ML18219D112
| ML18219D112 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/07/1978 |
| From: | Tillinghast J Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co) |
| To: | Case E Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18219D112 (14) | |
Text
REGULATORY IN NATION DISTRIBUTION DISTRIBUTION FOR INCOMING MATERIAL i~
~
REC:
CASE E
G ORG:
TILLINGHAST J NRC IIV 8( MI PWR SYSTE
.RIDS)50-3i6 DOCDATE: 03/07/78 DATE RCVD: 03/i4/78 DOCTYPE:
LETTER NOTARIZED:
NO COPIES RECEIVED
SUBJECT:
LTR i ENCL 1
RESPONSE
TO NRC"S I TR OF 03/03/78... FURNISHING ADDL INFO CONCERNING THE STEAM GENERATOR SUBCOMPARTMENT PRESSURE
RESPONSE
ANALYSIS AND THE STAFF REVIEW OF RFSULTS OF AUDIT OF ENVIRON QUALIFICATION RECORDS'OR UNIT i... WfATT INFO CONCERNING CABLE TESTlNG.
PLANT NAME: COOK UNIT 2 REVIEWER INITIAL:
X, DISTRIBUTOR INITIAL
+++++++++++++++>+ DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS NOTES:
i.
SEND 3 COPIES QF ALL MATERIAL TO IS(E PSAR/FSAR AMDTS AND RELATED CORESPONDENCE (DISTRIBUTION CODE BOOl)
FQR ACTION:
INTERNAL:
EXTERNAL:
ASST DIR VASSALLO++LTR ONLY PR l ZAK44W/ENCL RFG F ILE~~~W/F L
K'4'2 NCL P.
COLLINS++WfENCL HELTEMES++W/ENCL MIPC44LTR QNLY BOSNAK++W/ENCL PAWLICKI4~~W/2 ENCL NQVAK++W/ENCL CHECK+4W/ENCL BENAROYA++W/ENCL IPPOLITO++M/ENCL GAMMILL++W/2 ENCL BUNCH+4W/ENCL KREGER++W/ENCL LPDR S ST.
JOSEPH MI++W/ENCL TIC++W/ENCL NSIC++W/ENCL ACRS CAT B~~W/i6 ENCL BR CHIEF KNIEL+wLTR ONLY LIC ASST LEE44LTR ONLY NRC PDR44W/ENCL OELD+~LTR ONLY HOUSTON%+W/ENCL CASE4l+LTR ONLY KNIGHT44LTR ONLY S IHWEIL++W/ENCL ROSS+4LTR ONLY ROSZTOCZY+4W/ENCL TEDESC044LTR ONLY LAINAS+%W/ENCL F.
ROSA4+WfENCL VOLLMER~~+LTR ONLY J.
COLLINS4+W/ENCL KIRKWOOD++W/ENCL DISTRIBUTION:
LTR 54 ENCL 44 SIZE:
2P+5P+2P CONTROL NBR:
780750053 Elk Q THE END
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'IRUPtoj(It It@I'(I: )E
)OPI INDIANA IL MICHIGAN POWER COMPANY P. O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 March '7, 1978 Donald C. Cook Nuclea'r PlaTLt Unit 2 Docket No. 50-316 DPR No.
74 Mr. Edson G. Case, Acting Director Office of Nuclea'x'eactor Reg'ulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Case:
This letter is in response to Mr. Karl Kniel's letters of March 3, 1978 wherein the NRC Staff requested additional information con-cerning the Steam Generator Subcompartment Pressure
Response
Analysis and the Staff Review of Results of Audit of Environmental Qualifi-cation Records, for Donald C, Cook Nuclear Plant Unit 2.
The staff requires resolution of this issue prior to initial criticality of D.C. Cook Unit, 2.
Attachment 1 to this letter pres'ents an item by item response to your staff's request for additional information on the Steam Genexator Enclosure Analysis.
Attachment 2 to this letter presents'information'n cable tes'ting of Cerro and Continental instrument cable.'ery truly yours, JT:em ing a t ice President Sworn and subscribed to before m
on this ~ day of March 1978 in New York County, New York CC-(attached)
Notary Public A HL L+ 1'RY NOTARY VUIILIC, Stoto oi Now York I'~o. 41-4606 i92 Qrrofitied in 4ooenS Corrnty Cortificeto tiled in Now York C t
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Question 1
Mechanical En i:n'eery;ng Branch For the steam generator vertical support columns:
a)
Provide "Faulted Condition" design criteria utilized to provide'ssurance against column buckling under the loading resulting from the worst case postulated steam line break, i.e,, at the steam generator outlet nozzle.
b)
Provide calculated stresses for the columns in a) resulting from compressive/buckling load effect of this postulated break.
a)
The AISC-69 formulae used to establish allowable buckling load for the faulted condition are the same as those used for the normal, upset -and emergency conditions.
These equations are used directly without adjustments for normal and upset conditions, whereas factors 1.5 and 1.67 are applied to obtain allnwahles for the.emergency and fauled conditions, respectively.
b)
For the steam generator vert';cal support columns, the calculated stress in the column is 41.3 KSI. The allowable stress in the column is given by AISC Equation 1.5-1 times the faulted condition factor of 1.67 and was found to be 43.5 KSI.
- Thus, the actual stress is
- 95% of allowable.
Question 2
Mechanica'1 Zn in'eeri'n Branch Describe or demonstrate how material properties at operating temperatures of the steam generator supports were taken into account in designing the supports to AISC-69 standard, Res onse AISC-69 does not address properties of materials at elevated temperatures.
- Hence, these effects were not discussed in our
'previous submittal of February, 1978.
However, studies were performed to investigate elevated temperature effects on material properties based on tests<1).published by U.S, Steel which demonstrate a loss of tensile strength of approximately 7% for temperatures of 500oF to 600 F.
(1 R.L. Brockenbrough, B.
G. Johnstone, "USS Steel Design Manual, United States Steel Corporation, Pittsburgh, Pennsylvania July, 1968
1
Question 2
Mechanical Engineerin Branch Res onse (Cont'd)
The original design load capacity for the steam generator upper lateral support.
was 3000 kips.
For this load a maximum extreme fiber stress in the belly band was found to be 46 KSl or 8% less than the yield stress at room temperature.
For the 2858 kip load corresponding to the governing main steam line break load combination, a stress of 43.8 KSl was calculated, thus, achieving a margin of l2$, compared to the yield stress at, room temperature.
This margin more than offsets the decrease in strength at operating temperature.
The lower support columns have a normal operating temperature of 120 F.
- Hence, the properties are 'insignificantly changed by operating temperature effects.
The lower latexal support design is not governed by the main steam line break accident.
Question 3 -Mechanical Engineering Branch Provide assurance that under the postulated main steam pipe breaks resulting stresses in unbroken piping attached to the steam generator do not exceed the Faulted Condition Stress limits (Service Limit D) per ASME Section III of the Boiler and Pressure Vessel Code.
Res onse D.C. Cook Unit 2 was designed under the rules of ANSI B31.1 and not ASME III.
- However, a steam line break was considered in the design analysis.
The calculated (loop) stresses for the steam line break were" quite low and are well within the allowables for ASME Level D stress limits.
Question 4
Mechanical Engineering Branch Provide additional information or the results of analysis that demonstrates that utilizing the loads derived from the nine node cavity model to analyze for support integrity indeed envelopes, i.e.,
assures that worst case loading combination has been accounted for, any similar analysis performed with loads derived from the 17 node cavity model.
R~es onse A.
Summary of steam generator upper support loads (in Kips):
Item Fx Fz' node model (as previously trans-mitted.
2233 445 17 node model Qt=
.00991 sec 1639 253 17 node model Qt=
.01477 sec. 1617 362 17 node model Qt=
.01619 sec'521 351 Therefore it is concluded that the previous calculations using the 9 node model were more conservative as previously stated and do represent the upper bound steam generator support loads for postulated steam line breaks.
Question 1
Structural En ineerin Branch The calculation for the section factor of safety is based purely on moment and neglects the tensile forces in the section.
Your factor of safety cal-culation should be computed on the basis of the stresses in the reinforcing steel.
~Res onse We have recalculated the section factor of safety based upon the stresses in the reinforcing steel.
In no case does the section factor of safety fall below 1.5.
Question 2
Structural En ineerin Branch The allowable shear capacity should consider the effects of membrane tension per EQ 11-8 of ACI 318-71 (Section 11-4.4).
The applicant has used the equation as stated in Section 11.4.3 of ACI 318-71 which is for members subjected to axial compression.
Provide just-ification for your approach; show that such a deviation is not significant to the functional requirement of the wall.
~Res onse The Donald C.
Cook Nuclear Plant was designed using the design criteria of ACI 318-63 code.
The steam generator enclosure was reanalyzed for the new loading conditions using the design criteria of this code.
At the request of the NRC we have investigated the sections using the shear criteria of the ACI 318-71 code.
Under this new criteria one section has a factor of safety in shear below 1.5.
The location and stress mode is as follows:
Location T
e of Stress F.S; 3-2 (Perimeter Wall)
Hoop Shear 1.48 However, the factors of safety are computed based on a concrete 28 day design strength of 3500 psi.
The actual minimum 28 day concrete strength for the steam generator enclosure area is 4400 psi and the actual minimum 90 day strength is 5800 psi.
~Res onse Using the actual concrete 28 day and 90 day strengths, yields safety factors as follows:
Location 3-2
~28 Da s l.66
~90 Da s 1.90
REQUEST FOR ADDITIONAL INFORMATION The following provides additional information regarding environmental qualification'f'nstrumentation cable which has been requested by the Staff by a letter dated March 3, 1978 from Mr. Karl Kniel to Mr. John Tillinghast.
Cerro Wire a Cable Instrumentation Cable The documentation for this cable identifies tests performed on single conductor No.
12 AWG copper wire with 30 mils of crosslinked polyethelene insulation.
The test profile exceeds the worst anticipated accident conditions for both radiation and temperature.
The test conductor demonstrated more than adequate physical and electrical integrity following the test.
The cable supplied to the D. C.
Cook Plant is an instru-mentation cable consisting of 4 conductors.
Each conductor is, a No.
16 AWG copper wire with 30 mils of crosslinked polyethelene insulation.
The insulation material and thickness is identical to the.samples tested.
The 4 wires are grouped together and wrapped with a shield consisting of 2 mil copper tape backed by mylar.
A single No.
18 AWG (minimum) wire is added in contact with the copper shield to maintain shield continuity.
The shielded conductor bundle is then covered with a 45 mil jacket of hypalon (chlorosulphonated polyethelene) insulation for mechanical protection and electrical isolation of the shield.
The testing done on the single insulated conductor demonstrated the adequacy of the cable used in the containment.
The individual wires of the cable can withstand the containment environment without the additional protection provided by the shield and hypalon jacket.
Hypalon insulation and jacketing materials have demonstrated excellent resistance to containment environment and offer excellent protection from the containment environment to'he individual wires which are capable of withstanding the con-tainment environment without, this additional protection.
Continental Wire 6 Cable Co
. Instrumentation Cable The documentation for the instrumentation cable supplied by Continental Wire 6 Cable covered tests performed on samples of the insulating material subjected to a test profile which exceeds the worst anticipated accident conditions inside the containment.
These tests were not conducted on cable samples.
However, the only material subject to modification of physical and electrical properties is the insulating materials.
The copper wire remains
~
essentially unchanged except for minor physical property changes through the temperature excursions and other environmental exposures which occur in the worst anticipated accident condition.
The tests on the insulating material samples resulted in minor changes in the physical and electrical properties of the test samples but also demonstrates'their suitability for continued use following the environmental test.
Testing of the finished cable was done during the instru-mentation splice qualification tests performed in November 1977 at Conax Corp in Buffalo, New York.
Samples of Continental Wire
& Cable twisted shielded quad (4 insulated wires plus shield) on hand at the D. C.
Cook Plant and identical to the cable used in the containment were used to make the splices to the electrical penetration feedthroughs and were subjected to both test profiles at that time.
One set of samples was subjected to 250 F steam for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> followed by immersion in sodium borated water at a minimum temperature of 190 F for 194 hours0.00225 days <br />0.0539 hours <br />3.207672e-4 weeks <br />7.3817e-5 months <br />.
The same test samples were then subjected to a test. profile of 340 F steam for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 250 F steam for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> followed by an additional immersion in sodium borated water at a minimum temperature of 190 F for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
The second set of samples was subjected to a test profile of 340 F steam for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 250 F steam for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> followed by immersion in sodium borated water at 190~F or above for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
In all of the above tests no evidence of failure or serious degradation of the physical or electrical properties of the cables was detected.
Following both tests the cable splices and electrical penetration feedthrough passed high potential tests equal to that required for new cable.
The above tests are identified as IPS-316, IPS-317 and IPS-319.
Documentation reports of these tests identified as IPS-326, IPS-327 and IPS-329, respectively are in the possession of D.
W. Hayes of the Region III Office in Chicago.