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Exhibit 1 Pages 429 to 441, Need for Experimental Programmes on LOCA Issues Using High Burn-Up & MOX Fuels from NUREG/CP-0176 Proceedings of the Nuclear Safety Research Conference
ML033650274
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Site: Catawba  Duke Energy icon.png
Issue date: 05/31/2002
From: Monteleone S, Sandra Nesmith
Brookhaven National Lab (BNL), Office of Nuclear Regulatory Research
To:
References
50-413-OLA, 50-414-OLA, ASLBP 03-815-03-OLA, RAS 7184 NUREG/CP-0176
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Exhibit 1 NUREG/CP-0176 Proceedings of the Nuclear Safety Research Conference Held at Marriott Hotel at Metro Center Washington, DC October 22-24, 2001 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Proceedings prepared by Brookhaven National Laboratory

NUREG/CP-0176 Proceedings of the Nuclear Safety Research Conference Held at Marriott Hotel at Metro Center 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

AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS m.Jmo Da.,uas Ijetulel As of November 1999, you may electronically access Documents available from public and special technical NUREG-series publications and other NRC records at libraries include all open literature Items, such as NRC's Public Electronic Reading Room at books, journal articles, and transactions, Federal www.nrc.gov/NRC/ADAMS/index.html. Registernotices, Federal and State legislation, and Publicly released records Include, to name a few, congressional reports. Such documents as theses, NUREG-series publications; Federal Register notices;' dissertations, foreign reports and translations, and applicant, licensee, and vendor documents and non-NRC conference proceedings may be purchased correspondence; NRC correspondence and Internal from their sponsoring organization.

memoranda; bulletins and information notices; inspection and investigative reports; licensee event Copies of industry codes and standards used In a reports; and Commission papers and their substantive manner in the NRC regulatory process are attachments. maintained at-The NRC Technical Library NRC publications In the NUREG series, NRC Two White Flint North regulations, and Title 10, Energy, In the Code of 11545 Rockville Pike Federal Regulations may also be purchased from one Rockville, MD 208524738 of these two sources.

1. The Superintendent of Documents These standards are available In the library for U.S. Government Printing Office reference use by the public. Codes and standards are Mail Stop SSOP usually copyrighted and may be purchased from the Washington, DC 20402-0001 originating organization or, f they are American Internet: bookstore.gpo.gov National Standards, from-Telephone: 202-512-1800 American National Standards Institute Fax: 202-512-2250 11 West 4 2 4 Street
2. The National Technical Information Service New York, NY 10036-8002 Springfield, VA 22161-0002 www.ansi.org www.ntis.gov 212-642-4900 1-800-553-6847 or, locally, 703-605-6000 A single copy of each NRC draft report for comment is available free, to the extent of supply, upon written request as follows: Legally binding regulatory requirements are stated Address: Office of the Chief Information Officer, only in laws; NRC regulations; licenses, Including Reproduction and Distribution technical specifications; or orders, not in Services Section NUREG-series publications. The views expressed U.S. Nuclear Regulatory Commission In contractor-prepared publications In this series Washington, DC 20555-0001 are not necessarily those of the NRC.

E-mail: DISTRiBUTIONenrc.gov Facsimile: 301-415-2289 The NUREG series comprises (1) technical and administrative reports and books prepared by the Some publications in the NUREG series that are staff (NUREG-XXXX) or agency contractors posted at NRC's Web site address (NUREG/CR-XXX), (2) proceedings of www.nrc.gov/NRC/NUREGSrindexnum.html conferences (NUREGICP-XXX)c, (3) reports are updated periodically and may differ from the last resulting from Intemational agreements printed version. Although references to material found (NUREGAiA-XXXX, (4) brochures on a Web site bear the date the material was (NUREG/BR-)XOO), and (5) compilations of legal accessed, the material available on the date cited may decisions and orders of the Commission and subsequently be removed from the site. Atomic and Safety Licensing Boards and of Directors' decisions under Section 2.206 of NRC's regulations (NUREG-0750).

I.

DISCLAIMER: Where the papers In these proceedings have been authored by contractors of the U.S. Govemment, neither the U.S. Government nor any agency thereof, nor any U.S. employee makes any warranty, expressed or Implied, or assumes any legal fiability or responsibility for any third parlys use or the results of such use, of any information, apparatus, product, or process disclosed In these proceedings, or represents that Its use by such third party would not Infringe privately owned rights.

The views expressed In these proceedings are not necessarily those of the U.S. Regulatory Commission.

ABSTRACT Tiiza ivpJUIi UuIIiaIm. Cola utI duumaif 6-iety fnesbektu I uLmerenm titwdie Marnon al metro venter Hotel in Washington, DC., October 22-24, 2001. The papers for the Plenary Sessions are included first, followed by the papers presented in each of the five breakout sessions conducted over the course of the three days. They describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Japan, and Russia.

The titles of the papers and the names of the authors have been updated and may differ from those that appeared In the final program of the meeting.

An asterisk (') in lieu of a page number indicates the transcript of remarks or the final paper was not available in time to be included herein.

iii

PROCEEDINGS OF THE NUCLEAR SAFETY RESEARCH CONFERENCE OCTOBER22-24,2001 Contents Abstract . ....................................................................... iii Registered Attendees ............................................................. xi PLENARY SESSIONS - MONDAY, OCTOBER 22,2001 Opening Remarks ...... I A. Thadani, Director, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission (NRC)

Keynote Speech "Enhancing the NRC's Capacity to Meet New Regulatory Challenges" . ........ 5 Hon. R. Meserve, Chairman, NRC (RES/NRC)

Recent Accomplishments . ......................................................... 13 R. Zimmennan, Deputy Director Expert Panel: "Overview of International/U.S. Research Programs" C. Kessler (OECD NEA) ................ ................................... 17 M. Livolant (IPSN) . ...................................................... 21 W. Magwood (DOE) ................. ..................................... 35 T. Marston (EPRI) . ....................................................... 41 K. Soda (J.AERI) . ........................................................ 47 "New Direction in Research" .. ....................................................... 61 Hon. G. Dicus, Commissioner, NRC PLENARY SESSION - TUESDAY, OCTOBER 23,2001 Expert Panel: "Waste and Decommissioning" Hon. E. McGaffigan, Commissioner, NRC ..................................... 69 M. Boyd (EPA) . ........................................................... 73 T. Cardwell (CRCPD) . ...................................................... 77 R. Meck (NRC) . ........................................................... 81 C. Paperiello (NRC) . ...................................................... 91 A. Wallo (DOE) . .......................................................... 97 J. Wemer (Consultant) . ..................................................... 100 V

Panre Expert Panel: "Advanced Reactors" Hon. J. Merrifield, Commissioner, NXL; ....................................... 11D M.Fertel(NI) ................................. 119 E.Lyman (NCI) .................................. 125 P. Lyons (US Senate Staff - Senator Pete Domenici) .............................. 129 W. Magwood (DOE) ....................................................... 133 T. Marston (EPRI) ........................................................ 137 PLENARY SESSION - WEDNESDAY, OCTOBER 24,2001 Speech by Hon. N. Diaz, Comnissioner, NRC ......................................... 141 Expert Panel: "Communicating the RES Role and Programs" J. Brons (NE1)............................................................ 145 M. Federline (NRC) . ....................................................... 149 A. Kadak (MIT) . ......................................................... 153 P. Norry (NRC) . .......................................................... 157 T. Okkonen (STUK) . ....................................................... 161 MONDAY, OCTOBER 22,2001 SESSION 1: ADVANCED REACTORS T. King (NRC), S. Floyd (NEM), Co-Chalirs PBMR Pre-Application Review ............................................. 175 J. Flack (NRC)

NRC Preparation for Future Reactor Licensing .................................. 191 J. Lyons (NRC)

New Risk-Informed (NERI) Project on Risk-Informed Regulation .................. 201 A. Camp, F. Duran (SNL), G. Apostolakis, M. Golay (MIT), D. Finnicum, S. Ritterbusch (Westinghouse)

New Plant Regulatory Framework ............. ............................... 211 S. Floyd (NEI)

Summary of NRC Workshop on High-Temperature Gas-Cooled Reactor Safety and Research Issues ....................................................... 215 T. King (NRC)

R&D Needs for New Plants: Utility Industry Views ............................... 227 E. Rodwell (EPRI)

Vi

Page Hesearcn Neeas, Heiatea to tne vre-appication Activtes ot imii .................. 239 M. Carelli, S. Ritterbusch, C. Brinkman (Westhinghouse)

SESSION 2: DRY CASK STORAGE RESEARCH M. Cunningham, J. Guttman (NRC) Co-Chairs The Package Performance Study: A Study of Spent Fuel Transportation ... . ........ 251 A. Murphy, R. Lewis (NRC), J. Sprung, K. Sorenson (SNL)

A Risk Analysis of Dry Cask Storage ................. ....................... 259 C. Ryder, et al. (NRC)

Seismic Behavior of Spent Fuel Dry Cask Storage Systems ...... ................ 269 S. Shauket (NRC), V. Luk (SNL)

An Analysis of a Spent Fuel Transportation Cask under Rail Tunnel Fire Conditions . ........................................................ 279 C. Bajwa (NRC)

TUESDAY, OCTOBER 23,2001 SESSION 3: AGE-REELATED ISSUES AND RESEARCH N. Chokshl, W. Bateman (NRC) Co-Chair Stress Corrossion Cracking and Non-Destructive Examination of Dissimilar Metal Welds and Alloy 600 ............... ......................... 287 D. Jackson (NRC)

Regulatory Activities Related to Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles ........ ....................

A. Hiser (NRC)

Aging Evaluation of Cables in Japan .. . ..................................... 295 A. Yamaguchi (Japan Power Engineering Corp.)

Enhanced Airworthiness Program for Airplane Systems ....... ...................

M. Sadeghi (FAA)

Methods to Integrate Aging Effects Into Probabilistic Risk Assessments ..... ........ 301 A. Buslik (NRC)

Wii

Page Age-Related Degradation of Structures and Passive Components at

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........................................................ _s I1 s

J. Braverman, C. Miller, C. Hofmayer (BNL), B. Ellingwood (Georgia Tech.),

D. Naus (ORNL), T-Y. Chang (NRC)

SESSION 4: FUEL RESEARCH R. Meyer (NRC), R. Yang (EPRI), Co-Chairs Study of High Burnup Fuel Behavior under LOCA Conditions at JAERI:

Hydrogen Effects on the Failure-bearing Capability of Cladding Tubes ..... .......... 335 F. Nagase, H. Uetsuka (JAERI)

Overview of Test Results on Mechanical Properties of Unirradiated and Irradiated Zr-1%oNb El 10 Alloy Cladding .............................................. 343 L. Yegorova, E. Kaplar, K. Lioutov (RRC), V. Smirnov, A. Goryachev (RIAR)

High Temperature Oxidation of Irradiated Limerick BWR Cladding ..... ............ 353 Y. Yan, et al. (ANL)

Status of the CABRI International Programme and Preparation of the CIPO Test Series ............................................................. 373 J. Papin, et al. (IPSN)

FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENIFLO Thermal Hydraulic Code ........... ................................ 381 K. Valtonen (STUK), A. Hamaniinen (VTT Energy), M. Cunningham (PNNL)

WEDNESDAY, OCTOBER 24,2001 SESSION 4: FUEL RESEARCH (Cont'd)

Pulse Width During a PWR Rod Ejection Accident .............................. 397 D. Diamond, B. Bromley, A. Aronson (BNL)

Revised RIA Criteria for Bumup Extension Applications ................................ 409 R. Montgomery (Anatech Corp.), N. Waeckel (EdE), R. Yang (EPRI)

PosterPaper:Need for Experimental Programs on LOCA Issues Using High Bumup and MOX Fuels ................. ............................... 429 A. Mailliat, M. Schwarz (IPSN)

Viii

SESSION 5: RISK INFORMING REGULATORY PRACTICES M. Cunninpham. C. Carnenter (NRC. Cn-rhairr An Overview of NRC Research Activities in Probabilistic Risk Analysis ..... ........ 443 S. Newberry, P. Baranowsky, M. Cunningham (NRC)

Pressurized Thermal Shock Risk Assessment ................................... 461 D. Whitehead, V. Dandini (SNL), A. Kolaczkowski (SAIC), E. Thomsbury, N. Siu, R. Woods (NRC)

Circuit Analysis in Fire Risk Assessment . .....................................

  • S. Nowlen, F. Wyant (SNL), N. Siu, H. Woods (NRC)

Risk-Based Performance Indicators and the Inspection Process ...... ............... 471 H. Hamzehee (NRC)

EPRI Strategic Action Plan for Risk Technology ................................ 475 J. Gaertner, J. Haugh (EPRI) ix

Need for Experimental Programmes on LOCA Issues Using High Bum-Up and MOX Fuels Alain MAILLIAT, Michel. SCHWARZ Institut de Protection et de SOret6 Nucl6aire De6partement de Recherches en S6curit6 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-IrradiV (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

II

1. INTRODUCTION In France and in other countries, a permanent evolution of the light water reactors (LWR) is observed since the seventies. The evolution deals with the reactor designs (900 MWeI3 loops, I 1300MWe/4 loops, N4, future EPR). It is also related to the fuel management and bumup increase (3 cycles, 4 cycles, 39' GWdItU, 47, 52, 60 GWd/tU Inthe next future). This evolution affects the fuel itself (UO2, MOX, Gd fuel), the cladding (Zircaloy, Zirlo, M5) and the control rods (Ag-In-Cd, 1 4C). As a consequence of these modifications, there is a permanent need to reassess the reactor safety studies which implies improving the associated knowledge and 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 burnup increase and the type of fuel. In France, under safety considerations, It was 11 requested prior to any generic authorisation of discharge bum-up extension, that the high bum- i 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. I 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 the Code of Federal Regulations (10.CFR50, part 50.46) as HAcceptance 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 I

calculated performance of the I

cooling system under the most Temp-Whws severe loss-of-coolant accident ECa conditions. A summary of these conditions is given on the figure 1 against. These first two requirements address: the peak cladding temperature (PCT) which shall not exceed 1204"C and the maximum cladding oxidation rate, BAb i defined through an equivalent PM N C.40 ¶UALVU7UM i cladding reacted (ECR), which MOK "Mns imC VW?

shall nowhere exceed 17% of the so 1W MT- (=crd) cladding thickness before oxidation but after cladding swelling with or Figure 1 without rupture.

'Mean value per assembly 430 i

The third request addresses the maximum hydrogen generation, the total amount of which shall

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..-- *' -i 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 BEHA VIOUR An important progress in knowledge relative to irradiated clad behaviour has been obtained from the results of the French EDFAIPSN [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

1 2.1.2 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[4,51 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 . X rods experiencing LOCA t:2Z; conditions have shown an accumulation of fuel debris in *H the swollen region -called balloon- of the burst cladding which resulted from fuel u

t V.

fragments slumping from upper sT L ESM FE locations (see figures 2 and 3 . T.t, 2.2 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 bumup fuel.

Indeed, the process was observed in the FLASH-5 test with 50GWd/t fuel in spite of a rather low clad strain (not higher than 16%). It is thought that the cladding temperature ., ing increase combined with its ballooning suppress

-at less partly- the bounding making possible r End the fuel relocation. Lone 1 Finally, fuel relocation process is not specific of high bumup fuel. It was also observed for fuel _i rod having a burnup as low as 48MWd/t (LOC5-7B test).

AN INCREASED CLADDING DEFORMATION '

In PBF-LOC experiments 2 couples of rods (2 unirradiated + 2 irradiated) were . _ .4 simultaneously tested in the same test train.

Available data for comparison, although in very z limited number due to technical problems, -

clearly indicate significant differences in the Rod,F deformation behaviour of irradiated versus unirradiated rods.

Figure 3 432

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;

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imicaing iow 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.

o Filling ratio of clad balloon at high bum-up, with fragmentation of U02 rim or MOX dusters o 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.

a 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

2.23 COOLABILITY Related questions should be considered additionally, relative to flow blockage behaviour of highly deformed cladding with possibly relocated fuel and the embrittlement potentials associated 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 information 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.

a 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 IRRADJ& 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 assures 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 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

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  • con rig IIAQI I tests) in France; tests on irradiated cladding and integral type experiments (ballooning I burst I oxidation I quench) on irradiated rods at ANL (USA) [8] and JAERI (Japan) 19] 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 lrradidprogramme, 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, Burn-up),

o the fuel-clad thermomechanical coupling (i.e. fuel relocation) o thermal azimuthal gradients (main factor affecting cladding strain and blockage ratio)

  • thermal-hydraulic aspects (i.e. quenching, coolability of blocked arrays) 0 3.2 TEST DEFINITION RATIONALE The following analysis provides the rationale for the APRP-Irradi6programme 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

these aspects impact the strength of the rod during the quenching phase and the residual ductility of the rod after the LOCA transient.

During the blowdown phase of the LOCA transient, there is much less heat generation in the 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, 1000XC and an increase of both the pellet rim and clad temperatures from 300'C up to 1000"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 thermomechanical 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 thermomechanical loads of the rods are provided. Without such a power during 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.

3.2.2 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 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).

436

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  • 41 BURST MeMERATUREC 01 60 0 700 800 900 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 V7A 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 B51B3 experiments Figure 5

[101.

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 6r1W%,Cu- Uy L.le agud arid 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, m 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

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 83 111,14], 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.

PHEBUS LOCk TEST CONFIGURATION Figure 6 Furthermore, 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 115,16] 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

-91 Presently, fragment size less than 500 microns can be located (see figure 7). Further impr-verment of ihe 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 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.

440

REFERENCES

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M. MUHLHEIM Y. MURAO F. NAGASE Oak Ridge National Laboratory Nuclear Safety Commission Japan Atomic Energy Rasearch Inst PO Box 2009 3-1-1, Kasunigaseki, Chlyoda-ku Shirakata-shirane 2-4 Oak Ridge, TN 37831-8045 USA Tokyo, 100-8970 Japan Toal-mnura, Ibaraki 319-1195 Japan Phone: 865 574 0386 Phone: 81 3 3581 9949 Phone: 31 29 282 5955 Fax: 865 574 0382 Fax: tl 3 3581 9337 Fax: 81 29 23 6160 E-Mail: muhlheimmdtornl.gov E-Mail: ymnumo~op.caogojp E-Mall: nagseepopsvr.tokaijaeri.gojp R. NANSTAD M. NEIL S. NOWLEN Oakl Ridge National Laboratory Canadian Nuclear Safety SandiaNational Laboratories PO Box 2008 Commission PO Box S800 Oakl Ridge, TN' 37831 USA 230 Slater St. Albuquerque, NM 171IS USA Phone: 965-574-4471 Ottawa, Ontario KIP 5S9 Canada Phone: 505-845-9850 Fax: 365-574-0641 Phone: 6139967323 Fax: 505-344-2829 E-Mail: nanstadrkroml.gov Fax: E-Mail: spnoite=sandiapov E-Mail: neilmncnsc-ccsn.gc.ca xnI

Z. OGISO T. OKKONEN M. OMEALIA Nuclear Power Engineering Corp. Radiation &Nuclear Salhy Eagan &Associates, PLILC 3.17-1 Tranomon Minatoku Authority 7918 Jones Branch Dr., Ste 600 Tokyo, 105-0001 Japan Nuclear Reactor Reg, PO Box 14 McLean, Va 22102 USA Phone: 8 13 3435 3427 Helsinki, FN 00881 Finland Phone: 703 913 4945 Fax: 813 3435 3428 Phone: 351 9 75988299 Fax: 703 918 4943 E-Mail: ogisoenupec.orjp Fax: 353975988382 E-Mail: mkoeirehotuallcom E-Mail: timookkonensfi S. ORLOFF J. OSBORNE A. OSUGA U.S. Deparunent of Energy Constellation Nuclear Mitsubishi Hfay Industries America 1000 Independence Ave, SW 1650 CalertCliffms Pkoy PO Box 355, Suite 268Easn Washington, DC USA Lusby, MD 20657 USA Piosburgh, PA 15230 USA Phone: 202 781 6227 Phone: 410 495 2252 Phone: 412374 7395 Fhx: Fax: 410 495 2828 Fax: 412 374 7377 E-Mail: orloffsrinavsanavymil E-Mail: john.mnosbornesccnppi.com E-Mail: danarsesheplAlhel.net L OTT 0. OZER A. PAL Oak Ridge National Laboratory EPRI Entergy Nuclear Nortbeast PO Box 2008 PO Box 10412 440 Hamilton Ave.

Oak Ridge, TN 37831-8057 USA Palo Alto, CA 94303 USA White Plains, NY 10601 USA Phone: 865 574 0324 Phone: 6508552089 Phone: 914 272 3518 Fax: 865 574 8216 Fax: 650 855 1026 Fax 914 272 3382 E-Mail: otljeornl.gov E-Mail: Dozerepricom E-Mail: apalentergy.com C. PARK K.C. PARK SD. PARK Kepco NuclearFuel Co. Korea Hydro &Nuclear Power Co. Xorea Electric Powe Research nsm 150 Deogjin-Dong Yusung-Gu 167 Samseong-dong, Gannam-gu Moonji-Dong Daejon.305-353 Korea Seoul, 135-791 Korea Tmejeon, Korea Phone: 82 42 868 1850 Phone: 02 2 3456 2595 Phone: 82 42 685 5670 Fax: 82 42 863 4430 Fax: 02 2 3456 2559 Fax 2 42 865 SS0 E-Mail: coparkrnail.knfc.co.kr E-Mail: parkkc@kJhnp.coJbr E-Mail: spazl~keprijre.

Jii PETITCLERC L PHILPS L PIREK IPSN Consultant Westinghouse BP No.6 118 Monroe St, #705 2000 Day Hill Rd Fontenay mux Roses, 92265 France Rockville, MD 20850 USA Windsor, CT 06095 USA Phone: 33 1 4654 9060 Phone: 301 294 3069 Phone: 560731 1865 Fax: 33 1 4654 8925 Fax: 301294 0291 Fax 860 731 1824 E-Mail: jean-louis.petitclerclpsn& E-Mail: Ianyephillips551ce.com E-Mall:

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. POTIS N. RAMCHANDANI J. RASED Global Nuclear Fuel Marlon Anatech 102 Trornbay Dr. Ste. 1800, 1040S Jasper Ave. 5435 Oberlin Dr.

Wilmington, NC 284122 USA Edmonton, AB TSK 2WI Canada SanDiego, CA92121 USA Phone: 910 675-5708 Phone: 1 780 707 9651 Phone: 858 455 6350 Fax: 910 675 6966 Fax: 1780US9191 Fax: 8584551094 E-Mail: gerald.pons~gpf.com E-Mail: mandtanchandan@matrikon.com E-Mail: joeearnaechcom M. RESNIK R. REYNOLDS L RICI;ARD U.S. Department of Energy Framatome ANP, Inc. Westinghouse 1000 Independence Ave., SW 3315 Old Forest Rd., PO Box 10935 2000 Day Hill Rd.

Washington, DC USA Lynchburg.VA245060935 USA Windsor, CT 06095 USA Phone: 202 781 6108 Phone: 434 332 3659 Phone: 860 731 6289 Fax: Fax: 434 832 2932 Fax: 860 731 6708 E-Mail: resnikrneenavsea.navymil E-Mail: rrnVnoldsfainatecheom E-Mail:

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D. RISHER S. R1TMERBUSCH E. RODWELL Westinghouse Westinghouse EPRI 4350 Northern Pike 2000 Day Hill Rd. PO Box 10412 Monroeville, PA 15146-2886 USA Windsor, CT 06095 USA Palo Alto, CA 94303 USA Phone: 412-374-5774 Phone: £607316621 Phone: 650 855 8714 Fax: Fax: 560 731 6588 Fax:

E-Mail: risherdhwudtinghous-com E-Mail: E-Mail: erodwdllpepri.com stanley.Lftatrbuscheus.westinghousesco KI ROGERS T. ROSSEEL P. RUDLING KCR Oak Ridge National Laboratory ANT 6202 Perthshire CL PO Box 2008 Uppsala Science Park Bethesda, MD 20817 USA Oak Ridge, TN37S31-6138 USA Uppsala, 75183 Sweden Phone: 301 530 4489 Phone: 365 574 5380 Phone: 46 70 6696681 Fax: 301 5304033 Fax: 365 241 3650 Fax: 46 18 506685 E-Mail: kcrogosaolacom E-Mail: iosseeltm~oml.gov E-Mail: peterzudlingbartese M. SADEGHI M. SATTSON P. SAVEROT Federal Aviation Administration Idaho National Laboratory lAI Corporation 1601 Lind Ave. PO Box 1625, MS3870 4103 Chain Bridge RdP,Suite 200 Renton, WA 98055 USA Idaho FaIls, ID 3415-3870 USA Fairbx, VA 22030 USA Phone: 425-227-2117 Phone: 208-526-9626 Phone: 703 359 9355 Fax: 425-227-1100 Fix: 208-526-2930 Fax: 703 359 0842 E-Mail: mrassoud.sadeghiefaa.gov E-Mail: sbmeinel.gov E-Mail: psazVerotjaicorp.com A. SCHNEIDER D. SCHRXE W. SHACK NRCNIPhysicsIA2EC Studsvik Nuclear AB Argonne National Laboratory PO Box 9001 Nykoping, 61182 Sweden Bldg. 212 Beer Sheva, 84965 tsrael Phone: 46 155 221000 Argonne, L 60439 USA Phone: 972 8 6567019 Fax: 46 155 263156 Phone: 630 2525137 Fax: E-Mail: david schrircestudsvilks Fax 6302524798 E-Mail: schniconetvision netil E-Mail: vjshackanltgov V. SHAH R. SIMARD J. SIMPSON Argonne National Laboratory Nuclear Energy institute Oak Ridge National Laboratory Argonne, IL 60439-4825 USA 17761 St., NW, Ste 400 PO Box 2008 Phone: 630-252-5848 Washington, DC 20006 USA Oak Ridge, TN 37831-6210 USA Fax: 630-252-3250 Phone: 202 739 8128 Phone: 365 574 0422 E-Mail: vnshazbanlov Fax: 2022933451 Fax: 865 251 5005 E-Mail: rtsaei.org E-Mail: simpsonjjeomLjov W. SLAGLE C. SMITH W. SMITH Westinghouse ISL, Inc. Bewchtel Power Corp.

4350 Northern Pike, Energy Center 11140 Rockville Pike, Suite 500 5275 Wcstview Dr.

Site. MS ECE 4-24 Rockville, MD 20852 USA Frederick, MD 21703-8306 USA Monroeville, PA 15146-2S86 USA Phone: 301-255-2265 Phone: 301 228 6096 Phone: 412 374 2088 Fax: 301-468-0883 Fwx. 301 360 0239 Fax: 412 374 2045 E-Mail: csmitheislinercom E-Mail: wasmithtbechtelcom E-Mail: slagle0i&estinghouse.com B. SNYDER K. SODA P. STOREY Energy &Management Consultants Japan Atomic Energy Research Inst Nuclear bistallations Inspectorate, 11005 Hunt Clb Dr. 2-2-2 Uchisaiwalcho HSE Potomac, MD 20854 USA Chiyodaku, Tokyo 100-0011 Japan NSD4A, RPm 314, Si Peters House, Phone: 301 983 9526 Phone: 81 3 3592 2156 Balliol Rd Fax: 301 983 9526 Fax: Sl 3 3592 2159 Bootle, Merseyside, L20 312 UK E-Mail: bsnvderaioip.com E-Mail: sodaghemsjaeri.gojp Phone: 44 151 951 4172 Fax: 44 151 951 4100 VWiH

A. STRASSER V. STRIZHOV R. SUGARMAN Aquarius Services Corp. Nuclear Safety Institute (IBRAE) Morgan Lewis 17 Pokahoe Dr. B. Tulskaya 52 1800 M SL, NW Sleepy Hollow, NY 10591 USA Moscow, 113191 Russia Washington, DC 20036 USA Phone: 9143668875 Phone: 7 095 9580873 Phone: 202-467-3725 Fax: 914 366 8876 Fax: 7 095 9580409 Fax: 202-467-7176 E-Mail: Stras69@attn.et E-Mall: vfsibraeacaru E-Mail: sugarnanemorganewis.eom D. SUNDERLAND T. TAIANO S.W. TAM ANATECH Corp. Japan Electric Power Info Center Argonne National Laboratory 321 Main Mall 1120 Connecticut Ave, NW, Ste 9700 S Cass Ave Poughkeepsie, NY 12601-3116 USA 1070 Argonne, L 60439 USA Phone: 845-454-6100 Washington, DC 20036 USA Phone: 630 252 6602 Fax: 845454-2292 Phone: 202 955 5610 Fax: 630 252 3250 E-Mail: dion2anatcch.com Fax: 202 955 5612 E-Mail: tm~ani.gov E-Mail: thokua@epiccom R. TEJANI S. TETREAULT J. THORP Matrikon International Inc. Donrey Media Group Constellation Energy Group 90 90 Ridgedale Ave. 666 11th Steet NW. Suite 535 2559 Aspen Road East Hanover, NJ 07936 USA Washington.DC20001 USA Port Republic, MD 20676-2615 Phone: 201-681-0296 Phone: 202-783-1760 USA Fax: Fax: 202-783-1955 Phone: 410-495-2724 E-Mail: riaztejanigmatrikonrcom E-Mail: STeteauH~donreyck.eom Fx: 410-495-2828 E-Mail: john.horp~ccnppl.eom R. TREGONNG .S. TRUBATCH N. TYLER Naval Surface Warfare Center Foley & Lardner Calvert aifs Nuclear Power Plant 9500 MacArthur Blvd. 888 16thStreetNW 1650 Calvert iffs Parkway West Bethesda. MD 20817-5700 Washington, DC 20006 USA Lusby, MD 20657 USA USA Phone: 202-8354012 Phone 401495 6714 Phone: 301 227 5145 Fax: 202-835-8137 Fax:

Fax: 301 227 5548 E-Masl: strubatchfoleylaw.eorn E-Mail: EdmundMTylec@ccnppi.com E-Mail: tregoningrlnswccdnhvy.m1l R. UHRIG J. VALENTINO il VALTONEN Univ. of Tennessee, Nuclear lnternationS Access Corporation Radiation & Nuclear Safety Engineering 1015 Ift Street,NW.Suite504 Autrity 306 Pasqua Engineering Bldg. Washington. DC 20036 USA PO Box 14/Laippatie 4 Knoxville, TN 37996-2300 USA Phone: 202-223-7040 Helsinki, FIN 00881 Finland Phone: 8659743110 Fax: 202-296-5373 Phone: 358975988331 Fax: 865 974 0668 E-Mail: valentinoliacdc.com Fax: 3589 75988382 E-Mail: ruhrigsutkledu E-Mail: keUo.valtone stukfi T. VANTTOLA D. VOGT N. WAECXEL Vrr Energy/Nuclear Energy Lauvenec Llvennore National Lab EPR/EDF Tedniikantie 4C, PO Box 1604 PO Box 808, L-638 3412 HilIview Ave.

Espoo, FIN-02044 VTI Finland Livermore, CA 94550 USA Palo Alto, CA 94304 USA Phone: 358 9 456 5020 Phone: 925 423 7798 Phone: 650-855-1688 Fax: 358 9 456 5000 Fax: 925422 9913 Fax: 650-855-1026 E-Mail: Ttmo.VantolavtLfi E-Mail: vogt3@iln.gov E-Mail: nlcola waeckel~df.f J.A. WANG P. WEBSTER J. WEL OakJ Ridge National Laboratory Canadian Nuclear Safety McGraw-Hin POBox 2008 Commission 1200 G St, NW, Stre 100 Oak Ridge, TN 37831-6370 USA 280 Slater St. Washington, DC 20005 USA Phone: 865 574 2274 Ottawa, Ontario KI1H 7E3 Canada Phone: 202-383-2161 Fax: 8655744018 Phone: 6139472771 Fax: 202-383-2125 E-Mail: wanzia~ oml.gov Fax: 613 995 5086 E-Mail: jenny-ieiliplams.con E-Mail: websterpecnsc-eesn.gc.ca Xix

R V.EINER C. WHEAL E. WIESER KW Consulting, Inc. Acrofkx Akair Cybernetics Corp. Nuclear Waste News PO Box 101567 4201 Northview Dr., Suite 302 8737 Colesville Rd, Suite 1100 Pittsburgh, PA 15237 USA Bowie, MD 20716 USA Silver Spring, MD 20910 USA Phone: 412 635 7732 Phone: 301 805 8300 Phone: 301 587 6300 Fax: 4126873965 Fax: 301 305 8122 Fax: 301 587 1081 E-Mail: bobek-wconsulting.com E-Mall: cv-tcalaalawcom E-Mail: ewieserjbpinewswm R YANG S. YIM IC YOON EPR. Ballard Spahr Andrews &Ingersoil, Framatome ANP 3412 Hiliview Ave. LIP 3315 Old Forest Road Palo Alto, CA 94304 USA 601 13SL,NW Lynchberg,VA24501 USA Phone: 650 855 2481 Washington, DC 20005 USA Phone: 434-432-3280 Fax: 650 855 1026 Phone: 202 661 7615 Fax: 434-332-2831 E-Mail: rwngecpri.com E-Mail: kyom~fraatechcom E-Mail: yimsebablrdspahr.cm Y. YOSIUOKA L YOUNGBLOOD D. ZANNONI International Access Corporation ISL. kmc State ofNew Jrsey 1015 13th Street,NW, Suite 504 11140 Rockville Pike, Suite 500 9 Tudor CL Washington, DC 20036 USA Rockville, MD 20352 USA Ewing NJ 08623 USA Phone: 202-223-7040 Phone: 301 255 2270 Phone: 609 984 7440 Fax: 202-296-5373 Fax: 301 468 0883 Fx:

E-Mail: E-Mail: ryoungbloodtlslincom E-Mail: dzannonl@depstzteaJ.us J. ZHANG M. ZflIERMAN Trantebel Energy Engineering Paul Scherrer buitut Ave. Ariane 7 Villigen PSL CH S232 Sitmrland Brussels, 01200 Belgium Phone: 41563102733 Phone: 0032-2-7739843 Fax: 41 S63102327 Fax: 32-2-7738900 EMail: mardnzlrmermannepslA E-Mail: jinzhao zhang~aractebc1.com xxt

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