ML20033H263
| ML20033H263 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/02/1990 |
| From: | Poslusny C Office of Nuclear Reactor Regulation |
| To: | Carns N ARKANSAS POWER & LIGHT CO. |
| References | |
| TAC-74139, NUDOCS 9004190054 | |
| Download: ML20033H263 (7) | |
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. April 2, 1990
- Docket No. 50 368 l
Mr..Neil S. Carns Vice President, Nuclear Arkansas Power.and Light Company P. O. Box 551 Little. Rock, Arkansas 72203
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW 0F CEN.386-P(TACNO.74139)
On July 20, 1989, Arkansas Power and Light Company submitted to the NRC a request to review Combustion Engineering report CEN.386-P.which supports the raising of the fuel pin burnup limit to 60 MWD /Kg for Arkansas Nuclear One, Unit 2.
The staff is currently reviewing the report and has determined that-additional information is required in order to complete its evaluation.
Enclosed is a set of 7 questions related'to information included in the topical report. Please provide your response within 45 days to facilitate the completion of the staff effort.
The reporting and/or recordkeeping requirements contained in this-letter-affect fewer than ten respondents; therefore, OMB cicarance is not required under P.L. 96-11.
Sincerely,
/s/
Chester Poslusny, Jr., Project Manager Project Directorate-IV Division of Reactor Projects
- III, IV, Y and Special Projects Office of Nuclear: Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page l
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b April 2, 1990 4
Docket No. 50-368 Mr. Neil S. Carns Vice President, Nuclear Arkansas Power and Light Company P. O. Box 551 Little Rock, Arkansas 72203 J
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF CEN-386-P(TACNO.74139)
On July 20, 1989, Arkansas Power and Light Company submitted to'the NRC a request to review Combustion Engineering report CEN.386-P which supports the raising of the fuel pin burnup limit to 60 MWD /Kg for Arkensas Nuclear One, Unit 2.
The staff is currently reviewing the report and has determined.that additional information is required in order to complete its evaluation.
Enclosed is a set of 7 questions related to information included in the topical report.
Please provide your response within 45 days to facilitate the completion of the staff effort.
The reporting and/or recordkeeping requirements' contained.in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L. 96 11.
Sincerely,
/s/
Chester Poslusny, Jr., Project Manager
+
Project Directorate IV Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTION Docket File NRC PDR Local PDR PD4 Reading G. Holahan MS 13E4 F. Hebdon P. Noonan C. Poslusn OGC-Rockville E. Jordan MS MNBB 3701 ACRS(10)yMS P-315 PD4 Plant F.ile R. Jones J)
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WASHINGTON, D. C. 20056
' April 2, 1990 1
(
Docket No. 50-368 -
i i
Mr.'Neil S. Carns.
Vice President,- Nuclear Arkansas _ Power and Light Company
(
P. O. Box 551
?
Little Rock, Arkansas 72203
SUBJECT:
REQUEST'FOR ADDITIONAL INFORMATION FOR THE REVIEW OF CEN-386-P(TACNO.74139)-
On July 20. 1989, Arkansas Power and Light Company: submitted to the NRC a request.to review Combustion Engineering report CEN-386-P which supports the I
raising.of the fuel-pin burnup limit to 60 MWD /Kg for: Arkansas Nuclear One, Unit 2.E The staff is currently reviewing the report and has determined that-additional information is required in order to complete its evaluation.-
Enclosed is a set of-7 questions related to information included:in the topical--
~
s report.- Please provide your response within 45, days to-facilitate the completion of'the staff effort.
The reporting and/or recordkeeping. requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance-is not required:
under P.L. 96-11.
Sincerely, to hit Chester.Poslusny, Jr.,-Project Manager-Project Directorate.IV q
' Division of Reactor Projects - III, IV, Y and Special Projectst
-.i Office of Nuclear Reactor Regulation
Enclosure:
l As stated cc w/ enclosure:
See next page
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CC:
Mr. Early Ewing, General Manager Mr. Charles B. Brinkman, Manager Technical Support and Assessment Washington Nuclear Operations Arkansas Nuclear One Combustion Engineering, Inc.
t P. O. Box 608 12300 Twinbrook Parkway, Suite 330 Russellville, Arkansas 72801 Rockville, Maryland 20852 i'
Mr. Neil-Carns, Director Nuclear Operations Honorable Joe W. Phillips Arkansas Nuclear One County' Judge of Pope County
-1 P. O. Box 608 Pope County Courthouse Russellville, Arkansas 72801 Russellville, Arkansas 72801 Mr. Nicholas S. Reynolds Bishop, Cook, Perce11 & Reynolds 1400 L Street, N.W.
Washington, D.C.
20005-3502 Regional Administrator, Region IV U.S. Nuclear Regulatory Connission Office of Executive Director for 1
Operations l-611 Ryan Plaza Drive.' Suite 1000 l
Arlington, Texas 76011 1
. Senior Resident Inspector U.S. Nuclear Regulatory Connission
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1-Nuclear Plant Road Russellville, Arkansas 72801 Ms. Greta Dicus, Director Division-of Environmental Health Protection Arkansas Department -of Health 4815 West Markam Street Little Rock, Arkansas 72201 Mr.-Robert B. Borsum Babcock & Wilcox
-Nuclear Power Generation Division 1700 Rockville Pike, Suite 525~
Rockville, Maryland 20852 l
V L
Enclosure; 7.V L
i QUESTIONS FOR TOPICAL REPORT CEN-386-P' I
- VERIFICATION OF THE ACCEPTABILITY OF A-c 1-PIM BURNUP LIMIT OF 60 Ed/kg FOR COMBUSTION ENGINEERING'16X16 PWR FUEL"
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1.
The burst strain data from Fort Calhoun cladding with local burnup levels.
between 55 to 63 mwd /kgM, presented in Section 4.1.5.4 of the: topical t
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- report, show very low cladding strains between 0.03 to 0.1%.
i Section 4.2.II.A.2(g) of the. Standard. Review Plan (SRP)_(Reference 1),
which addresses " acceptance criteria" to preclude pellet / cladding inter-l
. action (PCI) failures, states that uniform strain (elastic plus plastic)
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l-of the cladding should not exceed.1%.for. normal operation and anticipated operational occurrences-(A00s). This-strain limit has also traditionally
'been applied as a limit for cladding; strain lin Section 4.2.II.A.1(a) of.
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~
the SRP. The cladding burst data froml Fort Calhoun suggests that Combustion Engineering (C-E) fuel cladding may fail.at_ unifom strains-significantly below the 1% strain limit recommended in the SRP. There-l fore, several questions arise from this data:-
a)
How applicable are the burst tests and measured strains-from the Fort Calhoun cladding to failure mechanisnis'from. normal operation and A00s in-C-E commercial reactors? This response shouldl address those failure mechanisms identified in Section 4.2 of the-SRP.
If this data is applicable to these failure mechanisms-or if:their -
applicability is unclear,.please address the following additional; questions.
b)
Should the unifom strain limit _ of C-E cladding for normal operation and A00s be decreased to a level below 1% when. local burnups exceed 55 mwd /kgM7 c) Will the fuel cladding become even more embrittled at local burnups.
above 63 mwd /kgM7 d) Will fuel failures become more frequent as the number of fuel rods that exceed local burnups of 55 mwd /kgM increases from commercial operation?
1
2.
Please discuss how the standard deviation, a, is calculated for oxide 3
thickness at rod average burnup levels of 60 mwd /kgM from the 14x14 and 16x16 fuel. rod data in Section 4.1.2.1.a.
Should this data be separated because there are inherent differences between-the corrosion behavior;of j
14x14 and 16x16' fuel rods or should they be combined because the differ-ences;in corrosion are not uniquely design dependent? Also,_it appears that the estimates of Y + 3a, provided in this s,ection,. assume that a is f
independent'of burnup, while the corrosion data in Figures 4.1.2.a-1 and 4.1.2.a-2 suggests that~a becomes larger at higher burnups. What impact would e more variable and larger calculated a have on-the performance _
analyses of~the 16x16 fuel rods (e.g., cladding stress) at 60 mwd /kgM7 Please provide the effect as a percentage change from the condition of no cladding-wastage due to corrosion.
3.
The results of the cladding collapse calculation in Section 4.1.4.a are based on an assumed finite " hot" axial gap length in the fuel. column of i
modern C-E designs.
In this-analysis it is implied from postirradiation examination (PIE) data that " hot" axial gaps greater than this assumed size have a low probability of existing in C-E's 16x16 design, but no-probabilities are calculated based'on this PIE data. What is the probability of the C-E.16x16 design having an axial gap of this assumed.
size or greater? The probability may be calculated using PIE data of i
" cold" axial gap sizes measured from modern fuel designs other than C-E's 16x16 design, but these designs should be-comparable in fuel length, densification characteristics, and density. A correction between
" hot" and measured " cold" gap size is permissible but assumptions made in this correction should be stated.
The calculated probabilities may also take into account that today's fuel designs typically fom smaller axial i
l gaps in the fuel column than previous " older" fuel designs because of changes in fuel fabrication. For example, three separate populations of axial gap size can be identified according to the following fuel characteristics 1) older densifying fuel, 2) older nondensifying ' fuel with low fuel densities (i.e., <94% theoretical density), and 3) newer nondensifying fuel with higher fuel densities (i.e., >94% theoretical 2
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. density); with fuel with the latter characteristics displaying the smallest axial gap sizes following irradiation.
4.--
From the-small number of FATES 3B predictions of low temperature fission gas release data provided in' Table.4.1.6.a-3, it appears that.this code t
may be underpredicting this. data by a small amount,(i.e., I to 2% release when rod average burnups exceed 54 mwd /kgM). What is the effect, if any, on fuel performance calculations at low temperatures if the FATES 3B code predicts 1% release when 2.5% release is the actual amount released?-
q l
S.
In Section 4.2.1.a. on guide tube wear, it is stated that References 4.2.1.a-2 and 4.2.1.a-3 have verified the conservatisms in C-E analytical; predictions of guide tube wear for these assemblies.
Please provide a comparison of measured-and predicted maximum wear for 16x16 unsleeved assemblies along with their burnup levels.- What is the maximum wear predicted for the C-E 16x16 assemblics.at the' maximum l
residence times expected for the burnup levels requested?
6.
Section 4.2.2.a addresses the shoulder gap between the top of the fuel
~
rods and the bottom of the upper-end-fitting, but does not address the possibility of the assembly hold-down spring bottoming out due to assembly growth.
Please demonstrate that the assembly hold-down spring does not bottom-out at the assembly burnup level requested. Also, what is the predicted gap margin for preventing bottoming-out of the hold-down spring?
7.
What is the maximum boron-carbide content, in weight percent, of the alumina-boron carbide pellets in the burnable poison rods for the 16x16 design? Have additional postirradiation examinations been performed on-burnable poison rods (such as for helium release, internal void volumes, pellet swelling, etc.) since those examinations presented in CENPD-269-P, Revision 1-P7 REFERENCE 1.
U.S. Nuclear Regulatory Commission. July 1981.
"Section 4.2, Fuel System. Design." Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants--LWR Edition. NOREG-0800, Revision 2, U.S. Nuclear Regulatory Connission, Washington, D.C.
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