ML19220A884
| ML19220A884 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/29/1977 |
| From: | Meyer R Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904250062 | |
| Download: ML19220A884 (9) | |
Text
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DEC 2 9 1377 MEMORANDUM FOR:
D.B. Vassallo, Assistant Director for Light Water Reactors, DPM THRU:
P.S. Check, Chief, Core Performance Branch, DSS FROM:
R.O. Meyer, Leader, Reactor Fuels Section, CPL, DSS
SUBJECT:
SER SUPPLEMINT CONCERHING BURNABLE POISON MATERIALS Plant Name:
Three Mile Island, Unit 2 Docket number:
05000320 Milestone Nunber:
24-24 Licensing Stage:
OL Responsible Branch LWR-4 and Proj ect Manager:
Systems Safety Branch Involved:. E.
Silver Core Performanca Branch Description of Review:
, SER Supplement Input Requested coupletion Date:
December 16, 1977 2eview Status:
Couplata The Reactor Fuela Section of the Core Performance Branch has prepared the attached supplement to the TMI-2 SER.
The supple-describes our review of the Gd 0 -UO ment 23 2 demonstration. fuel rods and 3 C-C hurnable poison rods to be irradiated during 4
cycle 1.
We conclude that the propossd irradiation of a few test rods poses no Jafe ty.. concern f or 1.'tI-1.
We thus approve the irce.diation testa subj ect to the coniition that surveil-Lance and post-irradiation examinations ha performed and reported to the NRC.
We nota in. passing that B&W would be required 40 submit a more comprehensive safety analysis.for these materials before including videscala.uses in future applications.
a Ralph 0. Meyer, Leader
. Reactor Fuels Section Core Performance Branch Division of Systems Safety Enclosure.
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and 3 C-Graphite 3urnable ?oison Rods Gd20 3-UO3 4
Materials, Mechanical and Thermal Design Evaluation Descriotion Two new burnable poison materials are proposed for limited use in test rods to be irradiated in TMI-2, Cycle 1.
These burnable poisons, while new to 35W pressurized water reactors, have seen extensive use in other thermal reactor designs.
Specifically, 3 C-C (graphite) burnable poison rods are used 4
in high temperature gas cooled (HTGR) reactors of General Atomic fuel r ds have design, such as Fort St. Vrain, and Gd 03-UO2 2
been used in BWRs for several years and have also seen limited use in PWRs as well.
Eight Zircaloy-clad burnable poison rods (BPRs) using 3 C 4
in graphite matrix (3 C-C) pellets will be employed in place 4
C pellets.
The boronated of eight standard BPRs using Al2 03-34 graphite (3 C-C) consists of 3 C particles dispersed uniformly 4
4 in a pelletized graphite matrix.
Approximately 9 7 *. of the pellet is matrix material.
Nominal dimensions of the BRPs are provided C 3RPs shows the dimensions in Table I.
Comparison with A1 0 3-3 4 3
to be identical.
f"*1 Four fuel assemblies, each containing four Gd 0 3-UO 2 2
rods (total of 16) are also to be irradiated in TMI-2, Cycle 1.
Enrichment in the gadolinia-bearing rods was reduced from 1.93 to 1.80" (by weight) U-235, to compensate for a reduction in 0 -UO; pellet thermal conductivity relative to pure U03 Gd 3 3
The Gd ;0 3-UO ; fuel pellet dimensions, total colann length, stack weight, cladding material and dimensions, grids and end r z 4 o; s m t.1
+#
, fittings are identical to those in non-gadolinia bearing MK-34 fuel assemblies.
Desien Evaluation 3,C-C 3 prs 4
The principal performance considerations of concern for 34C-C involve irradiation-induced swelling, gas release, and compati-bility with cladding and coolant.
Measurements of irradiation-induced swelling of 3 C-graphite at General Atomic (Ref. 1) 4 indicated that the dimensional change in baronated graphite is controlled by fast neutron damage to the graphitic matrix and B-10 fission product damage in the graphite binder matrix; not by swelling of 3 C during irradiation.
The curve used by B&W for 4
their 3 C-C swelling model is based on data from irradiation of 4
20-30% (by weight) 3 C in graphite (Ref. 2), whereas %3% (by 3
weight) 3,C in graphite will be used in TMI-2.
The lower 3,C 4
4 content results in a somewhat lower expected swelling contribution from 3-10 fission fragment damage.
Hence, the use of the swelling curve for 20-30". (by weight) 3 C in graphite to predict the 3
swelling of the 3 C-C BPRs, is, in this sense, conservative.
Like 4
the pellets irradiated in references 1 and 2,
the 3&W pellets for the demonstration BPRs are fabricated by extrusion.
This pro-duces an anisotropy in the crystallographic orientation of the graphite matrix.
The degree of anisotropy is important because it affects the swelling behavior.
Purchase specifications on
,6 b
- the 3;C-C pellets limit the anisotropy to values below those used to develop the swelling curve in the General Atomic tests.
C-C B&W calculations yield tl% cladding strain (resulting from B4 swel_ing) at end of cycle-l.
This is well within their 3% design limit, which is based on thermal-hydraulic calculations of the amount of coolant flow reduction required to cause slug-flow boiling.
A1 0 -34C BPR, the 34C-C BPR accumulates helium Like the 2 3 and lithiu= daughter products as a result of the absorbtion of thernal neutrons via the B-10 (n, 2) Li-7 reaction.
In its model for gas release, B&W assumes that all de helium produced is released from the B4C-C co= pact to the available plenum volume.
B&W's calculation of the resultant BPR pressure shows that the internal rod pressure is always below system pressure, even for the most limiting moderate frequency transient.
This satisfies a design criterion for burnable poison rod pressure by preventing a condition under which ballooning of the cladding could occur.
In terms of pellet / cladding chemical compatibility, the main concerns involve the poteatial for carburization and hydriding of the Zircaloy, since fornation of :irconiu= carbide or hydride embrittles the cladding and increases the likelihood of cladding breach.
From thermal-hydraulic design calculations, using the most conservative values for irradiated thermal conductivity, thermal expansivity, and pellet and cladding tolerances, the maximum pellet surface te=perature and cladding inner surface temperatures were 314 and 663*F, respectively.
Since the W.? 4 9
-4, minimum surface temperature for carbide formation is approximately 1450*F (Ref. 3), there is ample margin to preclude carbide forma-tion.
With respect to hydriding, 35's uses the same fabrication procedure, including the same drying process, as that used in 34 C-alumina rods.
In this procedure, the rods are dried after loading, just prior to sealing, in the belief that removal of trace amounts of moisture reduces the available source of hydrogen required for hydriding to occur.
Hydriding and resultant cladding perforation has been identified as having led to a power ancmaly (Ref. 4),
which occurred when the 34C in some 34C-Al2 03 burnable poison rods was leached out by the primary coolant.
By a similar process, if the 34C-C BPR Zircaloy cladding were perforated, either d te to hydriding or some other process, the B4C would be rapidly lost through a two-step chemical reaction involving (1) the reaction O to form 3203, followed by (2) formation of boric of 34C with H2 acid by contact of the B 203 with additional water.
However, even if this were to occur so that all the 3-10 were to be dissipated, the major result would be an increased power peaking in the surrounding rods in the fuel assembly.
Analyses have been per-formed which show that neither at 30L when the poison worth is greatest nor later in life when the poison worth is lower but assembly power is greater, is there inadequate margin between the accident conditions and the peaking limits.
In summation, the major performance considerations for 34C-C, involving irradiation-induced swelling, gas release, and chemical 6
8. #,
.s o compatibility, have been addressed.
We thus b e ' '. a v e there is reasonable assurance that cladding integrity wi.1 be maintained throughout cycle 1 operation.
However, even if all the 3-10 in the BPRs were rapidly removed via primary coolant ingress through cladding perforations, the neutronic ef:tets would be unimportant because the number of BPRs (3) is sttall and they are well-spaced in one 3PR spider.
Therefore, we conclude that no significant safety concern exists regarding the proposed irradiation of the eight 34C-C test 3 prs.
TA3LE 1 BPR Descriotions
- Cladding Standard Al 0 3-34C 34C-C 2
3RP 3RP Zircaloy-4 OD 0.430 in.
0.430 in.
1D 0.362 0.362 in.
Min. Wall Thickness 0.0325 in.
0.0325 in.
Pellets OD 0.340 in.
0.340 in.
Stack Length 126 in.
126 in.
- nominal dimensions are given O
t.1
.s e, 4.
r
- Gd 0 3-UO2 Fuel R;ds The principal performance concerns for Gd 0 -UO 1""
1"*
3 3 2
the effects of the gadolinia additions on =aterial's properties such as thermal conductivity and irradiation-induced densifica-tion.
To compensate for a decrease in Gd 0 -UO P*11*'
- h*#'*1 3 3 2
conductivity (relative to pure UO 2), the gadolinia-bearing rod enrichment was reduced from 1.98 to 1.80" (by weight) U-235.
With this lower enrichment, 3&W has stated that the maxi =um gadolinia pin power remains less than 81 *. of the peak pin power in the core at all times during the cycle.
3&W has also stated that there is no discernable difference between UO and Gd 0 -UO densification.
Recent experimental 3
2 3 3
evidence (Ref. 5) suggests, however, that Gd2 03-UO3 rods will densify more in reactor than pure UO3 This tendency is further exaggerated, GE reports, by changes in fuel cracking and relo ation behavior.
Thus, whereas the Gd 0 -UO r ds will 3 3 2
have a lower linear heat generation rate (LHGR) than pure UO 2' the enhanced densification of the gadolinia-bearing rods, couplad with their reduced thermal conductivity,will have the effect of increasin pellet temperatures.
Therefore, despite the sub-e stantial reduction in rod power, offsettir-effects cause stored energy in Gd 0,-UO fuel to be about the same as in UO fuel.
3 ;
3 3
Nevertheless, taking into account the lower enrichment of the demonstration Gd 0 -UO ds and t 'n e i r location in the core, 2 3 2
the hi-power CO rods are probably more limiting.
[#
(.1
.s t-
O 9
1-relatively'few in number and In summation, because they are are not located in peak power assemblies, we conclude that no significant safety concern exists regarding the proposed irradia-tion of Gd;0 3-UC 2 demonstration fuel rods and that there is reason-able assurance that cladding integrity will be maintained through-out cycle 1 operation.
Summarv There is no appraent safety concern regarding the proposed irradiation of 3 C-C and Gd2 03-UO; test rods.
Furthermore, we 4
encourage irradiation testing of fuel system design inovations prior to their introduction into widescale use.
However, to reap the potential benefits to future plant safety analyses, observations of the performance of the test rods must be made.
None has been proposed to the NRC.
The irradiation of the test rods in TMI-2 is thus approved on the condition that (a) aurveillance and post-irradiation examinations (PIE) be performed, including destructive PIE if rod perforations occur, and (b) results are reported to the NRC.
REFERENCES 1.
0.M.
Stansfield, " Irradiation Induced Dimensional Change in HTGR Contrel Materials," Gulf General Atomic Report GA-12035, April 23, 1972.
2.
0.M.
Stansfield, " Neutron Irradiation Effects in Boronated Graphite, Hafnated Graphite, 3;C, and HfC - Summary Report on the SG-1 and 3G-2 Experiments," G t. l f General Atomic Report GA-10643, June 13, 1971.
3.
3.
Lustman and F.
- Kerze, Jr.,
The Meta 11urev of Circonium, McGraw-Hill, 1955, p.
449.
" Report to Congress on Abnormal Occurrences," NRC Report, 4
NUREG-0090-3, March, 1977.
5.
G.A.
Potts, " General Electric Densification Program Status,"
Ceneral Electric Company Proprietarv Reoort NEDE-21232-P (N'n-Pr'prieta'ry Version, NEDO-2128z)
Rev.
1, April 1977.
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