ML20153F437
| ML20153F437 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 05/02/1988 |
| From: | Brey H PUBLIC SERVICE CO. OF COLORADO |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| P-88154, TAC-62026, NUDOCS 8805100329 | |
| Download: ML20153F437 (6) | |
Text
h Public Service ~
l% h 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 May 2, 1988 Fort St. Vrain Unit No. 1 P-88154 U. S. Nuclear Regulatory Comission ATTN:
Document Control Desk Washington, D.C.
20555 Attention: Mr. Jose A. Calvo Director, Project Directorate IV Docket No. 50-267
SUBJECT:
Additional Information on FTE-2 Post Irradiation Examination
REFERENCES:
(1) PSC Letter, Brey to Berkow, dated 7/18/86(D-86468)
(2) PSC Letter, Brey to Berkow, dated 9/11/86 (P-86545)
Dear Mr. Calvo:
In a telephone conversation on March 25, 1988 among the staffs of Public Service Company of Colorado (PSC), General Atomics (GA), and the NRC, the post irradiation examination (PIE) reports for fuel element FTE-2 were discussed. These reports were submitted to the NRC in References 1 and 2.
The NRC requested that PSC submit additional information sumarizing the results of the PIE in the context of the Fort St.
Vrain licensing basis and outstanding regulatory comitments.
The additional information, which sumarizes the matters discussed in the telephone conversation, is enclosed.
If you have any questions or coments regarding this information, please contact Mr,. M.H.
Holmes at (303) 480-6960, i
8805100329 880502 Very trtly yourt, 0gi PDR ADOCK 05000267 P
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P-88154 Page 2 May 2, 1988 Attachment ec: Regional Administrator ATTN: Mr. T.F. Westerman, Chief Projects Section B Mr. R.E. Farrell Senior Resident Inspector Fort St. Vrain I.
I
Attachment to P-88154 Page 1 FTE-2 POST IRRADIATION EXAMINATION ADDITIONAL INFORMATION The results of the destructive post irradiation examination (PIE) of fuel test element FTE-2 were described in References 1 and 2.
The FTE-2 PIE was conducted to fulfill two purposes regarding fuel examination.
The first purpose pertained to the performance of H-451 graphite under irradiation.
Beginning with fuel Segment 9 (Cycle 4) in May
- 1984, the Fort St. Vrain (FSV) fuel elements originally fabricated with H-327 graphite were replaced with elements fabricated with H-451 graphite.
General approval for use of H-451 graphite in FSV reload fuel segments was obtained from the NRC in 1979 as the result of a series of submittals to the NRC by General Atomics (GA). One of the requirements for use of H-451 listed in the NRC Safety Evaluation Report (Reference 3) was that PSC conduct destructive PIE on fuel test elements FTE-2, 4 and 6 to confirm that the performance of the H-451 graphite observed for these elements is consistent with the performance described in the GA submittals.
Examination of these elements was required by the NRC because they were inserted into the core at the beginning of Cycle 2.
- Hence, they serve as lead irradiation assemblies and provide confirmation of graphite performance at intermediate exposures prior to the fuel in Segment 9 reaching those exposures.
The second purpose was to examine one of the 88 fuel rods in FTE-2 that contain FSV reference fuel particles.
(The remaining fuel rods in FTE-2 contain a variety of experimental fuel particles which were, at the time of FTE-2 preparation, being considered for use in future HTGR applications.)
Fuel particle examinations are conducted to confirm that particle coating failures are not excessive, that fuel kernel migration rates are consistent with those assumed in deriving the core themal safety limit, and that no unanticipated coated fuel particle degradation is occurring.
l The results of the PIE reported in References 1 and 2 indicate that the performance of FTE-2 in Cycles 2 and 3 was very good and j
consistent with expectations.
The H-451 graphite and the FSV reference fuel particles were in very good condition.
H-451 Graphite Performance Reference 1 Section 1 (Item 1) notes that the H-451 graphite was in excellent condition with no observed cracks.
The measured tensile strength and Young's moduli were found (Item 3) to be in agreement with the values given in the GA Graphite Design Data Manual.
These j
values are the same as those presented in Reference 4, which i
transmitted to the NRC in one package the GA licensing documentation pertinent to use of H-451 graphite in FSV. Thermal er.pansivities were found (Item 4) to be about 19% lower than expected, but this j
Attachment f
to P-88154 L
Page 2 deviation was noted to be conservative with respect to graphite i
stresses. Subsequent evaluation of the silicon carbide temperature i
monitors in FTE-2 resulted in the conclusion that the differences l
1 between the measured and predicted values of all H-451 graphite l
properties, including themal expansitivity, was less than 10% at 650 degrees C (Reference 2),
i Based on these results, it is concluded that the performance'of the r
H-451 graphite in FTE-2 was consistent with the expected graphite behavior that served as the licensing basis for use of H-451 graphite in FSV reload fuel segments.
Fuel Particle Performance As noted above, fuel particle examinations are coaducted to confirm that particle coating failures are not excessive, that fuel kernel i
migration rates are consistent with those assumed in deriving the
~
1 core thermal safety limit, and that no unanticipated fuel particle i
degradation is occurring.
Fission product source items used as initial conditions in FSV accident analyses are derived by assuming a core average particle failure fraction of 5%, with up to 10% particle j
failure at the end of life of a fuel segment.
Failure is defined as i
a breach of all four particle coatings.
e
]
I Kernel migration is a phenomenon in which carbon in the particle i
buffer coating diffuses from the hot side of the kernel, through the kernel, to the cold side, resulting in migration of the kernel toward the particle coatings.
Excessive kernel migration results in kernel-l 1
coating interaction and coating failure.
Kernel migration rates j
increase with increasing temperature and themal gradient.
The kernel migration behavior of a coated fuel particle is an inherent characteristic of the fuel kernel material.
The FSV Reactor Core Safety Limit (Technical Specification SL 3.1) is derived assuming that during nomal reactor operation a kernel may migrate up to 20 microns ssee Updated FSAR Section 3.6.8).
Hence, particle i
I examinations are conducted to ensure that kernel migration in excess
{
of this amount has not occurred.
The smallest kernel migration i
distance observable using metallographic techniques is 3 to 4 i
1 microns.
Therefore, an observation of no kernel migration implies i
migration of less than 4 microns. Kernel migration is discussed in Section 3.6.7.6.2 of the Updated FSAR.
a l
With regard to the perfomance of the FSV reference fuel particles in 1
FTE-2, Reference 1. Section 1 notes (Item 9) that the reference fuel i
was in excellent condition. Small percentage failures (1.1% or less) for individual coating layers were observed (Item 10), but there were i
no cases of total coating failure.
This perfomance is better than j
the 5% core average failure assumed in the FSV licensing basis.
l No kernel migration was observed (Item 11) in the fuel particles.
Thus, kernel migration was less than 4 microns during the two cycles j
that FTE-2 resided in the core.
Based on linear extrapolation.
l
Attachment to P-88154 Page 3 kernel migration over six cycles of residency would be expected to be less than 12 microns, which is well within the assumptions made in deriving the core thermal safety limit.
Linear extrapolation is conservative, since fuel temperatures generally tend to be higher in earlier cycles than in later cycles.
Limited chemical interaction.was observed (Items 11 and 12) between the fuel kernels and some of the coatings in four of the 459 particles examined.
This int.iction occurs as a result of dispersion of the kernel material into the particle coatings during fuel manufacture.
Kernel dispersion occurs as a result of a reaction between the kernels and hydrogen chloride gas generated during the sic coating process.
Control over the coating process minimizes this effect, and the fuel product specifications limit the amount of dispersion allowed in the final product.
Accordingly, this phenomenon is not as limiting as kernel migration and is therefore not reflected in the core thermal safety limit.
Kernel-coating interaction resulting from kernel dispersion has been seen in a previous PIE and reported to the NRC (Reference 5).
_A larger amount of kernel-coating interaction was observed in Segment 2 fuel removed at the end of Cycle 2.
However, QC records indicate that Segment 2 fuel exhibited kernel dispersion at the upper end of the range allowed by the fuel product specifications.
It was, accordingly, anticipated that follow-on PIE inspections of fuel produced for later segments would show a reduction in kernel-coating interaction.
This anticipation has been confirmed by the results of the PIE of FTE-2.
Based on these results, it is concluded that the performance of the FSV reference fuel particles in FTE-2 is very good and is conservative with respect to the fuel performance assumed in developing the licensing basis of Fort St. Vrain.
Referenccs e
1.
PSC letter, Brey to Berkow, dated 7/18/86 (P-86468);
Subject:
"Destructive Examination of Fort St.
Vrain Fuel Test Element FTE2," GA Document 908909, Issue N/C, 7/11/86, 2.
PSC letter, Brey to Berkow, dated 9/11/86 (P-86545);
Subject:
"Thermal Diffusivity of H-451 Graphite and Temperature Monitor Analyses in FSV Fuel Test Element FTE-2," GA Document 908928 Issue N/C, 8/22/86.
3.
NRC letter, Ganmill to C. Fisher (GAC), dated 5/21/79 (G-79238);
Subject:
"Evaluation of Topical Report GLP-5588."
4.
PSC letter, Lee to Collins, dated 12/2/83 (P-83391);
Subject:
"Technical Specification Amendment Request Regarding Fort St.
Vrain Use of H-451 Graphite."
5.
PSC letter, Brey to Collins, dated 1/3/84 (P-84001);
Subject:
"Meta 11ographic Examination of a Fuel Rod from Segment 2 FSV Fuel Element 1-2415," (GA Document 906968 Issue A, 11/9/83),
1 4
.