ML19339A763
| ML19339A763 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 11/07/1975 |
| From: | Vandenburgh D YANKEE ATOMIC ELECTRIC CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WYR-75-126, NUDOCS 8011040750 | |
| Download: ML19339A763 (12) | |
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YAl!KEE ATOMIC ELECTRIC CORIPANY x%
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. United States I;uclear Regulatory Commission
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C Attention: Office of 1:uclear Reactor Regulation 3
Referencc:
(1) License !;o. DPR-3 (Docket IJo. 50-29) 4 (2) Proposed Change !!o.125 (July 14,1975).
h?,k (3) Proposed Change !;o. 125, Supplement No. 1 (October 10, 1975).
(4)
Proposed Change 1:o. 125, Supplement 1:o. 2 (October 28, 1975).
(5) Letter from R. A. Purple to R. H. Groce, dated October 30, 1975 regarding questions on Proposed Change No. 125.
Subject:
Core XII Analysis
Dear Sir:
The attached information is provided in response to your letter (5) regarding Proposed Chango 11o.125.
Answers to questions 2 and 6 require tran. e ttal of information of the type which Exxon !!uclear maintains in confl.tence and withholds from public disclosure. The inferr.ation has been handled and claurified proprietary by Exxon 1;uclear in accordance with their procedures and standards, and we hereby make application for withholding from public disclosure this information in accordance with the provisions of 10 CFR 2.790(b) for the following reasons:
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It reveals certain distinguishing aspects of fuel Q, >'
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TN design vhere prevention of its use by any of Exxon e
N0'/ / $975 1:uclear's competitors without licc r se from Exxon
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q 1;uclear constitutes a ccmpetitive economic advantage
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over other companics.
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Its use by a competitor would reduce his expenditure k\\
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or resourecs or improve his competitive position in Y.. x I.
the design and manufacture of a similar product.
Q The Yankee Atomic Elcetric Company has a. proprietary agreement with the Exxon 1;uelcar Company and has handled this information in accordance with that agreement.
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United States Nuclear Regulatory Commission November 7, 1975
' Attn: Office of Nuclear Reactor Regulation Page Two For yo' convenience both a proprietary response (Attachment A, fout copies) and a non-proprietary respont.e (Attachment B, forty copies) are provided. All pages containing proprietary information are clearly designated as such and should receive no public disclosure.
This supplement to the Proposed Change No. 125'has been reviewed by the Nuclear Safety Audit and Review Committee.
We trust you will find this information satisfactory; however, should you desire additional information feel free to contact us.
Very truly yours, YANKEE A':OMIC ELECTRIC COMPANY D. E. Vandenburgh Vice President COMMONh'EALTH OF MASSACHUSETTS)
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COUNTY OF WORCESTER
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Then personally appeared before me, D. E. Vandenburgh, who being duly sworn, did state that he is a Vice President of Yankee Atomic Electric company, that he is duly authorized _to execute and file the foregoing request in the name and on the behalf of Yankee Atomic Electric Company, and that the statements therein are true to the best of his knowledge and belief.
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Armand R. Soucy Notary Public My Commission Expires September 9, 1977 0
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'N, Question 1 show either by analysis or by previous plant experience that the dissimilar materials between tie rods and fuel rods will behave satisfactorily regarding differential thermal expansion.
Response
The EUC Yankee Rowe fuel was designed and f abricated using a 304L stainless steel skeleton and Zircaloy-4 fuel cladding.
The arrangement utilizing dissimilar metals for structure r 1 cladding is essentially the same arrangenent that is used by exi 31 in Yankee Rowe, which is in its second cycle of operation with
.nt performance difficultice.
This material combination is also ti s previously used in two ENC fuel assemblies for the Ginna react 3inna assemblies have performed satisfactorily to date after ar aly 13 months of reactor operation.
Column buckling of the tic rods (guide bars) due to differential thermal expansion between the stainless steel skeleton and Zircaloy-4 fuel cladding was also considered in the Yankee Rowe, fuel design.
Conservative snalyses show that a minimum safety factor of 2.8 exists between the critical buckling load and the load that is generated by frictional forces due to dif ferential thermal expansion.
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4 Question 2 Provide a detailed drawing and description of the spacer grids, including the connecting scheme between rode (fuel-rods-to-spacer or tie-rod-to-spacer).
_ Response i
The following proprietary drawings show the connection be, tween the tie rods *
(guide bars) and the spacer grids:
XN-302010, spacer assembly, type A XN-302011, spacer assembly, type B XN-302014, guide bar (corner)
XN-302015, guide bar (side)
XN-302004, fuel bundle skeleton assembly (type A)
XN-302005, fuel bundle skeleton assembly (type B) l The spacer grid assemblies are attached to the guide bars in the skeleton asserbly by weld ' joints made by tungsten inert gas ' arc welding methods.
In addition to quality control requirements, thp welding process is controlled to assure that each joint exhibits a minimum of 500 lbs ultimate strength in chear. Conservative analysis and testing demonstrates a joint strength requirement of no more than 145 lbs.
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-Question 3 Discuss out-of-pile proof tests, if anyrin support of' fuel assembly design verification.
Response
Listed-below are the out-of-pile proof tests which were performed to verify.
the adequacy of the Yankee Rowe fuel design:
A., Yankee-Rowe locking system strength test This test was performed to verify that the locking devices used to attach the upper nozzle assembly to the guide bars was sufficiently strong to withstand design handling loads.
Test results indicated that minimum safety factor of 2.,0 exists between the yield point load and the design load with only four of the eight total guide bars carrying the load.
B.
Spacer grid springs and dimples Tests were performed on production spacer springs to verify
'orce deflection characteristics.
In addition, tests were.
conducted on spacer dimples to determine the support stiff-ness which was used-in design calculations for contact stress in the fuel cladding.
C.
Cladding to Spacer contact friction Tests were performed to determine the friction loading between the spacer grid contact points and the fuel rod cladding.
This information was used to verify the adequacy of assumed friction coefficients in calculations to establish the spacer grid and guide bar loads due to differential thermal expansion.
D.
Spacer Assembly strength test Tests were performed on a typical Yankee Rowe Spacer Assembly to verify its structural integrity.
Data obtained frem these tests were used to verify the adequacy of the spacer design from the static and fatigue standpoints.
E.
Lower Nozzle Strength Test Strength tests were performed on the lower nozzle assembly simulating ~ loading during reactor operation.
Test results showed -that the design was adequate by a large margin.
F.
Upper Tie Plate Strength Tes,t Strength tests were. performed on _a Yankee Rowe production upper tie plate. Data frca this test confirmed that the upper nozzle design was adequate.
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Pressure Drop Tests Tests were performed on a prototypic Yankee Rowe fuel assembly utilizing production parts to verify calculated component pressure loss coefficients.
Loss coefficients were obtained for the inlet hardware grid spacers, bare rod friction, and outlet hardware over a Rey'nolds Number rang,e of %100,000 to
%450,000. _ Test results provided the basis fbr determination of the assembly and assembly component pressure, drops which was used to confirm the thermal-hydraulic compatibility of the ENC fuel and the existing fuel.
H.
Fretting Corrosion Tests A test was performed on a prototypic fuel assembly as in section G above to demonstrate the adequacy of Yankee Rowe fual design with respect to corrosion, fretting corrosion,_and mechanical wear under reactor hydraulic conditions for approximately 276 hours0.00319 days <br />0.0767 hours <br />4.563492e-4 weeks <br />1.05018e-4 months <br />. Test results showed no evidence of corrosion or fretting corrosion.
Some nochanical wear was observe'. on the upper nozzle hold down springs and bolts due.to 1.teral vibration of the hold down springs. ' Further testing was directed at evaluation of this wear and resulted in the determination that the wear was self limiting at approximately.008 inch total which is not detrimental to fuel assembly performance.
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Question 4 Discuss the recponse of fuel ~asserblies regarding scismic and LOCA conditions.
Response
The column buckling strength and other structural strength characteristics of the E! 0 assemblics are at least equivalent to the existing asserblies.
Dynamic loading tests simulating IDCA loads using a prototype E!!C assembly are scheduled for the first half of 1976.
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g Question 5 Provide input values and,.results for Che' cladding creep collapse calculation.
Response
Listed below are the input values used in the creep collapse calculations for the Yankee Rowe Fuel o
Design exposure (Assembly Average) 30,000 MWD /MTU Design basis power history Table 1 o.
o Design basis fast flux history Table 1 o
Fuel rod propressurization level (Helium) 125 j;5 psig o
Initial pellet density 94.0 f; 1.5% T.D.
Final pellet density (after densification) 96.5% T.D.
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Pellet dish volume 1.0 f;.3%
o Cladding thickness (minimum 2a)
.0232 inch o
Cladding outside diameter
.365 f;.002 inch o
Cladding inside diameter
.317 f;.0015 inch o
Cladding material Zircaloy-4 o
Coolant pressure
,2015 psia o
Axial' location of collapse 80% above bottom of core-o Length of fuel column.
91.0 inch o
Fuel rod plenum volume _-
.2348 inch 3 o
Initial cladding ovality (2a)
.00094 inch o
nelium Absorption-10% initial o
Cladding Temperature increase in gap due to radiation from pellet ends considered Results of the~ creep collapse calculations are summarized in Figure 1 which shows ovality as a function of time.
Ovality at the end-of-life (21000 hours is calculated to be approximately.060 inches. Clad collapse is not predicted for the residency time of the fuel in the core.
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Design Case:
.C50 Pressure = 120 psig
,i Initial Ovality = 0.00094" Power History per 3
Table 1
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TABLE 1 DESIGN BASES VALUES FOR ROD PRESSURI276 TION 9
l TIME PERIOD (Hours)
LINEAR HEAT RATING (kw/ft)
FAST FLUX (n/cm2 - sec > 1 Mev)
O Rod Pellet Rod Peak 0-1000 6.93
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5.29 (10) 1000-3153
- 6.93 8.67 5.29 (10)13 3153-6827 6.88 8.34 5.29 (10) 5.29' (10)13 6827-10500 6.79 7.81 10500-15750-5.73 7.24 4.51 (10)13 i
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Question 6 Descrite the fuel pellet' design (chamfer, dish, etc.) and compare with Corb XI pellets and any sinilar Exxon design for which applicable experience is available.
Pesponse The response to this question is proprietary to Exxon Nuclear.
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NRC DISTRIBUTION FOR PART 50 DOCKET MATERI AL
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CONTROL NO:12 9 30 i
FILE:
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FROM: Yankee Atomic Elce Co DATE OF DOC DATE REC'D LTR TWX RPT OTHER Westborough, if ass D E Vandenburgn 11-7-75 11-11-75 XX
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50-2y DESCRIPTION:
ENCLOSURES:
Ltr te cheir 7-14-75 tech specs change Suppl 93 to Proposed tech specs. change F125 sub=tecal & our 10-3.0-75.ltr... o.....
..........(40 cys enc 1 rec'o)
ADAN NbEDGED
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(N N-PR PRImRY VERSION)
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