FR2 experiments, i.e..2.7 mrii. - . -.. *. . *,. ' ,. . :. .,
Figure 5 displays the evolution of the fuel centerline and clad outside temperatures the cld outside temperature reaches a maximum level around 11008 C while the fuei centerline temperature ienains below 1200"C during the heatup phase.
Figure 6 shows the evolution of the oxidation rate ECR as calculated with Cathcadt-Pawel rate law, for the ruptured mesh and the two neighbor meshes. It appears that the oxidation rate rises from 9.2%
to near 18% on the rupture/relocation mesh, while the-increase in ECR does not exceed 2% on the neighbor meshes. Figure 7 displays the evolution of ECR values at ruptured mesh; as calculated with Cathcart-Pawel and Baker-Just rate laws; ionpared to the corresponding curves for the calculation -
without fuel relocation (figure 3) a clear distinction.can now- be made betv'en both -evolutions; corresponding to the increase in clad temperature: The nAiimufin value ofECR calculdted with Baker-Just rate law is 19.4%, thus exceeding the curreni acceptance limit. .-
Figure 8 displays the evolution of the remaining thickness 6f 'the clad P-Zr layer,.showing thess-me sharp drop as in reference case corresponding -to clad ballooning up to rupture,. followed by the decrease corresponding to high temperature oxidation, with a fial' thickness just below 250 pm.r Since the maximum oxygen content at this temperature level remains below 0.9 wt %, it appears that -
the ChungtKassner criterion ivould be satisfied'for the thermal shock limit but not for the handling limit ' . . . . -
3.5.3 Sensitivity to the balloon filling ratio Finally, comparative calculations were performed with filling ratio values of 40% and 70%, the latter value corresponding to the fuel void fraction measured by gainma. decay counts in some. PBF-LOC...
experiments. The particle diameter was keptat th'sam6 value a in the previous cilcilation (2,7 rm)).
Figure 9 displays the evolution of clad putside temperature'wvi'tflincreasing valie of Ihe balloon filling.
ratio: for 40% filling the temperature level is siidilar.to that of reference case without fuel relocation whereas peak clad temperature reaches 1i44C with 70% filling; .- ,.
I Figures 10 and 11 display the evolutions of the oxidation rate ECR;as calculated with Cathcart.
Pawel, and of the remaining thickness of the clad #-Zr layer respectively,iwith increasing filling ratio.
011024 ;
U"XLT -y -Wa
i I I
I i
I I
values. For highest'filling value, the total oxidatiQh rate ECR reaches 19.7% (22%'with the Baker-Just rate law) while the ft-Zr layer remaining thickness remains near 230pim at the end bf LOCA transient.
4
SUMMARY
AND.CONCLUSIONS:
LOCA transient calculations have been perfohned with an adapted version of the French code CATHARE-2 in order to evaluatetoe possible impact of crumbling and relocation of irradiated fuel iu the ballooned region of a cladding after burst.
Focus has been put on the sensitivity ofTeak clad temperature and final oxidation rate on the filling ratio of the ballooned cladding with fuel crumble.
The calculations do not intend to give a bestcestirnate view of the detail behaviour of high burnup fuel.
rod under LOCA transient. In particular, the thermo-rechanical.properties of irradiated.zircaloy were not available for-tSe calculation of cladding defdrmation and bdrst with irndiated material.
The results indicate that for fuel rel6cation in the ballooned region with a filling ratio up tothecvalues obtained m FR2 or PFLOC expeiiments,'the peak clad temperaiure itay increase significantly, but still remains belo .the ECCS acceptance limit (1200TC) on PC*
- On the other hand, the .maximum.cladding oxidation'rate exceeds the 17% acceptance limit when the initial (in -service) oxidation rate is cumulated 'with the transient oxidation rate and 'when the. initial oxide layer is 'assumed 'no more protective for 'transient oxide. growth..'However,'altemative'
- mbrittlerneiit criteria based on residual thickness of ductile metal, such as the Chung and Kassner
-criteria, indicate a fair remaining margin' to the thermal embrittlemeni lirnit, whereas the.
ebshock
.handling mbdtilefmentligtiappears exceeded.-
The results of the preset .stiudy give some insight into'the possible impact of t&e'cruinblini and .
relocation of high burnup U0 2 fuel in a LOCA transient, aipheuiomena that was observed previously in in-pile experiments and which inight significantly affect the late evolutionr of accident transient and associated safet issues. It must be pointed out that results of corresponding calculations with low burnup UO2 or high'.burnup MOX fuels would have been more severe with regard to acceptanice limuts. . .
The results of the present calculation study-give some'support to the need for further experimental data,' to be provided by irradiated fuel LOCA experiments involving fuel relocation. A .best lepresentativity should be.obtiined with in-pile cxperimientim so as to maintain heat geneation jn fuel,..
fragtments Whatever their displacement may be. during'the: relocatida process: Such experiments aire' currently under planning bylHalden Reactor Pr9jeci'and by IPSN. .
- .OX.T. .S. .
- .
- 11025 i
I I
I REFERENCES 1*. JABROUGHTON eta PBF. CA Tests Series. Tests LOC3 and LC5 Fuel- Behavior Report.
NUREGICR 2073, June 1981.
- 2. JIM. BROUGHTON et a], PBF LOCA Test LOC6 Fuel Behavior Report.
- * * .*NREGCR 3184, April 1983. t h H
'..EJl.Rpepriet.:'
KARD et- W., LWR Fuel Rod Behavior in thi FR2 in-pile Testg Siitltn h etu
.Phase of-a LOCA.--
KFK 3346, March 1983.
- 14. M. BRUEG et ida, High BrnupFuel BehavioT durineg a LOCA Tpe ALOCdent The FLASH5
- IAEA Technical'Commttee Meeting Behavior of Core Material and F.P. Release in Accident Cotiditions in LWIRs, Cadarache, France, March 16-20,1992.
- P.PARSONSE/
U.5. D.CHNDLE, CA. MANNRTe Donitation, Oxidation and Embritt191ent of PWR Fuel Cladding in a Loss-of-C olant AccidenlAState-of-the-Art Report.
. NUARECD-CSNI 31 eport 129, Decemb8.986
'6. P..JERGQUIST, A Parameter -Studye-oncerning the Im on the Calculate PeHk Clad
.Temperature of a.Redist n bfte Fuel After Cladding Swelling and R-PpatLre.
- FV-79-0O017/2 Cndteion Reato Safety CInormation Meeting.
Marchesa-d USA Otobr2 196
- .7. *T.R,. ACKLE. "Steadystate.Fuel Rubble Thermal Analysis", HJZ-317-80 co~rrepondec from ofPRFdcldigcna-den~oat~cdn.
H.I. Zeile, EG&G, Idaho, to RE. Tiler, DOE/ID, Idaho Falls, tt-o-h-r Report, 29, 1980.
Septefinbei C. GRAN3 JEAN,:Rc C C EBJFFE, N. W'AECKEL-, French Investigations of High
- .-- .Buru Effect on LOCA TheeMeoiinchanical Behavior. Pare Two 'Oxidatid and Quenching
-. . - .P ExpeRiments under Simulated LOCA Dorinationswith Higf Bunup Clad Material.
- *.- :.9,*- H.M. CHUNQ. T.F. KASSNER, Embrittlenenti Criteria for Zircaioy ]Fiel Cladding Applicable to Accidt Situations in LightPWater Reactors: Summaoy Rpr - C
-Argonnea National Laboratory, NUREG/CR-1 344. AN1-79-48, January 1980. ---
- .- S. IMTRA, TAKeiSHItEffect- of Gas Pressure onCthe Effective Thermal Conductivity of of.
Paed Beds.
Heat Transfer Japanese Research, Vol 3, No.4, p.13,1974.F2
- 011 026 krlA .s.
i I
1-I I I - -
rMO24 .... ............. .... ............. ~iMc IN" 11............
U .............. ................
sm.- ...........
7.
- 6. ..............
300. ... ........... .... ...... .. ...............
S.
03..... ..... 4-4 ........ ........ .........
4.
- 3. ... . .................
3,,...
- 2. ..............................................
L-I h A 10. M 0 .M . 140.1W.130.3D. .ai 3L4. o0. a 2W 2i 14. la 1323D.io Iarm fedLWOb1Hoct WU= 57 MW CAT-NU ld,6A it c9mow me!Je of kmn 1ala. No tdd rdoaMWL No g~obecii efflec of Wtelacf jx VLX Puv Figure 1. *Figure 2.
I
, mud wn 8J4 BUIA24 I
I ....-........ ACA=A 03). I I mu-IV
.. . ..... . .. . . . 30 .... ... .... . .
S.........7... .
3D. .................. I I
.. .... .. . I
- 10. II
.2.. ... .. . .. . .t I:. . . . . . . . .
I I
& a0 40. o0. C i.la 120 140.W12 0 Lary Bmde LOCA IlbtRod. YW=97 (Vj/IU PkWyB M~LOc. Hot W W = 9MADWa NOt ~L dcctm~ No PMW M efeof k~iM ad& j 1-3 No Weatocabon No prFrc effc of Aa u d& III P&
Figure 3. Figure 4:
011027 OXLTA. we
I I
,- t - I a-mw 1EU .I.tfJ 7M24 I
X~
IjM2=
- a..
I 200..
I:
14 6z
.4 2
'D. 2D. il M _.IM 0. 140. -IM. 180. 2M.
-c. 0. 4o. a). I].- 2w. am. Ila. )W.. Im. zu.
. t WU, , CA1HAM LB. LOCA. t~ IbdP, Cd:
LB:LOCA.- HoLRod.. Oz" =.9 M*G VII PEN
. . . Figure 6.
- Figure 5: ::
i1,W-
. Et-2 ErAZ4-
.22
,uEb&.
- 1. 12)
.16.
t.
331 Ii2 to.
8.
- 6. .2501&~
.1 I
- D..3~0~3.3 2J3.4 313~
TI'4M lo2m I t.Z .T. - . lt9. 10,20 LB LOCA. M-Rod.hjmW.S7 GVdQl 'AF ~LOXA. U odBimw,=9GVd/W-lC1U
.Sot i-~Lm,6. f . .wlc~ ja ofbdb vI ti k
__________________&IM Figure 8.
... . . igui. tI-?
1n 1p0 MOXLrA .S*O i
I
I i4 ICEM- 221M EDU EDIo& Cl)
U) rn 2i 1TL24 I UL FMlael 15 ..
V. ,:
10.
Ti
.7 ... :
71S(EM Im. 14U. IUL Im. za U. W. 41L, 61 IUU. 2UEL'laX i4a.. 195. 195. in LB LOCA.-Ht Rod. imip= 57 GVd/LJ fCAMARE2 ILB LOCA. Ikt- Rod. &irrup =57 W~dltU Fuel 1210ciU2 inripUms resh. WLO=r. of fillkig ratio jVIA. PE FW. w.tlba kmnobarn veh. kamice of 1IkbV tajo
- .. Figure 9. Figure 10.
kts LXX iudress (fs1am EDETA24 SM.
4m2. . . : . . . I . .
4W. I I
Z00.
I 1SO. i
.... I.. I .
U. 20. 40. 21. 21. 1D. 120. '140. 150. 210. 2M.
LB LOCA. Ho.Rod: mup =57 GVd/tU CA11WM Fia31 vlcainin rlixza ffedLk ilnfL e of tilrg nbo V i.1Pvv Jigurb II.. -. -
011029 maOxLYA.$
I I
I
4 ., -
. . . . I as I 1 1. . .
I . I
.. I
'I I.
- A 1, U
. : C. GRAG.DJEAN, G HACHE; C. RONGIER
- . *... ... .PSN, C da cdhe4Fratice - I
',OECD Tpici Meeting 6n OCA Fuel Safity Criteria,
..* ::. - Ax-en-Proveimce; March22-23, '2001
.I
.1 .0 ,
High Burnup Fuel LOCkCalculatlons to Evalanate the Possible Impact of Fuel relocation after 3hirst
- : ;.;.'BACKGROUND (1):
.: ,X .. ;' '
Mainfmindi gs were provided by thr gesults of:.PBF-L£C,FR2, FLSII enmenb: .
- .:fuel relocation was observed ini. all irradiated r6ds as a slumping 6f fuel fragments from upper:
- lcations into the'swollen region i .fuel m'ovemnbht initiation occurred at burst in E3. aid E4 FR2 tests
°fiul mnotion; (fav'oted in FR2 due to non closure. of gap) is supposed to be counteracted by a.tight
.fuel-clad bonding ..
. . .. :. .bonding was not obseirved on FLASHS (50 GWd/t) despite loO clad ballooning (16%o)
- Important issue
- :. . . ...
Fuel relocation i increases local power and reduces drastically pellet-clad gap
- t impact On Peak'CladTemperatu're ;:and Oxidation.Rate?
- imporiance
- U'O2 at BOL, MWOX:at EOL : .: .
- C. CGrtnda . * . .* * * . * . * . .. OECD TbplcadMeeting on LOCA FuelSafetyCriteria A*.en-Provence. 22-23 MarcA, 20J
s: t
. hi Id 0 39
' I-,.
F I
l:
- 4
- I
~. 0Z I
1~ 1'n . - ...
,xJ L ,.1..... 1 .. *
- Pie.Trinslant. P Tralent .
Heutton. radiographs of rod Fl .(burnp.20 OOO MWd/tj,):
- I I
Compari§on
.p etwe-e~n e
, statuspre'translent andd post-transient
.: I i
iII
. . . I t
II t
i
I-'
- .igh Burnup Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel relocation after Burst
- BACKGROUND (2)
- BERGQUIST (Sweden, 1978-79): Parametric transient calculations with TOODEE-2 code
-impact of fuel relocation, after clad ballooning and burst / referenme case without relocation
- ma'in sensitivity to peaking.factor(Fq) and den'ity of rel6cafed fuel. (Preloc)
Results:*. r~ef case Aq2.32). :w/o relocation - PCT. - 2000TF = 1093°C.: . *
- I 0
relocation, Fq =2.09, Preoc;50% Piheor -> PT -2050°F 1121C. .1 ielocation,'Fq =2.09, Preloc = 60% Ptheo~r - Tcad 71 above 2150 0 Fand subsequent escalation I- W
- YACKiE (INEL, 1586): Stadgy. stateltlermal analy.sis of a fuel rubble in clad balloon
. - fuel relb6ation ratio.e xirapolated from'PBF-LOC experiments
- relocated fuel modeled as a serie' 6f 7, concentric nodes with stagnant steam gaps
.powier' ANS deca heat at. 100 s;Tflht radial power profile ;
. Results: \vorst case.: 89% ciadding strain 4160% fuel redistribution
> TCiad 1320 K (+ 225 ,) and _Tent iuel = 2450 K'(+1270 K)
- Conclussion: eftul relocation- not -aproblein since boih T are wvell betow melting points!!
C GraiiJean * . , .: . . , . OECD Topical Meeteng on LOCA Fuel Safety Criteria
, . ,; , ! .Aix-en-Provence. 22 23 Marck,2001
-Igh BurnupFuelLOCA CIlculations td Evaluate.
- ._., .. the Possible Impact of Fuel relocation after Burst..:
', ' , A .. . . . , .. '
.IPSN Calculations: Large Break LOCA calculations with irradiated fuel rods
- Codeversion' CATHARE2:y1.3L with.specific modifications to: . .
> simulat6. fuel accumulation in the.ruptured mesh after burst
. c.alculate oxidation rate ECR and P3-Zr.remaining thickness C
GCalculationProcedure I- C
.C: . 15 step whole CATHARE2 LOCA computation run withbut fuel relocation '
- .W provides the hydraulic conditions for folloWing c0ulations s
+ step stand-alone fuel modu1e (CATHACOMB) calculations.
- > unddr imposed hydraulic conditions retrieved. from previous step . i
.* . i'without fuel rlocati6h (reference case)
- - . > with simiulationi of fuetrelocation.after.burst, according to uier's input.characteristics for I i
- <. the filling of the tlad balloon . . . .. .. :.
C.GrandJeah . . * . . ,. OECD topicalMetting on LOCA FulSafet Criteria.
- . , . .* - *
- A tix-enProvence, 22-23 March, 2001 .
. 4 1 - . - .
I . . . . .
ml . . .
F-,
.1.Hlgh Burnup FIeA LOCA Calculations to Evaluate I ,. the Possible Impact of Fuel relocation after Burst saiM O.UEE Vhole.CATHARE2 standard chlculation.i.
> large break LQCA (double endedbreak on cold leg)' : , ', i
- I t > mean core rod: fresh fuel o
- hot assembly rods.:. irradiated fuel 57 GWd/t I f w Photrod fPmeancorerod 1.28 (1.55 it BOL) Fq -1.94 (2.3 at BOL) ,I, , 1
> at accident initiation:*. :
- core at '102% of nominal p6wer,
- . cosine axial power profile,. . .**. .
- pellet-clad gap closed in irradiated rods,
- . rod internal pressure Thn= 15.6 Mpa C. GrandJcan . OECDToplcal Meeting onLOCA Fue[Saf'ryCriteria AIx-en-Prownee, 22.23 Marchi 2001
'I.
II a
- High Burnup Fuel LOCA Calculations to Evaluate
. . the Possible Impact of Fuel relocation after Burst
- an.'
Stand-alone .CATHACOMB calculations:
- siuppression of the protectivY cffect 6f initial oxide scale
.(ccording to the results of TAGC1R experimentson irradiated cladding)
- >. hoinogeieousfilling oftIheballoon' g i,, . Pilling*ratio (.1- void ratio):
. .:4. +asecase: 61.5% (.va1ifik'easuredinFR2 pement E5)
- I
.* . sensitivity.stuly 40% anidJO%
.fuel fragmenits issimijated to spherical.particles in cojntact with the claIddingvwall (res. gap - 1pm) .
I
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- I IMURA/YAQI model, validated against the DCi eperiment (SNI.4 C. Grandjean , . , :.. O!ZCD Topical Meeting on LOCA FuelSafety Citeria
- .* , ,>. , *. * .. . , *......' ' AIx-en-Provence, 22.23 March, 2001
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4 EXHIBIT F
d £ SEGFSM Topical Meeting on LOCA Issues Argonne National Laboratory, May 25-26, 2004.
USE OF CATHARE2 REACTOR CALCULATIONS TO ANTICIPATE RESEARCH NEEDS V. Guillard', C. Grandjean, S. Bourdon, P. Chatelard Institut de Radioprotection et de Suret6 Nucleaire IRSN/DPAM: BP3 - 13115 St Paul-Lez-Durance, France valia.muillardairsn.fr ABSTRACT To analyze the consequences of the introduction, in Nuclear Power Plants, of advanced fuels at high bum-up, decided by most of the utilities in western countries in order to reduce the fuel cycle costs, IRSN has initiated a research program focused on the study of such PWR fuel behavior in LOCA conditions.
A first step of this program, comprising analytical and experimental parts, has been to identify the main physical phenomena, linked with thermomechanical behavior of irradiated rods in bundle geometry, to be taken into account in reactor safety analysis: cladding deformation and flow section restriction in bundle geometry, mechanical interaction between neighbor rods or structures, axial extension of balloons ; cladding oxidation and secondary hydriding ; fuel fragmentation and relocation, balloons filling rate and FP release ; fuel rods thermal behavior in bundle geometry during reflooding conditions, rewetting of the claddings around ballooned regions with fuel relocation mechanical resistance of irradiated claddings in post-quench conditions.
This paper summarizes an analysis of sensitivity calculations performed with CATHARE2 "Best-Estimate" code, used in France in the frame of realistic methodology to evaluate safety margins.
The objective of these calculations is to point out, among parameters affecting last-mentioned phenomena, those for which taking into account basic uncertainties lead to important uncertainty on global code response (Peak Cladding Temperature, oxidation rate ...). That is the case of fuel relocation phenomena, whose impact is highly dependent on parameters such as, in the example of LB LOCA transient, cladding radial and axial deformations in bundle geometry, burst criteria, balloon filling rate, thermalhydraulics around balloons. A lack of knowledge on theses parameters for irradiated U02 and particularly MOX fuel may lead to reduce safety margins.
This study may provide some elements to identify future research needs to complement present experimental data base, reduce uncertainties and develop more realistic calculation models, which may better fit the thermomechanical behavior of advanced irradiated fuels.
' Corresponding author
ESMQPCAL MEE.TIN ONQA ISSU.ES ~
Argnne ..N'ationI abroyMa. 5 27, 2004 USE OF CATHARE2 REACTOR CALCULATIONS TO ANTICIPATE RESEARCH NEEDS V. GUILLARD, C. GRAND JEAN, S. BOURDON, P. CHATELARD Institute for Radiological Protection and Nuclear Safety (IRSN)
Major Accident Prevention Division I Fuel in Accident Situations Department BP3 - 13115 St-Paul-Lez-Durance Cedex - France Content of the presentation
> Introduction
> Main physical phenomena to be modeled
> Main hypotheses of the calculations
> Main results of the LB LOCA calculations
> Conclusion and perspectives calcuclations ,,to,anticipate";re~searchneed bAIIR2-ecto q~j.~ ~TD.~ei.,~~OCA~issues;.
IR I ANL, May 25-27, 2004
El General Background El CATHARE2 ... j.! -',.,;'
,f<,"4
> French thermaihydraulics system code with CATHACOMB fuel module 'i .
¢ A. A%A.
'z
- _ 1; _W an El Objectives s'r1-ltj- .' " '>4Qs'
. F
> Identify future research needs for new generation of fuels ¢
> Improve knowledge, models and calculation methodologies E
uApproach used to reach this objective
> Identify physical phenomena involved in thermo-mechanical behavior of advanced irradiated rods in bundle geometry (State Of the Art by C. Grandjean & G. Hache)
> Take them into account in the modeling
> Quantify basic uncertainties (CIRCE tool and sensitivity calculations)
> Evaluate global uncertainty on CATHARE2 code response (SUNSET tool) 4 d ATAe. sdtbou ndary conditions for further more precise nationsp and analysis under LOCA conditions.
Dlg~biNS,". Use..PI.GAofHA.? r Jiotns anticipate researtch needso t o valia .0UI Iard(~irsn Jr - - -ANL, May 25-27, 2004
b;; - '.' . "' , ! -, , , .. . :
BE MODEl
! A.
Reference:
C. GRANDJEAN &.G-HACHE:
"LOCA Issues Related to Ballooning, Relocation; Flow Bl3 `6ckago.fand abi ity.
Main Findings from a Review of Peast ExperimentalP~rograms.>
Temperatures ('C)
, - Fuel I in -Clad T)tPICAL LB LOCA TRANSIENT EASY NOT SO EASY Oxidation. kineticsA....
Fuel fragmentation and secondary'hy~idridng.
(debris size, 1206 granulometry, / *. ......... ff-., {
porosity)
.PI Clad coolability 800 around ballooned regions with fuel relocation Clad Fuel rods thermal behavior ballooning ...
. , Flow res during ref looding conditions Mechanical
- interactions 40 between neighbor0 rods or structures Axial extension
_X .J~.
u] LUUI IU JI Burst criterion Fuel relocation (instant of occurrence, balloons filling ratio) :
I Wi HEBUS Post-quench
., residual ductility 50 100 150 Time (s) 1 Fuel 1l Clad X Bundle tY t.
"tciaeresarchn eeds~ SEGSM K Meek olLOGCA Issues
>P...' , _
vali .ciil ard(irsn.frANLI May 25-27, 2004
El Calculation of several transients including Large & Intermediate breaks LOCA 0 Use of standard CATHARE2 versions (V1.3L, V2.5)
El Basic uncertainties on CATHARE2 models taken into account
> Consistent with PIRT implications El Zry-4 cladding without hydrogen uptake effect on mechanical properties
- .. C Hypothesis linked with the use of irradiated fuel at high burn-up
> Modification of thermal properties
- conductivity and thermal capacity laws from SCANAIR code
> Deletion of protective effect of initial oxide layer on transient oxidation
. outcome from experimental ANL and TAGCIR program analysis on irradiated Zry
> Introduction of Baker-Just correlation to calculate oxidation rate El Initial state of irradiated rods given by METEOR code
> Gap width and pressure, radial and axial fuel power profile, external cladding oxidation profile, cladding thickness ...
Aai Example: l s o impact of irradiated fuel relocation Larue;Brak'conditions on a PWR900Mwe
~L~i4E valia~uilard~iirsn.fr; -::ANL, May 25-27, 2004
El Main hypothesis Balloon filling rate by relocalized fragments
- Relocation mesh 90 I
= burst location mesh filled by fuel debris 80O rund value from upper meshes 70-
- Filling ratio of the balloon
= function (balloon size) 60* 0 0 x A A
- Balloon size derived from - 50 CATHACOMB calculation of 0 mean elongation ' 40' o PBFILOC-gammascanning a-
- PBFILOC-micrographies Basic calculation value = 61.5 %
- Modification of the power factor = 30 A FR2 and the fuel mass in the 2L -Upper bound relocation mesh 20-
- EdF calculation X IRSN reference calculation
- Considering relocated fuel as 10 + IRSN sensitivity calculation a porous medium (Imura-Yagi model for the conductivity calculation) 0 20 30 40. 50 60 70 80
- Flow blockages: impact on T/H and fuel cooling not taken into account Ballooning (%)
c at SIt c eSEGFSMT T M ee Issues
,ANL, May 25-27, 2004
High sensitivity of the relocation impact on PCT
> Burst at 24 s on stress criterion deduced from EDGAR experiments (s at rupture = 57.5%)
LB LOCA - Irradiated U02 Hor rod claddinS temperaiture tiburst nesh 1200.0 1100.0 ationhimpact depends on 100.0 ie>which determines the energy 900.0 uo.'in the burst mesh izea1nd filling ratio, 800.0 IF :e~rjminies relocated fuel mass 700.0 600.0 I-500.0 400.0 H CLAD BURST FUEL RELOCATION 300.0 200.0 Withlut relocatbin Wih relcatin (filing ratio: 61.5%)
Wth rlocatbn (flling ratio: 70.0%)
2, 0 0Cto 150 0C on PCT 100.0 i fiding tbaitoonfilling ratio:61.5 to 70%)
- 0.0 . . . . . . . . . . . . . .
0.0 20.0 40.0 60.0 80.0 100.0 120.0 14Q0 160.0 180.0 200.0 Time is]
u-rDiC Ue; ~ A HA E?x a torbcalcflla tionsi tobbc~ at 51 I~irs~frA ~ L~~ va~a~cui~ard(
~~~ rW niedds' { SEGFSM p Meet.'- n.I$.L OC~ s u s NL, May 25-27, 2004
RE AU' CA
.Main resu1tsJforhicj burnsuU2ie g'z2z2 7 El Fuel relocation impact on ECR and residual p-layer m Burst at 24 s on stress criterion deduced from EDGAR experiments (s at rupture = 57.5%)
LB LOCA - Irradiated U02 LB LOCA - Irradiated U02 Hot rod o.xidal tion rate tzburst inesh Hot rod beta-layer thickness at burst incsd 2U.0 . - -. E00.0 . , * , * * '
18.0 550.0 01 500.0 t'- ~~160 vS0.0 450.0 14.0 W 400.0 C,
- c10.0 L) 200.0 5.0 150.0 4.0 - Without relocation
- With relocation (filling ratio :61.5%) 100.0 - Without relocation
-- With relocation (filling ratio :70.0%) - With relocation (filling ratb: 61.5%)
2.0 50.0 -- Wiith relocation (filling ratb:70.0%)
0.0 0.0 20.0 0.0 . I . . t .
40.0 60.0 80.0 100.0 120.0 140.0 160.0 100.0 200.0 0.0 20.0 40.0 p
60.0 50.0
. I . p .L 100.0 120.0 140.0 160.0 180.0 200.0 Time 15] Time [s]
% t % on ECR r i674m,,on residual.p-layer thickness
,ofilling ratio: 61.5 to 70%) p ding balloon filling ratio: 61.5 to 70%)
.H . Use o,-C, IA4Etreactorcalculati-A_ s to anticspatesreearch needues
. . LU.I valia.guillard ' . '.' : ANL,May
-'. 25-27, 2004
O Fuel relocation impact on PCT
- Burst at 18 s on stress criterion deduced from EDGAR experiments (£ at rupture = 58.7 %)
LB LOCA - Irradiated MOX Hot rod clidding £eJnpetrature gjrburstinsh A 'nia J-Ir'I .
lb lI*ZXU . Enf'npg~ffuel relocation impact CLAD BURST -
FUEL RELOCATION J
- Without relocation I
- With relocation
.1.1.Ii. 16OC on PCT Time [s]
- - E I ' valia.Quillar (i~~rANL,: May 25-27, 2004
El Fuel relocation impact on ECR and residual P-layer
> Burst at 18 s on stress criterion deduced from EDGAR experiments (Eat rupture = 58.7 %)
LB LOCA - Irradiated MOX LB LOCA - Irradiated MOX Hlot rodl oxtiation rate titburstmnerh Hot rod betw-ltyer residul thticknerr CSat burtt Inesh
. . I-
. L
. . L Time is] Time [s]
r arge ugnecertainty g ,h,er.impact .bonhydrogen uptake effect "'Need to reduce Nn,&ECR,,and residual ,B-layer -an'd'iie'hanicaI properties untcrtainties 2I !,U ;> .;
pend FlTa. g on I .s~~si Ingalloy o ... clad
-A.c'ricaucuaaticnsn:. t-o a.n - .. EG2S M AtNM2ipt27, va~i~pui~ar(~irn~frANL May25-7, 2004
El Use of CATHARE2 code for high burn-up fuel analysis under LOCA conditions
> Example of Large Break transient
> Emphasis on fuel relocation phenomena impact
> Main results: PCT and ECR increased El High uncertainty on global code response due to identified lack of knowledge
> Instantoffuel movement - HALDEt?
- imposed as burst time in the simulation
- depends on clad ballooning/deformation and burst criterion
> Balloon size - HALDENSW 4
- which is also linked with ballooning/deformation model
- Filling ratio FR2, PBF, HALDEN? ,
> Relocated fuel properties
. fragments size, granulometry, porosity, conductivity, .. ANL, HALDSVEN-)
> Bundle effects Need of:integrts Axial extension of balloon Flow blockages Clad coolability around ballooned regions with fuel relocation T,,.,THARE2 an"iiea,.researchpebd<zEGFSM To6MCs, L1Ivai.uIRd rn~ . .ANL, May 25-27, 2004
- V
~I~ii; SPETIVES.'
AC Perspectives
> Modification of clad mechanical properties to take into account hydrogen uptake effect
> Study to be complemented by calculations using CATHARE I FRETA A rods 3D thermo-mechanics
- rod-to-rod interactions models
- cooling and reflooding models for overall bundle
> Use of NEPTUNE 3D local module for flow blockage cooling calculations
> IRSN plans to develop a new code for fuel LOCA calculations
! - GG: H ache) + Analytical studies l gep mprovements.
- -pn'rradiated rod .behavior-under.- KLOCA conditions De opm ewgeneration.of,fuel and bundle models
- .
. 'educe uncertainties !andjgain ne~w rmargins . .
. vaia ANL, May 25-27, 2004
I f EXHIBIT G
I NUREG/CP-0176 i
Proceedings of the Nuclear Safety Research Conference l
l Held at Marriott Hotel at Metro Center l Washington, DC October 22-24, 2001 Manuscript Completed: March 2002 Date Published: May 2002 Compiled by: Susan Monteleone, Meeting Coordinator S. Nesmith, NRC Project Manager Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Proceedings prepared by Brookhaven National Laboratory
Need for Experimental Programmes on LOCA Issues Using High Burn-Up and MOX Fuels Alain MAILUAT, Michel. SCHWARZ Institut de Protection et de Sfjret6 Nucl6aire DMpartement de Recherches en Securit; C.E.. Cadarache, France Safety studies performed in IPSN and elsewhere pointed out that high bum up might induce significant effects, especially those related with fuel relocation during LOCA situations. Uncertainties exist regarding how much the existing safety margins associated with peak clad temperature, clad oxidation, core coolability, clad residual ductility can be reduced by new fuels like the MOX one, bum up increases, the arrival of various alloys for fuel rod cladding. A better knowledge of specific phenomena associated to fuel effects is required in order to estimate the new margins and to resolve the pending uncertainties related to the LOCA criteria.
Therefore, in addition to the programmes currently planned in the Halden reactor, IPSN is preparing the so-called APRP-irradie (High Bum up fuel LOCA) programme. One of the important aspects of this programme is In-Pile experiments involving bundle geometries in the PHEBUS facility located at Cadarache, France.
429 I
- 1. INTRODUCTION 1t .'I In France and in other countries, a permanent evolution of the light water reactors (LANR) is
,rs observed since the seventies. The -evolution deals with the reactor designs (900 MWe/3 loops, 1300MWe/4 loops, N4, future EPR). It is also related to the fuel management and bur-nup increase (3 cycles, 4 cycles, 39' GWd/tU, 47, 52, 60 GWdItU in the next future). This evolution affects the fuel itself (U0 2, MOX, Gd fuel), the cladding (Zircaloy, Zirlo, M5) and the control rods 1,
(Ag-In-Cd, B4C). As a consequence of these modifications, there is a permanent need to reassess the reactor safety studies which implies improving the associated knowledge and 1" upgrading the corresponding calculation tools. Such a need is not specific to the French
¶ situation. For the studies associated with the continuous evolution of the reactor operation, the safety authorities requirements are both related to the design basis accidents and the severe accidents. They have to appreciate to which extent their analyses and criteria might be modified by the bumup increase and the type of fuel. In France, under safety considerations, it was requested prior to any generic authorisation of discharge bum-up extension, that the high bum-t up fuel behaviour be validated, with the support of appropriate R&D tests results, under accidental conditions, particularly under Loss-of-Coolant-Accident (LOCA) conditions.
The current regulatory safety criteria for LOCA, still in use in most countries, are derived from the ECCS acceptance criteria that were issued by USAEC in December 1973 and published in 1' the Code of Federal Regulations (10.CFR50, part 50.46) as "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled-Nuclear Power Reactors".
The criteria are stated as 5 TYPICAL LOCA TRANSIENT requirements, concerning the calculated performance of the cooling system under the most Tempahtrcs I1&.'2 severe loss-of-coolant accident conditions. A summary of these 1am conditions is given on the figure 1 against. These first two 1/2:
ii 4" requirements address: the peak cladding temperature (PCT) which shall not exceed 1204'C and the I maximum cladding oxidation rate, 4W defined through an equivalent cladding reacted (ECR), which shall nowhere exceed 17% of the cladding thickness before oxidation but after cladding swelling with or Figure 1 without rupture.
- 1' .¶i
' Mean value per assembly 430
....... MVWA The third request addresses the maximum hydrogen generation, the total amount of which shall not exceed 1% of the hypothetical amount generated by the reaction of all the metal in the cladding surrounding fuel. Finally the last two requirements are related with core cooling. The calculated changes in core geometry shall leave the core amenable to cooling and after any operation of the ECCS, the core temperature shall be maintained at an acceptably low value and decay heat removed for the extended period of time required by long-lived radioactivity.
- 2. UNCERTAINTIES AND PENDING ISSUES In the aftermath of the AEC LOCA criteria release, numerous studies were undertaken world-wide in order to improve the basic knowledge of the physical phenomena intervening in LOCA transients, so as to allow a better prediction with realistic models. Beyond the numerous experimental investigations that were conducted on unirradiated rods or cladding, either in-pile or out-of-pile, there exists a few number of available results of such experiments with irradiated material. Following is a very short review of the current knowledge on clad and fuel rod behaviour gained from experiments on irradiated material, that will introduce the pending questions and critical issues for irradiated fuel behaviour in LOCA.
2.1 UNCERTAINTIES 2.1.1 CLAD BEHAVIOUR An important progress in knowledge relative to irradiated clad behaviour has been obtained from the results of the French EDFIIPSN [1,2] program (TAGCIR and HYDRAZIR tests),
addressing the oxidation kinetics and quench bearing capability of irradiated zircaloy. The main outcome concern:
- the protective effect of corrosion oxide scale;
- the oxidation kinetics of irradiated zircaloy;
- the resistance to quench loads of irradiated zircaloy;
- the effect of high hydrogen content, as a result of internal hydriding during LOCA transient.
Relative to oxidation kinetics and quench behaviour, a comprehensive understanding of all involved phenomena and of their inter-related influences is not yet achieved and leaves still pending questions, most of them being not specific to high BU fuel. One important question is the influence on clad quenching resistance of axial constraints that may result from differential contractions upon quench between guide tubes and a fuel rod blocked in spacer grids as a result of ballooning or metallurgical interaction. Such blockage consequences had been evidenced on past tests at JAERI [3] on unirradiated rods and should therefore be expected to some extent on irradiated rods.
431
21.2 ROD BEHAVIOUR There exists a few number of available results from experiments with irradiated fuel rods under LOCA conditions. The main outcome were found in results from the PBF-LOC tests[4,5] in the USA, the FR2 tests[6] in Germany, and the FLASH5 test[7] in France. They concern the fuel relocation process and an increased cladding deformation.
FUEL RELOCATION All the available tests performed with irradiated fuel rods experiencing LOCA conditions have shown an accumulation of fuel debris in the swollen region -called balloon- of the burst cladding which resulted from fuel fragments slumping from upper BodaEU locations (see figures 2 and 3 :z:
below from FR2 results). Figure 2 This process, here after called fuel relocation, is initiated at the time of the cladding burst, as demonstrated by the FR2-E3 and E4 tests. It is thought that the driving forces are both gravity and the pressure difference between the rod upper plenum and the channel.
This fuel relocation is not counteracted by the fuel-clad tight bounding, which exists with high burnup fuel.
Indeed, the process was observed in the FLASH-5 test with 5OGWd/t fuel in spite of a rather low clad strain (not higher than 16%). It um is thought that the cladding temperature Spnng increase combined with its ballooning suppress -M
-at less partly- the bounding making possible j -.
the fuel relocation. Hpedzrne Finally, fuel relocation process is not specific of high bumup fuel. It was also observed for fuel l rod having a bumup as low as 48MWdIt (LOC5-76 test). _ =
AN INCREASED CLADDING DEFORMATION In PBF-LOC experiments 2 couples of rods (2 -
unirradiated + 2 irradiated) were simultaneously tested in the same test train.
Available data for comparison, although in very z limited number due to technical problems, r-clearly indicate significant differences in the Z Rod ft deformation behaviour of irradiated versus unirradiated rods.
Figure 3 432
2liiiETI o A higher circumferential rupture strain for irradiated rods (a factor greater than 2 relatively to unirradiated rod strain for maximum values) and more axially extended; o A wall thinning affecting almost all the circumference of irradiated rods, thus indicating low azimuthal temperature differences as compared to unirradiated rods.
These differences in behaviour have been attributed to the lower temperature differences on the clad of irradiated rods, circumferentially and axially, as a result of the pellet-clad gap reduction due to clad creepdown during rod irradiation.
2.2 THE PENDING ISSUES A better understanding of the specific phenomena shortly mentionned above leads to raise a list of some complementary questions related with rod behaviour, fuel relocation process and coolability issue during LOCA transients.
2.2.1 ROD BEHAVIOUR The question mark about rod behaviour is related with the influence of hydrogen pick-up and other irradiation effects on ballooning, burst behaviour and embrittlement during reflooding which were not considered when 10CFR50, part 50.46 was released.
2.2.2 FUEL RELOCATION Several questions are induced by the relocation process. The first ones concern the process itself. The needed data are the following ones.
o Instant of fuel movement at high bum-up, with possible delay due to fuel-clad bonding.
a Filling ratio of clad balloon at high burn-up, with fragmentation of U02 rim or MOX clusters a Impact of the relocated material on steam access inside the balloon and hydrogen uptake rate.
The second set of question marks concerns the consequences of the relocation process.
o Which are the effects on peak clad temperature and final oxidation ratio of the local increase in lineic and surfacic powers and of the local decrease in fuel-clad gap resulting from fuel accumulation?
'Note that these last issues are particularly important for end-of-life MOX fuel for which power generation is not reduced, unlike for U02 fuel.
433
2.2.3 COOLAB /TY Related questions should be considered additionally, relative to flow blockage behaviour of highly deformed cladding with possibly relocated fuel and the embrittlement potentials i iassociated to fuel fragmentation. The 90% value for flow blockage still coolable, as derived from results of flooding experiments (FEBA, SEFLEX et al) on unirradiated rods arrays is questionable since these experiments did not take account of any fuel relocation and associated effects. The needed infomnation are the following.
o What is the maximum flow blockage ratio that leaves coolable an irradiated rods bundle?
o Does the maximum flow blockage ratio attainable with an irradiated rods array remain below the maximum coolable value indicated above?
There is presently a complete lack of data allowing to answer these questions.
o what flow blockage configuration would be worst coolable with occurrence of fuel relocation?
In other words, is the coplanar flow blockage still the worst coolable case?
- 3. THE IPSN APRP IRRADIE PROJECT For many years, IPSN and several other safety organisations have applied a three-tier method for their reactor safety researches. The first step consists of computer code developments from the existing data bases. The second step involves small-scale, out-of-pile experiments, which provide the additional data bases requested by the code developments and their preliminary assessments. But, as the reactor phenomenology cannot be totally reproduced in such small
[- scale experiments, a third step consisting of integral in-pile experiments using real materials is essential for comprehensive accident analyses. Their results allow the final code assessment in terms of reactor applicability and simulation completeness. This in-pile part of a programme 4t Iassures that the investments done for code developments and small scale experiments will produce profits in terms of reactor safety. This three-tier method is applied by IPSN for the
- ', 1,various research programmes devoted to reactor safety, design basis accidents including RIA and severe accidents programmes.
434
Regarding the LOCA issue, the current testing programmes dealing with irradiated material only involve out-of-pile experiments: separate effect quench tests on irradiated cladding (TAGCIR tests) in France; tests on irradiated cladding and integral type experiments (ballooning / burst /
oxidation I quench) on irradiated rods at ANL (USA) [8] and JAERI (Japan) [9] with the support of an important programme of mechanical tests. In addition, OECD has planned an in-pile programme consisting of some single rod geometry tests with irradiated fuel. The programme should be conducted in the Halden reactor and should provide information about the relocated fuel characteristics.
But these programmes will not solve all the previously mentioned uncertainties because these ones are mainly associated with the combined behaviour of fuel and cladding under representative conditions of the reactor evolution during the LOCA transient. Based on the long fruitful experiences of a three-tier method, the so called APRP Irradi6 programme, providing the in-pile experiments third tier, should provide the missing part of the data bases required for code assessments in terms of reactor applicability and simulation completeness. This programme is prepared in a coherent way with the present international efforts in order to validate, and possibly update, the results obtained from separate effects tests and previous limited in-pile tests.
3.1 THE MAIN EXPERIMENTAL OBJECTIVES The main objectives of the in-pile experiments will be to investigate the behaviour of fuel and cladding with conditions representative of the reactor during LOCA sequences. The main factors that will be accounted for are:
o the nature of fuel (U02, MOX, Bum-up),
o the fuel-clad thermomechanical coupling (i.e. fuel relocation) o thermal azimuthal gradients (main factor affecting cladding strain and blockage ratio) o thermal-hydraulic aspects (i.e. quenching, coolability of blocked arrays) 03 3.2 TEST DEFINITION RATIONALE The following analysis provides the rationale for the APRP-Irradi6 programme characteristics. It is shown that the conditions for having representative data for reactor applications are both in-pile tests and bundle geometry.
3.2.1 NEEDS FOR IN-PILE TESTS The in-pile test need results from three reasons.
Neutron flux provides the unique way to produce the correct heat generation in the fuel fragments, corresponding to the residual power, whatever are the relocations induced by the ballooning and/or the burst of the rod. Both. the exact amount of heat generation in the balloon and the heat exchanges with the rod channel depend on the characteristics of the relocated fuel fragments, their size, shapes, and compaction ratio. This heat generation correctness is one of the main conditions for having realistic estimates of the relocation consequences in terms of equivalent clad reacted, peak clad temperature and hydrogen uptake inside the balloon. All 435
i these aspects impact the strength of the rod during the quenching phase and the residual gP1i Xductility of the rod after the LOCA transient.
During the blowdown phase of the LOCA transient, there is much less heat generation inthe fuel and the clad coolant heat transfer is drastically reduced. Therefore, the fuel-stored energy is redistributed in the pellet and the cladding. Simultaneously, within a few seconds, this redistribution produces a decrease of the pellet center-line temperature from 1500'C down to, say, 10000 C and an increase of both the pellet rim and clad temperatures'from 300'C up to 10000 C. Due to these temperature transients, the.central part of the pellet will experience a contraction while the rim and the clad will undergo an expansion. Fuel mechanical stresses and fragmentation could be induced by these adverse effects. It has to be kept in mind that during usual experiments, for which a blowdown phase is not reproduced, clad and fuel temperatures are simultaneously increased or decreased without producing any comparable therrnomechanical transient. In-pile tests including a blowdown phase provide the way to get a definitive answer regarding the additional fuel fragmentation prior to the relocation and how much this refragmentation process affects the amount and the characteristics of the relocated fuel.
Finally, during reflooding and quench process studies, in-pile tests allow to maintain the heat generation in the fuel corresponding to the residual power. By this way, more representative conditions of the thermrnomechanical loads of the rods are provided. Without such a' power during i,7 . the reflooding phase, steam production and cladding oxidation are reduced; the temperature transients experienced by the rods are less severe. Consequently, under estimates of core embrittling during reflooding could be obtained.
6.2.2 NEEDS FOR BUNDLE GEOMETRY In addition to the requirement associated to heat generation mentioned above, bundle geometry is a second important condition to produce realistic data. Relocation being closely associated with the volume which is made free by the rod burst, it is clear that a correct amount of relocated fuel will be produced only if the sizes of the balloons are representative of the reactor conditions. Such balloon sizes can be obtained -as explained below- only with bundle geometry. This is the reason why these tests are essential and complementary of single rod tests.
During the early stage of the LOCA transient the fuel rods experience the ballooning and burst zt processes. For such phenomena, bundle geometry is a necessity to get a correct azimuthal temperature field around the fuel rods since this field is crucial to produce a realistic balloon size. An illustration of the impact of the azimuthal temperature field on the stain at burst is given by the results of single rod tests with heated or non-shroud. A uniformly heated shroud reduces the azimuthal temperature variation around the rod. For such tests a higher rod deformation is obtained. Conversely, an unheated shroud tends to increase the azimuthal gradient and, therefore, leads to small rod deformation at strain (see figure 4 below).
,i*t '-;
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- 0. :..436
120 -, A +
Red sym=L .
A heated shroud BURST STRAIN % ..
100 - ' 7ARM.,
STMkD 4
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A A. ......__
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.. ... ...... .. . < ^ribols SMALL
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- +AmbaIs n*n heated sh-od BURST TEMPERATURE OC 0 4-600 700 800 900 1000 Figure 4 Additional reason for a bundle are the radial interactions between adjacent fuel rods that need to be taken into account because they modify the size and shape of the balloons. Such kind of balloon interactions are clearly illustrated with the side picture (figure 5) from PHEBUS LOCA test 215. Having in mind that the amount of relocated fuel is associated with the size and the shape of the balloon, the picture demonstrates that realistic data will require bundle geometry. This bundle geometry requirement to ensure representative mechanical interactions with neighbour rods was stated in early 80ies - several years ago- in consideration of ORNL MRBT B5/B3 experiments Figure 5
[10].
During core reflooding, a bundle is an obvious requirement for reproducing, on one hand, the correct flow blockage induced by the ballooning of the rods and their radial interactions and, on the other hand, the excess of heat generation at the blockage location due to the fuel fragments relocated in the balloons.
437
Finally, such bundle geometry is also necessary to represent the axial and radial stresses induced by the grids and the adjacent rods, which might restrain the rod contraction during quenching.
3.3 EXPERIMENTAL CONFIGURATIONS Since it is hardly conceivable to carry out one type of experiments that will address all pending questions with any chance to provide some usable results, it appears more appropriate to perform two kinds of in-pile experiments, namely separate effects tests and integral tests.
3.3.1 SEPARATE EFFECTS TESTS The objectives of these tests are to address phenomenological aspects, in order to confirm or correct and extent the previous results relative to:
o rod deformation, o fuel relocation, o the resulting resistance to thermal shock loads, with or without effect of clad axial constraining.
These tests should be realised with one irradiated rod within a ring of 8 fresh fuel rods, which will provide a representative thermal environment in order to ensure representative strains and subsequent phenomena. In addition, these in-pile separate effects tests should include a blowdown phase. As mentioned before, this phase will provide representative conditions for the temperature transient inside the fuel to study the consequences in terms of thermomechanical pre-fragmentation during blowdown.
3.3.2 3.3.2 INTEGRAL TESTS.
This kind of tests will address the aspects of:
o impact of blowdown phase o flow blockage o quenching behaviour and coolability.
These tests should allow to check the absence of unexpected phenomena or unexpected coupling between foreseen processes, and finally provide data for the validation of reactor computational tools.
These tests should be realised with 9 high bum up rods with a ring of 12 or 16 fresh fuel rods which will provide a representative thermal environment in order to ensure representative strains and subsequent phenomena. A blowdown phase will be simulated depending on its importance as observed in the previous studies. Finally, additional axial stress during quenching due to rod blockage in the assembly should be simulated during these tests.
438 I,
I 3.3.3 EXPERIMENTAL FACILITY Such a programme is envisaged by the IPSN in the PHEBUS facility where some twenty LOCA tests were run between 76 and 83 [11,14], see figure 6. By this way IPSN would take advantage of the know-how accumulated when the previous LOCA programme with fresh fuel was run.
'I PHEBUS LOCA TEST CONFIGURATION
<'Vt Figure 6 Furinermore, a new LOCA programme in the Phebus facility would take advantage of the R&D efforts made for the subsequent programmes in terms of high activity material measurements Tomography technique [15,16] is one of the examples, which can be given how such efforts provide practical applications for the new LOCA programme. This technique provides the 3D location and the nature of the material fragments everywhere in a bundle. The exact geometry of the bundle at the end of the test can be reconstructed and explored from the inside. Fuel relocation studies and code validation will be made easier through this technique. Figure 7 439 I
[if i Presently, fragment size less than 500 microns can be located (see figure 7). Further improvement of the existing technique will increase the resolution providing several points inside the clad with an oxide/metal discrimination.
- 4. CONCLUSIONS Studies performed in IPSN and elsewhere pointed out that high bumup may induce specific effects under LOCA conditions, especially those related with fuel relocation. Uncertainties exist regarding how much these effects might affect the late evolution of the accident transient and the associated safety issues. IPSN estimates that a better knowledge of specific phenomena is required in order to resolve the pending uncertainties related to LOCA criteria. IPSN is preparing the so-called APRP-lrradi6 (High Bumup fuel LOCA) programme. One of the
- '5- important aspects of this programme is in-Pile experiments involving bundle geometries in the PHEBUS facility located at Cadarache, France. A feasibility study for such an experimental programme is underway and should provide soon a finalised project including cost and schedule aspects.
e .s; iiz .
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444 A'..
I'
'I Or, a.440
-N EXHIBIT H
N NUREG/CR-6744 LA-UTR-00-5079
'.I.
Phenomenon Identification and Ranking Tables (PIRTs) for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High Burnup Fuel Los Alamos National Laboratory U.S. Nuclear Regulatory Commission A Office of Nuclear Regulatory Research Washington, DC 20555-0001
Simulation of fuel relocation Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) rPhenomena (Parameter)
Specimen selection Burnup I Definitlon and Rationale (nmportance, Applicabilityand Uncertainty)
The amount of burnup to which the fuel rod used for the specimen was exposed.
I H(4) Fuel morphology (fragmentation, rim characteristics, bonding, etc.) is important. The nature of the bonding between the pellet and the cladding changes with the burnup increase. It will affect the potential for fuel relocation. The segment bumup level can determine the extent of pellet-cladding bonding and corresponding susceptibility to fuel relocation during ballooning and rupture. Fuel and rim-zone microstructure and the state of bonding with cladding are strongly influenced by fuel bumup.
M(O) No votes.
L(O) No votes.
Fuel: Y(l): MOX agglomerates Clad: N Reactor. N Burnup: N K(l): Data PK(3): Data, Judgement UK(O): No votes.
Appendix D-66 I
Table D-1. PWRk and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) I Phenomena (Parameter) I Definition and Rationale (Importance, Applicability, and Uncertainty)
Simulation of fuel Specimen selection: Composition of the fuel, I.e., a specified MOX composition.
relocation Fuel type (MOX)
H(2) May affect the amount of fine grain material after relocation. Fuel structure and mechanical properties are Influenced by fuel type.
M(1) The consequence of fuel fragments relocation (higher local decay heat and higher cladding temperature) could be more effective with MOX fuel than with U02 fuel. Nevertheless the viscoplastic properties of the MOX should impair the fuel fragments relocation at high burnup.
L(1) No significant differences in pellet-cladding bonding behavior or pellet cracking behavior are anticipated or have been observed with MOX fuel, and therefore no significant differences in relocation behavior are anticipated.
Fuel: N Clad: N Reactor: N Burnup N K(l): Data PK(2): Data, Judgement UK(I): Judgement Appendix D-67
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) l Phenomena (Parameter) lDefinition and Rationale (Importance, Applicability, and Uncertiainty)
Simulation of fuel Specimen selection: Composition or designation of the metal utilized In fuel-rod fabrication relocation Alloy type H(2) May affect burst (beta favoring or alpha favoring additions). Ductile burst and brittle failure by thermal shock and post-quench forces are influenced strongly by cladding alloy type.
M(1) In general, compositional differences have not been observed to significantly affect cladding burst behavior. However, If significant differences In burst behavior occurred, the relocation characteristics could be similarly significantly altered.
L(l) Data show no significant impact of alloy type on the balloon size that could influence the fuel fragments relocation.
Fuel: N Clad: N Reactor: N Burnup: N K(2): Data PK(1): Data, Judgement UK(1): Judgement Appendix D-68 I
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) I Phenomena (Parameter) I Definition and Rationale (Importance, Applicability, and Uncertainty)
Simulation of fuel Specimen selection: Chemical and mechanical bonding between the fuel pellet and the cladding prior to the relocation Chemical and 11test.
mechanical bonding 1-1(4) Fuel morphology (bonding) Is important. It will affect the potential for fuel fragmentation relocation. It is speculated that bonding could significantly affect the relocation characteristics by impeding pellet fragment movement.
However, this effect has not been demonstrated. Major factor that Influences fuel slumping and potential release of fuel particles upon burst and subsequent fragmentation.
M(O) No votes.
L(O) No votes.
Fuel: N Clad: N Reactor: N Burnup: N K(O): No votes. i PK(3): Data, Judgement I UK(I): Lack of data I
i i
I I
I I
I I
Appendix D-69 I
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) l Phenomena (Parameter) I Definition and Rationale (Importance, Applicability, and UncertanLty)
Simulation of fuel Specimcn selection: Crack pattern and crack density of the fuel pellets prior to the test.
relocation Cracking H(2) Controls the rubble bed characteristics after relocation. Degree of fuel cracking directly Influences the potential for fuel relocation and release.
M(O) No votes.
L(2) Beyond a given bumup the number of cracks Is stable. In general the macroscopic fuel pellet cracking pattern develops early In life and does not change significantly with elevated exposures. Therefore, this contribution to fuel relocation susceptibility Is not expected to be a dominant parameter during this test series.
Fuel: N Clad: N Reactor: N Burnup N K(l): Data PK(3): Data,Judgement UK(O): No votes.
Appendix D-70
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) I Phenomena (Parameter) I Definition and Rationale (Importance, Applicability, and Uncertainly)
Simulation of fuel Conduct of Test-During Determination of whether blowdown processes must be simulated in the test.
relocation With or without blowdown H(0) No votes.
M(l) During the blowdown phase of the LOCA transient, fuel stored energy is redistributed in the pellet and the clad. This redistribution produces a decrease of the pellet centerline temperature and increases the pellet rim and clad temperatures. Due to these temperature transients, the central part of the pellet will suffer a contraction while the rim and the clad will experience an expansion. These adverse effects could induce fuel mechanical stresses and fragmentation. The expansion and contraction Inside the fuel pellet may affect bonding and fuel debris sizes.
L(2) Vibration loads occurring during the blowdown phase may cause additional pellet fragment movement. In general, pellet fragments are relatively constrained within the fuel rod by the column geometry, as evidenced by characterization of fuel column geometry in hot cells. Therefore, this effect is not considered to significantly contribute to relocation susceptibility later during the cladding heatup and rupture phases. Fuel thermal contraction and cladding heatup during the blowdown phase increases the pellet-cladding gap and possibly facilitates pellet fragment relocation. Cladding heatup rate and temperature, either with or without a blowdown, are the primary factors that influence burst shape and dimensional changes.
Fuel: N Clad: N Reactor: N Bumup: N K(O): No votes.
PK(2): Data, Experience, judgement UK(2): Judgement Appendix D-71
I Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) I Phenomena (Parameter) I Definition and Rationale (mportance Applicability, and Uncertainty)
Simulation of fuel Conduct of Test-During Simulation of the temperature response of the fuel and cladding during the blowdown relocation Blowdown temperature phase of a large-break LOCA.
transients for fuel and cladding H(2) Important parameters that Influence cladding burst and dimensional changes.
M(O) No votes.
L(I) Pellet fragment movement. In general, pellet fragments are relatively constrained within the fuel rod by the column geometry, as evidenced by characterization of fuel column geometry in hot cells. Therefore, this effect is not considered to significantly contribute to relocation susceptibility later during the cladding heatup and rupture phases. Fuel thermal contraction and cladding heatup during the blowdown phase increases the pellet-cladding gap and possibly facilitates pellet fragment relocation. Cladding heatup rate and temperature, either with or without a blowdown, are the primary factors that influence burst shape and dimensional changes.
Fuel: N Clad: N Reactor: N Burnup: N K(l): Data, Judgement PK(2): Data, Experience, Judgement UK(O): No votes.
Appendix D-72 i
I Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategor (Test type) l Phenomena (Parameled I Definition and Rationale importance, Applicability and Uncertainty)
Simulation of fuel Conduct of Test-During Look at the Impact of fuel fragment relocation on the cladding temperature during the relocation Pre- and post-burst test high temperature oxidation phase and the quenching phase.
phases (2)
H1l) Data of fuel relocation determines the impacted phases.
M(3) Needs in pile test to be prototypical (heating source should come from the fuel). If the objective is as speculated above, this test would help to characterize at which point in time the bulk of the relocation occurs.
However, most rods that balloon also burst and it is not clear that a separation in time would significantly affect the LOCA performance (i.e., whether relocation occurs instantaneously to fill the ballooned region as opposed to instantaneous relocation on burst). Burst shape and dimensional changes are influenced by clad phase at the time of ballooning and burst.
L(O) No votes.
Fuel: N Clad: N Reactor: N Burnup: N K(l): Data,Judgement PK(3): Data, Calculation, Judgement UK(O): No votes.
Appendix D173
I Table D-1. rWR and BWR LOCA. Category D-Separate Effect Testing (continued)
Subcategory Crest type) l Phenomena (Parameter) I Definition and Rationale (Importance, Applicability, and Uncertainty)
Simulation of fuel Conduct of Test-During The amount of gas in the rod upper plenum, for a given initial pressure in the test rod.
relocation Internal pressure and moles of gas H(3) Driving force for relocation, together with gravity. It Is crucial to have a pressure evolution representative of a full-length rod. Internal gas pressure is the driving force for fuel fragments relocation. To be prototypical the amount of gas within the rod prior to the test has to be maintained constant. The internal pressure is a'measured parameter, not an input data. initial pressure is the primary factor that determines the burst temperature and shape and potential release of fuel particles from rim zone at burst. Plenum gas Inventory is a secondary factor.
M(l) If gas flow is the primary relocation mechanism, then an accurate simulation of that gas flow would be needed to obtain the most meaningful results. However, it is anticipated that similar relocation behavior would be obtained over a relatively wide range of gas flows.
L(O) No votes.
Fuel: N Clad: N Reactor: N Bumup- N K(O): No votes.
PK(4): Data, Calculations, Judgement UK(O): No votes.
Appendix D-74
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) lPhenomena (Paratneter) l Definition and Rationale (ImEortance, Applicability! and Uncertainty)
Simulation of fuel Conduct of Test-During During ballooning and after burst, the fuel rod vibration induced by the flow can favour relocation Flow induced vibration crumbling of the fuel pellet stack.
H(O) No votes.
M(2) Fuel column axial gaps have been observed to form and continue during normal reactor operation. This results suggests that fuel column shakeout is not likely with normal flow-induced vibration even over very extended periods. It is further noted that with cladding perforation, steam ingress will promote fuel pellet oxidation that has been observed, with failed fuel during normal reactor operation, to cause effective blockage within the fuel rod to preclude fuel downward fuel pellet fragment motion, again overriding the effects of flow induced vibration. Secondary driving force.
L(2) Potential impact of rod vibration Is expected to be small. Ballooning and burst occur after blowdown, and steam-flow-induced vibration during and after blowdown would be insignificant.
Fuel: N Clad: N Reactor: N Burnup N K(O): No votes.
PK(2): DataJudgement UK(1): Judgement Appendix D-75
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) l Phenomena (Parameter) l Definition and Rationale (importance, Applicability, and Uncertainty)
Simulation of fuel Conduct of Test-During Manner In which test specimen isconstrained bysurrounding rods to simulate potential relocation Exterior rod constraints in-reactor behavior.
H(1) Prior ballooning experiments have shown that coplanar ballooning is not likely, and therefore balloons may not be constrained by adjacent ballooned sections. However, the constraints provided by adjacent non-ballooned rods can still provide a significant restriction on the amount of cladding ballooning and corresponding fuel relocation.
M(1) Rod constraints during ballooning may affect the fuel distribution at the relocation site.
L(2) The purpose of these tests Is to analyze the separate effect of fuel fragment relocation. Exterior constraints Influence ballooning shape to some extent.
Fuel: Y(1): Most modern BWR fuel designs use part-length fuel rods resulting in zones where there Is a significant gap between adjacent rods (because rods in certain lattice locations terminate at a lower elevation). This design feature may permit greater ballooning and relocation at those elevations. However, the fuel rods at those peculiar locations would correspondingly experience a clrcumferential temperature gradient, which is known to reduce the resulting I
burst strain.
Clad: N Reactor: N Burnup: N K(O): No votes.
PK(4): Data, Judgement UK(O): No votes.
Appendix D-76
Table D.1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcatego!y (Test type) I Phenomena (Parameter) I Definition and Rationale rImportance, Applicability and Uncertainty)
Simulation of fuel Conduct of Test-During Determination of the dimensions of the ballooned area and the cladding breach during relocation Balloon size and burst the test.
size H(4) Affects the amount of relocated fuel In the balloon. The balloon and burst size represents the maximum potential volume for relocation. Directly Influence the potential for fuel relocation, slumping, and release at and after burst.
M(O) No votes.
L(l) Balloon size and burst size are measured after the test. No need to measure It on-line Fuel: N Clad: N Reactor: N Burnup: N .
K(l): Judgement PK(2): Data, judgement VK(O): No votes.
Appendix D-77
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) I Phenomena (Parameter) I Definition and Rationale (Importance, Applicability, and Uncertainty)
Simulation of fuel Conduct of Test-During Longitudinal dimension of the fuel rod segment to be tested.
relocation Length H(2) The driving force for fuel fragments relocation is the internal gas pressure in the plenum. For high burmup fuel rods the axial gas transport is significantly impaired. A short rod would favor the plenum gas participation The rod length has to be prototypical to avoid experimental bias. At the least, the length between two grids must be tested.
M(l) The amount of fuel above the ballooned/burst section defines the potential fuel volume to be relocated. However, the size of the ballooned/burst region defines the maximum possible relocated fuel volume. Therefore, If the ballooned/burst location can be defined with reasonable certainty, sufficient length can be provided above that region to enable prototypic relocation.
I L(l) Length more than about 15 times of the pellet length (6 Inches) is sufficient.
Fuel: N Clad: N Reactor: N Bumup N*
K(l): Calculation PK(3): Data, judgement UK(O): No votes.
Appendix D-78 I I
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) I Phenomena (Parameter) I Definition and Rationale (Importance, Applicability, and Uncertainty)
Simulation of fuel Conduct of Test-PTE Granularity of dispersed fuel fragments is measured to get relevant information on the relocation Granularity of dispersed fuel density In the relocated fuel fragments zone.
material H(3) The equivalent fuel density of the relocated fragments allows codes to simulate the local overheating of the cladding. Major factor that Influences the potential for fuel relocation and release.
M(l) Smaller pellet fragments would be expected to result In easier fuel movement and possibly a higher density of relocated fuel. However, pellet cracking patterns are established early in life and do not vary greatly with Increased exposure, so a widely varied granularity of material, prior to dispersal, is not expected.
L(O) No votes.
Fuel: N Clad: N Reactor: N Bumup: N K(O): No votes.
PK(4): Data, Judgement UK(O): No votes.
Appendix D-79
Table D-1. PWR and BWn LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) Phenomena (Parameter) I Definition and Rationale (importance Applicability and Uncertainty)
Simulation of fuel Conduct of Test-PTE Non-intrusive measurement of the temperature differences of the tested fuel rod.
relocation Thermography H(l) Provides the fuel distribution in 3D.
M(O) No votes.
L(2) Low added value Fuel: N Clad: N Reactor N Burnup N K(O): Data PK(2): Data Judgement UK(I): Judgement Appendix D-80 i
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Testtype) l Phenomena (Parameter) lDefinition and Rationale (Importance Apicability and Uncertainty)
Simulation of fuel Conduct of Test-PIE Self defined.
relocation Thermal diffusivity of rubble bed H(l) Output parameter.
M(I) Probably difficult to do, but would be useful in quantifying the effective thermal properties of the rubble mass (This assumes that in the ballooned/burst region if the material Is still there - It may be worthwhile to capture this just prior to burst although there may not be significant relocation at that time If gas flow is the primary relocation mechanism), otherwise this is best done analytically.
L(l) No rationale provided.
Fuel: N Clad: N Reactor: N Burnup: N K(0): No votes.
PK(2): Data, judgement UK(I): Judgement Appendix D-81
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) I Phenomena (Parameter) I Definition and Rationale (Importance, A2plicability! and Uncertainty)
Simulation of fuel Conduct of Test-PTE Measure the shape and the size of the ballooned area of the tested fuel rod.
relocation Strain profile of cladding as f(6,z) H(3) The purpose of this test is to assess the amount and characteristics of relocation. A determining aspect of that process Is the amount of ballooning (free volume to which the fuel may relocate), and therefore this volume should be known intany assessment of relocation characteristics. Note that the circumferential variation of cladding strain should also be determined. Axial variation of clad circumferential strain Is a parameter that directly influences the potential for fuel relocation and slumping.
M(1) Will give some indications on potential impact of the balloon the shape (magnitude and extension) on the amount of relocated fuel.
L(O) Noivotes.
Fuel: N Clad: N Reactor: N Bumup: N K(l): judgement, Calculation PK(3): Data, judgement UK(0): No votes.
Appendix D-82
Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcalegory (Test type) I Phenomena (Parameter) I Derinition and Rationale (ImEortance, Applicability, and Uncertainty)
Simulation of fuel Conduct of Test-1E1 Size of the opening in the cladding after the burst.
relocation Burst size H(3) This Is taken to be the effective surface area of the bulged region that was removed as a result of the burst. Similar to the preceding item, the hole size will be a determining factor in the amount of relocated fuel retained within the ballooned region. Burst opening size and burst circumferential strain are parameters that directly Influence the potential for fuel relocation and release at and after burst.
M(O) No votes.
L(1) If the internal pressure is not maintained (prototypal case) the opening is small.
Fuel: N Clad: N Reactor: N Bumup: N K(2): Data PK(2): Data Judgement UK(O): No votes.
Appendix D-83
IC Table D-1. PWR and BWR LOCA. Category D- Separate Effect Testing (continued)
Subcategory (Test type) l Phenomena (Parameter) LDefinition and Rationale (Importance, Applicability, and Uncertainty)
Simulation of fuel Conduct of Test-PTE Determination of the mass of fuel remaining within the tested fuel rod and the mass that relocation Material balance (in-rod left the fuel rod through the rupture.
and dispersed)
H(2) *This Is the primary result to be quantified In this test series, to be correlated with the ballooned region and burst size. It is the amount of lost material that Is of Interest as It could possibly contribute to such effects as flow blockage, etc.
M(O)
L(2) This information is covered by the local measurement of the fuel density.
Fuel: N Clad: N Reactor: N Burnup: N K(I): Judgement PK(2): Data, Judgement UK(I): Judgement Appendix D-84
2-I
..al -
EXHIBIT I
I _.
Unclassified NEAICSNI/R(2003)9 Organisation de Coopiration ct de Developpement Economiques (C Organisation for Economic Co-operation and Development NUCLEAR ENERGY AGENCY 05-Mar-2003 English - Or. English CONI ITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS Dz
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%0 ONGOING AND PLANNED FUEL SAFETY RESEARCH IN NEA MEMBER STATES Compiled from SEGFSM Members' Contributions October 2002 0
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Document complet diponible sur OLIS dans Ion format d'origine
- e. Complete document available on OLIS In Its original format
NEA/CSNI/R(2003)9 Abstract This report is in response to an action placed on SEGFSM members to compile ongoing and planned fuel safety research in NEA member states with the aim of providing CSNI an overview on related R & D international programmes and projects, along with the identification of current and future needs and priorities.
The report is based on replies to a questionnaire distributed to SEGFSM members requesting them to identify fuel safety research programmes and to provide information on achievements and future plans. The report is confined to the replies received, as a consequence it cannot be viewed as comprehensive; programmes may well be in progress in addition to those detailed here.
The report is organized in topic sections relating to: fuel and clad studies, integral fuel rod tests and PIE, LOCA and RIA studies including whole rods and bundles as well as single effects studies of fuel and cladding, code development for both steady state and transient fuel behaviour, thermal hydraulics, reactor physics codes and finally severe accident studies.
5
NEA/CSNI/R(2003)9
- 10. Severe Accident Studies 10.1.1 Fuel Moderator Interaction (MFMI) Experiments, (Canada)
These MFMI experiments will be performed at the Chalk River Laboratories of AECL. The CANDU Owners Group (COG) is sponsoring these experiments under Joint funding from Ontario Power Generation, Hydro Qudbec, New Brunswick Power, and AECL.
The MFMI experiments are being performed to improve our understanding of the energetics of the interaction of molten material with the heavy-water moderator under CANDU single-channel accident conditions, and to provide data for use in reactor safety code validation. The stagnation feeder break and flow blockage scenarios both lead to fuel channel failure and have the potential for injection of small quantities of molten U0 2-Zircaloy mixtures into the heavy-water-filled moderator vessel. The results of the MFMI experiments will be used to verify the assumption currently made in CANDU safety analysis that classical steam explosions do not occur under these conditions.
The main feature of the MFMI facility is a robust confinement vessel (5.5 m tall, 1.5 m in diameter) located inside a concrete pit enclosed in a concrete building. The experimental plan provides for incremental increases in the molten material loading to help manage this risk. The MFMI experimental program has been defined and the facility is under construction; No experimental results have been obtained to date.
- 11. Conclusions Normal Operations; U0 2 and MOX fuel Over the last twenty years there has been a gradual increase in discharge bum-up from commercial power reactors. With the original 3 .cycle operation, the discharge bum-up was of the order 30 MWd/kg but now, most countries have increased that to a level approaching 60 MWd/kg, see Table 11.1. This has been accompanied by intense R&D both in test reactors like that operated by the OECD Halden Reactor Project and R2 at Studsvik and LTA irradiation in power reactors.
76
rvniiCSN1/R(2003)9 It is anticipated that with the current research pro grammes there are sufficient data to develop and validate fuel performance codes to sul )port these higher levels of discharge bum-up.
The main goal for MOX fuel is for it's safety issuesato be treated indistinguishably from those of U0 2 . Research to date has shown that MOX pellets have a slightly worse thermal conductivity but similar degradation with Iiurn-up. At the same time, it exhibits slightly greater fission gas release, although the onset of release as a function of temperature and bum-up is little different from th; at of U0 2 . FGR from MOX at high bum-up is exacerbated by a higher reactivity than VU0 2 due to its neutronics characteristics. Improvements in MOX are being Ipursued by vendors by investigating the effect of homogeneity on thermal performance aand FGR. However, MOX pellets are naturally more compliant than U0 2 because of their higher rate of thermal creep.
LOCA The 1980s saw great attention paid to the LOCA s;cenario and much data are relevant today. However, the data on high bum-up behavio ur are rather scarce. In this respect, aspects such as possible fuel relocation or 'slumpingg' into the ballooned-area leading to higher clad temperatures need addressing as well as the effect of axial constraints during quenching. Most important is the need to re-visit the 17% Equivalent Clad Reacted (ECR) criterion in the light of new alloys and ne:v geometries, (clad diameters and thickness, see Table 11.2).
Table 11.2 Examples ofPWER andl9WTRfuel designs Assembly type PWR PWR BWR BWR 14 x 14 17 x 17 8x 8 1Ox10 Cladding outer diameter, mm 10.72 9.50 12.52 9.62 Cladding inner diameter, mm 9.48 8.36 10.79 8.36 Cladding wall thickness, mm 0.62 0.57 0.87 0.63 Cladding cross section. mm' 19.70 16.00 31.67 17.80 Fuel pellet diameter, mm 9.29 8.19 10.57 8.19 Fuel pellet cross section, mm' 67.80 52.68 87.75 52.68 Rod pitch, mm 14.10 12.60 16.30 12.40 Rod-to-Rod distance, mm' 3.38 3,10 3.78 2.78 Water cross section (subchannel) mm' 108.5 87.9 142.6 . 81.1 Total clad cross section in assembly mm' 3861 4624 2027 1780 Cladding surface to fuel volume ratio mm"' 0.50 0.57 0.45 0.57 Cladding-to-fuel cross section ratio 0.29 0.30 0.36 0.34 Water to fuel cross section ratio, subchannel 1.60 1.67 1.63 1.54 Notes:
- Total dad cross section Increased in PWR from 14x14 to 17x17. decreased 10% in BWR from 8x8 to 10x1O.
- Cladding to fuel cross section ratio relatively constant.
78
NEA/CSNI/R(2003)9 Two aspects of fuel slumping are the impact of fuel-clad bonding on the propensity of slumping and FGR in the slumped region leading to increased local pressure. Another aspect is the impact of axial gas flow through a 'tight' fuel column on ballooning behaviour. From section 5.2 it is clear that there are several research programmes addressing the properties of high exposure cladding to LOCA but not so much on separate effects fuel studies. Halden have carried out axial gas flow studies in fuel rods over a range of bum-up and test have shown a severe restriction in volume flow at high bum-up thus restraining the rate of clad ballooning. What is lacking therefore is evidence for or against slumping and the internal pressure generated by FGR within the slumped region during the temperature/time envelope of the transient.
RIA Early experiments on low bum-up fuel showed that fuel failure by rapid reactivity insertion only occurred after energy depositions around 200 cal/g. It was not until the first CABRI REP Na test on high bum-up fuel with severely oxidized cladding that it was realized that under these conditions a much lower energy deposition caused fuel failure and extensive fuel dispersal. This result initiated a renewed interest in this type of accident with integral and separate effects tests initiated as outlined in section 6. The main issues with respect to the cladding are its mechanical response during high rates of strain and the effect of hydides on the mechanical properties. Regarding the fuel, tests have shown that the clad strain was greater than that expected from thermaI expansion of the fuel pellet. Consequently, there would appear to be a new loading force imposed by high bum-up pellets, so the goal is to explain this new force and quantify it. In this respect, the high bum-up structure at the pellet rim is under intense separate effects study as this is anticipated to be the root cause of the increased clad loading. The reduction in acceptible energy deposition at high bum-up is considered therefore to be a result of degraded clad mechanical properties and increased strain from restructured regions of pellets. It is clear that the several research programmes both separate effects studies on fuel, section 6.2 and on cladding, section 6.3 should in the near future lead to a better understanding of this type of accident. As a separate but parallel study, it is important to improve reactor physics codes and calculation to see whether or not it is possible to deposit energies as high as those required for fuel failure.
Resolution of Safety Issues As a final comment on the inter-play between phenomena influencing fuel safety, the criteria used to assess compliance and the experimental database on which such criteria can be derived, Figure 11.1 shows a 'road map' linking these three components. From this it is easy to identify where supporting data already exist and where new data will be generated by ongoing programmes or programmes already in the planning stage.
This report has addressed safety issues as they are applied to current reactor systems, the so called 'Generation II' designs. There are now advanced reactor designs categorised as generations III, III+ and IV. A common element of these is the introduction of passive safety features. Thus one question which requires consideration is whether or not these new designs will operate within the safety envelope of the current design. If this is the case, then the scenarios currently being addressed should apply without extension to these new systems. Hence, future R&D will concentrate on 79
NEA/CSNI/R(2003)9 compliance of new materials to the current or a reduced safety envelope and not the consideration of new scenarios.
It is clear that the currently most important issues to the international nuclear industry are high bum-up performance both in normal operations, LOCA and RIA conditions.
The survey of international research programmes outlined above demonstrates the large element of activity to address these issues. When put together, the individual programmes add up to a tremendous effort in both time and money and will ultimately lead to a much better understanding of materials and component behaviour in a wide range of postulated scenarios.
It is very important therefore, that these activities are well supported and that their results should be made available to the widest possible audience. Thus ensuring a common culture of safe and economic production of electricity from nuclear power generation.
80
EXHIBIT J IUnclassificd NEA/NSC/DOC(2004)8 Organisation de Cooperation et de DNveloppement Economiques Organisation for Economic Co-operation and Development 06-llay-2004 English - Or. English NUCLEAR ENERGY AGENCY NUCLEAR SCIENCE COMMITTEE
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INTERNATIONAL SEMINAR ON PELLET-CLAD INTERACTIONS WITIH VATER REACTOR FUELS 9-11 March 2004 Aix-en-Provence x
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MI JT00 163584 0 Document complet disponible sur OLIS dans son rormat d'origine la Complete document avilAble on OLIS In its oritinal format
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International Seminar on PELLET-CLAD INTERACTION IN WVATER REACTOR FUELS organised by CEA Cadarache/DEN/DEC in co-operation with OECD/NEA, IAEA, EDF, FRAAMATOME ANP, COGEMA 9 - 11 March 2004 H6tel NOVOTEL PONT-DE-L'ARC I avenue Arc de Meyran 13100 Aix en Provence, France Introduction This was the third in a series of three seminars that started with the seminar on ((Thermal Performance of High Bum-Up LWR Fuel)> at Cadarache, France, 3 to 6 of March 1998, followed by ((Fission Gas Behaviour in Water Reactor Fuels)), also held at Cadarache, from 26-29 September 2000.
The aim of this third seminar was to d raw up a comprehensive picture of our current understanding of pellet clad interaction and its impact on the fuel rod, under the widest possible conditions.
Pellet-Clad Interaction In PWRs and BWRs, once the fuel-clad gap has closed, I to 3 years after irradiation started (depending on the materials), the compressive stress experienced by the cladding and due to the primary fluid pressure is reversed to a tensile stress induced by continued fuel swelling.
Enhanced clad stress is likely to occur in the region of the pellets' ends, especially when the fuel rod is submitted to power ramps, in relation for instance with incidental transients in the operation of the reactor.
In the presence of aggressive fission products (e.g. iodine typically) released by the pellets, this situation can lead to stress corrosion induced failures resulting in primary water contamination.
This risk is an important industrial challenge to demonstrate that margins are guaranteed for the different current situations and for classes of transients encountered in reactors operation, and justifies the development of so-called PCI-resistant fuel products.
2
mechanisms, or by reduction of the power manoeuvrings responsible for PCI. Would therefore such studies be useless, in the possible context of reduction of PCI frequency? Without being pessimistic, the history of the nuclear industry has shown us that unexpected behaviours are ready to occur when pushing the components to a higher duty, a longer life or reduced operational margins. A scientific knowledge of the fuel behaviour beyond what is strictly required to avoid any PCI failure, and we are unfortunately far from such a scientific knowledge, will not be a waste of time or money, but will allow us to react more efficiently in case of such events occurring.
Overall Recommendations and Open Questions Fuel MfaterialBehaviour in PCISituation I) More efforts are needed to develop "clever" devices able-to provide data on the evolution of fuel mechanical properties with local bum-up and temperature.
- 2) Reliable experimental data are needed to better characterize fuel gaseous swelling kinetics (including irradiation induced gas atom re-solution) under different conditions of temperature, stress and fission rate, including the relative importance of intra-granular swelling; Attention should be paid to evaluating the gaseous swelling driving force and its contribution to the local mechanical loading of the clad.
- 3) Despite considerable improvements of the pellet mechanical modelling in the last decade, further improvements are required. A better characterization of the local stress (stress tensor against hydrostatic pressure) might be necessary for the comprehensive modelling of the different ways stress affects the pellet progressive additional cracking, the viscoplastic flow and the fission gas behaviour. This may necessitate 3D mechanical modelling.
CladdingBehaviourRelevant to PCI
- 1) Stress corrosion cracking, especially in iodine atmosphere, is known to be responsible for PCI failures. Despite many efforts and good analytical work, the need still exists for developing mechanistic models able to reproduce the mechanical tests performed on pressurized tubes as well as to predict the clad crack propagation in true transient conditions. Knowing that the SCC cracks preferentially develop at the pellet-pellet interface and in front of pellet cracks the need for developing duly validated 3D models becomes clear.
- 2) In order to better simulate potential clad damage due to power transients, further work is also recommended concerning the migration of potentially aggressive chemical species such as I, Cs, Cd. Since the papers presented in this session focused on the behaviour of I and, to a lesser extent, that of Cs, it is recommended that the experimental efforts be now directed toward understanding the role of Cd.
- 3) It is recommended that microhardness measurements be pursued to better quantify the evolution of microhardness as a function of burn-up. Such data would be useful not only in understanding PCI SCC but also other phenomena such as secondary damage in failed fuel rods.
19
In Pile Rod Behaviour I) The reasons why MOX fuel and Cr-doped fuel appear to behave better with respect to conventional U0 2 under PCI conditions must be tackled further: is it fuel cracking propensity by itself, and/or is it enhanced viscosity reducing the hour-glass effect by dish filling and perhaps favouring peripheral cracking? How does gaseous swelling act, with which kinetics? So there is a need for new experiments in which the contribution of the individual phenomena is evidenced .
That is the case for gaseous swelling for zero hold-time ramp tests which have been proposed on Cr-doped fuel. It would be the case for gaseous swelling plus creep at higher ramp rates. In parallel, there is a need for experiments to study the high temperature phenomena not far from fusion conditions without any cladding damage (columnar grain growth, central void formation, etc.).
- 2) Relevant comparative analytical data with differences in pellet geometry (e.g. short pellets) could contribute as well to this attempt of varying the relative weight of different phenomena.
- 3) The question of the concurrent cladding improvement might be asked. What kind of benefit can be expected from a new cladding concept?
AModelling of the AfechanicalInteraction between Pellet and Cladding
- 1) The development of de-cohesive models versus diffuse crack models looks promising for the treatment of pellet cracking.
- 2) Mechanical phenomena are assessed differently by ID/1.5D and 3D models, only the latter having the potential to approach the phenomenon with accuracy. Nevertheless, running times are long and the results are still dependent on materials data and interaction prediction. So, it seems that both models should be developed, fast-running ID/1.5D models taking profit of the comprehensive view available from 3D ones
- 3) The developments on pellets mechanical models to cope with cracking, and the provision of an accurate description of heterogeneous products, should be used as inputs of PCMI codes and could help to understand the differences exhibited by doped fuels and MOX
- 4) The ultimate goal of all fuel vendors should be a failure free operation, with no limits imposed on operation. PCI resistant products could contribute (see above), but their good performances have to be demonstrated more widely, by modelling and by complementary experiments.
- 5) It is recommended to extend the use the existing fuel performance databases (e.g. IFPE) for model improvement and validation, and in particular to evaluate PCMI effects on gaseous swelling and vice versa. A FUMEX-I11 exercise devoted to PCMI/PCI effects should be considered.
- 6) Predictive PCI modelling should be presented, not just explanations after the event.
20
EXHIBIT K v- N-t April 21, 2004 Mr. David Modeen, Vice President Nuclear Power Sector Electric Power Research Institute P.O. Box 217097 Charlotte, NC 28221
Dear Mr. Modeen:
In 1998, EPRI and NRC signed the Addendum on Testing High Burnup Fuel to the current Memorandum of Understanding on Cooperative Nuclear Safety Research Between EPRI and NRC. According to that addendum, NRC would conduct testing at Argonne National Laboratory at NRC's expense and EPRI would provide high-burnup fuel rods that were ready for the testing program at EPRI's expense.
Testing of Zircaloy-clad fuel rods, which EPRI provided, is well advanced and the Argonne program is producing important results. In our planning, however, attention has turned to the more advanced cladding alloys, ZIRLO and M5. Discussions have been on-going between NRC, Westinghouse, and Areva concerning the testing of high-burnup rods with these advanced cladding alloys.
There are several pressing reasons to initiate testing on high-burnup rods as soon as possible.
Proposed revisions to the current regulatory criteria for the loss-of-coolant accident (LOCA) will be based on data from Zircaloy-clad fuel that is in the laboratory now. In the absence of high burnup data from other alloys, performance of ZIRLO and M5 under LOCA conditions will be based on data from Zircaloy. However, parallel testing at Argonne on unirradiated ZIRLO and M5 tubing has shown significant differences compared with Zircaloy. NRC's high burnup planning document highlights these differences and describes the need to confirm high-burnup behavior with tests on the alloys in question.'
Industry and NRC share the desire to use pre-hydrided material for future testing in lieu of the more expensive testing of irradiated material. The NRC/EPRI program at Argonne will address this issue; however, the use of pre-hydrided material as a surrogate must be demonstrated for an appropriate range of alloying elements. This development is extremely important in order to simplify testing with even newer alloys.
'Memorandum from W. Travers to the Commission, 'Updated Program Plan for High-Burnup Light-Water Reactor Fuel," August 21, 2003, Adams ML031810103.
4 & I'l Mr. David Modeen A recent user need memorandum from the Office of Nuclear Material Safety and Safeguards requested additional testing on high-burnup spent fuel cladding.2 Testing with irradiated ZIRLO and M5 was specifically requested in that memorandum because the data are needed for cask licensing.
Timeliness in addressing the above issues is important. The work on Zircaloy-clad fuel rods at Argonne should be completed in late 2005, and testing on the irradiation behavior of advanced alloys should begin at that time. Considering that it has taken approximately one year to make arrangements to obtain rods, to prepare specimens, and to ship them to Argonne, plans for shipping irradiated ZIRLO-clad and M5-clad fuel rods should be made now.
The program at Argonne has been a model of cooperation and productivity. We would like very much to continue in that spirit, and I request your support to ensure timely availability of irradiated ZIRLO and M5 cladding.
Sincerely, IRA!
Ashok C. Thadani, Director Office of Nuclear Regulatory Research Distribution:
SMSAB R/F DSARE R/F JStarefos DET R/F C:\ORPCheckout\FileNET\ML04I 130490.wpd
- See previous concurrence OAR in ADAMS? (Y or N) v ADAMS ACCESSION NO.: M ln41 1fn4qQ TEMPLATE NO. RES-f08 Publicly Available? (Y or N) Y DATE OF RELEASE TO PUBLIC SENSITIVE? L To receive a copy of this document, Indicate In the box: '"C = Copy without enclosures "E" = Copy with enclosures "N" = No copy OFFICE SMSAB l C:SMSAB D:DSARE l D:RES l NAME RMeyer:rm:mb JRosenthal FEltawila AThadani DATE 04/20/04* 04/20/04* 04/20/04* 04/21/04*
Jack Stronider concurred on 04/21/04*
2M. Virgilio memorandum to A. Thadani, "User Need Memorandum - Assessment of High Burnup Fuel Cladding Integrity Performance Under Accident Conditions," March 5, 2004, Adams 040650621.
-I%- 11
- ll 4 ,
EXHIBIT L
LOCA Issues Related to Ballooning, Fuel relocation, Flow Blockage and Coolability Main Findings from a Review of Past Experimental Programs Claude GRANDJEAN, Georges HACHE IRSN, CE Cadarache, France SEGFSM Topical Meeting on LOCA Issues ANL, May 25-27, 2004
URSL]3 IRSN State-of-the Art Review on LOCA The S. o.A. review has been divided in 3 parts:
o 1st Part: Clad Ballooning and Rupture. Flow Blockage. Fuel relocation.
0 2nd Part: Coolability of Partially Blocked Regions in Rod Bundles after Ballooning.
El 3rd Part: Cladding Oxidation. Resistance to Quench and post Quench -Loads. Safety Criteria.
SEGFSM Top. Meet, on LOCA Issues, ANL May 25-27, 2004 2
I RSI] IRSN State-of-the Art Review on LOCA Clad Ballooning and Rupture Flow Blockage Fuel relocation SEGFSM Top. Meet. on LOCA Issues, ANL May 25-2 7,2004 3
URS[]n CLAD BALLOONING and RUPTURE / FLOW BLOCKAGE OUT-OF-PILE TESTS IN-PILE TESTS SINGLE ROD MULTIROD SIGLE ROD MULTI ROD EDGAR (CEA)
PBF-LOC (INEL) PHEBUS (IRSN)
KfK (REBEKA) KfI (REBEKA)
FR2 (KfK) NRU-MT (AECL)
EOLO-JR (Ispra) TREAT FRF (USA)
JAERI JAERI FLASH (CEA)
ANL BCL KWU KWU UKAEA UKAEA Westinghouse Westinghouse SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 4
Alab lips W1 BALLOONING / BLOCKAGE. Main Findings (1A) 9 i. 11 5 L",11- J-Out of Pile Sinhgle Rod Tests Lgleadiiig iniflueice of azinlutlilalAT 4+ Direct heating (EDGAR) or internal heating and heated shroud (KfK, ORNL...)
0> azimuthally uniform temperature + large circumferential burst strains (up to > 100%)
- Internal heating with unheated shroud
>non uniform temperature " low average burst strains process linked to the an isotropy of a-Zircaloy (" hot side straighteffect ')
120 -
A SMA U. AZIM UrA I. RW asyp -
100 - (RA64VWi i i rtw4 ""*
4 It d
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-P o TIPICAL UUNDLE VALUL Ai A A a
- CRNL A AR SSEi-<A 60 - A
- AA . .. *RCRNLBKI,8s 6* Z
-4
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0 .
600 700 800 900 1030 BU R1S' TrEM.PEsLV URE C(
SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 alonuthla :addmg tomptrature dftarance AT,. K 5
~~RS[] BALLOONING / BLOCKAGE. Main Findings (1B)
Out of PileSingle Rod Tests Olleadinginfluence of azinuthlal AT Modlelina of strain and burst
- 1D: mean strain, deduced from an "average" of experimental results (cx: NUREG-630) burst criteria: average stress = f(T) (- CATHARE versions up to V1.4,...)
110°h%
- 2D : strain azimuthal profile low/0 FREIAWCIao(EDGAROta
- REPBA Wd(arve)
- allows to take account of ATaz burst criteria: maximum local strain, (eFRE TA, .....)
40%/
30°h 0 10 20 30 40 50 60 AMnwUhal teaepamWe ciffeence (K SEGFSAI Top. Meet, on LOCA Issues, ANL May 25-27, 2004 6
iIRSI BALLOONING / BLOCKAGE. Main Findings (2)
Multi Rod Tests Z inflientce of thermal and niechzanicalinteractionsbetween rods Tests: ORNL MRBT (4x4, 8x8) and JAERI (7x7) q large deformations of inner rods + contact on peripheral rods before clad burst qk axial extension of straining and blockage 100 B-5 80 2 w
00 22
.\a z4 0 CC I 0a 20 20 o 20 40 60 so 100 DISTANCE ABOVE BOTTOM OF HEATED ZONE (cm)
ORNL Reconirnandation(Cliapman, 1982) :bundle tests : at least 2 ring's of t'uard rods aire required to reproduce representative conditions of the inner rods in a reactor assembly SEGFSM Top. Meet, on LOCA Issue~s, ANL May 25-27, 2004 7
IiRS[] BALLOONING / BLOCKAGE. Main Findings (3)
Multi Rod Tests
- r400. . .. . . . . . ... .
C iYinfluence ofguide tubes -------------- R i Test REBEKA-4 (5x5 bundle with unpressurized *300 _____--.. - --- L:
.2 3 '
rods in outer ring) .200 ,s,~~'W.C -- NR---
N.
- Test R-4: Ema~x 79% on a rod neighbor of G.T. ,8i
- o--------------
-H.R iSTR: 5.0 g/tm2 4 -t/. - ---- ...
x Z
Q JAERI Tests 21 to 24 ( 7x7 bundle) with 4 G.T. NQ 20 , ,
I-W/O CON.RP. j q presence of G.T. does not reduce, and even -j
- ST.: 5.4g/crnm TM" hi SR.0° ;
increases strain on neighbor rods, despite large ATaz (57°C in REBEKA-4, 71°C in JAERI-24) 4i. so I
- W/
.
- Coi oN.R0;0i .
,400 ... . . .. . . ..... ... ........
- HNE BOKC I X I KfK explanation: stop of "hot side straight effect" due to contact of deforming rod with GT (for s<20%),
then gap re-opening on hot size
-+ ATaz reduction -* increase of burst strain i Modelimg: importance to describe thermal (radiative) and mnechanical (contact) interactions betveen rods and rods and structures SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 8
BALLOONING / BLOCKAGE. Main Findings (4)
Multi Rod Tests X inflitence of thierin al-hydraulic conditions partially illustrated by NRU MT-4 vs. MT3 tests (32 full length rods, 12 inner rods pressurized)
MT-3. early reflood, clad rupture under 2p cond. / MT-4 : late reflood, rupture during heat-up under steam
, I MT-4 MT-
.0 9 s * . IS am ; .3 '3 90 fS *0 Axial distribution of strain in NRU MT-3 and MT-4 tests (x unit= inch)
Deformations appear significantly coplanar, due to the homogenizing effect of grids and not much different (maximum value, axial spread) in MT-3 / MT-4 (....as opposed to REBEKA 6 / 5 results)
Two-phase TH influences on blockage are complex anid difficult to foresee and transpose at
- scales SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 J
RS[XJ, BALLOONING / BLOCKAGE. Main Findings (5A)
In pile tests with irradiatedfuel rods injluence ofpellet/cladgap reduction or closure duringpriorirradiation ( 1 azimuthal AT) i El PBF-LOC Tests:+ increase of circumferential burst strain on irradiated rods (<16 GWj/t) /unirrad.
(- x22). larger axial extension and clad thinning over whole circumference I4oc V
0 0
0 I-D U
C U
c,1 SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 10
BALLOONING / BLOCKAGE. Main Findings (5B)
In pile tests with irradiatedfuel rods (cont.)
El FR2 Tests:-+ no apparent effect of irradiation on clad deformation nor apparent sensitivity to azimuthal AT 140 - 0 unirradiated rods
- irradiated rods 120 A fuel rod simulators may result from particular irradiation 100 conditions in FR2 0 80 REBEKA burst criterion l T and P -> no clad creepdown)
O 6
- with additional effects of axial al 6 constraint due to spring, limiting
' nhoop strain, and of the closeness of Li 0* 8 0 7 shroudtotestrod,limitingrod 20 -bowing 0 4337-625o 0 20 40 60 80 100 [K]
Maximum azimuthal temperature difference at burst elevation.
SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 11
R S [] BALLOONING I BLOCKAGE. Remaining Needs.
In pile tests with irradiatedfuel rods (coilt.)
O No bundle test with irradiated rods carried out up to now wvhat could be the cumutlative effects of irradiationand bundle size oil blockage ratio?
On the basis that:
vBurst strainfor PBF-LOCfresh rods ~ ORNL Single Rod, unheated shroud
>Burst strainforORNL Multi-Rod >> ORNL Single Rod, unheatedshroud
>Burststrainfor PBF-LOC irradiatedrods >> PBF-LOCfreshrods I t INEL recommended to perform bunidle tests of sufficient bullndle size with irradiatedrods (J.M. Broughton, Sin Valley, 1981)
According to the lukown effect of H charged during irradiation on clad mechanical properties (shift of trantsus temperature Talac+, - significant reduction of burst strainfor high BU irradiatedZircaloy (- 600 ppmn H), see EDGAR results oil pre-IzydridedZy )
ve expectedl the issute to concern imore specifically irradiatedfuel rod claddinigs with low H w
uptake under irraliation(low burn up Zy4, B WR alloys or advancelPWR alloys at high BRU)
Bitt recentt results (see presentationi by N. Wdeckel) indicate this may also concern high B U Zy4 SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27. 2004 12
WI FUEL RELOCATION. Main Findings (1)
In pile tests with irradiatedfuel rods (cont.)
XJfilel relocation observed in PBF-LOC,F?2, FLASH-5, ANL-Lilnerick tests with irradiatedfuel rods (2.5 < BU < 56 GWj/tU)
I. AU.I'd II
. S I"',
I I
I, I.
I I
II l:'t I FRESH FUEL ROD Tcst B 1.1 I~IIi~il;?Ih~ JllII 35 GWd/tU ROD Test G 3.2 SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004
[ [K]R FUEL RELOCATION. Main Findings (2)
I Main parameters:
- instant of fuel collapse likely near tburst at least for low BU (FR2: E3 & E4)
- fragments granulometry heterogeneous at high BU (see ANL test ICL2) q will enhance filling ratio
- filling ratio may be altered during post test handling (PBF-LOC)
Impact
+ local heat generation and transfer to clad t raises T and ECR (for observed fill. ratios)
+ may affect secondary hydriding conditions FR2 rod F1 (20 GWj/t)
SEGFSM Top. Meet. on LOCA Issues, 1st objective of Halden LOCA tests ANJL May 25-27, 2004 14
SE]
Ai FUEL RELOCATION. Main Findings (3)
Fuel chunks relocation Filling ratio as a function of burst strain 90 80 -
70-0u A 60 0 50' o PBFILOC-gammascanning(7SMIRT) O a-
- PBFILOC-tnicrographies A FR2IE5(KfK-3346)-slow depress.
r 40 iz +e FR2IE5(KfK-3346)-divergent cone
-experimental upper boundary 30 o El-Shanawany(Glasgow88)-base case la El-Shanawany-sensitivity 20 0 Edf X IRSN-base case 10 0-0 10 20 30 40 50 60 70 80 Burst strain (%)
SEGFSM Top. Meet. on LOCA Issues, ANL May 25-2 7, 2004 15
Im mm No I L"Al IRSN State-of-the Art Review on LOCA Coolability of Flow Blockages Due to Clad Ballooning under LOCA Transient Conditions SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 16
I RS [] NEffect of Clad Swelling Upon Assembly Cooling 0 investigations on cooling under LOCA reflood conditions of a rod bundle containing a pre-established partial blockage region OL Common Experimental Characteristics Bundle of electrically heated full length rod simulators, with a group of rods bearing a pre-shaped deformation over a given axial length (4 pre-determined blockage ratio )
o establishment of steady state initial conditions in steam (Tg 600 to 8000C) 0 heating power residual power NB: 1o power increase in the balloonedregion of heated rods, and largegap o run of a liquid reflood transient under forced or gravity reflood conditions iimpact offlow blockage on coolability evaluated upon comparison of clad temperatures in the blockage and by-pass regions SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 17
SE]; ,~ FLOODING EXPERIMENTS WITH BLOCKED ARRAYS Specific Experimental Programs
- FEBA (KfK, Germany)
. 5x5 rod bundle; Lheat= 3 .9 mr; "conventional" simulators ; forced reflood
- Blockage over 3x3 or 5x5 rods ; x=62% or 90% ; thick sleeves; LBma = 65 mm (90%)
- SEFLEX (KfK, Germany)
. 5x5 rod bundle; Lheat= 3 .9 m ; REBEKA simulators; forced reflood
. Blockage over 3x3 rods; xt= 90% ; thinned cladding (e 0.5 mm); LBmax= 65 mm (90%)
+ THETIS (AEA Winfrith, UK)
- 7x7 rod bundle; Lheat=3 .6 mr; "conventional" simulators ; forced or gravity reflood
. Blockage over 4x4 rods ; t=80% or 90% ; thin sleeves (e - 0.3 mm); LBma = 200 mm
- CEGB (Berkeley, UK) a 44 rod bundle; Lhe, 1 m ; blockage over 4x4 rods; x= 90%; LBmax = 147 mm; forced reflood
+ FLECHT - SEASET (W. USA)
- 21 and 163 rod bundles ; Lheat= 3.66 m ; forced or gravity reflood
- Short concentric sleeves, coplanar or not; long non-concentric sleeves, non-coplanar SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 18
RSL-,I,-'- FLOODING EXPERIMENTS WITH BLOCKED ARRAYS FEBA and SEFLEX Main Results iiisi,,,,,,,,,,,,i,...
00000 FEBA I 0000~
~(DO I0 n%
__0 FEBA Test section; Blockage 90% on 3x3 rods SEFLEX Ptot ii SEFiEX :estI . 32 He gap MOMiR red bund I "Sl4 e-IiLIt el.-
S te e A.i, C tUddie4 A
00000 00000 _i SEFLEX 0 000000 000 Ir Ptot C Ar gap SEFLEXtest U*. 33 "ItLUK tdied"Pi Air.n.f"IteE 04"8 Cl1004l l5,4i 0000
,s..d.., .... ,.3.8 cWe RAllo tentd2i25'
$Y$t- pressur 2.t bar REiSEAred bondLe Otiypes.ted Cetddlne SEFLEX Test section; Blockage 90% on 3x3 rods BYPIesOS. red Cted4fli
+ GL@Ck5@. ted CLaddl.0 mde"'~eath sLies' olt siege. red Ctoddle?
XILeCkIR9. MaIker Shoe K SEGFSM Top. Meet. on LOCA Issues, Temperatures at the blockage midplane ANL May 25-27, 2004 19
COOLABILITY OF BLOCKED REGIONS Main Findings (1)
The evolution of temperatures within and downstream of a blockage region results from the combined effects of:
t the by-passing of fluid flow towards the unblocked flow channels W significant reduction of flow, then of cooling capacity (under similar lineic heat flux)
> penetration of liquid droplets inside the blockage (due to inertia) impact of droplets on balloon walls, fragmentation, re-entrainment in finer droplets, increase in turbulence 71 liquid / vapor heat transfer (vapor de-superheating) 7 4 enhanced cooling of walls (at least for short blockages)
Žpossible fall of droplets at the blockage outlet (widening section) 0 dispersion and evaporation in steam jets 4 enhanced cooling (at least for short blockages)
SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 20
l RS§ U COOLABILITY OF BLOCKED REGIONS Main Findings (2)
( Blockage representativity (thin vs. thick sleeves) observed from SEFLEX / FEBA tests results SEFLEX: lower heat capacity of the balloon walls and low coupling with heater
'1/2early rewetting of the balloon, propagation of secondary quench fronts up- & downstream
> FEBA results conservative / reactor rod balloon with fresh fiuel ('notfor irrad.fuel)
(DInfluence of flow restriction and blockage length FEBA (90% / 62%; LBIoCk- 65/125 mm), THETIS (90% / 80%; LB1OCk= 200 mm)
- FEBA 62%: blockage always better cooled than by-pass region
- FEBA 90%: low penalty (+40'C on Tma at blockage outlet) for VfloOd mnn = 3.8 cm/s
- THETIS 90%: coolability limit for Vreflood < 2 to 3 cm/s 80% blockage ratio: better cooled than 90% for high Vreflood opposite for low Vrflood (2 cmls)
Influence of blockage length 4 linked to penetration and length of influence of droplets q highly dependent on flow blockage ratio and T.H conditions: flooding velocity and lineic power SEGFSM Top. Meet. on LOCA Issues, ANL May 25-2 7. 2004 21
COOLABILITY OF BLOCKED REGIONS Main Findings (3)
(DInfluence of the blockage configuration : coplanar / non coplanar
> FLECHT SEASET 21 rods, short balloons (low axial overlapping) under n0on coplanarconfiguration:
- the flow redistribution around one balloon increases local turbulence, then cooling of neighbor rods,
- but the isolated influence on droplets fragmentation in the adjacent channel is lower than in coplanar configuration Influence of the nature of reflood (forced / gravity)
> THETIS tests, 80% blocage ratio, forced / gravity reflood
- rapid oscillations of inlet flow and liquid level under gravity reflood, vanishing after 90 s
- temperature evolution, in blockage and by-pass, very similar to those in comparable forced flow (DInfluence of the presence of a by-pass region or not
> FEBA tests, blockage 3x3/5x5 ; FLECHT SEASET 21 rods (config B/ config C) tests without by-pass non representative : no flow redistribution > velocity increase in the blockage, thus enhanced cooling / configuration with by-pass SEGFSM Top. Mfeet. on LOCA Issues, ANL May 25-27, 2004 22
COOLABILITY OF BLOCKED REGIONS Pending Questions Cl Impact of fuel accumulation inthe balloon (fuel relocation)
- EXISTING: analytical tests on bundles of electrically heated fuel rod simulators with partially blocked regions bearing pre-shaped balloons, large heater/clad gap cheaters are fixed and heating axially uniform (inside / outside balloons) significant differences between results of comparable tests FEBA and SEFLEX
> underlines the large impact of thermal coupling between the heat source and the ballooned cladding b The impact of fuel relocation in fuel rod balloons, as was obsenred in all in-reactortests with irradiatedfuel, leading to an increase in local power (lineic and surfacic) as well as a very reduced fuel-clad gap, on the coolability of the blocked region, is still fully questionable and should be addressedby specific analytical tests with a simulation of fuel relocation.
SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 23
CONCLUSIONS. Pending Issues El Flow Blockage in a Bundle of Irradiated Rods
- No multi-rod burst test with irradiated fuel available up to now
( Such tests (with low H uptake clad material ?) would be of main interest LI Fuel Relocation
- Instant of fuel collapse, granulometry and filling ratio at high BU
@ Halden single rod tests (IFA-650)
LI Coolability of Blocked Bundles with Fuel Relocation
- Open question, particularly for long balloons and low reflood rate, for which the blockage ratio still coolable might be less than the widely accepted value of 90% derived from FEBAISEFLEX tests results.
V Need of specific analyticaltests with a simulation of fuel relocation (representativelineic power and gap)
SEGFSM Top. Meet. on LOCA Issues, ANL May 25-27, 2004 24