STUDSVIK/NF(R)-69/83 Studsvik Report

FAILURE DEVELOPMENT IN LEAKING LWR FUEL RODS — A LITERATURE tSURVEYJ

D O Pickman

Studsvik Nuclear CONTENTS

Page LIST OF TABLES iii LIST OF ILLUSTRATIONS iv 1. INTRODUCTION 1 2. CAUSES AND CHARACTERISTICS OF DEFECTS 2 3. EFFECT OF DEFECTS ON FUEL PERFORMANCE 6 4. DEFECT DETERIORATION MECHANISMS 8 4.1 Clad Bore Surface Oxidation 10 1.2 U0_ Oxidation and Gap Conductance 11 4.3 Secondary Hydriding Mechanism 14 5. TYPICAL DEFECT BEHAVIOUR 17 5.1 Small PCI Defects 17 5.2 Fretting Defects 22 5.3 Hydride Defects 25 5.4 Other Influences on Defect Behaviour 26 6. DEFECT DETECTION 32 6.1 Measurement Techniques 35 6.2 Activity Release Measurement Requirements 37 6.3 Activity Measurement Systems 37 6.4 Activity Monitoring Systems in Use 38 6.4.1. French PWR Practice 39 6.4.2. FRG Practice 43 6.4.3. U.S. Practice 45 6.4.4. Canadian Practice 48 6.4.5. Swedish Practice 49 6.4.6. U.K. Practice 53 6.4.7. Fiscellaneous Practices 54 Page 6.5 Activity Release levels 56 7. DISCHARGE CRITERIA 61 7.1 U.S. Practice 62 7.2 French Practice 63 7.3 U.K. Practice 6*4 7.4 Japanese Practice 65 7.5 Practice Elsewhere 65 8. DESIGN ASPECTS 65 9. OPERATIONAL ASPECTS 69 10. REMEDIES 72 10.1 Reduced Rating 72 10.2 Gap Atmosphere Control 73 10.3 Hydrogen Getters 74 10.4 Barriers 76

10.5 U02 Pellet Design 77 11. PROP^.ILS FOR FUTURE WORK 78 12. SIP / G 80 13. C

14. FffSFiNCES 93

^ 1£S 101

1 .USTRATIONS 104

/. PENDIX A

A'.PENDIX B

/fl'ENDIX C

APPENDIX D

BIBLIOGRAPHY

ii TABLES

Table Title Page

1. Radio-Isotopes of Interest for Activity Release Measurements. 101

2. Cycle Burn-Up and Primary Coolant Activity; Status of EDF's 900 MW (e) PWR Units in Commercial Operation as of June 15th, 1981. 102

3. Behaviour of Some Unusual Defects in the SGHW Reactor 103 ILLUSTRATION'S

Figure Title Page

1. Examples of fretting damage: (a) 0.25am deep hole at fixed stop in spring grid design (b) spacer pad wear at grid, (c) smooth clad wall thinning from contact with ring grid. 2. Small PCI crack believed to have re-sealed. 3. Primary hydride sunbursts, (top) early stage but stress effect on hydride orientation already noticeable, (middle) major sunburst with only minor leak, formed in 85 days irradiation, peak LHGR 520W/cm, (bottom) sunburst with large leak showing beginning of hydride re-conversion to Zr. 4. Massive hydride sunburst showing lamination effect resulting from ZrH, , re-conversion to Zr in inner i .o region of cladding. 5. Axial PCI cracks. 6. Bulk off-gas and iodine release data for individual rods with small to moderate defects over first 100 days. 7. Small oxide nodule on bore of cladding opposite radial crack. 8. Deltoid hydriding at mouth of partial penetration PCI crack. Note hydride re-orientation in plastic zone at crack tip. 9. Proposed failure limits (surface heat flux/time to failure) for Zircaloy clad LWF fuel. 10. Bulk off-gas and iodine release data for individual rods with severe defects over first 20 days.

IV 11. Transverese clad fracture near end of BWR fuel rod

ascribed to fatigue failure in hydrided region. Incipient

defects seen at both ends of a pellet in rod 2 below.

12. Axial and 45° cracks in defect rod 999-H2 from Ringhals 1.

Probably primary defect, 598mc from bottom end.

Original x 6 (See also Figure 14).

13. Spalled bulge in defect rod 999-H2 from Ringhals 1.

Probably secondary hydride defect 2100mm from bottom

Original x 6 (See also Fig. 15)

14. Section of defect in Ringhals 1 rod 999-H2 at 598mm

from bottom. Believed to be primary defect.

Original x 53

15. Section of defect in Ringhals 1 rod 999-H at 2100mm

from bottom. Probably secondary hydride defect.

Original x 53.

16. Tihange 1 parametric study. Correlation of gap escape

rate constant with isotope ratios.

17. Correlation between inferred diffusion constants and

linear heat generation rate for intact and defect rods.

18. Relationship between fractional release and decay

constants for fission gases and iodines.

19. Fission gas and iodine spiking effects on shutdown and

start-up.

20. Activity release following a shutdown and during power

ramping.

21. H_ flux vs. time to saturate a Zircaloy-2

surface, (calculated)

22. Hydrogen absorption by Zircaloy-2 at 400°c. FAILURE DEVELOPMENT IN LEAKING LWR FUEL RODS

D.O. PICKMAN

1. INTRODUCTION

Leaking fuel rods, although undesirable, are tolerable in LWR's provided the activity release is within acceptable limits and they are not subject to rapid unpredictable deterioration. This review of failure development is based largely on a literature survey backed up by some personal interpretation and judgement where conflicting evidence appears to exist.

In any review of this field it is necessary to recognise that there are several known causes of defects which lead to a leak of fission products in LWR fuel rods. Some are related to rod or system design, some to reactor operational aspects and some to defects in fuel rod or reactor manufacture. The type of defect and how it develops during continued operation is often dependent on the cause. A brief review of known causes of defects in LWF fuel rods is included.

Primary defects in fuel rods c-jy not be single, isolated, leak sites, PCI defects being a classical example where very many penetrating cracks can form in a single fvA rod.

However, all such leaks formed by the same event are primary leaks. The only secondary defect mechanism is internal i. -Iriding, although knock-on events can follow on the same, or neighbouring rods, from defects that cause an interference with local coolant flow.

- 1 - The major part of this survey comprises information on leak rates fro» typical defects, causes and rates of development of primary leak sites, mechanisms and rates of development of secondary hydride defects and analysis of activity release measurements in an attempt to characterise the number and severity of defects present.

2. CAUSES AND CHARACTERISTICS OF DEFECTS The early history of defects in LHRfS was reviewed by Locke. This showed that internal hydriding, crutf deposition, fretting (both by spacers and foreign bodies) and power ramping (PCI) have been the principal causes of defects in BWR fuel rods. In PWR's crud deposition has not been a cause of defects because of lower primary circuit copper content. Fretting experience has been similar, although one source of debris, pieces of stainless steel wire from steam separators, is absent. For various reasons PWR'S have also been freer from PCI defects, but have

experienced problems of clad flattening following U0? densification, rod bowing and rod lengthening. These latter two phenomena have not led to defects but to life limitation. Rod lengthening also led to some life linitation in a few BWR's. Except when enhanced by dense crud, waterside corrosion has not caused defects in BWR or PWR fuel, despite some early pessimistic predictions, nor has the associated hydrogen, pick-up. The move in recent years to higher burn-up is now, however, causing concern in relation to clad corrosion as a possible life limiting phenomenon, which could cause defects in both BWR's and PWR's.

In looking at the characteristics of defects, some can be forgotten as not relevant to modern BWR's and PWR's operating in Sweden. The causes of relevant defects are fretting by foreign bodies and PCI, with hydride defects also of interest because their mechanism is relevant to secondary

- 2 - hydriding. Fretting defects have the characteristic of rapidly enlarging from a small pin-hole leak to a size determined by the object causing the leak. If caused by a spacer for example they can become large leaks with exposed UO very quickly. The PWR baffle jetting defects and some in elements with early wire type sparer g»-ids in BWR*s were of this type

(Fig. 1). Fretting i-i caused by a tapping or tapping plus sliding contact between two components. It is probably due tc enhanced corrosion because of the continuous removal of the protective oxide layer which is trying to form, assisted by welding and tearing of asperities or fatigue (Pi of asperities. "' It is believed that a number of recent defects in

Swedish BWR's were caused by fretting between foreign bodies trapped in spacer grids and the fuel cladding. The size of penetration produced will depend on the geometry of the foreign body and its freedom of movement.

It is a characteristic of fretting damage that, once started, it progresses rapidly. However, fretting caused by foreign bodies trapped in spacer grids will have a random start time, depending on when the foreign body was captured.

PCT defects, unlike fretting, form almost instantaneously at the time of a power ramp and may be very minor single cracks or a number of large axial cracks of substantial width. The cause of PCI defects has been exhaustively studied. It is a form of stress corrosion cracking with iodine as the most probable corrosive agent. The stress level at which cracking starts is below the yield stress and reduces with irradiation damage to the cladding, typically reducing with burn up from around 500MPa to 200MPa. Stress concentrations at radial pellet cracks which open during a power increase and bi-axiality at pellet ends are also is-pcrtar.t factors. Some of the minor PCI defects no longer leak fissicr products when the power level is reduced and may be effectively sealed for the remainder of the fuel rod life (Fig. 2).(4'5f6f7) Some rods with PCI defects have developed severe secondary hydriding defects near both top and bottom ends in 15 days. (4) The more typical PCI defect probably lies between these extremes and may be capable of continued safe irradiation for a prolonged period.

Hydriding, which was the early classic defect in both BVffi and PWR fuel rods has been extensively studied/8'9'10'11'12'13'l4) The problem was soon solved by controlling the moisture and hydrocarbon content of fuel rods. The hydride problem was first reported by workers from Savannah River ' with UO tubular fuel in Zircaloy-2 cladding. The problem was compounded by an adverse orientation of ZrH, , precipitates i .0 and low clad operating tenperature, but shortly thereafter was also reported from Big Rock Point and other early BWR's. This problem, while it no longer exists for fuel rods manufactured under proper control, is of interest because of the light it can shed on the mechanism and kinetics of secondary hydridinp. The H_0 contained within a fuel rod when raised to power reacts preferentially with the UCL to raise the stoichiometry and to a lesser extent with the inner clad surface.

There is an equilibrum between U0_, H_ and H?0 which limits the amount of oxidation of UCL in the presence of hydrogen, (7) but radiolysis produces a HO radical which can oxidise UO further despite the presence of significant H. partial pressure. This reaction generates further H .

The consumption of the oxygen from the contained H?0 in the fuel rod then leads to a situation where the contained gas is mainly H_ + He (the fill ffas) and a small amount of fission product gases. In the absence of a significant amount of oxygen the inner clad surface is liable to suffer

-i» - rapid hydriding and in sone cases the fornat ion of a surface layer or patch. ' However, even thin films of ZrO are normally impermeable to H , but in the absence of oxygen they develop defects and eventually become permeable. The defects may be anion vacancies, dislocation networks, sub-grain boundaries, aicrocracks etc. It is also known that fast neutrons and fission fragnents can danage the ZrO layer and that fission product attack can damage Zr0o films and make them more permeable. Fluorine, which can be present from pickling solutions (HNO /HF) or from the UO is particularly effective at breaking down the ZrO_. Damajre to ZrO_ films has been confirmed by electrical resistivity measurement.'" In the primary hydride defects the initial hydriding appears to have occurred at only a limited number of sites but with a pre-dispositior. for hot spots (high power regions), crevices and heat-affected zones where oxide films are absent cf thinner than elsewhere. Once attack starts the H_ flux entering the surface must exceed the rate of diffusion to the colder outside surface for a massive local hydride inclusion to forn. These inclusions ("sunbursts") lead to high local tensile stresses in the outer part of the tube because of the I6t volume increase and the final leak path develops along radial fingers of ZrH c which penetrate to the outer surface, Figure 3. Cladding bulges or blisters, visible on the outside are almost always a sign of a hydride sunburst with a large volume change as the ZrH, , forms. Hydride

defects may be severe and release a lot of activity from exposed UO,,, or they may be relatively minor and deteriorate only slowly. (17) Once a hydride defect starts to leak then it nay no longer continue to develop and cases have been seen in which continued diffusion of H_ to the cooler regions on the clad o.d. results in the zirconium hydride near the bore, Mfi) which formed first, reverting to zircor.um. It may be at this stage that the cuter, now-hydrided, region of the cladding bulges and often

- 5 - breaks away leaving a defect riddled thin inner layer of cladding. Figure **.

The tiae to developoent of leaks caused by primary hydride defects is generally short, perhaps typically up to 100 days, but soae have taken much longer to develop (17) probably because sose additional stressing (from clad creepdovn, UO swelling, small power variations) is needed to propogate cracks. There could indeed be complex situations such as the diffusion of H fron a non-leaking patch to the colder outer surface producing high local bore surface stresses and leading to a stress-corrosion failure. One thing is clear, that if soae hydride patches form and disappear as a result of diffusion then a zone of weakness is likely to be left at the bore surface.

An unusual effect has been reported in secondary hydriding, (7 ' 19) but which could also operate in the primary hydriding case. This is the accumulation of H at cold spots such as pellet end positions or at non-heat generating objects such as insulating pellets, other non-fuel pellets or Zr spacer washers within the fuel stack. A similar effect has been noted in the end cap region ' where ZrH. , formed at a hot spot diffuses to the cooler end-cap regions.

3- EFFECT OF DEFECTS ON FUEL PERFORMANCE

When a fuel rod defects, ie. develops a hole or crack in the U0_ containment of cladding, end caps and welds, there will be a pressure equalisation between coolant and rod interior. For very small leaks a pressure gradient from the coolant at the entrance to the leak and the internal extremities of the rod may persist because the internal reaction

- 6 - rate rray exceed the flow rate through the defect. However, in most cases substantial pressure equalisation will take place within minutes, aided by the reduction in elastic 'compressive strain* in the cladding. In BWR fuel rods the water will boil during its passage through the defect and into the fuel/clad gap, so reducing the inflow rate. The coolant in the gap will be in the vapour phase, but some condensation is likely in end plena. The situation in PWR rods is less clear and it seems probable that the fuel/clad gap will contain a 2-phase mixture of water and steam with some condensation again in end plena. This intrusion of water and steam into the fuel/clad gap would be expected to increase the gap temperature drop. There are many indications that such is the case from (21 22) observations of UO structure, particularly grain growth. ' The increase in centre U0? temperature at ^OOW/cm rating has been quoted as (20 ?3 2If) 150-200°C ' 3) for rods with closed gaps. This is attributed to a change in UO stoichicometry, but the speed at which increased fission product release occurs makes this a dubious conclusion. It seems more likely that the gap temperature drop increase is initially responsible.

This »ap regime therefore has important consequences not only for UO temperature and hence P.P. release, but in the transfer of released F.P's to the leak site and in chemical reactions with both fuel and cladding.

The initial formation of the leak is generally followed by a burst of activity, following pressure equalisation, the exact mechanism of which is not clear, unless it is purely a diffusion process during which the stored inventory of fission products is released. Apart from escape by recoil or knock-out, where UCL is directly exposed to the fuel rod exterior, the escape of fission products by diffusion is not likely until the initial inflow of coolant to equalise pressure is complete. This time period is a function of leak size, pressure differential, clad

- 7 - internal volume and fuel/clad gap. From comparison with other Indicators

of leak development, particularly rod length change, it is known that the

first appearance of activity in the coolant may be within seconds. It

would seem very probably, however, on the basis of the known flow

impedance in the gap of some fuel rods that times of hours could be

involved for some defect rods.

Pressure pulsing or power cycling both result in flow reversals through

the defect and enhanced activity release from gap to coolant, especially

of the radio-iodines which tend to be trapped on the clad inner surface,

but can be washed out. Release characteristics are discussed later in

more detail, but there is abundant evidence that the steady state release

in a rod with a defect increases by one to two orders of magnitude as

compared with the release in a similar intact rod. Iodine release is

generally increased less, or takes longer to reach equilibrum and iodine

release from a defect PWR rod i? larger than from a defect BWR rod.

This higher iodine release is associated with the presence of a liquid

water phase in the fuel/clad gap.

H. DEFECT DETERIORATION MECHANISMS

Ths literature is full of information on the behaviour of defect fuel

rods and in particular on the formation of secondary defects by a

bydridinp; mechanism akin to that which causes primary hydride defects,

but with important difference. Deterioration can be measured by an

increasing release of activity to the coolant, although this may only

reflect a higher F.P. release rate from the U0? because of a higher

temperature, caused by the defect, leading to a larger source term in the

fuel/c'ad gap.

- 8 - As has been mentioned before, all defects do not deteriorate, some effectively close after the causative event (e.g. a power increase) is no longer operative. In fact, of all forms of defect, the PCI type in its less severe form is the most likely to re-seal. Hydride defects are most unlikely to do so because of clad strain relaxation on pressure equalisation.

Some defects deteriorate rapidly from the moment of formation, others (25 261 more slowly, ' such that on typical discharge criteria it may be (25) necessary to discharge them in less than 20 days or un-necessary to (171 discharge them in 389 days/ "

Deterioration i.s very much dependent on the size of a primary defect and to a lesser extent on the position and type of defect. Many experiments have been performed using fuel rods with artificial defects to study activity release, deterioration rate and mechanisms. Most such experiments have used pre-drilled small holes (0.15 - 0.5mm) as defects » » » » although Chenebault and co-workers also us^d a (oil) device to cut a small slit over the fuelled region at power. It is now widely recognised that most such defect tests are unrepresentative of the behaviour of operational defect rods. For some reason, possibly related to the inner clad oxidation or lack of clad creepdown it is rare for massive hydriding to be found, although accelerated hydriding is frequently found in such tests and indeed in ex-reactor capsule tests.

The only positive findings of massive hydride in an artifically defected (23) (29) rod were reported by Janvier £.nd Freshley.

Defect development nay be local or remote from a primary leak and is clearly influenced by internal reactions with the fuel and cladding, by

- 9 - temperature enhancement effects and by fission product attack as well as by the type of initial defect. Mechanisms of defect development have

(5 7 11 been proposed by many PeoPle ' ' '12,13.14.17,22,25,26,29.30,31,32,33)

following the early theory of Markowitz and the supporting work of Marshall '12) and Sawatzky. 3 The Markowitz theory has stood the test of tine, but needs fine tuning to explain some detailed difference in behaviour.

Before considering the whole picture of the mechanisms involved, some of the contributing factors in the process are discussed below.

4.1 Clad Bore Surface Oxidation The waterside corrosion of zirconium alloys in BWR's and PWR's is well documented and understood. Corrosion rates are higher in BWR's because of the higher 0 content and probably because irradiation damage in the ZrO film is greater at the lower surface temperature. Corrosion takes place by diffusion of oxygen ions to the metal/oxide interface and electronic conduction to the oxide/coolant interface where most of the corrosion H_ is liberated

and swept away in the coolant. Some H? does, however, also get transported to the oxide/metal interface and enters the metal. The

so-called H2 pick-up averages under 10$ in BWR's and over 20$ in PWR's. The ZrO films formed are protective (cubic rate kinetics) at least up to a thickness of about 50 microns, after which there is a transition to a higher rate, probably as a result of

recrystallisation in the ZrO? film or an allotropic transformation.

Some of the more general principles apply to oxidation of the clad bore surface, but there are important differences. Firstly, the H

- 10 - not taken up into the ZrO , and transferred into the clad remains

in the rod internal atmosphere. Any difference between BWR and PWR

will rapidly disappear with distance away from the leak and the H

content will increase as 0 is used up in reactions with the U0p

and Zircaloy. Other factors which will influence the internal clad

corrosion rate are the clad surface temperature, (a function of

surface heat flux) and radiolysis in the gap. It can be expected

therefore that the ZrO thickness on the clad bore will be at a

maximum close to a primary defect and will reduce with distance

away as the PH O/PH ratio reduces. Although the mechanism of

hydrogen pick-up is not fully understood, it would be expected to

increase still further as the PH O/PH ratio reduces in line with

the greater H_ pick-up in PWR compared with BWR, Such phenomena

are indeed seen.(5'21'22'23) Local ZrO thicknesses of 15-25

microns, decaying to 5 microns at an axial distance of 500mm, were (22) seen in defected fuel rods from Obrigheim ; of 20-40 microns in

rods with a 0.3mm drilled hole irradiated for 42 days at ratings

between 200 and 500 W/cm in French loop experiments (5-10

microns expected); and 50 microns close to a primary defect.

Enhanced H? pfek-up has been noted by Clayton and Neimark

and can be clearly seen in PIE photographs by Forsyth and (21) Jonsson (compare Appendix 7 with Appendix 18). The H?

concentration in the outer half of the clad thickness is greater

than in the inner half because H diffuses down temperature

gradients in zirconium alloys, an important property in the

secondary hydriding saga.

1.2 UO^ Oxidation and Gap Conductance

When water/steam gets into the fuel/clad gap it is also free to

- 11 - penetrate into interconnected porosity, rauial cracks and other discontinuities in the U0?. Some U0_ oxidation will occur, raising the stoichiometry to U0? with release of H_. Some H will also be released by oxidation of the cladding and the partial pressure of H in the gap atmosphere will build up. Further oxidation of

UO would normally be buffered by H as an equilibrum exist? between U02+. and PH2/PH20 for different temperatures. '

However, this equilibrum is disturbed by the presence of various other radicals or molecules produced by radiolysis and HO in particular is believed to be able to oxidise UO to much higher 0/U ratios even in the presence of high H partial pressures. There is thus a strong sink for 0 and in the absence of any corresponding sink for H (the ZrO? layer is normally protective) the PH /PH O ratio builds up.

There are very many observations of UO oxidation, although unfortunately not too many in which actual stoichiometry, structure and composition of second phases have been determined.

Hyperstoichiometric UO has a lower thermal conductivity than stoichiometric material, as have 0>.OQ and U_OQ. The observations of

U0? oxidation are unfortunately very inconsistent. In some cases a

Widmanstatten type structure with U^0q needles has been seen in the (29 34) central regions of U0_ pellets, ' in others there was little or no departure from stoichiometry at the U0_ centre, but an increasing 0/U ratio towards the pellet rim. It seems clear that in some cases both U^0q and U_0« have formed. Excessive U02 (7 22) oxidation in the neighbourhood of a clad leak has been seen ' with in some cases erosion of an apparently powdery or friable repior. of oxidised fuel. Such local oxidation can cause a local

- 12 - bulge in the cladding resembling a hydride defect. The most const n manifestation of UO oxidation is a surface reaction layer of a

rather structureless phase which often penetrates surface cracks and pellet interfaces, for example Appendix 4 in reference 21 or

Figures 13»I1* and 15 in reference 30. It is not absolutely clear, however, whether in some cases these are not compouno.i of U, Zr z.~d

fission products. Strange structures in U0_ have als? been (21) reported apart from the expected indication of higher operating

temperature. Such structures nay be artefacts, or rejections of

severe grain fall out at a particular radius observed in fuels from

some of the Risp bump tests.

The operating temperature increase seen in a number of experiments,

(inferred from UO grain growth or columnar grain formation) has

been attributed to an increase in 0/U ratio in the fuel rather than (23) reduced gap conductance. Janvier and co-workers claim an

increase of 150°C in centre temperature for a PWR rod with a defect

in the plenum, 180 micron gap, irradiated at 505W/cm. Schuster and

co-workers on the basis of effective fission gas diffusion

coefficient believe that the change in stoichiometry has resulted

in a 200° increase in UO temperature in both BWR and PWR rods, but

also observe that at the higher 0/U ratio F.G. release will be

higher even at the same temperature, so the 200°C may be an

overestimate. In both cases these temperature increases are said

to be for closed gaps so that gap conductance is not increased. It

would be very surprising if this is so, particularly for BWR fue]

rods with steam filled gaps. It is also unlikely that most fuel

rods will have closed gaps over a large part of their life. It

could well be that much larger temperature increases than 150 or

- 13 - 200°C can arise in defected BWR fuel rods with significant at - power fuel/clad gaps.

Secondary Hydriding Mechanism

The mechanism for primary hydriding has been discussed earlier.

Secondary hydriding differs in several important respects:

(a) The site of water/steam ingress is a fixed point, the primary

leak, whereas in primary hydriding the water is uniformly

distributed, mainly in the U0 .

(b) the rate of water/steam ingress is very variable, depending

on leak size and the quantity of HO is not fixed as in

primary hydriding.

(c) the ingress of water/steam is continuous and may increase

with time.

The mechanism is fairly clear and has been described in numerous (5,7,15,17,22,26,32,3^,35) . . . . , papers. ' ' ' ' ' ' ' ' TIt is caused by a low oxygen

availability, such that defects in the ZrO? film on the clad bore

cannot be repaired, and a sufficiently high H availability to

build up a layer of ZrH , on the bore surface in competition with i • o

the diffusion flux of H to the colder clad outside surface. A

sufficiently rapid flux can only be provided, according to

Markowitz by radiolysis (2H 0 ——> H202 +H ) and by liberation

of further H from reaction of the HJ3 with U0 . However, the

available H? must also depend to an extent on leak size, especially

for very small leaks for which pressure equalisation is slow.

Apart from this a high flux may be available at a low H2 partial

pressure if only a very small area of clad inner surface is

permeable. For example, Proebstle and co-workers state that at

an unprotected spot of 10 cm* area hydride can build up at the _4 bore if the H^ partial pressure exceeds 10 atm. The further requirement is that oxygen or H_0 partial pressure must *>e so low as to Inhibit the ZrCL film repair process.

It emerged from studies by Joon on primary hydride defects that the ratio of mass of HO to rod internal void volume was the most effective criterion for failure. This type of relationship appears to be borne out by capsule test studies of leaking rods in steam/H» atmosphere, where the H absorbed was greater the smaller the gap and by the French Crusifon 2 test. It is particularly relevant to small gap fuel rod situations.

It has been shovn by Marshall that exposure of oxidised

Zircaloy in H can lead to breakdown of the ZrO? film, dependent on

H partial pressure and temperature. Time to breakdown is reduced with increasing H pressure and temperature e.g. at 15mm Hg H_ pressure lOh at i»00°C, 4O-6Oh at 350°C, 100h at 100mm Hg at 300°C.

Lunde found autoclaved and autoclaved and scratched surfaces more protective in H? at 400°C than pickled or sand-blasted, and

J that sand-blasted tubes picked up H? (7.6x10 imn Hg) most rapidly.

She also found that contamination with HF led to very rapid sunburst formation on the pickled and sand-blasted surfaces. There is abundant other evidence that ZrO_ film breakdown is assisted by halides or halide containing compounds such as Csl.

Much of the early work on the influence of 0? starvation on (37) hydridinr was reported by Shannon. He found the oxidation rate il at 400°C was pressure independent from below 1mm Hg to 7.6x10 mm Hg 4 and catastrophic hydriding was inhibited even up to 5.17 x 10 mm Hg

- 15 - hydrogen pressure. However, he found that if the oxidation rate was s jpressed by depletion of H90 then catastrophic hydriding took place within a ftw days at .. pressures as low as 7mmHg and temperatures as low as 150°C. Shannon found the critical H 0/H ratio below which catastrophic hydriding occurred was !0~ .

Others have reported lower values, 10 at 320°C and 10~ at 313°C.

Zima thought the ratio was a function of H pressure and produced the expression (pH 0) protective £0.2pH . Proebstle believes that Shannon's work shows that it is the absolute pressure of the water which controls the potential for massive hydriding.

Before going on to consider typical behaviour for defects of different sizes and types, it must be noted that in real defects arising during reactor operation there is a dynamic equilibrum involving:

(a) flow into the rod interior under the influence of the

pressure differential.

(b) a post-pressure equalisation flow into the rod interior to

compensate for pressure reduction caused by 0_ and H

absorption by chemical reactions or solution.

(c) a diffusional mixing within the fuel/clad gap involving HJ),

H , residual He and fission products.

(d) diffusion into and out of the clad defect of coolant and

fission products.

The equilibrum can be disturbed by power changes or pressure fluctuations as evidence the F.P. release spikes due to power cycling. In this case for PWR fuel with some liquid water in the gap the spikes are richer in halogens than rare gases and for

- 16 - the longer half life halogens such as I and I release fractions are 5 to 10 times higher than for non-cycled fuel. The pressure pulsing effect was seen by Blackadder and co-workers in an experiment with artificially defected rods in the R2 reactor at Studsvik. Small pressure pulses (1 bar reductions in 78 bar every 2 hours) were caused by an electric heater in the loop pressuriser. They were sufficient to cause a 20-30% increase in measured activity in the loop with a delay time of 5 to 20 minutes.

5. TYPICAL DEFECT BEHAVIOUR

Defects will behave differently depending on size and type. The various processes which must now be considered in relation to particular defects have been described. Application is complicated by factors such as initial design and burn-up. Different types of defect are therefore considered in relation to enlargement of the primary leak site. Three typical types of defect are the minor PCI defect, the fretting defect and the primary hydride defect. More severe PCI defects are likely to behave similarly to fretting defects.

5.1 Small PCI Defects The most likely small defects in fuel cladding are short-narrow cracks formed by a power excursion and known as PCI defects. It is known that incipient non-penetrating defects can be produced in this way as well as a whole spectrum of penetrating cracks, ranging from ones which don't leak to massive splits, Figure 5.

The small leaks may not be detected in reactor ' or may produce small activity releases. ' Such defects have been

- 17 - characterised by PIE measurements as typically cracks of about 5mm length and 1-2 microns width, equivalent in area to an 8x10~ mm diameter hole. Assuming that rod internal pressure at the time the leak forms is significantly lower than coolant pressure then there will be a coolant inflow which will persist for some hours or even longer since pressure equalisation at these leak sizes is slow.

For example, Locke quotes calculations for flow through a round hole showing that pressure equalisation would take a few hours for a 1.5x10 mi", diameter hole and about 3 years through a 10 mm diameter hole. Times for pressure equalisation for narrow cracks of the sane area may be longer, perhaps by a factor of 3 or 4. The practical consequence is that pressure equalisation would never occur for the smallest leaks because internal reactions would maintain a pressure gradient. If the power increase that caused the leak was a short transient event the crack is likely to close and effectively stop leaking. In this condition re-sealing by oxidation of the crack surfaces is a strong possibilty. Figure 2 (7) shows a small PCI crack said to have re-sealed in this way.

Blackadder and co-workers have examined the opening and closing behaviour of such small cracks and conditions for permanent opening. If the higher power which caused the crack persists for some hours then creep of the cladding is likely to prevent re-sealing, or if it was initially high enough to produce local plastic deformation at the crack extremities then re-sealing is unlikely.

This type of defect will admit some coolant, either for a short time only, or at a very slow rate. It may then behave rather like a primary hydride defect if the rod is at low burn-up and there is

- 18 - ready transport within the gap for the water/steam admitted.

Secondary hydrides can be expected either at hot spots on the clad inner surface (peak rated position in BWR, upper half in PWR) or where inner surface oxide films are at a minimum, such as weld areas or other heat affected zones. For rods with narrow gaps, perhaps at medium to high burn-up, these very small defects may admit water/steam so slowly, and it may move within the gap to eo.ualise pressure so slowly, that low H_O/H_ ratios develop close to the initial leak site and accelerated or massive hydriding occurs in the vicinity of the primary leak. This scenario seems

credible as a mechanism for deterioration of a primary leak (17) site, but no well authenticated case has been found. These

small leaks have been said to enlarge by oxidation and

erosion. ~>J ' Evidence for erosion is tenuous, but internal

oxidation near where the crack meets the inner surface, together with some local U0? oxidation can lever the clad outwards (oxide

jacking) and so open the crack further. This is not likely to be a

rapid process, as although clad oxidation at the inner surface will

be accelerated by temperature and radiolytic effects, it is still

relatively slow. Locke shows release curves for small hydride

defects over a period of 100 days, Figure 6. The activity release

increases by about 50% from these defects in 100 days and this is

the sort }f increase that could easily be expected from local

oxidation. This represents fuel rods at stage 2 of the 4 stage

process described by Locke. ' It is interesting to note that

the noble gas and iodine activities both increase at a similar

rate.

These small PCI cracks can also result in secondary hydriding

- 19 - remote from the primary leak, a situation which is most likely to arise for saall leaks with relatively new fuel of high rating. In such cases there is ready penetration of steam/water along the rod with a rapidly reducing H 0/H ratio such that at soae point or points the ZrO_ is locally broken down and massive hydriding occurs. Heterogeneous nucleation of hydrides has been observed by Marshall . In practice a number of hydride patches are seen to develop, together with accelerated hydriding in the sane regions. Other observations are of hydride annul i at pellet ends and hydridinir at end caps, either in crevices or at hot spots where the end cap is in contact with the bottom UO pellet. Typical sunbursts at an early stage and after development of a leak are shown in Figure 3. It is of course possible for non-penetrating sunbursts to form and tc disappear due to lack of continued H supply at a sufficient rate, and diffusion to the outer surface. Such events may well lead tc residual areas of weakness, but there is a lack of evidence that such areas exist. Much must depend on whether local deformation of the cladding is needed to accommodate the 16.?f volume change. If it is, then the re-conversion of the hydride to Zr is likley to leave a porous or cracked area on the bore surface.

It has been suggested by Garzarolli and co-workers (22) that localised oxide patches up to 300 microns thick, which have been seen opposite radial pellet cracks, act as nuclei for hydride sunbursts. The relatively few observations of such oxide patches and the frequent occurrence of hydride sunbursts as secondary damage features casts doubt upon the validity of this suggestion. A pood example of such an oxide patch opposite a radial crack is

- 20 - (21) shown by Forsyth and Jonsson, Figure 7. The local deformation caused by the volume change, the local effect of stress on ZrH , 1*0 re-orientation, as well as the accelerated H? pick up and migration of H to the colder outer surface can all be seen in this figure. The appearance of such patches is very similar to that of hydride patches, but judged by this example the morphology is different, compare Figures 3 and 7.

In summary, small PCI or similar defects are likely to slowly increase in size, possibly with accelerated hydriding öf cladding close to the defect, due to local oxidation of cladding and U0_ leading to some crack opening. At some stage the rate of HO inflow to maintain the pressure equilibrum as H and 0 are absorbed is sufficient for the H partial pressure at a distance _2 from the defect (where the HO/H ratio is reduced to<10 ) to be sufficient (>7nm Hg) to cause H absorption at a rate exceeding the diffusion flux to the cuter surface. The locality, or limited number of localities, where the rapid H_ uptake occurs are those where the ZrO layer first becomes permeable by reduction by H_, by oxygen diffusion into the metal, by mechanical or irradiation damage or by volatile fission product attack. An alternative is that the first region to become permeable is one where the internal ZrCL layer was thin, say in the weld region. The attainment of the critical HO/H ratio depends on reaction rates of the oxygen with UO- and Zr exceeding the diffusion rate of fresh H-0 to the reaction front. The bottom end of a rod may act in a special way as a crevice, making penetration of fresh HO difficult. Partial penetration PCI cracks or other defects will also act as crevices and centres of hydriding, Figure 8. The effects of increasing rod

- 21 - rating on the development of these defects is expected to be: (a) a modest increase in rate of increase of the defect size (b) for a given defect size a higher reaction rate with UO and

cladding leading to an enhanced rate of H?0 inflow and an earlier availability of a sufficient flux of H . (c) a reduction in the H pressure necessary to break down the ZrO film. (d) more volatile fission products available to damage the ZrO» film. (e) more rapid diffusion flux in clad wall and higher H solubility.

All these factors except (e) suggest a more rapid deterioration of the rod at higher ratings in line with the Locke curve »33^ Figure 9, which has attracted fairly general support.(5 7 22)

5.2 Fretting Defects Fretting defects may be caused by trapped foreign bodies, usually either at inlet to the bundle or at spacer grids. They may also be caused by rod or bundle vibration or by excessive coolant cross flow as in the PWR baffle jetting defects. Some early fretting defects were caused by inadequate spacer design, particularly spring relaxation or fracture. The characteristic of fretting defects is that a substantial sized hole develops rapidly, the rate depending on the size, shape and attitude of the object that is causing the fretting. In this respect they are similar to large PCI defects. In such defects pressure equalisation will be rapid and flow through the defect will not limit the supply of H- for attacking the cladding. With such a ready availability of H-0 to

- 22 - the rod internals no hydriding local to the defect is at all likely, but internal oxidation of cladding and UO , both local to the defect and for some distance axially in both directions, is expected. The same factors govern secondary hydride defects as in the small PCI defeat case previously discussed, but here an increase in the size of the defect over a long period of time is not necessary in order to make H available at a sufficient rate.

In other words leak siie is not rate determining for massive hydride formation. If there is a significant fuel/clad gap at the tine of leak formation then the necessary low H 0/H_ ratio is less likely to be achieved than with a very small gap, because the diffusion of HO as steam or oxidising radiolytic species will be faster. For the critical H 0/H ratio (or critical 1 .. steam pressure) to be achieved means that a faster rate of consumption of oxygen is needed. Thus higher ratings will favour secondary hyiride defect formation and local rod internal conditions will have an influence. Such conditions are clad internal temperature and gradients (different for BWR and PWR cases) U0? temperature and connected porosity/cracks. Also the axial rating gradient may be of importance in terms of local gap variation. It is a combination of all these factors that determines whether secondary hydrides form and if so where in relation to the primary leak. Typical behaviour is that secondary hydride defects will form some long way away from the primary with a number of small hydride patches along the clad between primary and secondary leaks. Other factors that favour sunburst formation, as has been said earlier, are heat affected zones (welds or welded or brazed on appendages) and (22) crevices. Garzarolli found that secondary defects tended to

form at the position of maximum clad inner surface temperature

- 23 - which is in the upper half of the rod in PWR fuel and at the position of maximum power in BWR fuel rods. For the PWR rods (from

Obrigheim) the mean distance between primary (large PCI) defects and secondary hydride defects was 2.25m. In one case from the

SGHWR (pressure tube BWR) a defect caused by a broken plenum spring which fretted through the clad wall caused a hydride defect in the bottom end cap weld over 4m away which detached itself on a reactor (17) depressurisation after 25 days. There were numerous non-penetrating hydride defects along the whole length of this rod.

In summary, for these initially large primary defects deterioration rate may be large in terras of rate of increase of activity release because of the plentiful supply of CL to react with the UO close tc the defect. The formation of hyperstoichiometric UO , U^O. or

U,0Q over an increasing distance, together with local clad bore surface oxidation and degradation of gap conductance over a long length of rod leads to a large increase in fission product release over a short period of time. This effect was shown by some of the (25) fuel rods discussed by Locke, Figure 10. In such cases thick

(1000 microns) layers of t' 0o have been seen close to the 3 8 (22) defect and this friable material can easily escape into the coolant. A similar example with loss of fuel has been described by (7) Davies. This is the more severe end of the spectrum. The typical fretting or medium sized PCI defect probably deteriorates at a tolerable rate for some months and then develops large secondary defects by the hydriding mechanism. What remains something of a mystery is why some medium sized defects develop secondary hydride defects within a few days or tens of days, while others survive for perhaps 100 days or more. Locke has proposed a relationship to surface heat flux (Figure 9), but there

are exceptions to his rule which suggests that a rod with a peak

surface heat flux of 80W/cm* would survive without hydriding for

about 500 days. Reds with double this heat flux, l60W/cm*f would

fail by hydriding in 50 days. For a 17x17PWR, 80W/cm* is 206 W/cm

and for an 8x8 BWR it is 266W/cm. If the Locke curve is reasonably

correct, or conservative, then much fuel in modern PWR's and BWR's

is not at risk of secondary hydriding.

5.3 Hydride Defects

Primary hydride defects are as likely as any others to deteriorate

and to lead to the development of secondary defects. There has

sometimes been some confusion where widely separated defects have

been found, both with the characteristics of secondary hydride

defects. In principle one could be a primary hydride and the other

a secondary hydride, but if the burn-up is substantial as in the (21) Finghals 1 defects reported by Forsyth and Jonsson, then it is

extremely unlikely, since an incipient primary hydride defect would

be eliminated by diffusion, and although a defect could eventually

form at the site because of a residual weakness, the defect would

not resemble a hydride sunburst. It may be that in the Ringhals 1

defect rod H2 there was a small undetected PCI defect.

How then does a primary hydride defect develop after leak?

Firstly, because of the immediate access of H?0, hydriding will no

longer occur, but rapid oxidation of some of the hydride in the

sunburst, plus diffusion of some H? towards the outer surface and a

spread of H radially away from the leak as a hot spot develops is

likely. These processes can cause a lamination to form which

- 25 - produces the typical blister and the outer half can spall off as

(OQ) seen in the Ringhals-1 defect rod R1-1015-A8. A classic

/ 4 O \

example of lamination is shown in Figure 4 from work by Bain.

While this deterioration at the site of a primary hydride defect is

proceeding, there is the same probabilty of secondary defects

forming as with other primary defects of a similar size.

Although hydride defects may look dramatic, activity release is not

always large, nor deterioration rapid. All the data in Figure 6 is

for hydride defects with linear heat ratings in the range 236-449

W/cm (47 to 90W/cra2). Their behaviour therefore is consistent with

the Locke curve. It is interesting that the Locke curve delineates

areas of successful operation (time/surface heat flux) of defected

fuel rods with respect to secondary hydride failure. Deterioration

of the primary defect site or some set level of activity release

are not necessarily cause for falling outside the successful defect

operation area. However, it may be that there is indeed a

correlation between activity release rate from a primary defect and

liability to form a secondary hydride defect. If this were so it

would tend to suggest that large leaks would always lead to

secondary hydridin? provided the surface heat flux was high enough

and that ready access of 0 at all positions would not prevent it.

5.4 Other Influences on Defect Behaviour

In some cases defects have failed abruptly for reasons that are not

related only to local deterioration processes for primary defects

or to the formation of secondary leaks by hydriding.

Depressurisation, waterlogging, power ramping or repeated power

cycling can all cause non-penetrating secondary damage to fail by cracking, despite the fact that delta - ZrH does have some ductility above about 16O°C. Fuel rod vibration may also lead to fitigue failure at local stress raisers such as pellet end positions.

After a secondary hydride defect has formed it is possible that a local internal coolant flow between the two leak sites may develop because of the pressure differential in the coolant along the element. Such a flow could limit any further hydriding between the two leak sites.

The state of stress in the fuel rod cladding can have a marked effect on the failure mode in the presence of non-penetrating hydrides or possibly even with enhanced H content as a result of accelerated hydriding . Normally fuel rod cladding will have a compressive hoop stress as a result of the coolant pressure, but it can also have an axial tensile stress, especially in the lower half when there is no bottom end gas plenum. Such stresses can arise even in the absence of true diametral interference as a result of a mechanical jamming process. Stresses of this type relax as a result of creep and irradiation growth, but can be regenerated following power changes. A number of cases of transverse cracking close to the bottom end of fuel rods have been reported (17) Figure 11. Such failures are usually linked to local hydride which cracks under the axial tensile stress. The other main source of stress is from power ramps especially following gap closure by creepdown and

U0? swelling. Such ramps can cause longitudinal cracking originating at positions where hydride patches have formed. There is also the possibility that delayed H cracking could extend

- 27 - partial penetration or even penetrating cracks and that eventually fast fracture could occur if the critical crack length (probably about Ucm) was exceeded. Such phenomena were the cause of cracking in Zircaloy-2 pressure tubes in CANDU reactors, but it seems unlikely that they will have an important effect in clad cracking because the high stress levels are rapidly reduced by creep of cladding and fuel.

It is unusual for a secondary hydride defect to fail other than by a local bulge or blister, with some spalling, unless influenced by some applied stress which can cause axial or transverse cracking.

Axial cracking is frequently seen, presumably reflecting a tensile hoop stress in a region with a hydride patch or patches present.

However, such axial cracking is often reported close to rod (UQ) ends where the pellet/clad interaction should be low. It (4) seems likely that the axial cracking reported by Garlick was caused by secondary hydriding in a region with tensile hoop stresses present from a power ramp which caused the primary defect.

Thi.» raises the interesting pos5ibility of a special type of secondary hydridin^ in which a radially penetrating type of hydride attack occurs rather than the typical hemisperical nodule. Such a mechanism could be the origin of very major axial splits. An axial crack seen near the bottom end of the Ringhals 1 rod R1-1O15-A8

could be of similar origin, but it is not known to have experienced a power ramp. However, this rod had been irradiated for a long

tine with defects present and interaction, leading to a clad

tensile hoop stress, could have built up by clad bore surface oxidation and U0? swelling, plus enhanced thermal expansion due to

oxidation causing a reduction in thermal conductivity.

- 28 - There are isolated cases of rapid deterioration which would seem to

fall well outside the Locke curve, Figure 9. One of these is the (in rod reported by Garlick referred to above. Figure 10(a) shows

the off-gas activity released by this element PEF which had 6

leaking rods. The peak rod burn-up was about 17 GWd/tU. The activity release per leaking rod is high and the rate of increase

is abnormally high considering its modest rating of 368 W/cm (74

W/cm*). Yet t\ is element developed extensive secondary hydriding

and cracking at the bottom end and within 0.5 to 1.0m of the top

end in all 6 leaking rods during 15 days further irradiation. The

PCI defects near the peak rated position were large and numerous

and this rod would have a very tight fuel/clad gap during the 15

days post ramp irradiation. It seems most probable that the many

secondary defects were the result of the clad being in a highly

stressed condition at the time that a H_ flux from the PCI defects

penetrated to the secondary defect regions. This is advanced as an

explanation for the rapid development of secondary hydride induced

defects at this lov; rating. It is interesting to note that a

second element subjected to a planned power ramp experiment, UGA

(Figure 10(a)), gave even higner off-gas activity from 9 defect

rods. It was removed after 2\ hours at 450 W/cm. No defects or

hydriding were found at or close to the top end, unlike the rods in

element PEF, and small hydride defects were only found in 2 rods at

their bottom ends. Some of the non-penetrating cracks had deltoid

shaped hydriding at their mouths, Figure 8, and some of the wider

penetrating cracks are reported to have had more uniform hydride

layers along their 'whole length'. This is a most strange

observation, unless this hydride formed before the crack by the

mechanism referred to above for the axial crack formation in PEF

- 29 - rods. It could therefore mean that secondary hydriding has occurred in 2\ hours in rods with a peak rating of about 450W/cm

(90W/CIT11). The difference between UGA and PEF could be primarily one of rating, the secondary hydriding in UGA, the higher rated occurring more quickly and closer to the primary leak because of the more rapid reaction rate of HO with the clad and UO».

Nevertheless the behaviour of rods in both these elements requires some explanation when compared with, say, UFY and CRT (Figure 6) which were both primary hydride defects.

The ability of rods with very severely deteriorated defects to survive prolonged irradiation without excessive activity release is well illustrated by two rods from Ringhals-1, RI-999-H2 and

R1-1O15-A8. Figure 12 shows a defect 600mm from the bottom end of rod H2, believed to be primary defect and Figure 13 a defect 2100mm from the bottom end, believed to be a secondary. On visual examination before sectioning the former showed up as a small crack and the latter as a hole. The rod A8 had 3 major defects, two axial cracks about 2cm long at 200mm from the bottom, almost certainly the primary PCI defects, a hole from which virtually all the clad material had gone, about 5mm in diameter, at 1500mm and a cracked bulge at 2600mm where the rod broke into two during handling. The defect rod H2 had the most intensive (21) post-irradiation examination at Studsvik. There were numerous other eddy current signals between the two major defects and a few above the upper one. These mostly appeared as small bulges and were clearly hydrided regions. In the vicinity of the major defects there is very considerable damage to the cladding, more so in the upper defect where spalling from the clad outer surface has

- 30 - occurred.

Interpretation of the sequence of events in rod H2 is difficult

because of a lack of phase identification, but both defects at

first sight have the appearance of severly degraded secondary

hydride defects. The surface of the UO under the centre of the

lower defect is fairly regular with no major disintegration, but

there are limited areas of reaction product eating into the surface

and intruding into small cracks. The m ximum depth of penetration

is about 200 microns and there is some volume increase, Figure 14,

suggesting a density in the range 7-8g/cm3. This seems likely to

be a compound containing a volatile fission product such as Cs.

The very much broken up nature of what is left of the inner region

of the cladding (about 20% of the clad thickness has reacted)

suggests quite clearly that this region has been hydrided (the many

partial penetrating redial cracks support this) and that hydride

has oxidised, the H diffusing away to form some further hydride at

the colder outer surface, (See reference 21, Appendix 3» etched

section in defect region). The clad inner oxide thickness at this

axial level, but away from the major defect is about 10 microns,

and there is accelerated H? pick-up with diffusion towards the

outer-surface.

The defect in rod H2 at 2100mm (Figure 15) shows some differences,

although the gross damage has general similarities. The light

coloured reaction phase with the U0_ is missing and the surface is

regular, except for a small area under the major defect where the

U0? is partly broken up and shows a grey phase. The clad is again

severely broken up over about 20$ of the thickness on the bore

- 31 - side, with another 20X spalled away from the outside, giving the

defect the appearance of a hole from the outside. Some of the

broken up debris in the inner clad reaction zone has the appearance

of oxidised Zr. It seems as if the sequence has again been massive

hydride formation followed by migration of H to the outside,

forming massive external hydride which b~.s spalled off. Internally

the re-converted Zr has probably been so defect ridden, because of

the shrinkage, that it has oxidised extensively, or possibly as

suggested for the lower defect the internal hydride has at least in

part oxidised directly. Away from the massive defect the bore

surface oxide layer is about 14 microns thick.

In both these defects in rod H2 there will have been local hot

spots, and in both defect regions, as well as over the whole length

between the defects, there are extensive structural changes in the

UO related to changes in stoichiometry and increased temperatures.

Some of these changes are so unusual as to warrant further

investigation.

It is not possible to be sure which defect in rod H2 was the

primary, neither of the big defects seems likely to have been and

there may be a small primary PCI defect between the two.

6. DEFECT DETECTION

A defect is a penetration of the U0_ containment through which coolant

may enter the rod and/or fission products may be released to the coolant.

This release process may be by diffusion, by wash-out or by

recoil/knock-out if the defect is large enough. All defects do not

- 32 - release detectable fission product activityK "' and the only evidence nay come from eddy-current testing following the finding of an excessive internal H content. The exact size at which leaks become detectable is probably at an area of about 1C~ mm*. Assuming at least near pressure equalisation between coolant and the internal free volume of the rod, then the activity release rate depends on the fission rate in the UCL

(birth rate of fission products), the F.P. release rate by all the known processes, the diffusion rate in the fuel/clad gap and the size of the defect (escape rate coefficient).

The release rate from the UCL, although partially athermal, is mainly controlled by thermally activated processes, although at typical PWR pov?er reactor ratings there is little such release in intact fuel rods.

Release rate is enhanced in defect fuel rods by two mechanisms, a reduction in gap conductance, especially for the larger open steam filled gaps and by changes in the 0/U ratio of the fuel, either increasing the hyper-stoichiometry or producing phase changes to U^CL or U,Og all of which reduce thermal conductivity. The increase in release rate has been cuoted as two orders of magnitude by Schuster and co-workers for BWR ard PWR fuel (calculated 200°C temperature increase); as 10 to 50 times (24) higher by Chenebault and co-workers (calculated temperature rise only

150°C, so H?0 in gap nust also increase release); as one to two orders of (H1) magnitude by Warlop and co-workers for PKR rods. These latter workers make much of the presence of liquid and vapour phase mixtures of

H_0 in the fuel/clad gap and a step increase in gas release when the UCL surface tenperature reaches the saturation temperature of the pressurised water in the gap. This v/as at 170 W/cm in their SILOE experiments.

Fractional iodine release from UCL has been shown to be the same as that

- 33 - of the noble gases, xenon and krypton. However, release through a defect

is more complex because of trappir-g in the gap, probably as compounds with caesium and H?. Release of I is particularly prominent as spikes on

reactor shutdown ir both BWP's and PHP's. The spiking effect ' ° is

more pronounced in PVR's than BWR's because it is a leaching process

which occurs when liquid phase HO forms in the gap. There is a delay in

the spike peak in both reactor types. Some Cs release spikes are also

seen due to a leaching process.

Various measurement? of activity of isotopes, mainly of Xe, Kr and I, and

ratios of different isotopes have been used to detect the presence of

defects, to assess their severity, to attempt to assess the number of

leakers and to attempt to locate them. Location is easy in reactors with

separate fuel channels. In conventional pressure vessel reactors

location is difficult, but some information can be obtained by variation

of activity release when control rods are moved, or from measurements (HU) which indicate the burr-up or rating of the leaking rod. In deciding

or. the isotopes to measure ar.'i where in the circuit to measure them,

there ?re differences between PWR's and BWR's (noble gases largely go

vith separated steam in BWR's) and between the noble gases and

radio-iodines which are subject to trapping and greater delays in release

i.e. a" apparently smaller escape rate coefficient.

Ii all reactors there is some activity detectable, because there is

always some tramp uranium on the fuel or in the circuit. In assessing

activity releases this background level must be taken into account. In

general terms the higher the activity release rate the more severe the

defeot, or the greater the number of defects i.e. number of leaking rods.

However, sone guidance can be obtained from the ratio of short to long half-life isotopes as to whether a defect with a large escape rate coefficient is present, always provided the total release is well above background. There are other methods which are discussed later.

In assessing activity release, clean-up factors, losses and stripping out of noble gases to steam or cover gas must be taken into account.

The isotopes most commonly used as indicators of activity release from defective fuel are 131I (half-life 8.04d), 133I (20.8h), 133Xe (5.25d), 135Xe (9.10h), 138Xe (14.1m), 87Kr(76m), 88Kr (2.84h). Other isotopes are used for special purposes, particularly 134I (45) for following U contamination (easier to measure by y-spectroscopy than 138Xe ) and 239Np, also an indicator of U contamination.

6.1. Measurement Techniques The most common radioactive decay process is when a nucleus increases its proton to neutron ratio by converting a neutron into a proton plus an electron and ejects the electron (a beta particle - this is beta radioactivity) forming an atom of a new element. The proton to neutron ratio can also be increased by the direct emission of a neutron. Alternatively the proton to neuton ratio may be reduced by converting a proton into a neutron by emission of a positively charged particle, a positron. This is another form of beta radioactivity, again resulting in the formation of a new element with an atomic number less by one. The other process of some importance in activity monitoring is the emission of an alpha particle (a He atom) comorising 2 protons and 2 neutrons. This is known as alphc^ radioactivity and only occurs in the heavier elements where the reduction of total mass by 4 mass units gives a

- 35 - more stable element.

The direct neutron emission referred to above is usually subject to 87 some short delay, up to a maximum 56s half-life for Br. This leads to a further possibility for detection of severe defects.

All the above radioactive decay processes are accompanied by a small loss of mass and hence a release of energy. This appears as kinetic energy of the particles and often partly as gamma rays, short wavelength electromagnetic radiations. Alpha, beta and gamma radiation is emitted with one or more characteristic energy levels, usually expressed in MeV. The radioactive atoms also have characteristic disintegration rates (A) which determine their half-lives (t,)according to the relationship: At, = Ln 2 = 0.693 Z Total activity has been expressed in the past as Curies (1 Curie = 3.7 x 10 disintegrations/second). Nowadays, however a smaller unit (SI unit) the Becquerel (Bq) equal to 1 disintegration/second is more widely used.

All radiation detectors are essentially ionisation detectors in which a direct current is induced in a gas or semi-conductor in ?. chamber fitted with an anode and a cathode to create an electric field. Higher-applied voltages produce further ionisation by collisions (multiplication). Pulse type counters, proportional counters and Geiger-Muller counters (increasing voltage on anode beyond proportional region) are variants of the simple ionisation chamber. The types of radiation-detector used in modern power reactors are more likely to be solid state semi-conductor devices

- 36 - based on the creation of electron-hole pairs and the collection of

their positive and negative charges on electrodes by applying a

suitable voltape. Silicon, Ge(Li) and high purity Ge are the

materials mainly used to measure the deposited energy from incident

radiation. With suitable electronics such detectors can produce an

energy spectrum of all the isotopes contributing to the radiation

field at any tine. From these spectra the amounts of the various

isotopes present can be determined and hence release rates can be

calculated.

6.2 Activity Release Measurement Requirements

Information on activity release is needed to prevent excessive

activity release from a reactor site and to protect operating staff

from excessive doses. Important subsidiary requirements are to

prevent a major release of active material to the circuit which

will cause problems with maintenance and fuel changing and also to

provide some information on the location of leaking fuel rods.

Ideally the detection system would provide information on

(a) the number of leaking rods

(b) the severity of the leaks in the leaking rods

(c) the location of the leaking rods.

Some of this information can often be obtained, but there are

usually uncertainties and the value of the data obtained depends on

the measurement systems fitted.

6.3 Activity Measurement Systems

In PWR's all activity measurement, for noble gases, radio-iodines

and other fission products are made on the recirculating primary

- 37 - circuit water. Measurements are made either on samples or on small

diameter by-pass pipes, they are likely to be continuous, using a

counter for total beta and gamma activity, and with periodic

gamma-spectra measurements on samples.

In direct cycle BWR's more options are available as the fission

gases Xe and Kr are largely stripped off in the steam phase while

soluble fission products such as iodine recirculate in the water

phase. Insoluble solids, such as particulate UO and knocked out

atoms also recirculate in the water phase. The measurements made

on the recirculating water have to be interpreted in terms of

position in the circuit (delay time) and clean-up system type and

capacity. The fission gases going with the steam phase are removed

in the condensor off-gas system. Activity can be measured anywhere

between the condensor and the stack depending on the detailed

design, bearing in mind any clean-up system (charcoal beds) delay

syster, and re-combiners.

6.H Activity Monitoring Systems in Use

With the "umber of options available it is not surprising that

ireasurements of different isotopes or isotopic ratios are used by

different operators for diagnostic purposes. The most useful

radio-isotopes are listed in Table 1 together with their

half-lives. Alpha, beta and gamma radiation energies can be

obtained from a chart of the nuclides.

Generally some measure of total activity release to the coolant,

and in the BWR case from the stack, plus some indication of whether

a severe leak (with exposed U0?) is present, is required.

- 38 - Rate of release, rate of increase of release with time, and increments in release which indicate a new leaking rod appearing are all helpful. The presence of some of the very short half-life isotopes is indicative of a high escape rate coefficient, provided the total activity release is greater than that due to tramp uranium.

Meaningful interpretation of activity release measurements can only be obtained in the steady state because of spiking effects of power disturbances such as cycling, ramping, shut-down and start-up.

The techniques used and reported in the literature surveyed are summarised below:

6.4.1. French PWR Practice

Contributions by authors from CEA, EDF and Framatome in the

(2ll I 5 i 6 l>7 48 49 50 51) period 1979.1983 ' * ' * ' ' ' ' ' represent the

most authoritative work on the interpretation of activity

release from defective fuel rods. The basic French

approach was well described by Pelletier, Beslu ?nd

co-workers from CEA ' ' , although the logic employed

in arriving at an estimate of the number of leaking rods

present in a reactor stretches the credibility. The method

is based on observations of activity release in US and

French reactors which were correlated with PIE findings on

number and size of defects. The method is semi-empirical.

Various radio-isotope ratios are measured Xe/ Xe

{5.256 /9.1h), 87Kr//135Xe (76m/9.1h), 138Xe/135Xe

(1i).1ir/9.1h), 138Cs/135Xe (32.2m/9.1h) and compared with

- 39 - ratios determined by the PROFIP code using experimental data from operating reactors (Tihange and Fessenheim). In a parametric study ratios were established for various average clad escape rates and failed fuel rod average power values. Comparison of measured ratio values with the results of this parametric survey (Figure 16) gives values for the average clad escape rate and average fuel rod power. From this information the activity calculated by the PROFIP code is compared with the measured activity to derive a number of leaking rods. The method is at best approximate, but for average type defects it gives reasonably accurate results . The model used to calculate ccolant activity uses mass-balance equations in the UO , in the Ras gap and in the primary circuit. The 235 239 variations in fission product generation from U, Pu and other heavy elements are taken into account. Release from the UO by recoil, knock-out and diffusion is taken into account, also trapping in fuel and on the clad surface. Release from the fuel/clad gap and removal from the primary circuit coolant by purification, losses etc. are the final steps in the calculation, and a correction is then applied for tramp uranium present at start-up with a

'clean' circuit, key equations used are: 1. Mass balance in fuel dNf = B - Np - F,. . B "dt fuel

2. Mass balance in gas gap: dNg = Ff 1 B-(A + ^ dt

3. Mass balance in primary circuit:

dNp =£v»Ng -XNp -/?Np dt 1. The activity A in the primary circuit, in disintegrations/second/unit mass of water is:

A - A " T 5. From 1,2,3 and 4 above:

6. Release fraction from a rod may be written: 0 N2 v F Fr = w-2& + v fuel Where: Nf = Number of atoms of an isotope in the fuel Ng = " " " " " " in the gap

Np = " « w n n « in tne coolant B = Number of atoms of the isotope produced in the fuel (no. of fissions x yield) \ = Disintegration rate of the isotope. F-. .. = instantaneous fractional release from fuel to gap A = gap escape rate constant, per unit time. (Tg = 2 Log 2, where Tg = gap delay time.)

S = Trapping coefficient for iodines. C - Number of leaking fuel rods. a = Clean up rate for the isotope in the primary circuit, per unit time. A = Activity in primary circuit in disintegrations per second per tonne of water. T = Mass of water in metric tons.

The number of leaking rods is obtained from equation 5, knowing B,/8and-^ (from PROFIP), but there is a big uncertainty as to what F value to use. For 17x17 fuel in Fessenheim B is normalised at 178 W/cm and A at 13.6 mVh. In using the method, an apparent number of defected rods C is defined so that

and with t = 1 equation 5 becomes

This activity A corresponding to an apparent defect is divided into the corrected specific activity for each isotope to give the number of defected rods. Table 2 shows some results of the application of this method.

As an indicator of strious fuel damage a short lived, easily measured, isoptope is also measured in French PWR's' . -I has been proposed (half-life 52.5 minutes) and in conjunction with Xe measurement (half-life 5.25 days) the extent and evolution of damage can be followed. Arain the problem is that tramp uranium emits the short lived isotopes, so a simple measurement of a short lived isotope is not conclusive, except that an increasing rate is evidence of deterioration.

An empirical approach for evaluating fuel rod defects, short of the use of the PROFIP code is to assume that freedom from iodine spikes during power transients is an indication that there are no leaking fuel rods present. This can be confirmed by checking that coolant activities remain below the following limits (normal purification - 42 - systems in use): 131I < 10~3/*Ci/g 133I < 10~2/

The iodine isotope ratio is useful in that it has a value of about 1 for very small defects and a value of less than 0.6 for larger leaks, or for tramp uranium only. This latter aspect requires that additional information on activity levels is needed for interpretation.

Delayed neutron detectors are used in loop experiments in France ( He detectors, -^He + n • • ^ ^He) but there is no mention in the literature surveyed of their use in commercial PWR's.

A very large amount of investigation of activity release monitoring has been reported from France, which is all aimed at use in PWR's. Some of the methods are best used when diffusional release of fission products is not involved, say for fuel at ratings below 200W/cm, but the enhanced temperature and release in the presence of a defect probably means effectively a ceiling of about 150W/cm.

6.4.2. FRO Practice Coolant activity monitoring in BWR'S and PWR's in the Federal Republic of Germany has been reviewed by Schuster

- H3 - and co-workers from KWU . In PWR's the activity of fission gases and iodine isotopes is measured in the primary coolant, but in BWR's iodine is measured in re-circulating water and fission gases in the off-gas system prior to the delay line. Measurements of Xe, mKr, Kr, I and I are made and compared using diffusional and non-diffusional models as a function of linear heat generation rate. Plots of RjX/B show that fission gas release in both BWR's and PWR's is diffusion controlled, with higher releases from PWR fuel. The releases for the two iodine isotopes differ widely and are an order of magnitude higher for PWR compared with BWR. When the same data were compared using a release relationship of RA^B the gas releases showed a wide spread, but results for the iodine isotopes were closely grouped, showing that the release was not diffusional, but probably escape from deposits in the free space. Again, PWR fuel released much nore iodine, which is explained by the fact that water enters defective PWR rods and iodine wash-out can occur. In BWR's water is only present as steam inside fuel rods at power.

Fission gas releases from intact and defective rods were also compared using inferred diffusion constants, Figure 17. The release from defective rods is about 2 orders of magnitude greater than from intact rods and the dependence on LHGR is less in defective rods. There is also a suggestion that the effective diffusion constant in defective BWR fuel is slightly lower than in the PWR case. Garzarolli(22) refers to the use of the 133Xe/131I ratio

(5.25/8.04 days). A high value indicates small defects,

typically small PCI cracKS in PWR fuel, for which activity

release measurements are quoted from the Obrigheim reactor.

As the total activity release increases, due to defect

enlargement and the opening up of secondary hydride

defects, this ratio reduces from values close to 200 to

about 50. This is said to reflect the easier release of

the gaseous isotope through a small defect and so this

ratio is an indicator of defect deterioration.

Spiking behaviour, mainly of iodine isotopes, has been (42) described by Neeb who proposes a model which explains (43) the PWR/BWR difference and by Eickelpasch who presents

typical numerical values for the evolution of iodine

release during and following a reactor transient.

6.4.3 U.S. Practice

Locke has summarised U.S. practice in activity

monitoring which is designed to ensure that limits on I

and total activity imposed by NRC are met.

Westinghouse calculate the activity levels of I and I

for a defective fuel fraction of 0.01 using an iodine

escape rate coefficient of 1.3 x 10~ /second. They then

compare these values with the measured activities after

correction for tramp uranium. This gives a figure of a

certain percentage of average power density fuel rods (or a

percentage of total rod power) contributing activity to the coolant ?.t the specified escape rate through cladding defects. Westinghouse prefer a total I activity concept to one of a certain number of fuel defects. They prefer to report to NRC on the basis of a percentage of the specified limit. The disadvantage of this approach is that it gives no indication of the severity of one or more defects. Such information could however be obtained from the 1/ I ratio if so desired.

G.E. use a decay constant exponent to give a measure of the amount of fuel exposed to the coolant. They note that the gaseous activity release is a fraction of the equilibrum content of a defect fuel rod. Data are obtained for 6 noble gas isotopes 133Xe, 135Xe, 138Xe, 85mKr, 7Kr and pp 1*51 15^

Kr, as well as for I to I, neptunium and some other

solids such as barium. GE have derived a predictive model

for 'expected leaker bundle range1 based on observed

off-gas values supported by shut-down sipping results.

Babcock and Wilcox have described a sophisticated method of

analysis to determine the quantity of defect fuel present.

Their monitoring system has been described (based on a

high purity intrinsic Ge gamma ray detector) as well as the (54) calculation method. The approach is similar to that

reported by Beslu , the major problem in predicting

coolant concentration of any isoptope is the release rates

froiD fuel and gap. B&W use only iodine isotopes, saying

that 'they remain the best choice because of their high

yield, relatively rapid diffusion, fairly predictable behaviour and wide range of half-lives.1 To set against this view is the intermittent nature of iodine release resulting from trapping and leaching.

The B & W method is to measure the equilibrum steady-power activity level of 131I, 133I and 135I each month. After correction for background, monthly averages are calculated for131I/133I and 131I/135I ratios and for each isotope's concentration. The ratios are compared with those expected for a large gap release rate. If they are abnormally low they are corrected for a recoil component. This could of course indicate a severe defect, and presumably a judgement is taken based or. not only the ratios, but also the absolute levels of release. If the ratios are abnormally low, then a very small leak is indicated and a value for the gap release rate is determined from the ratios and the activity levels are corrected accordingly. The number of leaking rods is then calculated from the isotopic concentrations. Verification of the model is necessary because of uncertainty in the release rate from the fuel and sipping campaigns have been used to show fair agreement with the iodine model. It will be noted that the number of leaking rods in this case is based on comparison of the levels of the three iodine isotopes compared with predictions for their release from a core average power rod with the plant purification system operating at a known, constant, efficiency. The use of the isotopic ratios is only to refine the accuracy of the activity values for comparison with the model calculations.

- 47 - B * W als? use 23°N? and total alpha activity as indicators of the presence of large cladding defects, but believe they are not reliable indicators of severe defects.

B 4 W also use the 13UCs/137Cs ratio (2.06y/3C.2y), the so-called Q value as an indicator of the burn-up of defective fuel.

6.4.4 Canadian Practice In the CANDU/PHW reactors the concentrations of isotopes "I, Kr, Xe and Xe are measured on continuous sample flows of coolant using a hyperpure Ge gamma detector. Defects can be traced to 1 of 2 (or 4) coolant loops immediately. They are traced to an individual channel by delayed neutron detectors (BF,) using 137I (22s) and Br (56s). Neutrons from N (4.6s) and photoneutrons from N (7.5s) must be discriminated against.

Cuttler and Girouard have described the use of a hyperpure Ge diode detector on-line in Bruce A. The gamma radiation is measured at a pair of sample holders through which coolant from the 2 coolant loops circulates. Xe (5.25c!) builds up to give the largest signal and a plot against time of the 133Xe/88Kr concentration (5.25d/2.84h) gives a good indication of fission product release and fuel Op deterioration (more Kr seen).

Maotonald and Lipsett studied FP release from 3 defected fuel rods (1.2mm drilled hole) in the X-2 loop of

- 48 - NRX. Two rods were pre-irradiated to 6 GWd/tU, the other

was new. The pre-irradiated rods were irradiated for 370

and 850 hours and the new rod for 580 hours all at 480W/cm.

While the purpose was to study FP release, it was noted

that UCL pellets were oxidised in all cases and that some

U0 had disappeared from the pellets opposite the holes in

the cladding. It is perhaps significant that in these

short rods (180mm pellet stack) no large concentrations of

zirconium hydride were found in the cladding. Fractional

releases of individual fission products were plotted

against their decay constants, Figure 18. The decay

constant exponent is always greater than 0.5 (slope greater

than 0.5) showing that fission products are released at a

slower rate than by diffusion. The release of iodines is

slower than that of the fission gases confirming the

trapping of iodine in the gap. Rod RPP which shows more

rapid iodine release had 2 holes drilled inadvertantly and

would appear to give a neat confirmation of the iodine

waihout effect when multiple leaks are present. The fact

that release of fission gas is also slower than for a pure

diffusion process is not unexpected when it is realised

that the released gas originates from all parts of the rod

and has to travel within the gap or within cracks in the

fuel to the leak site. MacDonald and Lipsett also studied

spike effects on shut-down and start-up for iodines and

fission gases, Figure 19.

6.JJ.5 Swedish Practice (32) Hallstadi''s J has reported leakage rates for the 7 radio-isotopes 133Xe, 138Xe, 131I, 133I, 134Cs, 137Cs and 23° Np in 8 Swedish and 2 Finnish reactors before their yearly outages and the corresponding number of leaking rods found at shutdown. The rod average leak rates were found to vary by factors of between about 40 and 600, and

Hallstadius concludes that the number of leaking rods in the core at a given time cannot be determined, even approximately from steady state leak rates. This conclusion is in contradication to views held in France and by some workers in the U.S.A. It indicates that such methods are susceptible to a large error where very different types of defects are responsible for the leaks

(e.g. fretting and PCI) and also where large rod power variations, or reactor core average power differences exist.

The practice described by Hallstadius is to sample the

recirculating primary circuit water and to analyse the gamma-spectrum with a high energy resolution semi-conductor

detector. This system is said to give accurate estimates

of leak rates for non-volatile elements such as Cs and Np,

but not for iodine which distributes itself in both water

and steam because of its volatility in a ratio that depends

on reactor design. This problem is not mentioned by other

workers, the general assumption being that iodine is highly

water soluble and that the big majority remains in the

water.

The fission gases are determined either by sampling in the

- 50 - off-gas immediately after the recombiners, or at the stack»

Samples are analysed by the same semi-conductor detector system as those in solution.

Direct measurement of fission gases on a by-pass flow in the main stack is said to be attractive but is not applicable to Xe isotopes for reactors with charcoal filters.

Hallstadius notes the usefulness of the 133Xe/13 Xe and

" 1/ I ratios as indicators of defect severity and that 138 23Q a high proportion of J Xe (14m) or D Np (2.4d) may indicate core surface contamination (background) or the presence of fissile material circulating in the coolant. 134 137 He notes the possible use of the Cs/ Cs ratio as a Lieasure of the burn-up of defect fuel, but when applied to all the data no correlation was found.

Hallstadius has compared the defect release rates for several radio-isotopes with linear heat generation rate

(LHR), end of life burn-up and fuel density and stability. 133 137 239 He claims good correlations for Xe, Cs and Np, release rate increasing with LHR and reducing with burn-up, but he finds no positive correlation with density or U0_ stability. It would have been interesting to look for a correlation with connected porosity in UCL. The correlations with LHR for full and restricted data sets are shown in Pef. 32, Figures 9 & 10 (133Xe), Figures 16 and 18

(137Cs) and Figures 22 and 24 (239Np). They are far from

- 51 - convincing and other isotopes, such as I, show no correlation at all up to 200W/cm.

(57) Tor Ingemansson described the measurements made on the

Forsmark reactors. Alpha and gamma detectors are used to measure activity in the primary circuit recirculating water with gamma spectra being determined periodically. On the off-gas line 2 Nal (scintillation counters) and 2 ion chambers (to detect very high releases) are fitted downstream of the re-combiners. Gamma spectra are produced at 2 week intervals and there is continuous monitoring of total beta and gamma activity. Ingemansson preferred to use mainly 133Xe (5.25d) and 239Np (2.35d), but the ratio

1/ Xe was used to measure tramp U and an increase in the ratio indicated a defect. It was intended to try He detection as a method of confirming the onset of a new leaker since the change in the off-gas release was too gradual. Experience in the U.K. has shown that changes in off-gas activity are good indicators of new leakers.

Ingemansson also said that defects could be located, at least to one super-cell, by control rod movement (flux tilting) except when they were very severe. He noted that early leakers (presumably fretting defects) gave high releases of Xe and that secondary (hydride) defects started to leak about 6-8 weeks later. Such secondary defects occurred even in very low power rods.

Blaokadder and co-workers have reported typical coolant activity changes in the presence of leaking rods after

shutdown and accompanying power ramps, Figure 20. This

shows a secondary iodine spike, about 6 hours after

shut-down.

6.4.6 U.K. Practice

With currently only one LWR in operation (SGHWR - a

pressure tube BWR) experience in the U.K. is limited, but

has provided useful information on defect deterioration

effects. The reactor is equipped with a bulk off-gas

gamma-counter, downstream of the main condensor. A 1J

sample of the off-gases is passed through the counter

within 2 minutes and the count is converted to a CiMeV

(gamma plus beta) per day figure. Radio-iodines in the

re-circulating water are measured by a high resolution

gamma-ppectrometer, the data being expressed as release

rates iny^Ci/h. A beta proportional counter is used to

locate the leaking channel on a delayed dry steam sample at

100 psi, taken directly from each channel exit.

Moderate defects were found by Locke to show a doubling

time for off-gas activity of about 1*10 days with little

effect of rating in the range 250-450 W/cm, but severe

defects (e.g. severe PCI failures, leaking centre

thermocouple?) showed high initial release rates and

doubling" times of 15 to 20 days. Rods at very high ratings

(>600 W/cm) were certainly found to deteriorate more

rapidly. The decay constant exponents found varied from

0.57 to 0.84 for gaseous nuclides and 0.1 to 0.9 for

- 53 - radio-iodines. The release of gases is therefore slower than for diffusion alone, as found by other workers, while the large variability for iodine suggests a combination of trapping and spike release due to coolant or power perturbations.

Locke also reports on the value of the 1/ I ratio (more usually the inverse is used) in diagnosing the presence of serious defects. Values of the ratio in excess

of U when the I release rate exceeds 400yHCi/h are taken to imply that defects are severe and should be discharged as soon as possible.

6.^.7. Miscellaneous Practices In the CANDU reactor at Rajasthan, India, the state of the fuel is assessed by a once per shift determination of

Normal background is about 30-40/«.Ci/l and the level can rise as high as 4000/<.Ci/l with defects present. Leaking fuel channels are detected using delayed neutron counters which show a normal background of 200 cps with values up to 800 cps when defects are present.

In the Tarapur BWR (59) complex methods of analysis have been followed. Six gaseous isotopes Xe, Xe, Xe, Kr, Kr and Kr are measured as well as the five iodine isotopes I to I. For the gases the total release for each isotope at full power is equated to a release component for recoil, diffusion and equilibrium release. From the 6 equations with 3 unknowns the recoil, diffusion and equilibrium release components for each isotope can be derived either mathematically or by an iterative graphical (59) method which is described by Nandwani . The values obtained are extrapolated to rated power and the values are used for following trends with time. If this method can be shown to be soundly based and reliable then it could be of considerable value in following deterioration.

A second approach used at Tarapur is to evaluate the so-called 'B' factor, the slope of a plot of A/YÄ and log Å for each isotope (A-activity; Y-yield; A- disintegration rate) and also the 133Xe/135Xe ratio (5.25d/9.1h). A 'B1 value close to zero is indicative of recoil, close to 0.5 of diffusion and close to 1 of equilibrium. A drop in the

•B' value or a peak in the Xe/ Xe ratio suggests a loss of clad integrity. (A peak may seem anomalous, but this is a change from a predominantly tramp U situation to one with a small release of predominantly Xe).

An unusual approach to control of iodine spiking magnitude is the use of FG/I. This is the ratio of the fission gas source term (A/YA) for fission gases to that for I. The magnitude of the maximum possible spike after shutdown is said to be proportional to the FG/I ratio and a guide line of FG/I and I values is used to prevent I concentration ever exceeding 5/*Ci/l.

A sophisticated technique has been developed by

Kalfsbeek' for use in the Dodewaard BWR. The 133Xe

- 55 - release rate from a known single defect at a known rating

is used to calculate release rates at other ratings. A

number of leaking rods can then be derived from the Xe

release rate at any time using a core average rating value.

Comparison with sipping test results is good.

6.5 Activity Release Levels

The maximum activity a defect fuel rod can release is the total

fission products generated. The maximum rate of release is the

rate at which the fission products leave the UO to a free surface

connected to the fuel/clad gap, be it by sweeping, diffusion,

recoil or knock out. In the knock-out process other atoms,

including U, U, Pu and the whole range of fission products

are ejected into a space connected to the coolant. Most of these

atoms are non-volatile, non-gaseous, non-coolant soluble and will

only appear in the coolant if there is direct line-of-sight from

their point of emission into the coolant flow. This condition

would be met for a small diameter drilled hole, but not for a minor

PCI crack or hvdride defect.

Rods do not release all the fission products generated for other

reasons. Some do not undergo any systematic diffusion to free

surfaces where they are released. Mo and Nb are examples of such

fission products which are not 'released'. There is, however,

evidence of substantial mobility in that some of these high melting

point metallic fission products are found as small metallic

inclusions both inter- and intra-granular and in low temperature

regions of U0-. They have often been reported as acting as nuclei

for grain boundary bubble formation, implying that they had

- 56 - precipitated before sufficient fission gas had reached the boundaries to form bubbles.

Apart from the non-mobile fission products, all the gaseous and volatile fission products are not released from the U0?, because, apart from the recoil and knock-out which may account for 0.5? release, the controlling process is thermally activitated diffusion. Hence, despite the increased temperature in a defect rod (caused by reduced conductivity of the hyperstoichiometric UO and reduced gap conductance) it is highly unlikely that release of gaseous and volatile fission products will ever exceed about 50? in a defect LWR fuel rod. This would correspond to a volume average fuel temperature of about 16OO°C, an impossible figure.

In addition, all the released fission products do not escape to the coolant, depending on the shape, size and location of the cladding defect, the size of the gap, the gap atmosphere/environment, the

UCL surface temperature and the presence of chemically reactive fission products. The maximum release that can occur is that from

UO in a peak rated rod with a maximum size steam filled gap and (25 32) no gap impedance or trapping. Experience ' suggests a factor of up to 200 between release from the smallest detectable leak and the maximum credible release from a leaking rod.

While there is an enormous body of information on fractional fission product release, R/B ratio (release/birth), clad escape rate coefficient and trapping coefficients, there is limited information on fission product release rate from a single leaking rod. Data is often expressed in coolant activity concentration

- 57 - which can be converted knowing circuit flow rate, water inventory, clean-up capacity etc. It would be helpful for interpretation if more data was expressed directly as release rate to the coolant.

Some of the few release rates given in the published literature are by Locke and Hallstadius from SGHW and the Swedish BWR's. In the

former case, because of single channel monitoring, and subsequent

PIS, the fact that releases come from a single rod is fairly

certain. In the case of the Swedish BWR's there is a small uncertainty. Comparing the two sets of data for I and I

release (the Swedish BWR figures are an average of the release

rates before the yearly outages for each reactor for each year) the agreement is reasonable: 131-,-

SGHW

Moderate defect 2-90 mCi/h 4-190 mCi/h

Severe defect 100-150 mCi/h 900-2000 mCi/h

Swedish BWR's

All defects 20.2 mCi/h 90.2 mCi/h

The SGHW figures cover release rates found for about 30 small to

moderate defects and about 6 severe defects.

Lipsett and MacDonald give some data on release rates

(atoms/second) for an inventory of Q atoms in a rod and an escape

rate coefficient £(per second);

R =

They quote the R values for a series of experiments on defect

elements in the X2 and X6 loops of NRX. The data quoted is only

- 58 - for rare gases, not iodines, but comparing values of (gas escape rate x disintegration rate) for the Lipsett and MacDonald work with Swedish BWR releases for 133Xe (t£ = 5.25d) and 138Xe (tj = 14m) shows the following results:

Swedish BWR's 133Xe Average 4.76 MBq/s Range 0.1 - 17.3 MBq/s 138Xe Average 57MBq/s Range 3.8-150 MBq/s (F.2 1984 omitted) CANDU Experiments 133Xe Average 2.5 MBq/s Range 0.05-24.6 M3q/s 138Xe Average 4.8MBq/s Range 0.22-10.4MBq/s (1 MBq = 2.7 x 10~5Ci) The range in the CANDU experiments is certainly in part related to rating. A rod at 32OW/cm was consistently the lowest and rods over 600W/Cm consistently the highest. The CANDU experimental rods were short (17-23cm long U0 stacks) which would be expected to reduce release relative to the Swedish BWR rods, but the rating effect will have affected those releases also. Since the Swedish BWR ratings were very low compared to the CANDU rods it is surprising that the range of Xe releases is so close. The lower 3 Xs release in the CANDU rods is even more surprising. It may ref lee i. the small effective fuel/clad gaps at power.

The question of the effect of rating on activity release is of considerable interest. It has been referred to in several places in this report and is both complex and to an extent contentious.

- 59 - The literature surveyed in preparing the report does not give a lot of detailed information on the relationship between leak size, rating and activity release. Because of the complexity of the subject it is addressed in more detail in Appendix B.

It has been noted earlier that there is a limit to the amount of activity that can be released by a leaking rod, that there is an interest in knowing how many (and which) rods (or elements) are the leakers and that leak rate, up to the maximum possible, is a function of linear heat rating, defect size and location, some aspects of fuel rod design (length, gap and UO density and stoichiometry for example) and mode of operation. The mode of operation can account for some very large release spikes which may be more pronounced in PWR's than BWR's.

It has also been noted earlier that there are certain tell-tale indicator? of defect severity, regardless of total activity release. These are basically:

(a) the presence of radio-isotopes with very short half-lives

e.g. 13 Xe (14.1 minutes) 87 (b) the presence of delayed neutrons e.g. from Br (56 seconds) (c) the presence of Q€-radiation from heavy metal atoms 239 (d) the presence of Np, a^ emitter (2.39 days). Formed by 238 239 neutron capture in U —•> U and /? emission.

All these features are present in a new, clean, core because of a small anount of tramp uranium and uranium in the Zircaloy cladding.

It is necessary therefore to consider their presence in relation to total release and, particularly for the radio-iodines and some of the gases it is common to look for variations in the ratio of two

- 60 - isotopes of differing half lives.

The smallest leak that can be positively detected will depend to an

extent on the reactor design and detection systems. Since in all

reactors background activity can be measured then there is no

question of the sensitivity of measurement being limiting. The

problem becomes one of knowing when an activity increase is

significant against a varying background. It is helpful to

operators to be able to determine the effect on activity

measurements of defects with varying escape rate coefficients. The

smallest significant leak will have an escape rate coefficient of -9 -8 between 10 and 10 /second. In the UK SGHW reactor a release •pi rate of 3/*Ci/s for " I would be regarded as a clear indication of

a small leak.

(59) For all gaseous radio-isotopes in a BWR, Nandwani has quoted a

release rate (background) of 0.3 mCi/s. In SGHW a minor defect has

a fission gas release of about 15 mCi/s.

(?f\} As for the actual physical size of defects, Locke has quoted a

size equivalent to a 10 mm diameter hole as representing about the

practical limit of detection. This compares with a size of 10 mm

diameter as the minimum size of hole that will be detected in

manufacture by the conventional He mass spectrometry leak testing.

7. DISCHARGE CRITERIA

All Regulatory Authorities have some interest in the activity levels

present in licenced reactors, as well as in levels of activity released

- 61 - to atmosphere. Criteria are therefore laid dew. which limit certain activity levels. Such erit?ria require action to be taken to prevent the limits being exceeded. Reductions in power may allow continued operation until a convenient time for shutdown and discharge of leaking elements, some authorities also require an estimate of the number of leaking rods and subsequent confirmation following sipping tests.

7.1 U.S. Practice

Operational standards for U.S. reactors are laid down by NRC. The

limits specified are for dose equivalent I and total beta and

gamnia activity in the primary coolant. For PWR's these activities

can all be measured in the re-circulating water, but in BWR's most

of the fission gases, and perhaps a little of the iodines, is

seoarated into the steam phase and is measured at the SJAE (steam

jet air ejector). Measurements at the stack, after clean up in

charcoal beds and any delay imposed to allow short half-life

fission products to decay, are a separate central feature. The

limits imposed by NRC for the primary circuit are:

Steady state: PWP £ 1 *

BWR £0.2/*Ci/g, 131I dose equivalent

Short Term Spike: PWR < 60/*Ci/g, 131I dose equivalent

BWR Si»/iCi/g 131I dose equivalent

(1/tCi = 0.037 MBq)

The steady state limits must not be exceeded over a 48h period

Total^ + fc": PWR and BWR <. 100/E/«.Ci/g

C? = average of f> and fl'energies in MeV)

U.S. practice is important since many small countries tend to

- 62 - follov? NRC recornmendatior.s ar.d criteria. Recent U.S. reports on

methods for estimating the number of failed fuel rods and defect

sizes show that this has become a matter of increased importance.

They also indicate that there is some likelihood of success, even

if no better than a factor of 2, in estimating the number of

leaking rods. Tne work includes methods for assessing the burn-up

of defect rods. It is centred on INPO and ANS Working Group 5.3

with EPRI as contractor for INPO. The method used for BWR's is to

calculate monthly a "Fuel Reliability Indicator" based on the

steady state diffusion contribution to the sum of the six fission

gas activity rates (JiCi/s) for each month. ( Kr, Kr, Kr,

133Xe, 135Xe, 138Xe). A simplified version based on activity

release rates of Xe, Xe and Xe is used in some

circumstances. For PWF fuel the reliability indicator is based

only on the steady state activity of I in the primary circuit 131 (/*Ci/g) corrected for I from tramp uranium.

7.2 French Practice

French criteria and limits on permissible activity in their PWR's

have been described by Bournay and Roels . As in the U.S. 131 control is on both dose equivalent I and total noble gas fi+'rf.

The limits applied at this time (1981) were:

131I (dose equivalent) < 1 Ci/t

Noble gases < 300 Ci/t

(both at equilibrum)

However, for continuous operation, the Safety Authority must be

informed and certain actions must be taken when the equilibrum

concentrations reach the lower levels of:

I (dose equivalent) 0.12 Ci/t

- 63 - Noble gases 10 Ci/t

These are, in addition, in France certain limitations (said to be

temporary) on primary circuit coolant activity after a refuelling

shutdown, these are:

131I (dose equivalent) 0.03 Ci/t

Noble gases 2.5 Ci/t

(equlibrura values in continuous operation)

These regulations also involve informing the Safety Authorities

when certain higher levels are reached and shutting down within a

stipulated time thereafter. For example, if the dose equivalent 131 I exceeds 0.5Ci/t then the reactor must be shut down within 48 hours.

7.3 U.K. Practice

Locke has explained the discharge criteria used on the SGHW

reactor in the U.K. which differ from practice in the U.S.A. and

France. The limits applied are:

131I > 500 mCi/h release rate

133I/131I ratio >4 when 131I>400 mCi/h

Bulk off-gas 5,000 CiMeV (ytf+jr) per day in continuous operation

10,000 CiMeV (fi +?f) per day peak.

These limits are assumed to apply to one leaking rod and if none

are exceeded then the rod is in a safe condition. Monitoring for 239

Np in the recirculating water is also practiced in SGHW and has

given variable results. A level of 10/tCi/litre has generally

indicated a severe defect and a level of 50/tCi/litre has always

been a reliable indicator. 7 .'4 Japanese Practice In Japan MITI requires a reactor to be shut-down and for all fuel bundles to be sipped if the I content in the primary circuit coolant exceeds 10 ACi/cm3 and/or if the spike release or shut-down exceeds 0.1/tCi. These are stringent limits, but all BWR's have kept below them in recent years and virtually no sipping has had to be performed. (Only 2 defect BWR assemblies in 5 years). More PWR defects are experienced from baffle jetting and KEPCO are reported (ft?) to have some technique for failure location.

7.5 Practice Elsewhere Cuttler and Girouard report that for the Pickering CANDU reactors there is a warning level of 400Ci I in the coolant at which action to reduce the contamination must be taken and a level of 1000Ci requiring immediate shutdown. These are quoted as concentrations, presumably total amounts in the steady state in the 2 coolant loops.

Nandwani (59) quotes a total off-gas activity of 150m Ci/s at the SJAE as a control level at which preparation for shut down must be started and power reduced to keep release within this level. In fact the site licence would permit 6 times this activity level.

8. DESIGN ASPECTS

Although the occurrence of secondary defects would appear to be an inevitable consequence following the development of a leak in a fuel rod, there are such wide differences in reported behaviour that the question of design dependance must be addressed.

- 65 - From this literature survey there is no evidence for a difference in susceptibility between BWR and PWR fuel rods, but it is not possible to quantify this view with the data available. There are differences, notably higher clad bore surface temperature in PWR and higher UO centre temperature in BWR which both affect 0 absorption rate and which might be expected to influence defect behaviour.

No strong dependence of activity release on linear heat rating

(proportional to UC centre temperature) has been observed in the range (26) 250-450 W/cm although a drop in rating will inevitably reduce activity release and vice-versa. (See Appendix B) This means that within this rating range other factors are more dominant than rating alone.

Such fac; >ra are:

(a) size and position of defect

(b) characteristics of UO such as density, grain size, stoichiometry

(c) fuel rod length and fuel/clad gap

(d) burn-up

It is a common observation that artificial defects rarely lead t> the formation of secondary hydride defects. There could be a number of reasons for this, all of which have been advanced as the explanation;

(a) a good protective oxide film is formed on the clad bore surface

early in the irradiation

(b) there is a larger fuel/clad gap because there is no creep down.

(c) the defects are usually large allowing easy access to coolant to

replace absorption losses and maintain a sufficient CL partial

pressure to preserve bore surface oxide film integrity.

(d) rods used in these experiments are often short and it may not be

possible to get sufficient 0 depletion within the rod length.

There are arguments against (a) and (d) on the grounds that autoclaved

- 66 - (17) clad bore surface have not proved helpful and hydride defects were plentiful in early CANDU fuel rods . This leaves little scope to adopt features of artificial defect rods, but the possibility that an easier axial leak path for coolant could be beneficial is worthy of some consideration. This could provide more effective replenishment of steam/water where otherwise the 0 partial pressure would be so low as to prevent ZrO surface layer protection.

A fuel designer does not have freedom to increase fuel/clad gap beyond a certain limit without other penalties becoming unacceptable. Fuel temperature is increased and hence fission gas release, fuel bundle pressure drop is also increased or fuel weight reduced, and fuel/moderator ratio is changed. There are, however, ways of increasing the axial flow capability while minimising these disadvantages as with

the rifled bore cladding that has been described by Mogard .

While it is known that ZrO films are rendered non-protective in the

presence of some minimum H partial pressure, which is strongly

temperature dependant, there could be other barrier materials which are

protective. Alternatively there may be trace additions to Zircaloy which

would make the oxide films more protective.

The majority of the 0 from steam or water that penetrates through a

defect is absorbed by chemical reaction with the UO because, despite the

build-up of a H partial pressure, radiolysis of the H_0, particularly to

HO, enables continued oxidation even up to U_0Q to take place. This

oxidation rate is faster at higher temperatures so it is advantageous to

have a minimum of connected porosity in the U0?, although some HO will

get to the high temperature regions of the pellet through cracks.

- 67 - The role of fission gas plena is uncertain, but there are two possible beneficial effects. Secondary hydride defects have often been found at the bottom end of rods with only an upper gas plenum, but apparently never at the top end. This may be because the plenum represents a reservoir of 0_, or because local high axial stresses are avoided. A division of the gas plenum between top and bottom of the rod, as already practiced by some designers could therefore be of benefit.

It emerged from work by Jocn that a crucial parameter in determining whether primary hydride defects formed was internal rod volume which was effectively a measure of maximum H partial pressure. This same principle could apply to some extent to secondary hydriding. Assuming a primary leak is large enough to permit pressure equalisation then the volume of steam/water in a rod determines the amount of 0 that must be absorbed by chemical reaction before the critical condition for secondary hydriding can be reached. It follows therefore that unless the conditions for secondary hydriding are very local, then a larger internal volume could well be beneficial.

The specification of UCL pellets is an important aspect of design and is mainly determined by the requirements of rating, burn-up, PCI resistance, fission gas release, swelling etc. From a secondary hydriding point of view perhaps the most important feature is in minimising access of steam to the hot centre of the pellet. A minimum of interconnected porosity is desirable in helping to achieve this, say less than 1$, and it is tempting to favour flat-ended pellets with a relatively low 1/d ratio to minimise access to the central regions at pellet ends. This pellet design would have the further oenefits that it avoids cold spots on the clad inner surface (such regions with dished pellets have been found to

- 68 - act as H? sinks) and large local fission product releases during power increases because of lack of restraint at the dish surfaces.

9. OPERATIONAL ASPECTS

The influence of power fluctuations, start-ups and shut-downs on activity release have been described earlier. In general release increases, especially of iodine isotopes and especially in PWR'S. The effect is caused because U0- expansion and contraction causes a pumping effect and this helps to release trapped iodine, especially in PWR where there is some liquid phase HO in the gap. The spikes on shutdown are again partly because the rod de-pressurises, sweeping out a large part of the

gap inventory, and partly because the U0? contraction makes release from remote parts of the gap easier. A secondary iodine (and caesium) spike is often seen after shut-down, starting when water condensation occurs

inside the defective rod. Such spikes occur later in BWR than in PWR reactors and may last for several hours (6 '4 2 '43 ).

The pumping effect of power cycles can be reproduced by pressure pulsing the primary circuit, an effect demonstrated by Blackadder and co-workers

This pumping action brought about by power or pressure cycling and in its most extreme form by shutdown and start-up will have a marked effect on the composition of the HO, 0-, H and radiolytic radical mix in the fuel/clad gap. For situations where secondary hydride defects develop over a long time as a critical H or 0_ level develops somewhere in the gap, this type of disturbance may well prevent such defects forming.

- 69 - Experience shows that no operational measure can eliminate the deterioration of defects or the formation of seconday hydrides. However, the evolution of a primary defect can be followed by activity monitoring so that a planning decision on whether and when to remove it can be taken in good time. There is no proven way of anticipating the sudden onset of a secondary leak, but there are sensible precautions, particularly avoiding any power increase which could cause a rupture centred on a hydride patch. In fact, in the absence of any operational manoeuvre which could stress such a patch, secondary hydride leaks are likely to start as very small leaks and to grow slowly. However, there is a real possibility that once a second leak site is established then an internal coolant flow may result in a greater increase in iodine release than for the noble gases. Such an effect was commented on by MacDonald and (28) Lipsett and can be inferred from data presented by Garzarolli, Manzel (22) 133 131 and Stehle . In fact a reduction in the ratio of J3Ie/ J I as

Drooosed by these workers may be a good sign that a second defect has formed and an internal coolant flow (steam in the BWR case) has been established. This may not be as effective in BWR's as in PWR's but is worth loovine- for.

Apart from some such technique a pattern recognition process may be the best guide. The Locke curve (Figure 9) may be used, but despite fairly widespread agreement it is not always reliable. At least it does not predict some cases of very severe deterioration, possibly not caused by secondary hydrides, which it does not purport to do. This is a weakness in the concept of this curve, since people are inclined to regard it as a guide to a severe leak developing.

Some defects have led to very rapid deterioration, although at quite

- 70 - nodest ratings, others have survived beyond expectations at exceptionally high ratings. Some examples from SGHW experience are shown in Table 3« Element B which gave a very high off-gas activity and was discharged within a few hours had 8 leaking rods, but even so the release per leaking rod was well above the discharge criterion of 5000 CiMeV/day. In several of the rods, however, there were a large number of wide-mouthed cracks (Figure 5) so that the stored fission product gas inventory would be released rapidly. Perhaps, therefore the period of a few hours following a leak appearing should be treated as a spike for purposes of decision on discharge. No secondary hydride patches (incipient sunbursts) were seen in these rods, but a large number of incipient cracks were hydrided in a few hours, Figure 8. The rate of increase in activity release from element C is shown in Figure 10. Even allowing for the 6 leaking rods, and that the release per rod is below the discharge criterion (not known while in-reactor), the rate of increase indicates a defect at a late stage of deterioration.

Element D was discharged after 26 days in reactor for reasons other than activity release, no release above background having been detected. On discharge several U0 pellets were found in the bottom end fitting, the end cap having become detached due to severe local hydriding. The primary defect wa3 a fretting penetration in the plenum region caused by a broken plenum spring. Hydride patches, some severe were found throughout the lower half of the rod. These had formed in this element at the modest rating of 325W/cm (65W/cm*)in a period of less than 26 days. Element E, a very high rated element was almost certainly a primary hydride defect which was close to the peak rated position. The rod broke on handling about 30cm from the bottom end at a secondary hydride and no other significant defects were found. The rod that failed

- 71 - had a peak rating of about 600W/cm (element peak 74OW/cm). Finally,

Element F, had an instrumented rod which leaked early in life. It had

the highest single rod activity release of any of these elements and a

rapid deterioration rate.

Because of these variations in behaviour, the use of an empirical guide

such as the Locke curve as a pattern recognition aid cannot be relied on

without qualification.

It seems then that the best operational preventive measure is to ensure

steady power operation with a minimum of local power increases, and to

watch the off-gas activity (gas or iodines in a PWR) to establish a rate 133 131 of increase and particularly the iodine release rate and 1/ I ratio

for the appearance of new leaks and onset of a severe defect,

respectively.

10. REMEDIES

The best possible remedy for secondary hydride defects is, of course, to

avoid the primary defects which cause them. Accepting the inevitability

of some fuel failures however, then there are some ways in which the

onset of secondary defects may be avoided or delayed. If they occur,

then the initial activity release will not necessarily be much greater

than from the primary leak, but the fact that the secondary is at a

severely hydrided region makes it susceptible to major enlargement from

events such as power changes or natural development a3 described in

Section 5.3«

10.1 Reduced Rating

- 72 - It has been said elsewhere in this report that activity release from defects in fuel rod cladding is not very sensitive to rating. It has also been said that some low rated rods with large defects form secondary hydride defect rapidly. Despite these statements, which are supported by good and reliable evidence, there can be no doubt that in general a reduction in rating will reduce the likelihood of secondary hydride defects forming, or at least forming quickly. The Locke curve supports this view, based on a lot of reactor data and supported by many other authorities.

Reduced rating is perhaps more a design choice than a remedy, but the lower LHGP needed in smaller rod diameter assemblies must be beneficial in resisting; secondary hydride defects better than assemblies with fewer, larger diameter, rods.

10.? Gap Atmosphere Control Zircaloy cladding is immune to severe localised hydriding in boiling or pressurised water environments, and in steam, but in the fuel/clad pap a critical H /0 ratio can be exceeded which leads to

local breakdown of the ZrO? film on the clad bore and to rapid localised hydriding. This process depends on U0- temperature,

since the reaction of oxygen with the UO is the main cause of O?

depletion, on clad bore temperature because this influences ^ breakdown and on the internal free volume of the rod because for

any given H? volume this influences the partial pressure. In situations where there is a large enough gas gap to allow free flow of gas and vapour, then interdiffusion will help to avoid the

critical H-/0o ratio beine exceeded. The situation will also be helped by maximising the internal free volume and by splitting the

- 73 - gas plenum between the top and bottom ends, so providing an oxygen reservoir at both ends and avoiding an accumulation of H_ at a dead end. This dead-end, or crevice, effect has often been evident at the bottom end of fuel rods in the form of hydriding at the bottom end-cap or weld. Whether a bottom end plenum would have avoided it is not established, but it has rarely been found at the upper end of fuel rods with the more conventional top end gas plenum.

In rods with small fuel/clad gaps, either by design or as a result

of clad creepdown or UO swelling, then the critical H2/0 ratio may be exceeded at some distance away from the primary leak because diffusion in the snail gap is too slow to maintain an equilibrum concentration. This is the classic situation which leads to secondary hydridin?. If the effective leak path between UCL and cladding can be increased then there is a better chance of maintaining an equilibrun composition. This will delay the onset of the critical H_/CL ratio, or avoid it altogether if the primary leak size is large enough. Ways of achieving this without the disadvantages of an increased fuel/clad gap are either by a device such as rifled bore cladding or axially serrated U0_ pellets proposed by Pickman . This latter device would help additionally by increasing the number of short radial cracks, giving a further increase in the path for axial gas flow and, incidentally, reducing clad stress concentration in a power ramp.

10.3 Hydrogen Getters The reason that secondary hydride defects form is that in the right circumstances the cladding is acting as a getter. If a more efficient getter could be used, then perhaps the problem would be solved. G.E. have used getters to prevent primary hydride defects Min for many years . A getter known as HIPALLOY is located in the gas plenum and has been proved to be effective. A fuel rod with such a getter that had a minor PCI defect has been described by

Davies (7). The getter was heavily hydrided, but the leak appeared to have sealed up. No abnormal clad hydriding was found although a

10 micron thick layer of ZrO_ on the bore was further evidence of the leak. Clearly, for a large leak the capacity of a getter to absorb K? would soon be exhausted, but for a range of leaks a getter may be a worthwhile addition to delay or prevent secondary hydridin£. The location cf the getter in the gas plenum is not ideal, it would be better distributed, certainly between top and bottom plena, preferably as a coating on pellets or cladding.

In this connection the effect of the use of zirconium liner clad on

."econdary hydridng is of interest. The Zr surface layer will have lower oxidation resistance than Zircaloy 2 or 1, increasing the oxygen depletion rate and resulting in a thicker ZrO? layer on the bore surface. While there will be greater hydrogen pick-up because of the more rapid oxidation, the nature of the ZrO» film with respect to breakdown and hydrogen permeability is not known.

Moreover, if the junction between liner and clad alloy is for any reason not readily permeable to diffusing hydrogen, or contains imperfections within which molecular hydrogen can precipitate, then the possibility exists of large pressure build up and cracking.

The possibility of a delayed hydrogen cracking phenomenon has been referred to earlier. In this situation the stress build up required could result from the ZrO? layer growth and the additional possibility of exceeding the critical crack length with subsequent

- 75 - fast fracture cannot be ruled cut.

10.1} Barriers Barriers of various materials have been used in fuel rods as a remedy for PCI failures by preventing fission product attack or acting as low friction interfaces and minimising stress concentrations. The principle of a barrier to prevent localised H_

pick-up is attractive. It can either be a physical barrier to H? diffusion or one which reacts chemically.

All metals are permeable to atomic hydrogen to some extent following surface absorption, but in some, such as Cu and Al, diffusion rates are very slow. Some barrier materials were tested by GE ' in their search for a PCI remedy, including amongst the metals Hi, Cr and Cu. Only Cu was effective, but in the event the pure Zr liner clad now used in all GE fuel was preferred. It is euite possible that by slowing down the rate of H_ pick-up by the inner surface of the cladding, a thin layer of Cu (5 microns was used in the O.E. experiments) would be effective in preventing secondary hydride defects. There are of course penalties such as neutron absorption, possible LOCA problems and reprocessing difficulties as well as the practical problem that a complete inner surface coverage by Cu, including the weld region, is not possible, so attack could be concentrated in such regions. It has also recently been found that the effectiveness of Cu as a PCI remedy reduces with burn-up.

Non-metallic barriers, such as the graphite or siloxane coatings tested in Canada by AECL as PCI remedies (graphite coating

- 76 - designated CANLUB now used) are permeable to atomic and molecular hydrogen. Their effectiveness is probably mainly based on the lubricant effect, but some chemical effect of graphite has been claimed. At typical gap temperatures a graphite/steam reaction will take place as well as possible interactions with volatile fission products such as Cs and I. It seems unlikely that any such

reaction will remove H? from the system because of radiolytic decomposition, or will significantly affect the rate of H^ pick-up. / pg \ MacDonald and Lipsett^ irradiated 2 artificial defect rods (1.2mn.' diameter drilled holes) with 7.5 micron graphite layers on the clad bore in pressurised water for 370 and 580 hours and in both cases ths graphite layers disappeared during the irradiation. Siloxane layers may be more stable, but their high permeability means that they are unlikely to have a i-ate controlling function such as some of the metals.

10.5 U0c Pellet Design The possible benefits of using short, flat ended U0, pellets with low open porosity has been referred to earlier in this report. There are conceptual arguments for and against such a design and it must be said that no evidence has been found in the literature. In designing for flat-ended pellets the different dimensional behaviour has to be taken into account, especially in so far as it could affect the stressing of incipient or even well-formed hydride defects. Although there are some arguments in favour of particular features of the UCL pellet, these are so important in relation to the behaviour of normal intact fuel rods that it would be premature to seriously suggest a change from present designs based solely on s possible improvement in secondary hydriding response.

- 77 - 11. PROPOSALS FOR FUTURE WORK

There are some uncertainties associated with the development of secondary hydride defects and with the effectiveness of the proposed remedies.

Areas of uncertainty in particular relate to:

(a) the oxidation kinetics of UO at various ratings and the influence of burn-up, fuel/clad gap, distance from a primary leak and leak size. (b) the variation in gas composition in the fi.-.- clad gap as a function of distance from a primary leak, linear heat rate, leak size and fuel clad gap size and type, i.e. uniform or variable because of use of rifled cladding or serrated pellets. (c) the oxidation and hydriding behaviour of Zr liner cladding.

Item (a) could best be addressed by detailed ceramography, phase analysis

and stoiehiometry determination on the U0? in well documented defect fuel rods. Such an investigation should be made at intervals of perhaps 10 cm axial distance from the leak over the first 50cm and less frequently to the rod end. It would also be of considerable interest to find whether these were differences close to ends with and without gas plena.

Additions to this investigation would be (i) to measure ZrO2 thickness on the bore surface of the cladding at corresponding positions and to characterise its structure as fully as possible, e.g. density of anion

vacancies: (ii) to determine clad H? content and distribution, also at corresponding positions.

The data arising from such a study, especially from a comparison between

-78 - 2 or more leaks of different sizes, would go a long way towards the setting up of a quantitative model of the processes leading to the establishment of critical conditions for secondary hydriding.

Item (b) is in many ways complementary to (a), but in addition it would establish under what conditions of leak size, rating and fuel/clad gap secondary hydriding could be prevented because of adequate inter-diffusion of steam, H_, fission products, radiolytic radicals and He in the gap. The best prospect for such a study would be to perform a series of experiments in two steps. In the first step, unirradiated fuel rods (or rods pre-irradiated for 1 cycle in R2 with no defects) with a drilled hole in the plenum would be irradiated at 40 to 45 KW/m for a single 400h cycle in an in-pile loop in the R2 reactor. Design variations would be gap size (large and small, say 0.24mm and 0.1mm) and plain or liner clad. No secondary failure would be expected, but some secondary hydriding is likely and if there are design related differences these might be expected to show up. A quick examination of the rods could be made by neutron radiography.

A? a second step, a series of experiments in a "gas-flow" rig of a type currently under discussion at Studsvik could be performed. In this type of rig primary leaks of different sizes could be simulated in short-time experiments by the introduction of well defined amounts of steam into the fuel rods at one end and by analysis of the gas composition at the other end. By suitable control of the inlet steam/gas mixture the conditions at successive positions along a full length BWR or PWR fuel rod could be represented. Series of experiments with different leak sizes, different ratings and representing different burn-ups could be included as well as ones with rifled cladding, axially serrated pellets and similated end

- 79 - plena. Such experiments should show conclusively whether some of the design features referred to in this report would achieve the benefits suggested.

12. SUMMARY

1. Zirconium and its alloys have very high capability for absorbing H (only exceeded by Ti of the common metals) and form a series of hydrides. The solubility increases with temperature from less than 50 ppm (wt) at room temperature to about 700 ppm (wt) at 500°C. Hydrogen in solution diffuses to cooler regions in fuel rod cladding which has important practical effects. At a bare

zirconium alloy surface H absorption and solution occurs at low H? partial pressures. At higher H partial pressure absorption is followed by hydride layer formation e.g. in 10 minutes at 10mm Hg pressure; in 1 minute at 30mm H? pressure. (12) Figures 21 and 22.

The early problem of primary hydride defect formation was caused by local layer formatio i once the oxygen (from HO in the UO fuel pellets) had been reduced to a very low level by reaction with the cladding and U0-. In this situation the ZrO- layer rapidly becomes non-protective due to oxygen absorption in the metal substrate assisted by a H« partial pressure. Depending on the quantity of

H?0 (or other hydrogenous compounds) in a fuel rod, the bore surface temperature and the internal free volume, the conditions for local layer formation could arise. Such problems were resolved by improved U0 drying and in the case of G.E. by incorporating a

H2 getter.

- 80 - 3. The phenomenon of secondary hydride formation is related to the primary hydride problem, but with the important difference that there is a continuing supply of H.O (liquid or vapour or a mixture depending on reactor type) from the primary leak. Because ZrO films are protective when sound, i.e. only slightly permeable to H as in normal corrosion, secondary hydriding can only occur when locally the oxygen partial pressure is very low and the H partial pressure is high enough to cause layer formation. At lower H partial pressures accelerated hydriding with no layer formation can occur. The H flux in this case does not exceed the rate at which

the H? can diffuse away from the hot surface. Because following, or in the course of, pressure equalisation there is a flow of HO into the rod interior, secondary hydriding is usually some distance away, axially, fron the primary leak.

Primary leaks vary a lot in size depending on the cause. Some are so small as to be undetectable in reactor (say smaller than 10~ mm* in area) and normally only reveal themselves if subjected to PIE by a high H content or a hydrided getter. Some accelerated hydriding and a thicker than usual bore surface ZrO film may also be found. These very small leaks may seal up by oxidation at the leak site and show no deterioration. The more common type of small primary defect such as a small PCI crack is likely to have an area of about 10- 2 to 10-h mm* (such as a 10 micron diameter hole). Such holes are large enough to allow pressure equalisation within minutes to hours. At the smaller end of the spectrum the inflow of HJD may not be rapid enough to provide a large enough H flux to develop secondary hydride patches, but only to give accelerated

hydriding. The rate of oxygen reaction with U0? and cladding is

- 81 - crucial in determining whether a high enough H?/0_ ratio can occur at some distance from the leak to give conditions favouring ZrH. , layer formation (the familiar sunburst). This oxygen reaction rate is UO temperature and hence rating dependent as exemplified by the Locke curve. The influence of clad bore surface temperature, which can vary by about 50°C between BWR and PWR rods is less important because of a reduced irradiation effect at the higher PWR clad bore temperature. Ratine has two secondary effects; firstly it is the radiolysis of HO in the gap which provides the HO. radical which is mainly instrumental in oxidising the UO ; secondly sorre of the volatile fission products are effective at breaking

down a protective ZrO? film.

5. If, in a rod with a primary leak, conditions are such that secondary hydride patches (sunbursts) cannot occur, then there may be a slow, progressive, deterioration during which the leak size increases and the activity release rises and changes its character, e.g. isotope ratios. Since it is very rare for hydrides to form at or very clcse to a primary leak, or to continue forming if the primary defect is a hydride one, then the source of deterioration has to be sought elsewhere. Deterioration is often said to be caused by oxidation and erosion. Erosion may occur at a late stage if there is brittle powdery material formed at the leak site, but the earlier stages of deterioration are caused by increased clad oxidation close to the leak site and by UO» oxidation and expansion (temperature rise) in the region. It is well established that the

fission product release rate from U0? increases by one or two orders of magnitude in a defected rod because of changes in UO, stoichiometry and reduced gap conductance. The stoichiometry

- 82 - change is time dependant and part of the increasing activity release is caused by a gradual increase in UO temperature. There will nevertheless, even in severe cases, be a considerable volume of UO below 1200°C which will release very little of its fission gas and of course the fission product birth rate (the source term) will only change marginally. A lot of evidence suggests that the gap conductance effect is less important than the change in UO stoichionetry. The deterioration, i.e. increase in primary leak size, caused by local clad oxidation and U0? oxidation and expansion is effectively an oxide jacking effect which levers the defect open. At a certain defect size the situation changes in that conditions favourable to hydride layer formation will arise because the leak rate has reached a critical value. The deterioration time for small leaks may be very slow and a large number of leaks have been experienced which have survived periods of operation between 100 and 20C days without leaking secondary defects being formed. (Activity release doubling times of 100 -

150 days.) This is quite consistent with the Locke curve for rods with a surface heat flux of 125W/cir. or below. Some initially severe defects, such as major PCI defects with multiple open cracks, or some fretting defects in which the leak size may incresse rapidly, show a very rapid increase in activity release 2 even at modest surface heat fluxes e.g. 70W/cm . Such rods generally show a doubling time of 10 to 20 days, presumably because leak size is not rate-controlling for the internal reactions. Such rods frequently develop leaking secondary hydride patches within a few days. This class of defect doss not behave as predicted by

Locke. 6. The activity release from a leaking rod is determined by the rating (R/B ratio), by burn-up, by trapping in the gap (especially of Cs and I) and by the escape rate coefficient, a function of leak size. This subject is a complex one since most release is by back diffusion against an in-flow of coolant generated by internal reaction and pressure equalisation. There is presumably also a Bernoulli effect of the coolant flow across the leak which may result in some constant pressure differential. The other complication is that if there are large leaks with exposed UO then direct release into the coolant by recoil and knock out can and

does occur. Fission gases, volatile and soluble fission products and insoluble fission products such as 238U , 239Np are released to the coolant and can be detected and measured by ionisation detectors or garana - spectrometers. In BWR's the fission gases are separated with the steam and measured in the condensor off-gas while the other fission products are measured in the recirculating water. In PWR's all are measured in the primary circuit water. Measurements can be made on-line, usually on a by-pass flow, or on samples. Measurements can be made at specific energy levels of required gamma or beta emissions, or a whole spectrum can be measured over a period of time.

The measured activity release from a defect rod is a function of its rating and the defect severity. Since location of defects and asessment of numbers of defects contributing to measured activity are imprecise in pressure vessel reactors it is important to have some continuous measurement which detects increments of activity corresponding to new leakers and which measures rate of change of leak rate. The severity of a defect can be assessed by the ratio of short to long lived isotopes and 1/ I is commonly used for this purpose. Activity is often quoted in MBq/1 or mCi/g or cmJ, but it is more meaningful to quote it in release rate from the defect or defects, purification and leak rate from the circuit being corrected for. Other aids to the identification of severe defects are delayed neutron detectors (the longest half-life 87 delayed neutron emitter is Br at 56 seconds) oroc-detectors which are indicative of heavy elements in the release, i.e. of recoil or knock-out. There are methods of estimating the number of leaking rods based on an assessment of escape rate coefficient from a nunber of measured isotope ratios. (This is a semi-empirical method used in France based on a comparison of activity measurements and PIE results on fuel from Tihange and Fessenheim). The activity release for any isotope is then calculated assuming one leak with this assessed escape rate coefficient. Comparison of this number with the actual activity release gives an estimate of the number of leaking rods. The principle of this and other methods is that for a large leak in one rod then a large proportion of short half-life isotopes would be released because they would not be held up to decay in the gap. If for the same total activity release the ratio of short to long lived isotopes measured is low, then it indicates more than one rod contributing. The use of and usefulness of such methods depends to an extent on the type of release involved. In present day PWR's, even with the enhanced release in defect rods (about 150cC - 200°C increase in centre temperatures has been quoted) most of the release is athermal, i.e. by recoil and knock-out from free surfaces. T. e decay of short lived isotopes is therefore less than for fuel rods of higher ratings where most of the release is by thermally-activated processes and much of the

- 85 - short-lived fission products has decayed before it is released. There is a small complication that some activity is released by tramp uranium on the cladding surfaces, in the cladding alloy or in other components of the core. The activity release from such sources has the characteristics of activity release from a large defect, but of course the amount present is very limited, perhaps, typically, 3 to 6g in a reactor core.

7. Reactor operation can influence activity release and defect deterioration. Power or pressure changes influence flow into and out of a defect rod and particularly release of the soluble fission products I and Cs which tend to be trapped in compound form in the fuel/clad gap. Large spikes of I and to a lesser extent Cs are found on shutdown as the rod depressurises and again on start up as the rod heats up and expels water from the gap. PWR's show larger spikes than BWR's because water is more persistent in the gap. Fission gases show smaller spike effects especially on shutdown. Fission products such ar I and Cs which tend to be trapped in the gap may be 'washed out1 by an internal circulation between leaking deferts at different axial positions in a rod. This affect again is more pronounced in PWR's than in BWR's. For rods that have secondary hydride defects, either leaking or partially developed, stressing is bad in that it may lead to complete transverse rupture or a large axial split. Any form of power ramp or power cycling is therefore to be avoided if at all possible when defects are present.

8. There is no ma^ic way of anticipating a secondary hydride defect appearing and starting to leak, but the initial leak rate of a new

- 86 - hydride defect is usually quite small unless it is caused by some form of power shock which stresses the region severely. Secondary (or primary) hydride defects do have a special mode of deterioration which can lead to them becoming severe leakers. This is not associated with continued hydriding which does not occur because of the 0- availability, but with a diffusion of H from the hydride sunburst region to the cc-.'er outer surface with the corresponding volume changes. -vch a process was responsible for the severe defects examine ' ^y Forsyth and Jonsson(21 ) . The best anticipatory action •.-.-. can be taken is one based on experience and recognition r: >immon factors. A view should be formed on the likely cause <*.ny defect, based on experience, and its early release characteristics should be followed closely and compared with previous defect histories. Within reason, provided the initial doubling time of the activity release does not look like exceeding about 50-100 days, then the Locke curve can be used as a guide to the likely time available before secondary hydride failure occurs.

9. The susceptibility to secondary hydride defects can perhaps be reduced by some design measures short of reducing rating. Such measures include: a) increased internal free volume b) split gas plenum between top and bottom of rod c) facilitation of axial steam flow by rifled bore cladding or serrated pellets d) incorporation of a getter, preferably axially distributed. e) use of a barrier such as a 5 micron layer of Cu on the clad bore surface

- 37 - f) use of periodic small power or pressure fluctuations to

disturb gas gap atmosphere and encourage mixing.

13. CONCLUSIONS

1. The defects most of interest in Swedish reactors are likely to be

caused by PCI or fretting. Primary hydride and crud defects are

likely to be things of the past and corrosion defects may be things

of the future if very high burn-up targets are pursued.

2. All defects which admit coolant to the fuel/clad interspace may be

expected to show increasing activity release with time because of

leak size increase and increasing U0? temperature (oxidation and

reduced gap conductance). They will in general have a higher

steady release rate the higher rated they are (but see Appendix B)

and have a potential for the formation of secondary defects.

3. Many fuel rod? will show little increase in activity release for

100 day? or more.

i*. There is no known way, other than by the use of experience and

comparative studies, to predict precisely if and when a secondary

defect will form, but they can and will form quickly (say in the

first 20-50 days after leak) if the initial leak is large and the

leaking rod has a rating of 300W/cm or above.

5. Ways to handle and analyse activity release data to derive a

probable number of leaking rods and their burn-up (a guide to

location) have been described. Their accuracy is problematical,

- 88 - because release depends on so many factors, and their value is doubtful. A guide to burn-up ( 134Cs / 137 Cs) can be useful and a 239 severity index based on short half-life isotopes, Np, delayed neutrons or oc-emitters is important if serious circuit contamination is to be avoided. A selection of the best method must depend on the equipment installed at each reactor and the preferences of the operation's staff.

6. Some useful indication of the location of leaking fuel assemblies can be obtained in BWR's by control rod movements, but a similar technique has not been reported by PWR operators. Unnecessary control rod movements are undesirable so the use of this technique should be sparing, and only when the benefit clearly outweighs the risk.

7. The effect of power or coolant pressure changes on activity release (spiking) is well documented. The absence of iodine spikes in transients is a good sign of freedom from leakers and the magnitude is an indicator of total leak size, although nobody has attempted to use it quantitatively as a measure of the escape rate constant. In assessing the state of a core with respect to fuel defects it is important to make measurements in steady state conditions.

8. Obviously the best remedy for defect deterioration and propagation (including secondary defect formation) is to avoid the primary defects. The closest approach to achieving this aim has been in Japan in recent years, fuel defects (apart from some baffle jetting causing fuel vibration and defection in the PWR's Takahami 3 and

- 89 - Ohi 1) have been so few as to nake the problem of deterioration and secondary hydrides irrelevant. If the Swedish problem is confined to fretting then the best approach could be to look at inlet nozzle design to trap debris or at spacer design to minimise the risk of trapping foreign bodies.

9. There would seen to be no certain way of reducing the rate of deterioration of a primary defect, apart from reducing its rating and hence the local oxidation rate of cladding and UO , but the Cu barrier could be effective. However, it should be emphasised that many defects in LWF fuel rods, certainly at ratings up to 450W//cm, deteriorate only slowly and that apparently quite severe defects may release only small amounts of fission products.

10. There are large variations in reports of conditions for the formation of secondary hydride defects in LWR fuel rods. Variations are in time to form, rating and design features. If the secondary hydriding problem is assessed as ?. continuing serious matter then there are some design and operational changes which could, help. None of these is proved, but in the light of the well established mechanism they could be worthy of further evaluation or testing. One problem is that short rod behaviour is not likely to be representative of how full length LWR rods behave. Suggested items are:

a) reduce impedance to axial flow in the fuel clad gap without increasing average gap by use of rifled bore cladding or axially serrated pellets. b) increase interne! free volume by enlarging fission gas plenum.

- 90 - c) SDlit fission gas plenum between top and bottom ends of the fuel rod.

d) apply a thin H? impermeable barrier to the bore surface of the cladding. Copper is indicated as likely to be the most effective. e) modify clad alloy composition, or surface composition to increase protective quality of surface oxide layer. f) maintain reverse flows in the fuel/clad gap 'oy gentle power or pressure cycling to encourage mixing and so avoid the critical H /0 ratio being exceeded.

11. The response of different cladding alloys to secondary hydriding (10(e) above) is not represented in the literature available for survey. Clearly Zircaloy 2 and *l are not significantly different and there is no reported evidence for ZrNb alloys used in the U.S.S.P. or for the range of alternative alloys that have been tested world wide, e.g. the SCANUK alloys. Since there is now a new impetus for alloy development, because of the waterside corrosion problem with ultra-high burn-up fuel, it would be interesting to include tests of hydriding potential of alternative alloys in the presence of the oxide film.

12. The value of pre-autoclaving the bore of fuel cladding has been dismissed, partly as a result of UK experience, but it may be timely to review this matter more scientifically with alternative treatments prior to autoclaving, a range of thickness and different zirconium alloys.

13. Only the responsible authorities in Sweden can make the

- 91 - politico-economic judgements on whether any, and how much, further development is justified in an attempt to improve the behaviour of defective LWR rods.

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2. D.O. Pickman "Interactions Between Fuel Pins and Assembly Components". Nucl. Eng and Design 1975, Vol. 33, p. 125.

3. A. Garlick. "Stress-Corrosion Cracking of Zirconium Alloys in Iodine Vapour." Proc. BKES Conference. "Effects of Environment On Materials Properties in Nuclear Systems." 1971. Session 1, Paper 3.

1. A. Garlick et al "Power Ramp Experience in the Winfrith SGHWR". J. Br. Nucl. Energy Soc. 1977 Vol. 16 p. 225.

5. J.C Clayton "Internal Hydriding in Irradiated Defected Zircaloy Fuel Rods - A Review" Westinghouse Report No. WAPD-TM-1604, October 1987.

6. W.H. Elackadder, R.S. Forsyth, K. Malen and G. Ronneberg. "Fission Product Release to Loop Water from Operating Intentionally Defected Fuel" Proc. IAEA/IWGFPT Specialists Meeting on The Behaviour of Defected Zirconium Alloy Clad Ceramic Fuel in Water Cooled Reactors. Chalk River, Canada, 17-21 September 1979. Paper 3.2.4.

7. J.H. Davies "Secondary Damage in LWR Fuel Following PCI Defection. Characteristics and Mechanisms." Proc. IAEA Sept. 1979, as in Ref.

17. Paper 3.1.1*.

8. G.F. Caskey, G.R. Cole and W.G. Holmes. "Failures of U0- Fuel Tubes by

- 93 - Internal Hydriding of Zirealey-2 Sheaths." Paper presented at Symposium on Powder Packed U0- Nov./Dec. 1961; USAEC Report, CEND-153, Vol. 2, 77.

9. R.P. Marshall and M.R. Louthan. "Tensile Properties of Zircaloy with Oriented Hydrides." Trans ANS 1963 vol. 56 p.693.

10. H.E. Williamson and J.C. Ditmore. "Experience with BWR Fuel Through September 1971." Report No NEDO-10505 May 1972.

11. J.M. Markowitz. "Internal Zirconium Hydride Formation in Zircaloy Fuel Element Cladding Under Irradiation." Westinghouse Report No. WAPD-TM-351, May 1963.

12. R.P, Marshall. "Absorption of Gaseous Hydrogen by Zircaloy-2." J. Less Comnon Metals 1967 vol. 13 p.45.

13. A. Sawatzky. "The Diffusion and Solubility of Hydrogen in the Alpha Phase of Zircaloy-2." J. Nucl. Mater. 1960 vol. 2 p. 62.

1M. R.A. Proebstle, J.H. Davies, T.C. Rowland, D.R. Rutkin, J.S. Armijo. "The Mechanism of Defection of Zircaloy-Clad Fuel Rods by Internal Hydriding." ANS/CNA Joint Topical Meeting on Commercial Nuclear Fuel Technology Today. Toronto, Canada, 28-30 April, 1975.

15. D.O. Pickman. "Internal Cladding Corrosion Effects" Nucl. Eng. and Design 1975 vol. 33 p. I'M.

16. K. Une "Kinetics of Reaction of Zironium Alloys with Hydrogen." J. Less Conaon Metals 1978 vol. 5? p. 93.

17. D.O. Picknan. "The Causes and Behavious of Fabrication Defects in BWR Type Fuel." Proc. IAEA Symposium on Fabrication of Water Reactor Fuel Elements, Prague, 6-10 November 1979. Paper IAEA-SM-233/12.

18. A.S. Bain. "Mechanism Causing Hydride Defects at End-Cap Welds of

Zircaloy-Sheathed U02" Trans ANS 1969 vol 1? p. 99.

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20. E. Schuster, F. Garzarolli, K.H. Keeb and H. Stehle "Release of Fission Products from Defective Fuel Rods of Light Water Reactors." As reference 6, Paper ^.2.

21. P.S. Forsyth and T. Jonsson. "Experimental Study of Defect Power Reactor Fuel." Report STUDSVIK/NF(P)-82/72 Dec. 1982.

22. F. Gsrzarolli, R. Manzel and H. Stehle. "The Behavious of Defective Fuel Rods Under Continued Reactor Operation." Kerntechnik 1978 vol. 120 p. 463-

23. J>C. Janvier, Y. Kauffmann, M. Chagrot and G. de Contenson. "Post-Irradiation Examination of a Defected Fuel Rod Irradiated in a Pressurised Water Loop." As reference 6 paper 3.1.3.

2U. P. Chenebault, G. Kurka, A. Harrer amd J.P. Stora. "Evaluation of Fission Gas and Halogen Release Out of Failed Fuel Running at Constant

- 95 - Power and in a Power Cycling Regiae." As reference 6 paper 3*2.5.

25. D.H. Locke. "Defective Fuel Behavious in Water Reactors." ANS Topical Meeting on Water Reactor Fuel Performance. St. Charles, Illinois. Hay 1977.

26. D.H. Locke "Mechanisms of Deterioration of Defected LWR Fuel". As reference 6 paper 3.1-1.

27. B. Houdaille, R. Atabek and G. De Contenson. "Various Types and Behaviour of Defected Fuel rods in Experifflental Water Reactors and Loops.r As reference 6. Paper 3-1*2.

28. R.D. MacDonald and J.J. Lipsett. "Behaviour of Defected Zircaloy Clad UO Fuel Elements Irradiated at Linear Powers of 48OW/cm in Pressurised Water." As reference 6. Paper 3*2.6.

29. M.D. Freshley. "The Defect Performance of UO_/PuO_ Thermal Reactor Fuel." Nuel. Technology 1972 vol. 15, p. 209.

30. L.A. Neimark. "Examination of Fuel Rod Segment from Point Beach Assembly D-03" As reference 6 Paper 4.3.

31. R.R. Hobbins, D.W- Croucher, S.L. Seiffert, B.A. Cook, D.K. Kerwin, A.S. Mehner and S.A. Ploger. "Behaviour of Defective LWR Type Fuel Rods Irradiated Under Postulated Accident Conditions." As reference 6 paper 3.2.2.

32. L. Hallstadius. "ABB ATOM Fuel Rod Failure Experience." ASEA-ATOM

- 96 - report UK 88-273- 19th April 1988.

33. D.H. Lcoke. "The Behaviour of Defective Reactor Fuel" Nucl. Eng and Design 1972 vol 21, p. 318.

34. A. Garlick. Private Coonunication.

35. L. Lunde. "Localised or Uniform Hydriding of Zircaloy: Sone Observations on the Effect of Surface Conditions." J. Nucl. Mat. 1972 vol. 1», p. 211.

36. K. Joon. Reaktortagung, Haaburg (April 1972). Also "Probability of Fuel Failures Caused by Moisture." Paper to Enlarged HPG Meeting, Ustaoset, Norway, March 1971.

37. D.W. Shannon. "Role of the Oxidation Rate on the Hydriding of Zirconium Alloys in Gas Atmospheres Containing Hydrogen." Report HW-76562 Revised (1963).

38. G.E. Zima. "A Review of the Zr-2 Hydriding Problem." HW-66537 (1960)

39. T. Jonsson. "Study of Fission Products and Uranium Oxide Phases in a Defect BWR Fuel Rod." STUDSVIK/NF (P)-87/28, 15th July 1987.

40. T. Jonsson, A. Brisling, A. Holmer and T. Persson. "Defekta Stavar Fran Kraftreaktorer" STUDSVIK/K4-81/5, 18th February 1981.

41. R. Warlop, P. Chenebault and J.P. Stora. "Fission Gas and Halogen Release Out of Failed Fuel Rods." Proc. ANS/ENS Topical Meeting on

- 97 - Reactor Safety Aspects of Fuel Behaviour. Sun Valley, U.S.A. 2-6 August 1981. P. 2-434.

42. K.K. Neeb and E. Schuster. "Iodine Spiking in PWR's. Origin and General Behaviour." Trans ANS *978 vol. 28, p.650.

43. N. Eickelpasch, R. Seepolt and R. Hock. "Iodine release Mechanisa and Its Verification in Plant Operation." Trans ANS 1978 vol. 28, p. 652.

44. P. Beslu, C. Leuthrot, G. Frejaville, R. Behara, A. Dumont and Y. Musar.te "Description of a Method to Determine the Characteristics of Defected Fuels from Water Activity Measurements" As Reference 6 Paper 2.*.

45. J. Pelletier, J.P. Stora, P. Chenebault and C. Leuthrot. "Determination of the Fuel Rod Deterioration by the Means of the Uranium Contamination in Pressurised Water Reactors. Proc. OECD-NEA-CSNI/IAEA Specialist Meeting on Water Reactor Safety and Fission product Release in Off-Normal and Accident Conditions. Riso, Denmark, May 1983. p. 383.

46. P. Beslu, C. Leuthrot and G. Frejaville "PROFIP Code: A Model to Evaluate the Release of Fission Products from Defected Fuel in PWR" As reference 6, Paper 1.2.

47. P. Bourgeois and J.P. Stora. "Behaviour of Fission Products in PWR Primary Coolant and Defected Fuel Rod Evaluation." 5th SMiRT Conference, Berlin, August 1981. Vol. C. Paper C.1.1.

48. G. Kurka, A. Karrer and P. Chenebault. "Fission Product release from a

- 98 - Defected PVR Rcxi." Suclear Technology 1979 vol. 46 p. 571.

49. J.C. Janvier and M. Chagrot "Contamination of the Primary Circuit of a PVR by Fuel Rods with Ruptured Cladding." (In French) Proc. IAEA Symposium on Fabrication of Water Reactor Fuel Elements. Prague, Czechoslovakia, Novenber 1978. Paper IAEA-SM-233-2.

50. P. Bournay and C. Roels "Fuel Performance with EDF's PWR Reactors Related to the Behaviour of Defected Rods." Proc. ANS/NES Topical Meeting on Reactor Safety Aspects of Fuel Behaviour. Sun Valley U.S.A. 2-6 august 1981 p. 1-128.

51. J.P. Stora and P. Chenebault" Defected Fuel Rod Evaluation in EDF PWR Reactors". Proc. ANS/ENS Topical Meeting on Reactor Safety Aspects of Fuel Behaviour. Sun Valley, U.S.A. 2-6 August 1981, p. 1-119.

52. D.H. Locke "Ramifications of the Presence of Defected Fuel on LWR Power Reactor Operations." As reference 6, paper 4.5.

5?. K.E. Hughes and E.T. Chulick. "In-Line Fission Product Monitoring." As Reference 6, Paper 2.7.

54. J.T. Mayer, C.T. Chulick and V. Subrahmanyam. "B & W Radiochemical Analyses for Defective Fuel". As reference 6 Paper 4.4.

55. J.M. Cuttler and P. Girouard "On-line Detection of Failed Fuel in CANDU Power Stations." As reference 6 paper 2.2.

56. R.D. MacDonald and J.J. Lipsett. "Behaviour of Defected Zircaloy Clad

- 99 - UO Fuel Elenents Irradiated at Linear powers of ISKV/B in Pressurised Water". As reference 6 Paper 3-2.6.

57. T. Ingeaansson. Private communication. June 1989.

58. R.J. Rustagi and M. Das. "Experience with Modelling and Behaviour or Defected Fuel in RAPS-1". As reference 6, paper 2.1.

59. J.B. Nandwani. "Behaviour of Failed fuel - The Monitoring Methods and Operating Considerations at TAPS." As reference 6, paper 4.7.

60. H.W. Kalfsbeek. "The Abundance of Fission Gases in the Off-Gas of a Boiling Water Reactor." Nuclear Technology 1983 vol. 62 p. 7.

61. J.J. Lipsett and R.D. MacDonald. "Another View of Fission Gss Release from Defected CANDU Fuel Elements." As reference 6 Paper 1.3.

62. N. Oi. Private communication. August 1989.

63. N. Kiaer-Pedersen and H. Mogard. nIn-Reactor Performance of Fuel with Rifled Cladding." Proc. ANS Topical Meeting on L¥R Fuel Performance. Williamsburg U.S.A., 17-20th April, 1988. p. 284.

61». D.O. Picktoan. Letter to J. Stanbridge, BNF pic. 1.9.89.

65. J.H. Davies. "Accelerated Pe)let-Cladding Interaction Tests of Barrier Fuel". OE»P-25356, June 1981.

66. H. Mot^rrt. Private communication.

- 100 - I ABLE.

Hadio-Isotopes of Interest for Activity Release Measurement»

(2) Precursors (1) Half- Isotopes Lives Daughters

85Se(32s) —— 85Br(2.9m) 85Kr io.72y 85Rb (S)

•i 85Kr 14. 'i8h

87Se(5.6s) - 87Br(56s) 87Kr 76m 87R> (R)

83 88 88 88 Sed.5s) - Brd6.U3) Kr Rb (R) 131 131In (0.3s) 1J1Sb (23m) 131Te(32h) 8.0<4d (R)

132In (0.2s) 132Sn(H0s) —-- 132Sb (Urn) 132Te(78h) 132I 2.28h 132Xe (S)

133 133 133 133 I 5n (1.5s) Sb (2.5m) Te(55m) X 20.8h o 13USn ds) - Sb (10s) - 13"Te (42.) 13«X 52.5m 13*Xe (S) I 135 135j 135 135Sbd.7s) Te(19s) 6.58h x* (R)

133Snd.5s) 133Sb(2.5m) —— 133Te(55m) 133I(21h) 133Xe(5.25d) 133cs Staole -

133Cs(n) 2.06y (S)

137Te(Hs) — J ' T f Illm \ __ 137Xe(3.8m) 30.2y 137B. (S)

138Te (1.6s) 138I(6.Us) —- 138Xe(mm) 32.2m 138B. (S)

133Snd.5s) 133Sb(2.5m) —— 133Te (55m) 133I (21h) 5.25d

135Sbd.7s) 135Te (19s) —— 135I (6.6h) 135xe 9.1h (R)

138Ted.6s) (R)

239 U (21.5m) Pu

/ A \ All

STATUS OF EDF'S 900 MW (e) PWR UNITS IN COMMERCIAL OPERATION AS OF JUNE 15, 1981

B.U. (MED/MTU) PRIMARY COOLANT ACTIVIYY CHARACTERIZATION OF THE DEFECTS n (mCi/t) f!

ESTIMATE LEAKING F/i'B UNIT CYCLE E.F.P.D. CORE PEAK Xe-133 NOBLE 1-131 D.E.1-131 AVERAGE ASS. GASES ESCAPE RATE FAILED IDENTIFIED REJECTED COEFFICIENT RODS BY SIPPING V (s-*)

FESSENHEIM 1 EOC 1 400/403 14 640 17 46O 50 100 2.5 4 4-10-1' 1 or 2 2 ? EOC 2 284/276 19 680 27 380 130 200 5 10 3'10'VlO'* 2 1 1 EOC 3 260/285 20 590 35 220 240 400 8.4 17.6 10" 5

FESSENHEIM 2 EOC 1 44S/410 16 280 19 320 1 8 0.1 0.7 0 0 EOC 2 280/251 20 580 28 750 180 330 10 20 B-IO'VOO-* 4 4 4 3 182/274 18 520 33 800 2 7 0.2 0.7 0 o ro BUGEY 2 EOC 1 372/394 14 300 17 000 0 - 0 2 239/276 18 300 25 900 750 2 600 75 173 10-' 2

BUGEY 3 EOC 1 424/394 16 270 19 280 0 0 EOC 2A 111/244 14 620 23 550 40 - 3.6 5 10 •' 1 1 1 M BUGEY 4 EOC 1 387/393 14 870 17 660 50 60 10-* 1 0 0 K 2 176/267 16 240 24 400 93 200 4 5 N.C. 2 ^». Z BUGEY 5 EOC 1 418/393 \6 040 19 160 1 200 2 000 24 31 10 •" 20 8 2 BOC 2 7/248 10 650 18 800 200 700 5 10 N.C. N.C. HI W N.C. : non-calculated CIO 00

to TABLE 3 Behaviour of Some Unusual Defects in the SGHW Reactor Average Peak Irradia- Activity Release Number of fuel/ Internal Secondary Cause of Burn-up Rating tion tine Oefect Rods Clad Volune lydride Remarks when after after fleaent Oefect ... r 131. 133. )a»dge Oefected Defected Defected Off-Gas I I Gap (e.M (HMd/tu) (N/c*5 (Days) (CiMev/d) ( Ci/1) (••)

A(1) PCI 4500 386 85 600 5 25 1 0.13 45 Yes A second rod had leaked. A small FBR amount of secondary hydriding.

(3) B(2) PCI 5600 507 Few hrs 63,000 (5) (5) 8 0.165 72 Yes A ninth rod had leaked. A small UGA amount of secondary hydriding.

(7) (7 C PCI 11,800 3 68 18 4,300 650 2900 > 6 0.165 12 Yes Hydriding at bottom end and near PEF top in all rods.

Frettir g o (4) D in Plenum 400 325 26 (6) (6) (6) 1 0.165 72 Yes At bottom end and at many JYP intermediate locations.

Primary E Hydride 900 ca600 106 N/A 1 2 1 0.13 45 Yes One leaking rod, secondary CJE hydriding near bottom.

Leaking F Thermo- 650 550 4 Up to 97 1080 1 0.165 72 N/A Leak into centre of U0_. CKZ couple 22,600

(1) Rod A from reference 4 (6) No activity release detected pric»* to a (2) Rod C from reference 4 shutdown. (3) Confined to hydriding at the mouth of partial penetration cracks. (7) mCi/h. No separate hydride patches noted. (4) Estimated. (5) Not measured. STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

Figure 1

Examples of frettinq damage:

a) 0.25 mm deep hole at fixed stop in spring grid design

b) spacer pad wear at grid

a) c) smooth clad wall thinning from contact with ring grid

c)

Figure 2

Small PCI crack believed to have re-sealed.

- 104 - STUDSVIK NUCLEAR 3TUDSVIK/NF(R)-89/83

Figure 3 Primary hydride sunbursts a) early stage but stress effect on hydride orien- tation already noticable b) major sunburst with only minor leak, formed in 85 days irradiation, peak LHGR 520 W/m

c) sunburst with large leak showing be- ginning of hydride re- conversion to Zr

- 1.05 - l/J 58 Figure 4 2 C Massive hydride sunburst O showing lamination effect f resulting from ZrH-|6 re- conversion to Zr in inner region of cladding.

CO 10

Figure 5 Axial PCI cracks. STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

O

IODINE 1-131 OATA

aa 20 CO 60 BO 100 DAYS AFTER DEFECTION

x - 236 W/cm, Rod FFR • -256 W/cm, Rod BYM A - 355 W/cm, Rod CQW V - 402 W/cm, Rod BNA O - 413 W/cm, Rod DNT • - 440 W/cm, Rod UFY • - 449 W/cm, Rod CRT

Figure 6 Bulk off-gas and iodine release data for individual rods with small to moderate defects over first 100 days.

- 1.07 - STUDSVIK NlfCLEAR STUDSVIK/NF(R)-89/83

Figure 7 Small oxide nodule on bore of cladding opposite radial U0o crack.

X100

Figure 8 Deltoid hydriding at mouth of partial penetration PCI crack. Note hydride re-orientation in plastic zone at crack tip.

- 100 - IS

Irflhnq MtWrmi O Foiiur* -hydrided -KM op*rai>»9 C Op*'O

SUCCESSFUL DEFECT OPERATION PRTRIFERTFlooel

Soilon ETRittiioeo SKippmqpo" NPD 0'fsden i* oundrcmmingen EtR(LSBR) ?S BRP VBWR Gundrtmmingen o VD HYDRIDE FAILURE ETRmiiooclOEl MRX loop'WM] PRTRirERTFIooo) NRX looplWHI HWCTR DO-nt

MINI© — - z XX) w I oo vo oo

Figure 9 Proposed failure limits (surface heat flux/time to failure) for Zircaloy clad LVJR fuel. < M

1 IMHCOMIt DISCHARGI S0»r- n

IODINE I-13) DAT*

IMO

O ?00

O I

z w 1* >• DAYS AMIR OIMCIIOM I oo (*> ° PEF (Element C) 368 W/cm. • UGA (Element B) 450 W/cm. O CKZ (Element F) - 550 W/cm. X BYE 550 W/cm. • PEF (Element C) - 368 W/cm. X RLV - 497 W/cm. A RLV 497 W/cm. A BYE - --50 W/cm. • CKZ (Element F) 550 W/cm. Figure 10a Figure 10b JXX data for 8elected STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

Figure 11 Transverse clad fracture near end of BWR fuel rod ascribed to fatigue failure in hydride region. Incipient defects seen at both ends of a pellet in rod 2 below.

111 - 5TUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

Figure 12 i :r Axial and 45° cracks in defect rod 999-H2 from Ringhals 1. Probably fretting defect, 598 ran from bottom end (see also Figure 14).

Figure 13 Spalled bulge in defect rod 999-H2 from Ringhals 1 Probably secondary hydride defect 2100 mm from bottom (see also Figure 15).

- 112 - STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

Figure 14

Section of defect in Ringhals 1 rod 999-H2 at 598 mm from bottom. Believed to be primary defect.

1 mm

SECTION 1

- 113 - SECTION 2 STUDSVIK NUCLEAR 3TUDSVIK/NF(R)-39/83

Figure IS

Section of defect in Ringhals 1 rod 999-H2 at 2100 mm from bottom. Probably secondary hydride defect.

1 mm

SECTION 1

- 114 - at Idride

SECTION % STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

10' •>

vg 10T* iO1

vg 10•*

0 .

• ««••• i—i—r » i i t—;—i-^f -1 - .^-•—^^^.vg Iff;, vg 10

~- vg 10

1 vg o.{

" • —

• 1 i i t l 1 1 i l l 1 1—t—I—i—i \ i • •>-»•». • >-, vgiO*

D1 133XG/135XG

. 11 .,t.13Bv^/135v^ ' IÄXe/ aXe

Tcmpcrit'.'rc(0C)

I1 | | 800 1000 1200 1400 1600 1800

Figure 16 Tihange 1 parametric study. Correlation of gap escape rate constant with isotope ratios.

- 115 - 5TUDSVIK NUCLEAR STUDSVIK/NF(R)-8 9/8 3

IMMMUI I"

llM

••

M Malm

Figure 17 Correlation between inferred diffusion constants and linear heat generation rate for intact and defect rods.

a i "

SLOPE 0.S

H 1 1 Mil I-ll» T T •(•• HM'tx (kll. 'I Figure 18 Relationship between fractional release and decay constants for fission gases and iodines.

- 116 - STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

E O

5 z o

oe - z u u O u

10

TIME (h)

Figure 19 Fission gas and iodine spiking effects on shutdown and start-up.

- 117 STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

Figure 20a Avtivity release following a shutdown.

ROD 1319

CCRCNKOV SlCMAW

SO

So

10 LINEAR POtff * 10 w

Figure 20b Activity release during power ramping

- 118 - STUDSVIK NUCLEAR STUDSVIK/NF(R)-89/83

NI5 \Lxrc 35C \ o (300* c

o •A \ < £ ' -• - ^-— ——. — - o — • I», 10 20 30 40 50 60 70 80 90 Time, minutes

Figure 21

H2 flux versus time to saturate a Zircaloy-2 surface (calculated).

25 Immediote Surface Saturation ond Loyer Formotion 20 Homogeneous Sorption 8 15 followed by Hydride Loyer Formotion ot PointM X"

eneous Sorption ond Solution

10 20 Time, minutes

Figure 22 Hydrogen absorption by Zircaloy-2 at 400°C.

- 119 - APPENDIX A

DESIGN FOR IMPROVED GAS GAP MIXING

BACKGROUND

The thermal feedback effect in nuclear fuel rods first cane to general notice with the reporting of large fission gas release variations in near-identical fuel rods in Maine Yankee in the early 1970's. The effect depends on local degradation of gas-gap conductance by released fission gasest followed by a runaway situation as ever more fission products accumulate. The solution is to make axial flow of the released fission gases easier or to moderate the effect by He pre-pressurisation.

Work at Halden based on studies with gas flow rigs (eg. HPR 295» 1983; HPR 325, 1985; HPR 328, 1985; HWR 119, 1984; HWR 172, 1986) has shown that gas mixing in the fuel/clad gap of an operating fuel rod may be very slow. A segregation of lighter gases to the top end and heavier gases to the bottom end (the Clusuis Dickel effect) has also been reported (paper 117, Enlarged Halden Programme Group Meeting, Sanderstolen, Norway, March 1986). This effect depends on a cross gap temperature gradient.

It is now widely recognised that susceptibility to PCI defects is affected v.y the slow axial diffusion of released fission products.

The liability to develop secondary hydride defects is also highly dependent on gas mixing within the gas gap. If axial diffusion can take place easily, then the build up of a sufficiently high H-/CL ratio to permit ZrO.film breakdown

- 120 - and localised hydriding is less probable. There may be some special conditions where it can still occur, but certainly the incidence should be reduced.

Some design approaches for reducing the likelihood of secondary hydride defects have been discussed in the body of this report.

DESIGN TO IMPROVE AXIAL GAS FLOW

Increasing the size of the fuel/clad gap is the most obvious measure to improve axial gas mixing. This is, however, not an approach approved by most designers and fuel/clad gaps in use today mostly lie in the range 0.1 to 0.25 mm (diametral). Many experiments have been performed using fuel rods with large gaps and generally they have performed badly (e.g. rod BRi in INTER-RAMP; rods BG9 in SUPER-RAMP). These large gaps hava the disadvantage that fuel temperatures are higher and clad creepdown and fuel pellet re-location will gradually close the gap. There is a a neutronic disadvantage also in large gaps as well as an economic one.

As early as 1973, Mogard (Swedish Patent Nos. 7301524-0 and 7401150-3) suggested the use of cladding with an axially undulating bore surface as a remedy for PCI defects. The theory was that this design would reduce the circumferential stress concentrations caused by opening radial cracks in the U0- and the frictional resistance between fuel and clad once in hard contact. This design has been shown (Mogard, paper to Enlarged HPG - Meeting, Loen, Norway, 1988) to be beneficial not only in increasing the threshold for PCI defects, but also in reducing fission gas release (Studsvik Nuclear Technical Note, NF(R)~89/18). The style of cladding now in use is referred to as rifled and has 40 prismatic sides with the radial depth of the triangular channels at

- 121 - 0.02mm. The design ensures that even in a closed gap situation, e.g. a power ramp, channels for axial gas flow (and circumferential circulation at pellet ends which are chamfered) is still possible. The effectiveness has been shown by time-constant measurements in a ramp.

Following the logic behind the development of the rifled cladding, it is possible that the same effects could be achieved by the use of an axially serrated pellet. This ide* has previously been advanced by the author as a PCI remedy on the basis that an increased number of radial cracks could be developed in the UCL pellet. While this potential advantage remains, such a design would also assist axial gas flow and mixing without the disadvantages of a uniformly larger gap. Depending on the detail, gas gap conductance would not be reduced and there would be no, or a lesser neutronic penalty. Ideally such a design would be used with unground UCL pellets, but if necessary grinding could be used without a major adverse effect on the concept. In Table A, some conceptual designs have been considered based on serrations of depth 0.1mm, 0.05mm and 0.025mm. The volumes in the gap are independent of the number of serrations. The comparison rifled design is based on an ASEA-ATOM 8x8 BWft with a 40 sided bore, which gives a triangular gap of maximum depth 0.02mm at closure. For gap volume calculations the gap has been assumed to be triangular which truly it is not because the base is curved. Standard ASEA-ATOM 8x8 BWR designs with fuel/clad gaps of 0.120, 0.150 and 0.240mm are Included for comparison. The rifled clad cases have the largest cold gap volumes per unit length, because the minimum diametral gap is fixed at 0.2mm (see Studsvik Nuclear Technical note NF(R)-89/i8, Figure 1). This gap would only close, bringing the triangular passages into operation, at a very high rating (centre U0_ temperature over 2000°C), say over 450W/cm.

- 122 - In the serrated pellet cases chosen the one with a mean gap of 0.150mm would close the uniform gap at a rating corresponding to about 1075°C centre UO^ temperature (say 250W/cm). Also the residual radial gap at 0.05mm (8.6mm'/cm length volume) is more than double the gap and volume clearance of the 8x8 BWR rifled design. This design is indeed a very conservative one, the triangles formed between the apices of the serrations have a base of 0.685mm based on 50 serrations and a height of 0.05mm, an aspect ratio of more than 13, they would barely be visible to the eye! It could be that a larger number than 50, or a deeper penetration would be better. It is important that the loss of U0_ be minimised and that the span over the apices should be small enough to prevent clad creepdown or significant bending moments. However, as one of the claimed advantages of this design is the formation of additional radial cracks in the U0_, the notch effect must be great enough to ensure this. These radial cracks themselves will enhance the axial flow capability of this serrated pellet design.

INFLUENCE OF RIFLED CLAD OR SERRATED PELLETS ON FISSION PRODUCT RELEASE

It has been suggested that the designs proposed in this appendix, even if they are proved to be effective in raising PCI thresholds and reducing susceptibility to secondary hydriding, may result in higher steady state fission product release.

It is, indeed, most likely that slightly more fission gas will be released during steady state operation, since transfer within the fuel/clad gap will be made easier. More of the condensibles, iodine and caesium may also be released, especially from areas of the bore surface close to the leak, but this must be uncertain pending some actual measurement.

- 123 - Assuming that steady state releases do increase, then it is equally certain that spike releases during shut-down and start-up, or during other power manoeuvers, will be reduced. This could prove to be advantageous as the spike releases, especially following shut-down are operationally inconvenient and result in loss of time in opening up the vessel.

- 124 - TABLE A

Gap volume/cm Fuel Total Rod Clad I.D. (nun) Pellet O.D. (mm) Fuel/Clad Oap (mm dia) Length (mm1) Gap Design Volume UNIFORM/ >FLIET CLAO I.D. UN/MAX /cm AVERAGf J.D./M1N 'MAX. :i AD ID OR MAX MIN UNIFORM MAX N1N 'UNIFORM MAX MIN (mm1) :IAD I.D. >Elin OD UN/MAX PFllfTOD Swedish 8 x 8 BWR - - 10.90 - - 10.66 - - 0.240 - - 38.1 Standard Clad

n n - - 10.90 - - 10.75 - - 0.150 - - - 25.5

n ii - - 10.90 - - 10.78 - - 0.120 - - - 20. H

Swedish 8x8 BWR 10.91 10.90 - - - 10.70 0.2110 0.200 0.220 33.9 - 3.1 37.3 Rifled Clad

UK 57 Rod BWR - - 11.00 - - 10.875 - - 0.125 - - - 21.5 Standard pellet — UK 57 Rod BWR — 11.00 10.90 10.70 0.30 0.10 0.200 17.2 17.1 31.3 Serrated Pellet

« it - - 11.00 10.90 10.80 - 0.20 0.10 0.150 - 17.2 8.6 25.8

n n - - 11.00 10.90 10.85 - 0.15 0.10 0.125 - 17.2 1.3 21.5

17x17 PWR Standard - - 8.35 - - 8.19 - - 0.165 - - - 20.8

17x17 PWR - - - 8.15 0.2K 0.20( 0.220 25.9 - 2.6 28.5 Rifled Clad 8.39 8.35 17x17 PWR Serrated Pellet - - 8.35 8.25 8.15 - 0.20 0.10 0.150 - 13.0 6.5 '9.5 APPENDIX B

THE EFFECT OF LINEAR HEAT GENERATION RATE ON ACTIVITY RELEASE FROM DEFECTIVE FUEL RODS

One of the surprising findings from this literature survey is that over a certain rating range there does not seem to be much variation in steady state fission product release. The main sources of data have been Locke , Hallstadius(32) and MacDonald and Lipsett(56). The Locke data is for a limited number of naturally occurring defects in the SGHW reactor in the U.K. over a range of ratings fron 236 to 449 W/cm. The Hallstadius data is based on activity measurements in Swedish and Finnish reactors prior to shut-down for the yearly outage, together with the number of leakers found during the sipping campaign following the outage. Such data is of little value where there were a large number of leakers covering a wide range of ratings, but where there were only a few leakers over a narrow range then it can be of value and has been used. In both the cases of the Locke and the Hallstadius data there is a further problem in that the leak sizes are not known. This means that a higher release from a rod could be caused by a higher rating or a larger leak.

The MacDonald and Llpsett data is all on rods with artificial defects of known size and with accurately known ratings, but it may suffer to some extent from the known lack of reliability of the data from artificial defect rods. It may be, however, that fission gas release data ( Xe and Xe only) is of good quality, although allowance must be made for the fact that fuel/clad gaps were very small (80 to 100 microns) in half the rods so that at the high ratings of most of these rods gaps would be tightly closed.

- 126 - Some of the data fro» these sources is given in Table B. The rating range is very large, fro» 9KV/m to 70KH/m. Unfortunately truly comparative data is not available from the 3 sources, but the figures given in the Table, in ascending order of LHR are sufficient to show that even over this very wide range of ratings there is no clear correlation between activity release and rating. It •ust be presumed that leak size was responsible for some of the amjor deviations froc the general pattern, but the CANDU data does not show a very dranatic effect of the size of the artificial defects. For example rod MSZ at 58KH/m with a 10an* leak size has released less than rod 1PR-1 at 55KH/B with a 2mm* leak size. Both these rods were of identical design and irradiated under identical conditions in the X.6 loop of MRX. The six rods in this group that had the smallest leak sizes ( 0.1mmc) did not show abnormally low activity release.

Hallstadius (ASEA-ATOH Report, UK 88-273) has not analysed the 13 Xe data from Swedish reactors, but some figures are given in his Table 1. Similarly Locke 138 does not quote Xe data for the leaking SGHW rods, but only total ft + ¥ activity in ClMeV/day units. Looking at these sets of data in comparison with the CANDU data of MacDonald and Lipsett, both seem very high. The average Swedish release per rod is 57 MBq/s (range 3.8 - 150) and the total SGHW off-gas activity ranges from about 20MBq/s for a very small defect to about 200 for a moderate sized defect and 1500 MBq/s for a severe defect. Such 138 figures make the Xe releases from the short CANDU type rods at high ratings appear very small indeed.

There are a number of factors that can help to explain thia apparent insensitivlty of fission product release to rating. In general the release of fission gases and volatile such as iodine, bromine and caesium from UCL are thermally activated, apart from small quantities released from free surfaces

- 127 - by recoil and knock-out. The release from U0_ is therefore expected to be a function of the rating which determines U0_ temperature distribution. The main external factors which can effect it are U0_ thermal conductivity, b -n-up (which reduces conductivity and increases surface to volume ratio) and stress. High compressive stresses are known to impede the development of grain boundary gas bubble linkage and the formation of edge tunnels which link up to free surfaces.

Release from a leak in the fuel rod cladding to the coolant depends on the fission products available in the internal free spece and the impedance to their diffusion within the fuel clad gap to the leak site. In the case of iodine and caesium there is also a delay in release caused by compound formation in the gap, mainly compounds of iodine, caesium, hydrogen and oxygen, but also some containing Zr and U.

It now becomes clear that increasing rating, although it will increase U0_ temperature and fission product release from the U0?, may minimise the increase because of interactive stresses acting in the UCL. Similarly, release from the fuel/clad gap will be delayed because of the reduced gap at higher ratings and the greater likelihood of trapping reactions within the gap. These are the factors which can help to reduce the expected higher fission product release from higher rated defect fuel rods. The reduced release from the rod means that there is a higher inventory of released fission products within the rod which are expected to be released at or following power fluctuations and especially at shut-down and start-up. The net release from higher rated rods over a period of time is therefore higher, but this may take the fora of a moderate steady state release with larger spike releases for t.'ie higher ratings.

- 128 - If fuel rods are normally operating at ratings which correspond to centre U0_ teraperatures below the threshold for significant diffusional release (say below 1000°C) the activity release can be expected to be relatively insensitive to rating (because it is athermal) even allowing for a temperature increase of perhaps 200°C resulting from the defect.

In summary, under some conditions activity release from defective fuel rods may be insensitive to rating, but in such cases larger spike release effects are likely to be found on shut-down and start-up.

- 129 - TABLE B Activity Release from Fuel Rods as a Function of Rating

Activity Release (MBq/s) Source Rod or Reactor Linear Heat Reference Rating 133j 133Xe 138Xe (KV/m) Hallstadius R1-83 9 0.3 1.1 n R1-86 13 0.5 1 n R1-84 1J» 0.175 1.2 n F1-82 16 0.048 1 n F1-85 17 0.008 4

it F2-84 17 0.44 2

« TVO-2-86 17 10 M TVO-1-81 18 1.2 m B2-86 18 15 (1) it 03-86 20 0.4 40 Locke FFR 23.6 0.13 0.82 Locke BYM 25.6 3.14 12.6 MacD. t Lipsett CEV-1,(2' ) 32 0.014 0.09 Locke CQW 35.5 0.3 0.13 Locke BNA ca40.0 0.4 2.52 Locke DNT 41.3 0.15 0.69 Locke UFY 44.0 0.1 1.85 Locke CRT 44.9 3.28 8.11 (3) MacD & Lipsett RPP 47 1.15 3.19 (3) N LFZ 48 0.99 1.23 (3) •f RPL 49 2.91 3.52 (2) II CEV-2 50 0.5 0.51 (5) n RPR-2 54 5.05 12.26 (3) n RPR-1 55 3.52 4.26 n NSZ 58 1.53 2.38 (2) n CEV-3 64.5 1.99 2.87 (2) it CEU 65 0.67 1.96 n CEX(2) 69 6.58 2.21 N CEW(2) 70 6.43 4.12

1. Rating was higher in second half of cycle. 2. 0.33mm diameter drilled hole. 3. 1.6mm diameter drilled hole. 4. Slit 1mm x 10mm 5. 2 drilled holes 1.6mm diameter and 1 drilled hole 2.1mm diameter. APPENDIX C

BEHAVIOUR OF DEFECT FUEL RODS IN OFF-NORMAL CONDITIONS

Much attention has been devoted to the behaviour of fuel-rods in off-normal and accident conditions. While LOCA and RIA situations are outside the scope of this survey, a power/coolant mismatch, leading to clad dryout is a more probable occurrence, especially in a BWR for which stability margins are not large. In the dryout (or film boiling) regime there is a sudden large increase in clad temperature. In one of the rods irradiated in the PCM test series at Idaho Fall, at 670 W/cm peak, the clad temperature reached over 1400°C and the U0_ centre temperature over 2000°C. (D.W. Croucher "Behaviour of Defective PWP Fuel Rods During Power Ramp and Film Boiling Operation" NUREG/CR-0283, TREE-1267, February 1979). The behaviour of defective rods in this test series did not produce any dramatic contamination or rod break-up, although there was some loss of U0_, a lot of fuel melting and severe hydriding in the lower regions of fuel rods that did not go into film boiling. Three different types of defect (hydride, pinhole, axial crack) were tested and in no case was the rod behaviour seriously affected, compared with intact rods experiencing the sane film-boiling conditions. The extensive clad embrittlement was caused by 0. embrittlement more than by hydriding. This work is desclbed further by Osetek and King and by Bobbins et al in papers to the 1979 IAEA Specialist Meeting at Chalk River.

- 131 - APPENDIX D

A number of perhaps minor points of interest that have emerged during this survey are noted in this Appendix.

(a) Some Kr, and presumably also some Xe is released from irradiated fuel when it is heated in steam to about 500°C. This is about 0.7 to 1.5} of the total inventory according to R.A. Lorenz et al (1979 IAEA Meeting at Chalk River p. 169). This retention which is clad temperature dependent is believed to be in the surface layers of fuel and cladding. It has been used to estimate clad operating temperature in LMPBR fuel. (K.M. Swanson et al. J. Nucl. Mater 1969 vol. 33 p. 203).

(b) Cr- barrier clad, one of the PCI remedies tested by GE gave a very severe secondary hydride defect in a deliberately defected rod. This occurred in a 12h hold at 676W/cm, presumably because of a defect in the Cr layer. A complete ZrH. , shell formed at the axial location of the defect in the Cr layer.

(c) Hydriding under stress as a secondary phenomenon appeals to occur rapidly as pointed out in this paper. A rod in C3USIF0N 2 had a defect made during operation at U00W/cm. This rod had a very small gap (20 microns) and a secondary hydride developed over the whole cross-section in 'some hours'. There was a scram between forming the defect and the secondary damage appearing, so water will have entered the defect because of the static head of water. The defect was a fine slit a few millimeters long.

- 132 - (d) A rod which defected in film boiling during a clad ballooning test (Rod IB-019, Test IE5, LOC series in PBF. R.R. Hobbins et al. 1979 IAEA Meeting at Chalk River, p. 161) developed hydriding, up to 1000 ppm, said to be enhanced hydriding and which led to rod fracture in handling during only 60 seconds irradiation after the rod burst. There was a very wide gap during this time over the region of the rod in film boiling.

(e) J.C. Clayton reported on hydriding in the plenum of a fuel rod caused by contamination with Ni from the plenum spring.

(f) There are reports of U0- becoming very friable and powdery in defective fuel rods close to the defect, possibly due to grain boundary oxidation(56).

(g) Gas tagging of PWR fuel rods has been discussed as a means of identifying rods which develop leaks. (K.C. Cross et al. Trans ANS, 1975 vol. 22). A similar techniques has been used in some LMFBR fuel rods, but it is an expensive technique of uncertain success rate.

(h) The diffusion controlled release of isotopes of Br from U0- is very rapid, D s 1.9 x 10-1 7m 2s -1 . This is almost 100 times larger than the similar diffusion coefficients for I and Xe and 200 times larger than 87 for Kr. It is of some importance since Br is useful in detecting severe defects as an emitter of delayed neutrons with a half-life of only 56s.

- 133 - (1) Fuel rods with leaks In the gas plenum (upper) are said not to t 27) suffer secondary hydriding damage ' , although Janvier et alvtJ/ do report some enhanced uniform hydrogen pick-up and some hydriding near the end plugs in the cooler parts of the rod is referred to. It would be interesting to know whether the top end plugs were affected.

(j) In this report the possibility of redesigning spacer springs to minimise the risk of fretting defects has been referred to. In the same context, of course, it must be obvious that design changes to the inlet nozzles to help trap debris should be a priority. The problem will be to avoid changing the pressure drop, especially increasing it.

(k) There are some observations of acicular (or Widmanstätten) structures with ILO- needles in a U0- matrix, but little mention of how the composition was checked. There are some suggestions ' that these acicular structures may be caused by some fission products and are not w

(1) A report from NRC (NUREG 4485, Jan. 1986. M.F. Meller et al "The Impact of Fuel Cladding Failure Events on Occupational Radiation Exposures at Nuclear Power Plants") has examined the impact on occupational exposure of allowing higher activity release limits so that there are fewer re-fuelling shutdowns. No firm conclusions are reached, but it is clear that occupational exposure does not increase proportionally to radiation exposure rates in plant areas, and that airborne radiation levels do not increase pro rata to activity release from the fuel rods.

- 134 - BIBLIOGRAPHY

HYDRIDING

1. J.M. Markowitz. WAPD-TM-351 (1958).

2. K. Joon. "Primary Hydride Failure of Zircaloy Clad Fuel Rods" Trans ANS 1972 186-187.

3. D.W. Shannon. Corrosion 1963 vol. 19 PP. 114-420 (or HH 76562 Rev. Feb. 1963).

4. R.F. Boyle and T.J. Kisiel. "Hydrogen Permeation of Zircaloy 2 Corrosion Films". Bettis Technical Reviev WAPD-BT-10 1958 p.p. 31-48.

5. E. Hillner. "Hydrogen Absorption in Zircaloy During Aqueous Corrosion. Effect of Environment". WAPD-TM-411 (Nov. 1964)

6. S. Aronson. "Some Experiments on Permeation of Hydrogen through Oxide Films on Zirconium." Bettis Technical Review, WAPD-BT-19, 1960 pp. 75-81.

7. M.W. Mallett and W. M. Albrecht. "Terminal Solid Solubility of Hydrogen in Zirconium." J. Electrochem Soc. 1957 vol. 104 p. 142.

8. R.C. Asher and F.W. Trowse. "The Distribution of Hydrogen in Zirconium Alloy Fuel Cladding: The Effects of Heat Flux." J. Nucl. Hater 1970 vol. 35 p. 115.

- 135 - 9. K. Homma, J. Puruta and S. Kawasaki. "Behaviour of the Zircaloy Cladding Tube in Mixed Gas of Hydrogen and Steam." JAERI-71731 (June 1977).

10. D. W. Shannon and R.E. Westerman. "Hydriding and Thermal Re-distribution of Hydrogen in N-Reaetor Zircaloy Process Tubes." HW-77602 RD, May 1963.

FUEL FAILURES

1. J.A.L. Robertson. "Nuclear Fuel Failures Their Causes and Remedies". Proc. CNA/ANS Joint Topical Meeting on Commercial Nuclear Fuel Techology Today. Toronto, Canada. April 1975.

2. B.G. Hersey and H.B. Meieran. "Behaviour of an Intentionally Defect Fuel Rod Which Ruptured During Irradiation." WAPD-TM-628. July 1969.

3. D. Cordall, R.M. Cornell, K.W. Jones aand J.S. Waddington. "Fuel Failures in the Dodewaard BWR." Nucl. Technol. 1977 vol. 34 pp. M38-M48. t. F. Garzarolli, R. Von Jan and H. Stehle "The Main Causes of Fuel Element Failure in Water Cooled Power Reactors.1* Atomic Energy Review 1979 vol. 17 pp. 31 - 128.

5. F. Garzarolli, B. Boos, K. Guse and P. Niehoff "BWR Fuel Experience 2. Correlation of Fuel Performance to Manufacturing Variables". Proc. BNES Conference on Nuclear Fuel Performance, London 1973. Paper 71.

- 136 - 6. T. Hoshi et al. "Fuel Failure Behaviour of Unirradiated Fuel Rods Under Reactivity Initiated Conditions." J. At. Energy Soc. Japan. 1978 vol. 20 p. 651.

OXIDATION

1. R.E. Westerman. "High Temperature Oxidation of Zirconium and Zircaloy 2 in Oxygen and Water Vapour". HW-73511 (April 1962).

2. B. Cox "Some Factors which Affect the Rate of Oxidation and Hydrogen Absorption of Zircaloy 2 in Steam". AERE - R43*»8. Nov. 1963.

3. J.C. Clayton "Corrosion and Hydriding of Irradiated Zircaloy Fuel Rod Cladding" WAPD-TM-1M10 (Sept. 1982).

4. J.C. Clayton and R.J. Fischer "Corrosion and Hydriding of Zircaloy Fuel Rod Cladding in 633K Water and Reactor Environments." Proc ANS Topical Meeting on Light Water Reactor Fuel Performance, Orlando, Florida 1985. CONF 850 U01, vol. 1

5. J.M. Wanklyn. "The Properties of Oxide Films on Zirconium Alloys and Their Relevance to Corrosion and Hydrogen Uptake." Electrochem. Technol. 1966 vol. 4 pp. 81-88.

6. E. Hillner. "Corrosion of Zirconium Base Alloys - An Overview." ASTM -

STP633 1977 PP. 221-235.

- 137 - 7. J.M. Markowitz and J.C. Clayton "Corrosion of Oxide Nuclear Fuels in High Temperature Water". WAPD-TM-909 (1970).

8. J.C. Clayton. "Cladding Corrosion and Hydriding in Irradiated Defected Zircaloy Fuel Rods." WAPD-TM-1393. August 1985.

F.P. RELEASE AND DETECTION

1. K.H. Neeb, W. Schweighofer and R. Wurtz "Experimental Methods for Investigations of Internal Fuel Rod Chemistry of Light Water Reactor Fuels." J. Nucl. Mats. 1981 vol. 97 pp. 165-172.

2. R. Beraha et al. "Fuel Survey in the LWR's Based on the Activity of the Fission Products". Nucl. Technology 1980 vol. 19 p. 126

3. E. Schuster. "Escape of Fission Products from Defective Fuel rods of LWR's." Nucl. Eng and Design 1981 vol. 64 p. 81.

4. G. Kurka, A. Harrer and P. Chenebault "Fission Product Release from a Pressurised Water Reactor Defective Fuel Rod: Effect of Thermal Cycling." Nuel. Technol. 1979 vol. 46 p. 571.

5. J.R. Findlay, F.A. Johnson, J.A. Turnbull and C.A. Friskney. "Fission Product Release From UO. During Irradiation". Proc. IAEA/IWGFPT Specialist Meeting on The Behaviour of Defected Zirconium Alloy Clad Ceramic Fuel in Water Cooled Reactors. Chalk River, Canada. 17-21 Sept. 1979. Paper 1.1.

- 138 - 6. M.B. Hughes and E.J. Chulick "In-Line Fission Product Monitoring." As ref. 5 above. Paper 2.7.

7. G.C. Comley. "Release of Radio-Iodines from Defective Fuel in WSGHWR". Proc. BNES International Conference on Hater Chemistry of Nuclear Reactor Systems, Bournemouth, October 1977, p. **99«

8. W. Miekeley and F.W. Felix "Effect of Stoichiometry on Diffusion of

Xenon in U02" J. Nucl. Mater 1972 vol. 12 p. 297.

GAP CONDUCTANCE

1. M. Bruet and J.P. Stora "Fuel/Clad Heat Transfer Coefficient of a Defected Fuel Rod." Proc. CSNI Specialists Meeting on the Behaviour of Water Reactor Fuel Elements Under Accident Conditions. Spatind, Norway, Sept. 1976.

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