Boiling water reactor fuel rod

An improved nuclear fuel rod includes a plurality of cylindrical nuclear fuel pellets being disposed end-to-end in a stack, and an elongated cylindrical cladding tube providing a hermetically sealed chamber. The cladding tube includes a cylindrical wall having inner and outer spaced surfaces and a thickness "x". The stack of fuel pellets are contained in the chamber and spaced radially inwardly from the wall. The thickness "x" of the cladding tube wall between the inner and outer surfaces is the thickness required to generate sufficient heat internally of the wall between its inner and outer surfaces to satisfy the following relationship of a predetermined total fuel rod radiation intensity output, I, to a predetermined fuel rod radiation intensity generated by the fuel pellets contained in the fuel rod, I.sub.0 :I=I.sub.0 e.sup.-ux, where "u" is the attenuation coefficient which varies with cladding material and radiation type.

Reference is hereby made to the following copending application dealing 
with related subject matter and assigned to the assignee of the present 
invention: "Boiling Water Nuclear Reactor Fuel Assembly" by Rusi 
Taleyarkhan, assigned U.S. Ser. No. 729,602 and filed May 2, 1985 (W.E. 
52,509) now U.S. Pat. No. 4,678,631. 
BACKGROUND OF THE INVENTION 
Field of the Invention 
The present invention relates generally to fuel assemblies for a nuclear 
reactor and, more particularly, is concerned with an improved fuel rod for 
a boiling water reactor (BWR) fuel assembly which implements criterion for 
enhanced transient response and system stability. 
Description of the Prior Art 
Typically, large amounts of energy are released through nuclear fission in 
a nuclear reactor with the energy being dissipated as heat in the 
elongated fuel elements or rods of the reactor. The heat is commonly 
removed by passing a coolant in the heat exchange relation to the fuel 
rods so that the heat can then be extracted from the coolant to perform 
useful work. 
In nuclear reactors generally, a plurality of the fuel rods are grouped 
together to form a fuel assembly. A number of such fuel assemblies are 
typically arranged in a matrix to form a nuclear reactor core capable of 
self-sustained, nuclear fission reaction. The core is submerged in a 
flowing liquid, such as light water, that serves as the coolant for 
removing heat from the fuel rods and a neutron moderator. 
A typical fuel rod is composed of an elongated cladding tube having end 
closures or plugs attached to opposite ends of the tube so as to provide a 
hermetically sealed chamber in which a plurality of nuclear fuel pellets 
are disposed end-to-end in a stack. Both the tube and pellets are 
preferably circular in cross-section and the pellets have a length 
approximately twice their diameter. The diameter of a pellet is slightly 
less than that of the tube so that a clearance space or gap is formed 
therebetween to accommodate any swelling of the pellet during operation of 
the reactor. 
The trend in fuel rod design in all types of nuclear reactors has been to 
make the thickness of the fuel rod cladding tube as thin as possible for 
various reasons. For example, in U.S. Pat. No. 3,376,201 to Bain, it is 
mentioned that the protective sheath (or cladding tube) will, undesirably, 
absorb neutrons and must be made as thin as possible consistent with the 
aim of preventing distortion of the fuel element (or rod). With respect to 
a BWR, such trend overlooks the profound impact that fuel rod design has 
on nuclear coupled/decoupled thermal hydraulic transients which take place 
in the reactor. A nuclear coupled transient is one where the thermal 
hydraulic behavior is coupled to neutronic feedback due to the void 
reactivity feedback mechanism, whereas, a decoupled transient implies one 
where this void reactivity feedback is neglected. These transients refer 
to fuel rod responses to key pertubations, for example, neutron flux/power 
response to pertubations in the flow rate system pressure, subcooling, 
etc. 
Consequently, a need exists for redirection of attention to gaining a 
systematic understanding of the effect of fuel rod dynamics on BWR system 
transient/stability response, with an eye toward modification of the 
conventional fuel rod design to improve such response. 
SUMMARY OF THE INVENTION 
The present invention provides an improvement which is designed to satisfy 
the aforementioned needs. Underlying the present invention is the 
discovery that through only a minor change in fuel rod design, a profound 
improvement in BWR performance is achieved. Specifically, what has been 
uncovered is that making a minor modification in the thickness of the 
cladding tube to effectuate marginally higher (than what existed 
heretofore) internal heat generation would result in significantly 
improved power margins and transient/stability characteristics. Power 
margin is defined as the amount of bundle power (energy per unit time 
generated within fuel rods in a bundle) increase/decrease to arrive at the 
same transient response as for a base case from which a change was made, 
while transient/stability response merely indicates how a certain 
(perturbed) parameter will vary as a function of time. 
Accordingly, the present invention relates to an improvement set forth in a 
fuel rod for a nuclear fuel assembly containing fuel therein. The 
improvement comprises a cladding tube having a wall of a thickness "x" 
which is the thickness required to generate sufficient heat internally in 
the wall to satisfy the following (well known Beer's law for radiation 
attenuation) relationship of a predetermined fuel rod radiation intensity 
output, I, to a predetermined fuel rod radiation intensity generated by 
the fuel contained in the fuel rod, I.sub.0 :I=I.sub.0 e.sup.-ux ; where 
"u" is the attenuation coefficient which vaires with cladding material and 
radiation type. The attenuation coefficient "u" determines how much of the 
incident radiation, I.sub.0, will be allowed to "come out" (I); ie., 
I.sub.0 -I=amount deposited. As is clear from the Cases I-IV presented in 
the table later on, the attenuation coefficient is a constant for any 
given material. 
More particularly, an improved nuclear fuel rod, comprises (a) a plurality 
of cylindrical nuclear fuel pellets being disposed end-to-end in a stack; 
and (b) an elongated cylindrical cladding tube providing a hermetically 
sealed chamber and having a cylindrical wall of a thickness "x", with the 
stack of fuel pellets being contained in the chamber and spaced radially 
inwardly from the wall. The thickness "x" of the cladding tube wall 
between inner and outer surfaces thereof being the thickness required to 
generate sufficient heat internally of the wall between its inner and 
outer surfaces to satisfy the following relationship of a predetermined 
total fuel rod radiation intensity output, I, to a predetermined fuel rod 
radiation intensity generated by the fuel pellets contained in the fuel 
rod, I.sub.0 :I=I.sub.0 e.sup.-ux. 
These and other advantages and attainments of the present invention will 
become apparent to those skilled in the art upon a reading of the 
following detailed description when taken in conjunction with the drawings 
wherein there is shown and described an illustrative embodiment of the 
invention.

DETAILED DESCRIPTION OF THE INVENTION 
In the following description, like reference characters designate like or 
corresponding parts throughout the several views of the drawings. Also in 
the following description, it is to be understood that such terms as 
"forward", "rearward", "left", "right", "upwardly", "downwardly", and the 
like are words of convenience and are not to be construed as limiting 
terms. 
In General 
Referring now to the drawings, and particularly to FIGS. 1 to 3, there is 
shown a nuclear fuel assembly, generally designated 10, for a boiling 
water nuclear power reactor (BWR), in which the improvement of the present 
invention is incorporated. The fuel assembly 10 includes an elongated 
outer tubular flow channel 12 that extends along substantially the entire 
length of the fuel assembly 10 and interconnects an upper support fixture 
or top nozzle 14 with a lower base or bottom nozzle 16. The bottom nozzle 
16 which serves as an inlet for coolant flow into the outer channel 12 of 
the fuel assembly 10 includes a plurality of legs 18 for guiding the 
bottom nozzle 16 and the fuel assembly 10 into a reactor core support 
plate (not shown) or into fuel storage racks, for example in a spent fuel 
pool. 
The outer flow channel 12 generally of rectangular cross-section is made up 
of four interconnected vertical walls 20 each being displaced about ninety 
degrees one from the next. Formed in the spaced apart relationship in, and 
extending in a vertical row at the central location along, the inner 
surface of each wall 20 of the outer flow channel 12, is a plurality of 
structural ribs 22. The outer flow channel 12, and thus the ribs 22 formed 
therein, are preferably formed from a metal material, such as an alloy of 
zirconium, commonly referred to as Zircaloy. Above the upper ends of the 
structural ribs 22, a plurality of upwardly-extending attachment studs 24 
fixed on the walls 20 of the outer flow channel 12 are used to 
interconnect the top nozzle 14 to the channel 12. 
For improved neutron moderation and economy, a hollow water cross, 
generally designated 26, extends axially through the outer channel 12 so 
as to provide an open inner channel 28 for subcooled moderator flow 
through the fuel assembly 10 and to divide the fuel assembly into four, 
separate, elongated compartments 30. The water cross 26 has a plurality of 
four radical panels 32 composed by a plurality of four, elongated, 
generally L-shaped, metal angles or sheet members 34 that extend generally 
along the entire length of the channel 12 and are interconnected and 
spaced apart by a series of elements in the form of dimples 36 formed in 
the sheet menbers 34 of each panel 32 and extending therebetween. The 
dimples 36 are formed in and disposed in a vertical column along the axial 
length of the sheet members 34. Preferably, the dimples 36 in each of the 
sheet members 34 are laterally and vertically aligned with corresponding 
dimples 36 in adjacent sheet members 34 in order to provide pairs of 
opposed dimples that contact each other along the lengths of the sheet 
members to maintain the facing portions of the members in proper 
spaced-apart relationship. The pairs of contacting dimples 36 are 
connected together such as by welding to ensure that the spacing between 
the sheet members forming the panels 32 of the central water cross 26 is 
accurately maintained. 
The hollow water cross 26 is mounted to the angularly-displaced walls 20 of 
the outer channel 12. Preferably, the outer, elongated lateral ends of the 
panels 32 of the water cross 26 are connected such as by welding to the 
structural ribs 22 along the lengths thereof in order to securely retain 
the water cross 26 in the desired central position within the fuel 
assembly 10. Further, the inner ends of the panels together with the outer 
ends there of define the inner central cruciform channel 28 which extends 
the axial length of the hollow water cross 26. 
Disposed within the channel 12 is a bundle of fuel rods 40 which, in the 
illustrated embodiment, number sixty-four and form an 8.times.8 array. The 
fuel rod bundle is, in turn, separated into mini-bundles thereof by the 
water cross 26. The fuel rods 40 of each mini-bundle, such being sixteen 
in number in a 4.times.4 array, extend in laterally spaced apart 
relationship between an upper tie plate 42 and a lower tie plate 44 and 
connected together with the tie plates comprises a separate fuel 
subassembly 46 within each of the compartments 30 of the channel 12. A 
plurality of grids 48 axially spaced along the fuel rods 40 of each fuel 
rod subassembly 46 maintain the fuel rods 40 in their laterally spaced 
relationships. Coolant flow paths and flow communication are provided 
between the fuel rod subassemblies 46 in the respective separate 
compartments 30 of the fuel assembly 10 by a plurality of openings 50 
formed between each of the structural ribs 22 along the lengths thereof. 
Coolant flow through the openings 50 serves to equalize the hydraulic 
pressure between the four separate compartments 30, thereby minimizing the 
possibility of thermal hydrodynamic instability between the separate fuel 
rod subassemblies 46. 
The above-described basic components of the BWR fuel assembly 10 are known 
in the prior art, being disclosed particularly in the above 
cross-referenced patent application, and have been discussed in sufficient 
detail herein to enable on skilled in the art to understand the 
improvement of the present invention presented hereinafter. For a more 
detailed description of the construction of the BWR fuel assembly, 
attention is directed to the above-referenced application. 
Improved BWR Nuclear Fuel Rod 
The improvement of the present invention derives from work undertaken in 
order to understand fuel rod dynamics, particularly the effect of 
distributed internal heat generation fluctuation resulting from void 
reactivity feedback in a BWR. Void reactivity feedback refers to the 
change in reactivity (i.e. neutron thermalization rate) for a given change 
in coolant void fraction. That is, for lower void fraction increased 
neutron thermalization occurs, leading to increased power generated and 
vice versa. It was found that different fuel rod designs affect BWR 
system/fuel rod transient/stability characteristics. It should be noted 
that improved system transient/stability response, especially in BWR's, 
leads to improved thermal margins, lesser mechanical/structural vibrations 
and finally improved reactor control. 
Before describing the fuel rod design modification comprising the 
improvement of the present invention herein, an understanding of the basic 
parameter for evaluating transient/stability response, that being the 
so-called stability margin, is necessary. Briefly, the greater the value 
of the stability margin, the better the transient/stability performance. 
This implies that any modification that increases the value of the 
stability margin will lead to a faster decay of perturbations (flow, 
power, etc.) leading to lesser structural vibrations, better 
thermal/stress related margins and better control. 
FIG. 4 shows a standard fuel rod design in cross-section. It consists of 
two concentric elements. The central cylinder is the stack of fuel pellets 
52 and the outer ring is the cladding tube 54 of the fuel rod 40. The 
elongated cylindrical tube 54 provides a hermetically sealed chamber 56 
and includes a cylindrical wall 58 having inner and outer surfaces 60,62 
spaced by a thickness "x" determined in accordance with the improvement of 
the present invention, as explained below. The stack of fuel pellets 52, 
commonly composed of uranium dioxide, are contained in the chamber 56 and 
spaced radially inwardly from the wall 58 so as to leave a gap 64 
therebetween. 
In work leading to the improvement of the present invention it has been 
found that for a nuclear coupled stability/transient response, the amount 
of internal heat generation in the cladding tube wall 58 (which bears 
proportionally to its thickness, material density, gamma attenuation 
coefficient, etc.) has a pronounced effect. This is clearly demonstrated 
from FIGS. 5 and 6. FIG. 5 shows that including internal heat generation 
in the cladding wall 58 increases the stability margin since the energy 
content is directly proportional to the radiation intensity. The lesser 
the value of I/I.sub.0, the greater is the amount of energy deposited in 
the cladding for a given thickness "x". The y-axis in FIG. 5 represents 
the amount of energy generated/deposited in the cladding. The physical 
meaning behind this observation is demonstrated in FIG. 6 which shows 
nominal bundle power (energy per unit time generated within fuel rods 
within a bundle) as a function of stability margin. The stability margin 
is the distance from the origin at which the Nyquist locus intersects the 
real axis, the greater this distance the faster any pertubations will 
decay. As an example, consider the situation where one percent of the 
total fuel rod power is deposited/generated in the cladding tube wall. 
From FIG. 5, this indicates a shift of the stability margin from 348 to 
350. Again from FIG. 6, it will be noticed that this shift would 
correspond to a bundle power margin of approximately four to five percent. 
That is, the transient response with one percent power deposited in the 
cladding wall 58 at ninety-five percent power would be approximately the 
same as that with approximately zero percent power in the cladding wall at 
one hundred percent bundle power. This is a sizable gain since it is well 
known that power is a destabilizing parameter. 
The improvement in transient response depends on the amount of bundle power 
deposited in the cladding. This depends on the material attenuation 
coefficient and its thickness. A study of the actual improvement in system 
transient response (expressed as % Nominal Power) was conducted. From the 
four different cases considered, as outlined in the table below, it can be 
seen that the power margin (% Nominal Power) can vary from about 4.5% to 
more than 20%. 
TABLE 
______________________________________ 
Atten- 
Total 
Clad Clad tuation 
Power Change Power 
Mate- Thickness Coeffi- 
Depo- Stability 
Margin 
Case rial (cm) cient sited % 
Margin % 
______________________________________ 
I Zr 7.366 .times. 10.sup.-2 
0.1689 
0.8 1.8 4.5 
II Zr 0.1413 0.1689 
1.7 2.8 9.5 
III Fe 7.366 .times. 10.sup.-2 
0.4677 
2.2 3.3 12 
IV U 7.366 .times. 10.sup.-2 
1.4160 
6.7 6.5 22 
______________________________________ 
Case I: represents a nominal case for a given 
Westinghouse QUAD+ bundle design. 
Case II: illustrates that an increase in cladding 
thickness increases attentuation and hence 
energy deposition. Material of clad is 
still Zr. 
Case III: same as Case I, but material of cladding 
changed to steel (iron assumed as principal 
constituent). Due to increased attenuation 
coefficient, power generation in cladding 
increases, giving more power margin. 
Case IV: represents a situation wherein the cladding 
material is of very high attenuation coeffi- 
cient (eg. uranium is dispersed into the 
clad). 
______________________________________ 
In summary, it can be said that due to exponential (e.g. gamma) 
attenuation, marginal increments in cladding wall thickness would tend to 
substantially increase internal heat generation therein. Hence, an 
increase in wall thickness is suggested. Keeping the same outer diameter 
of the wall 58, a thickness increase means a lowering of pellet diameter, 
simultaneously increasing pellet density/enrichment. What the thickness 
"x" of the cladding tube wall 58 between the inner and outer surfaces 
60,62 should be is that thickness required to generate sufficient heat 
internally of the wall to satisfy the following relationship of a 
predetermined total fuel rod radiation intensity output, I, to a 
predetermined fuel rod radiation intensity generated by the fuel pellets 
52 contained in the fuel rod, I.sub.0 :I=I.sub.0 e.sup.-ux. Since 
different power plants are limited differently from transient standpoints, 
the exact value of "x" cannot be predetermined to give an optimized 
stability/transient response. However, for specific power margin 
requirements FIGS. 5 & 6 and the equation I=I.sub.0 e.sup.-ux would be 
used simultaneously. Also from the Cases presented above, it can be said 
that, in addition to change in thickness of the cladding wall, change in 
the material composition of the cladding wall such as incorporation of 
fissionable material into the cladding wall can increase cladding heat 
generation. 
The improved transient response of the type indicated above have several 
major benefits, since most reactor plants are usually limited by 
transients. Improved fuel/fluid coupled transient response characteristics 
would lead to improved CPR margins (i.e., better rod cooling), less fuel 
loading upon load changes, LOCA benefits, etc. 
It is thought that the invention and many of its attendant advantages will 
be understood from the foregoing description and it will be apparent that 
various changes may be made in the form, construction and arrangement 
thereof without departing from the spirit and scope of the invention or 
sacrificing all of its material advantages, the form hereinbefore 
described being merely a preferred or exemplary embodiment thereof.