ML18219D992

Response to Request for Additional Information on Exxon Nuclear Company Reports XN-76-25 and XN-75-39

ML18219D992
Person / Time
Site:	Cook
Issue date:	10/01/1976
From:	Tillinghast J Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co)
To:	Rusche B Office of Nuclear Reactor Regulation
References
Download: ML18219D992 (32)

Text

U.S. NUCLEAR REGULATORY ISSION DOCKET NUMBER NRC FOAM 195 I2.70 I FILE NUMBER NRC-DISTRIBUTION FoR PART 50 DOCKET MATERIAL FROM DATE F DOCU TO: Mr Rusche Indiana & Michigan Pwr Co E

$0 New York, NY J Tillinghast DATE RECEIVED ]P 5 76 ETTER NOTORIZED PROP INPUT FORM NUMBER OF COPIES RECEIVED

&ORIGINAL RV N C LASS I F I E D 1 signed DCOPV DESCRIPTION ENCLOSURE Ltr notarized 10-1-76....re our 9-3-76 Addi.info concerning EMEm Rpt XN 76-25 &

1tr.... trans the following:- XN-75-39 which furnished info concerning operation,at~:100% pwr........

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INDIANA II( MICHIGAN POWER COMPANY P. O. BOX 18 BOWLING GREEN STATION N E W YOR K, N. Y. 10004

~sf October 1, 1976 Donald C. Cook Nuclear Plant Unit No. 1 Docket No. 50-315 DPR No. 58 Mr. Benard Rusche, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Rusche:

This letter is in response to Mr. Dennis L. Ziemann's letter of September 3, 1976 which requested additional information concerning Exxon Nuclear Company Reports XN-76-25 and XN-75-39.

We submitted. these reports in support of our July 20, 1976 request for an amendment to the above captioned operating license to permit opexation at 100% of rated power for fuel cycle 2.

Attached you will find a xesponse to each question from Enclosure 1 of Mr. Ziemann's letter, except for 4.0/11/1, as well as responses to questions 1 (portion in brackets only),

6,7,11,12,22,23,26,33,34, and 35 of Enclosure 2. The balance of the questions in Enclosure 2 were identified as being generic in nature with no response necessary for Donald C. Cook Nuclear Plant Unit No. 1. The questions to be answered axe as called out in Mr. Ziemann's letter and subsequently modified at a meeting

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Mg B., Rusche October 1, 1976 held between the NRC Staff, Exxon Nuclear Corporation, and our-selves on September 14, 1976.

Very truly yours, o n 'llingh Vice Preside JT:mam Attachment Sworn to and subscribed to before me this /~day of October 1976 in New York County, New York Notary Public ViAs'Fi'f EEN BARRYr NOTARY I ~~..~, ar.'<" of N~e No. 4k -.-'-,i)2 Quafdreu in Queens County Cyrtificat~ fifed in New York Counttf.

grsrnrrussion kxprres rrIarch 30, p 7p cc: G. Charnoff R. J. Vollen R. C. Callen P. W. Ste'ketee R. Walsh R. S. Hunter R. W. Jurgensen Bridgman

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,QUEST S AND ANSWERS FRCN ENCLOS X 1 (concerning XN-76-25)

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2.0 4 'estion This section indicates that 1 or 2 fingers of control rod E-3 had bxoken off during rod timing tests; and that an inspection program is being developed to examine control rod E-3 and "possibly additional control rods during the upcoming refueling." Please indicate your plans for inspection of control rods; and, if you do not plan to inspect basis for this decision.

all rods, provide a Answer 2.0 4 The apparent failure of the contxol rod in core location E-3 is only the fourth such incident of failure known to

-our NSSS supplier and only the second occurrence to an RCC of the type in Unit No. 1. The RCC assembly, R-43, is located in a Region 1 fuel assembly which is scheduled for discharge during the refueling. Visual inspection will be performed to insure the fuel handling ecpxipment can latch onto and transport the assembly 'control rod unit to the spent fuel pit. Zt is premature, at this point in time, to decide on a definitive inspection program for that control rod since the program itself will depend upon our specific findings from the visual inspection. Xt is our intent, however,,to identify the type of failure and if possible the cause of failure.

With regard to the inspection of other control rods, the failure of R-43 is viewed as an isolated incident not indicative of a generic problem. This isolated event in itself is not considered significant enough to justify a major .on-site exam-ination program. There is the need to demonstrate-that all of the other control rods are operating properly and .are capable of performing their design function. With this in mind we intend to perform rod drop timing tests at the end of cycle 1 operation (damaged RCC R-43 excluded), then perform a drag test on each RCC and lastly perform the rod drop timing tests again as indicated in Section 10, page 70 of our submittal, XN-76-25.'f additional RCC assemblies are identified as suspect on the basis of anomalous end of cycle 1 drop tests, a visual examination shall be performed on those assemblies. The scope of any further examination of those assemblies depend upon the findings of the visual inspection.

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Questions 4.0/11/4 and 4.0 23 2 l

Xt is not clear that ANSYS or AXXBOW are NRC approved codes. please cite the document via which these codes were approved.

I Answers 4.0/11/4 and 4.0 23 2 XN-76-25 incorrectly, indicated that the NRC had approved the ANSYS and AXXBOW design codes. These codes have been used extensively and are considered by Exxon Nuclear to provide acceptable methods for evaluating fuel design.

uestion 4.0 19 1 Please clarify and justify the statement that "no signif-icance in reactor spring relaxation is expected" (what is the spring life expectancy and what is the amount of spring relaxation anticipated with timeP)

Answer 4.0 19 1 Pe calculated fast flux in t'e holddown spring region is 3.3 x 10 n/cm /sec ( 1 Mev) . For three cycles of operation corresponding to 7.6 x 10 secs, the fast neutron fluence is 2.6 x 10 n/cm . This. fluence can cause up to 20%%d relaxation of the springs, although the probable relaxation is less (see Figure 4.8, XN-75-39).

There is sufficient margin the spring design to cover more than 20%%d relaxation.

uestion 4.0 26.4 Please cite the reference in which data of Exxon fuel rod length changes wi.th burnup is reported.

Answer 4.0 26.4 ENC has measured fuel rod length changes on fuel from three reactors: Oyster Creek, Big Rock Point, and R. E. Ginna.

Approximately 100 rods have been measured, covering burnups from about 6000 MWD/MTM to 19,000 MWD/MTM. The results of these measurements. are shown in the attached Figure 1. Maximum fast flux fluence for D. C. Cook is calculated not to exceed 7 x 10 n/cm

( ) 1 Mev). Using the R. E. Ginna fuel rod length change as a basis, the fuel rod extension would be less than 8 x 10 in/in or 1.15 inches.

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Oyster Creek Puel Hods (Zr-2 Cladding)

)I Thin CladcLing-I Thick Cladding Ginna Puel Hods (Zr-4 Cladding)

Big Hock Point Puel F~s (Zr-2 Cladding)

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Question 5.0 55 5.2 Xt is stated that "That Neutronics design methods utilized to calculate the data presented herein are consistent with those described in Reference 5.1...". However, we note that Reference 5.1 has not been submitted.'ompletion of our review requires receipt of that Reference.

Answer 5.0 55.5.2 Draft copies of Revision 1 to XN-75-27 were given to the staff on September 15. The report will be submitted formally by October .1, 1976.

Question 10 70 A.

provide the acceptance criterion for each test and discuss how the measured parameter(s) relates to the values in the accident analysis.

Answer 10 70 A.

1. Rod Position Indication System The acceptance criterion is based on the technical specifications which require that all control rod position indicator channels are within 12 steps of their bank demand position. I

2. Rod Drive Mechanism Timing The acceptance criterion is that the control rod mechanisms function so as to move the rods in/out when the motion is called for by the operator.

3. Rod Drop Timing from the This test is acceptable fully withdrawn position to iftheeachdashcontrol rod drops pot entry in<1.8 seconds.

4. Xnitial Criticality The acceptance criteria are:

a) criticality shall be achieved within 75 ppm of the analytically predicted value, b) the neutron flux level at which nuclear heating (Doppler feedback) effects become measurable is established so that all zero power tests shall be conducted below that level, c) the reactivity measured using the reactivity computer shall be within 4% of the reactivity obtained from reactor period measurements,

P e Cr'itical Boron Concentrations The acceptance criterion is that the difference between measured and predicted boron concentrations shall be within 75 ppm,

6. Control Rod Bank Norths This test is acceptable if the integral worth of the control banks is within 10%%d< or + 100 pcm than 1000 pcm,

+ if measured worth is less

7. Isothermal Moderator Temperature Coefficient The test is acceptable if the isothermal temperature coefficients are determined to be negative,

8. Incore Moveable Detector Maps The acceptance criteria is such that the relative asserrihly powers are within + 10% for P ~

)<0.9 and + 15% for Pi >4 0.9.

The values of Fg, FgH and quadrant power tilts are to comply with the technical specifications,

9. Excore - Incore Detector Calibration The excore and incore detectors are to be calibrated to each other.

10. Power Coefficient The power coefficient shall be determined.

The startup physics tests will be performed to verify that the core can be operated safely and to make an initial evaluation of the expected performance as predicted by design calculations. Accident analysis configurations in general are not used, thus many measurement results do not relate directly to the values of the parameters used in the accident analyses.

However, the measured values of the parameters should relate to the values used in accident analyses such that they fall in the range of the values used. Further, a direct comparison of the measured and safety limit values can be misleading. In many cases the most conservative combination of reactivity coefficients is used in a transient analysis, even though the extreme coefficients used do not occur at the same conditions of lifetime, power level, temperature, and boron concentration assumed in the analysis.

r 10 70 B. 'uestion Describe in detail the control rod bank reactivity worth tests, the maximum deviations from calculated values expected, the criteria used. for determining these maximum deviations and the procedures to be followed i.f these devi.ations are exceeded.

Answer 10 70 B.

Three distinct rod worth tests are planned for the Donald C. Cook Nuclear Plant Unit 1, Cycle 2 startup physics test program. They will consist of rod worth measurements for control banks D and C under HZP conditions. The tests are:

a) D - bank worth b) C - bank worth with D fully inserted c) D + C worth in the overlap mode The reactivity worths of the control rod banks are obtained by monitori.ng reactivity changes while maintaining the reactor approximately critical through boron/control rod bank exchanges. Following the establishment of a constant dilution/

boration rate, the controlling control bank is periodically inserted/ withdrawn to compensate for the changing reactor coolant system boron concentration. The "reactivity change resulting from each control bank movement is recorded by a reactivity computer. The differential control rod reactivity worth i.s defined as the change in reactivity per unit change in bank position about an average bank position, and the integral worth is the summation of the individual reactivi.ty steps between fully inserted and fully withdrawn positions.

The results of the control rod worth measurements are compared. to the values obtained from design calculations.

The maximum- difference expected between the measured rod worths and design predictions is +10/.

The acceptance criteria used for comparison of measured results with design predictions are based on the experience that has been acquired as a result of the startup physics testing programs of the PNR's which are now commercially operating. However, these acceptance criteria are viewed as

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guides to possible measurement or design errors and failure to meet these criteria does not by itself constitute a safety problem. If observed differences between measured and predicted values are beyond the acceptance range, one or moxe of the following actions will be taken:

a) re-examining the measurement data for possible

.computational errors or unaccounted experimental perturbations b) verifying that the measurement conditions are consistent with the assumptions in the design prediction c) repeating the measurement or performing additional tests if the discrepancy remains unresolved d) evaluating the impact of the value measuxed with respect to the appropriate accident analyses sensitive to the measured parameter e) repeating the appropriate accident analyses using values consistent with the measured values of the parameter in question f) revising the plant operating procedures in a con-servative manner to permit operation while a xeview of the safety implications is being made.

uestion 10 70 C.

Minimum test requirements for certain tests are given below.

Control Rod Group Worth At least the first and second regulating banks.

If either of those measured values differs from the predicted by more than 10%, then measure the sum is off more than 10%,

the first shutdown bank. If then measure the worth of the N-1 configuration. (All rods inserted into the core except the rod pxedicted to have the highest worth).

Moderator Temperature Coef ficient At least two configurations:

The all-rods-out, and firct-bank in.

Power Distribution At least one configuration at zero power:

All-rods-out; at least 1 map at power before 75%

power is reached.

Power Coefficient Ir At least 1 measurement at a power level over 65%%d of rated power.

Answer 10 70 C.

We intend to incorporate into our startup program the requirements in 10/70 C.

uestion 10 70 D.

State your schedule for submitting to the NRC a .

brief summary report of physics startup test results. This report should include both measured and predicted values.

the difference between the measured and predicted value exceeded If the acceptance criterion, the report should discuss the actions that were taken and justify the adequacy of these actions.

Answer 10 7 D.

Following each of the startup physics tests that will be performed at the Donald C. Cook Nuclear Plant Unit 1 power station, a formal report will be issued. This report will describe in detail the physics tests performed as part of the startup program, and it also will- contain the specific results for each of the tests performed. The startup physics, test report will be issued to the NRC approximately 90 days following the completion of the testing program associated with the reload.

The procedure is to compare the measured value of the parameter(s) of interest associated with each test to their respective predicted values in this report. If the difference between the measured and predicted values exceed the acceptance criterion for the test, the report shall discuss the actions that wexe taken and justify the adequacy of these actions.

Please. note the following changes to our original submittal of XN-76-25, section 10.0, page 70:

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1. Change item 1 to read:

Rod Position Indication System'his test is performed prior to criticality in the hot shutdown condition to verify the system performs satisfactorily.

2. Change item 6 to read:

Control Rod Bank Worths To measure the differential bank worths at hot zero power i.n both the overlap and non-overlap conditions.

3. Change the number of the third item from 2. to 3.

4.In XN-76-25 p. 4la, drawing dimension for height of lower tie plate should be 2.720.

QUESTIONS A ANSWERS FROM ENCLOSURE 2 (concerning XN-75-39)

The introduction of the report should clarify how the report will be used with regard to reload licensing applications.

For example all Westinghouse plants may not have the same operational procedures and limits so that. the maximum power distributions reached during transients may differ from plant to plant. (Does the Exxon fuel design limits bound the worst 1S x 15 plant operational limits'P) The introduction and the design basis sections should include the bounding conditions considered and provi.de the justification for the values used.

Answer 1.

The design limits for 15 x 15 PWR fuel assemblies are as follows:

3 Loo Plant 4 Loo Plant Nominal Design Pressure, psia 2250 Maximum Design Pressure, psia 2500 Flow Velocity (average), ft./sec 15.0* 15.5*

Peak LHGR, kw/ft. 13. 5 Batch Average Burnup, MWD/MTM ~

33, 500 Peak Pellet Burnup, MWD/MTM 47, 500 Maximum Cladding Temperature, 'F SSO F ID

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6750F OD Maximum Coolant Temperature, 'F Inlet: 550 Outlet 642 Operational Transients Per Table 4.5 in XN-75-39 Maximum Full Power Hours 22,000

+Requires different holddown spring design for 3 and 4 loop plants.

Question 6.

The thermal conductivity of Zr-4 is based on data published by Scott (reference 1, XN-75-39) . There is a discrepancy in the equation for thermal conductivity given in your report compared to the equation given in the report referenced. please clarify.

Answer 6.

The equation given in XN-75-39 or the thermal conductivity of Zircaloy is based on the Nickel-fxee Zircaloy-2 data of Scott and not on the Scott equation for Zircaloy-4. The use of the Nickel-free Zircaloy-2 data gives conservatively low values for Zircaloy-4 thermal conducitivty.

Nhy is the equation for fuel densification for fuel densities between 85 to 91.9%%d T.D. for t ( 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> inconsistent with the Exxon model submitted for staff approval (reference letter.:

R. Nilson to J. F. 0 'Leary, December 13, 1973) .

Answer 7.

There is a typographical error in XN-75-39 for densities of 87 to 91.9/ TD. The correct term for t < 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is:

AP//max = [0.0148 tj Your desi.gn limits appear to be incomplete. Desi.gn limits for buckling (instability), fatigue, etc., as well as stress and strain, should be provided. Zn addition, there should be a specific value assigned for the design limits for the events catagorized in question 8.

Answer 11.

Design limit requirements are scattered through the report. For steady state conditions the mechanical design limits and stress or load intensities fox each component is summarized in the attached Table (Answer 11).

Table (Answer 11)

Stress Intensi.ty Desi.gn Limit Safety Com nent or Load Basis Ualue Factor Fuel Rod Primary Stresses:

BOL Cold -13,200 psi 1/3 UTS 26, 670 2." 04 BOL Hot 9,600 psi 1/3 UTS 16, 670 1;72 Secondary Stresses:

BOL Cold -13,950 psi 1/2 UTS 42, 500 3.05 BOL Hot -10,120 psi 1/2 UTS 25,000 2 47 Total Stresses:

BOL>> Cold 37F 560 psi UTS 85,000 2.26 BOL - Hot -34,360 psi UTS 50,000 1. 46 EOL Hot -40,740 psi UTS 50,000 1.23 Instantaneous Collapse BOL Cold 2485 psi. (1) 5002 psi 2.01 EOL - Hot 2485 psi (1) 2527 psi. 1.02 Fatigue Usage Factor 0.27 0.67 2 '8

Stress Xntensity Design Limit Safety Com onent or Load Basis Value Factor

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Side Plate Loading Cold YieM at 2560 ebs No yielding 35jog $5s 1+04 2 x Bundle Weight (wet)

Spaces Entersection Strip (2)

Weld Strength 126,000 psi (3) 55,900 pgi $ .29 Guide Tube to Spacer Weld Strength ) 300 lbs/tab (4) 150 lbs/tab 2.00 Crush Strength During Shipping 1890 lbs (5) 1740. lbs 1.08 Tie Plates:

Primary Load ) 8000 lbs (no yielding detected 2.5 x Bundle 3500 lbs = 2.28 in tests) Weight (dry)

Tie Plate Lockin Hardware:

Primary Load Sleeve 1200 lbs (6) 600 lbs 2.00 Primary Load Locking Rings 0 245 lbs (7) 175 lbs 1.40 Guide Tubes:

Primary Stresses 330 psi 1/3 UTS 11,670 psi, 35.4 Secondary Stresses 9250 psi 1/3 UTS 17,500 psi 1.89 Buckling 230 lbs Euler 465 lbs 2.02 Column

Stress Intensity Design Limit . Safety Com onent or Load Basis Value Factor Springs:

Spacer Cold 101,000 psi Y.S ~ 150,000 psi l. 49 plenum - Cold 95,370 psi Shear Stress 100,000 psi 1.05 Holddown Cold . 169,000 psi Y S ~ 180,000 psi 1.07 (1) Design basis is determined using Timoshenko and Gere equation for tube buckling (see "Theory of Elastic Stability", 1961).

(2) The 126,000 psi is the measured failure stress.

(3) Design basis is that weld to have strength equal to or greater than base metal.

(4) Minimum load .is ~ 2.5 times the guide tube load of 230 lbs assumed to be concentrated evenly on the four tabs of one spacer.

(5) Design Basis is established that a spacer shall withstand a lateral load equal to 6 times the bundle weight distributed evenly over 5 spacers (the upper and lower spacer do not carry much load during shipping).

(6) Minimum allowed load is ~ 2.5 times calculated guide tube load of 230 lbs.

(7) Minimum allowed load based on 2.5 times bundle weight evenly distributed over the 20 locking devices.

Discusa local primary membrane stress with examples.

Examples for general primary membrane stress, namely stress in the cladding due to pressure differential across the wall thickness, were provided as well as formulas showing how one would calculate such a quantity. The same type of discussion is needed to define and assess local primary membrane stress.

Answer 12.

'A local primary membrane stress is defined as the average stress across any solid section. It considers discontinuities but not stress concentrations and is produced by mechanical loads.

This limit is applicable to any local loading such as stresses due to fuel rod spacer support forces. Although these stresses were calculated and repoxted, i.e., Table 4.3, XN-75-39, the methodology used in the analyses is not described in the generic report.

Describe in more detail the extent of the surveilLance program. For example, identify the number of assemblies to be examined and the number of fuel rods to be inspected to evaluate the fuel rod parameters identified in Section 2.5.

Answer 22.

ENC surveillance programs are based on. need for confirm-ation of new designs. In the past, these programs have involved typically two to four assemblies wherein rods are removed for individual inspection. To date ENC has obtained performance data from at least 100 fuel rods. From these inspections, the trend in fuel performance has been established to the extent that future surveillance programs will be setup on a limited basis merely to ascertain that these trends are being maintained. Should some change~'in fuel processing or design occur, then a more extensive program would be instituted to evaluate the change.

For H. B. Robinson the fuel surveillance program utilizes two characterized fuel assemblies. Post-irradiation inspection at H. B. Robinson will involve the typical visual inspection plus dimensional measurements on bundle length and spacer grid envelope.

Both measurements are to determine irradiation growth of zircaloy.

The spacing between fuel rods and fuel rods and guides tubes will be measured on foux assemblies to ascertain row bow characteristics.

Question 23.

Fuel assembly lift off is not addressed in these reports.

If you choose to address Justify why this item was not discussed.

this in response to our question, we are specifically interested in the margin to lift-offas a function of'.percent of rated flow.

Answer 23.

1 liftoff force P

The fuel assembly is-a function of plant design, i.e., 4 loop plant's have a higher coolant flow velocity than 3 loop plants and, therefore, require a different holddown spring design. As a result, we have addressed lift-off in the documentation, provided for a particular plant. The holddown spring analysis for D. C. Cook is described in XN-76-25, Section 4.1.'3, and is generally applicable for Nestinghouse 4 loop plants.

A summary of comparable analysis for H. B. Robinson 42 was trans-mitted with CpsL's letter to Benard C. Rusche, Docket No. 50-261,-

dated October 17,, 1975, and is generally applicable for 3 loop plants.

Show the test results to support your spacer-grid analysis results in Section 4.8.2.

Answer 26.

Structural tests performed on grid spacexs have involved the following:

a) Side locking strength tests simulating loads during refueling b) Static crush tests c) Intersection weld strength tests d) Spacer to guide tube weld tests

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Dimple stiffness test, f) Lateral impact tests The results of these tests are discussed in the following:

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1) The side loading strength and crush tests results are discussed in Section 5.1.4 of XN-75-39.

2) Spacer intersection weld strength were evaluated through tensile tests on 10 intersection sample welds. The average load at failure was 138 1bs with a sigma value of 9 lbs.

On a three sigma basis, the minimum load expected would be 110. lbs.

For annealed Zircaloy-4 with a minimum tensile strength. of 655,000 psi, the load carrying capability at minimum cross-sectional area at an intersection is 48 lbs. Consequently, the weld design provides a safety factor greater than 2.

3) Spacer to guide tube weld strength have been evaluated through both weld qualification and, in-progress weld sample tests. The spot weld strengths have ranged from 300 to 400 pounds, consequently, the minimum load at. 150 pounds per tab is routinely met with a safety factor of 2.0 or greater.

4) Dimple stiffness tests were performed by compressively loading dimples on production spacers. The results show that dimple stiffness is 1760 lbs/in.

5) Impact tests are discussed as part of the seismic-LOCA analysis documented to be issued.

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What specific information will be derived from the fuel surveillance programs to provide the necessary confirmatory data to better predict PCI'P Answer 33.

PCI can cause lengthening of fuel rods via a ratchetting mechanism and can also enlarge the diameter of a fuel rod. ENC monitors both fuel rod diameter and length to determine performance..

Pellet stack length measurements and gamma scanning which are part of ENC's surveillance program, can provide additional information pertinent to PCI.

Provide justification, based on short-term fretting-corrosion tests, for assuming that no significant fretting corrosion or mechanical wear will occur over the lifetime of the fuel.

4 Answer 34.

ENC has run fretting-corrosion tests on 7 prototype PWR assemblies and 5 prototype BWR assemblies. Although these tests cover wide variations in design, the design philosophy and design details for fuel rod support are substantially the same in most cases. Test periods ranged from 500 to 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> at reactor conditions of temperature; pressure and flow. Fuel rod wear depths at spacer contact points.has typically ranged from 0.1 to 0.5 mils although wear of up to ~ 1.5 mils in depth has been occasionally observed.

Examination indicates that the wear is due primarily to fuel rod loading and unloading and not fuel rod-motion during the test. There has been little or no difference between observed wear for 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> and 1500 hour0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> tests.'o active fretting corrosion has ever been observed on any test rod in spite of the fact that spacer springs have been xelaxed up to 100%

in several test assemblies. We believe that it if conditions are will show up in a few hundred conducive to fretting corrosion hours.

Reload fuel representing 8 of the assembly designs tested have been loaded into a reactor. Five have been in a reactor 3 or more cycles. Detailed examination including profilometry of a number of rods in 3 reactors has not revealed wear significantly different than that observed in fretting corrosion tests. No.

.active fretting corrosion has been observed; uestion 35.

The onsite inspection of new fuel consists of the.

determination of f'uel rod axial position, rod-to-rod spacing and cleanliness. Is this adequate to insure mechanical integrity of the fuel rod assemblies and components after shipment.

Answer 35.

In addition to these measurements, one assembly per truck load is characterized by measuring the spacing or distance of the outer fuel rods to the spacer envelope. The purpose is to have a traceable dimensional check in case of an unpredicted occurrence.

ENC has performed extensive shipping tests to assure that the packaging and containers maintain the proper mechanical integrity of the fuel assemblies.

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