ML19269D540
| ML19269D540 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/29/1979 |
| From: | Utley E CAROLINA POWER & LIGHT CO. |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| GD-79-01401, GD-79-1401, NUDOCS 7906040221 | |
| Download: ML19269D540 (23) | |
Text
.'
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@ PAL Carolina Power & Light Company File: NG-3514(B)
May 29, 1979 SERIAL: CD-79-1401 Office of Nuclear Reactor Regulation ATTENTION:
Mr.
T... A. Ippolito, Chief Operating Reactors Branch No. 3 United States Nuclear Regulatory Commission Washington, D. C.
20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS.
50-325 and 50-324 LICENSE NOS. DPR-71 AND DPR-62 SEISMIC ANALYSIS OF SAFETY-RELATED PIPING
Dear Mr. Ippolito:
At our meeting on May 21, 1979, Carolina Power and Light Company committed to provide the NRC Staff additional information concerning our response to IE Bulletin 79-07 on seismic pipe stress analysis.
On May 23 and 24, 1979, the Staff identified to us, by telephone, and to a represent.ctive of United Engineers and Constructors, our architect engineer for the Brunswick Steam Electric Plant, several additional items that should be addressed in our respo nse.
The remainder of this letter and attachments respond to those requests.
1.
The analysis of the loads for the pipe supports for the first ten (10) lines reanalyzed for pipe stresses has shown that there were ten cases where the load exceeded allowable. Table 1-1 summarizes the data on the 98 pipe supports on these ten (10) lines. Table 1-2 presents the details of the ten (10) supports that were overstressed.
While evaluating these ten pipe supports, it was determined that the supports had been underdesigned initially.
In no case was the overstressed condition a result of the new load from the seismic reanalysis. As shown on Table 1-2, the new load actually decreased in five cases, increased less than 2.5% in four cases, and increased 19% in only one case (which was already over capacity by 14.5%). These ten supports were analyzed to datermine if their structural integrity would be maintained under the identified loads. Four of. these supports were found to maintain stresses less than yield and thus would maintain structural integrity.
When it was determined that structural integrity would be compromised for the six supports under the calculated loads, Carolina Power & Light Company decided to shut down both units and make necessary modifications to these supports to reduce stresses to less than allowable. These modifications have been initiated and the new capacity is shown on Table 1-2.
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l 2259 345
File: NG-3514(B) SERIAL: CD-79-1401 During this evaluation, it was noted that the overloaded pipe supports failed in two ways: either concrete anchors or in torsion. An investigation was begun to look at all pipe supports on safety related systems to determine if similar overloaded conditions may exist under the original load. The results of this investigation will be available on June 1, and all necessary modifications will be made prior to returning the units to operation.
2.
On May 24, 1979, the Staff informed vs by telephone that the seismic stress analysis should be based on absolute sum if a two-dimensional seismic analysis was ured, and that the square root of the sus of the squares (SRSS) was acceptable if a three-dimensional seismic analysis was made.
The Staff further stated that a stress from a two-dimensional analysis calculated using SRSS and multiplied by a factor of 1.38 would be acceptable. At the time that BSEP was licensed, two-dimensional SRSS seismic analysis was acceptable criteria, and it is not apparent to us that the back-fit of a two-dimensional absolute sum seismic analysis has undergone the necessary requirements of 10CFR50.109.
Although CP&L does not accept the Staff's position, we have prepared a revision to Table 2 of our letter of May 21 demonstrating the effect of multiplying the two-dimensional analysis results by the 1.38 f actor. We have also taken credit for conservatism that exists in the relationship between the OBE and the DBE.
The results of this exercise show that only one line of the first thirty-nine reanalyzed lines exceeds total allowable stress by 2%.
For this line, the total stress is still less than 0.9S.
y For the unreanalyzed lines shown in Attachment 3 of our May 21 letter (GD-79-1342), we have used the 1.38 factor to establish criteria for priority of lines to be reanalyzed. We do not plan to base our conclusions of acceptability on the use of the 1.38 f actor, since it is not the appropriate criteria for BSEP.
In determining the criteria for priority of reanalysis of the remaining lines, SRSS stresses were estimated on the basis of a factor of 1.5 increase, and this resultant was then multiplied by 1.38.
Credit for the conservatism of the OBE/DBE relationship was taken into account prior to applying the 1.5 increase. When this was applied to the 411 lines that have not been reanalyzed, 39 of the 411 exceeded allowable stress, and are tabulated in Attachment 2 to this letter.
Our reanalysis priorities have been changed to include these 39 lines in those to be reanalyzed the week of May 28, and the results of this reanalysis should be available on June 1, 1979. We still anticipate completing the total reanalysis in accordance with our previously stated completion date of July 21, 1979.
2259 j46
Mr. T. A. Ippolito May 29, 1979 3.
As a result of an I & E inspection at the Brunswick Steam Electric Plant to verify that the as-built dimensions were the same as the as-designed (as-analyzed) system, four deviations were noted. These are discussed in Attachment 3.
As stated in the meeting on May 21, 1979, and confirmed in our letter of May 22, 1979, Carolina Power & Light Company will verify as-built dimensions for all safety related systems at BSEP.
This verification is currently in progress for those lines outside' containment.
The lines inside containment will be verified at the next scheduled outage.
Due to the time constraints on reanalysis, the reanalysis is being conducted concurrently with the as-built verification.
If any discrepancies are identified between the as-built /as-analyzed configurations, an evaluation by a stress analyst will be made to determine if the line should be reanalyzed. This evaluation will be based on evaluating the magnitude of the computed stresses for the area in question, and the impact (increase or decrease) on the stresses expected for such deviation.
If it is determined that the line needs to be reannlyzed to determine the new stress level, we will promptly reanalyze the line.
4.
During our recent meetings, the relationship of IE Bulletins 79-02 and 79-07 has been discussed.
Some of the pipe supports analyzed in the first ten lines are anchored using concrete expansion anchors discussed in Bulletin 79-02.
In the 79-07 support reanalysis, these base plates were and will continue to be analyzed using IE Bulletin 79-02 as a guide.
The capacity established for the concrete anchors is 20% of the manufacturer's rated capacity. Using this criteria, two supports on the first ten lines had to be redesigned and now have sufficient capacity. As stated in item 1 above, the remaining supports using concrete expansion anchors are being investigated to determine their adequacy and will be reported on June 1.
A final report on all of our analyses and testing related to the concrete expansion anchors and IE Bulletin 79-02 will be submitted in compliance with the bulletin schedule.
5.
The Staff requested information on the location of the postulated pipe rupture for a LOCA relative to the point of highest stress. The BSEP piping design did not use the mechanistic approach of locating the pipe break at the point of highest stress.
The postulated break for doubled-ended guillotine or longitudinal split was analyzed for the pipe break to occur at any point on the pipe, inside or outside containment.
6.
We have been informed that during a meeting between NRC, another licensee and United Engineers & Constructors (UE&C),
some questions were raised by the RRC staff about the subject 2259 a47
Mr. T. A. Ippolito May 29, 1979 of valve operability.
In the event the staff may have any questions concerning this topic as it may apply to BSEP, we will be prepared to address this issue.
7.
Carolina Power & Light Company's criteria for determining if an overstressed condition is reportable is set forth below:
a.
Lines Yet To Be Reanalyzed The stress using new seismic data and revised analytical criteria are estimated for the lines that are yet to be reanalyzed. As stated previously, those with high estimated stresses are being analyzed first in the reanalysis program. We will not use estimated stress as a basis for determining overstressed conditions which are re po rtable.
b.
Reanalyzed Lines Those lines which have been reanalyzed and which show an apparent overstress condition will be evaluated in detail to determine if it is a reportable item. First, the known conservatisms will be removed from the analysis.
The line will be analyzed to determine if the stress at any single modal point exceeds FSAR criteria of 0.9S o r 1. 8 S e y
h whichever is the higher.
If the pipe remains overstressed, tLis will then be considered a reportable item and the NRC v. be informed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
Reanalyzed Pipe Supports When the reanalyzed pipe data is available, the pipe supports will be reanalyzed for the revised load.
If the lotd exceeds the apparant support capacity, the specific support will be analyzed in detail to determine if the stated capacity is the actual capacity without exceeding 0.9 S.
If the load still exceeds the capacity, a y
determination is made if the support will maintain structural integrity even if the allowable is exceeded.
If structural integrity is maintained, this is not considered reportable.
If structural integrity is not maintained, the support is taken out of the computer piping configuration, and the line is reanalyzed.
The results of this reanalysis are evaluated to determine if other supports and the pipe can take the additional lcad without exceeding their structural integrity.
If the system maintains integrity, the item is not reportable.
If the system does not maintain structural integrity, the item will be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I 2259 ;48
Mr. T. A. Ippolito May 29, 1979 In summary, CP&L has evaluated the data from the lines reanalyzed to date, and the estimates for revised stresses for lines yet to be reanalyzed, and it is our conclusion that the continued operation of the Brunswick Steam Electric Plant, Units 1 & 2 is warranted without undue risk to the public health and safety, while the reanalyses of seismic design continues. The problem associated with those supports that were found to be overstressed is a result of initial underdesign of those supports, and is not related to the use of algebraic, square root sum of the square, or absolute summation of seismic stresses.
The modifications of those supports which were originally under-designed will be completed in early June, and at that time, both units will be returned to power. As stated in our letter of May 15,1979, and in item 7 of this letter, 24-hour reporting criteria have been established if any piping or supports are determined to be overstressed during the reanalyses.
If you have any questions concerning this information, please do not hesitate to contact our staff.
Yours very truly,
/
E. E. Utley I
Executive Vice President Power Supply DLB/sg bec: Messrs.
D. L. Bensinger C. S. Bohanan D. B. Waters / File NG/3514(B)
J. M. Johnson W. B. Kincaid S. McManus A. C. Tollison, Jr.
C. W. Woods (LIS)
File: BC/A-4 File: B-X-0274 2259 49
GD-79-1401 ATTACHMENT 1 PIPE SUPPORT ANALYSIS An evaluation was performed on the pipe-supports of the first ten.
lines that were reanalyzed in the seismic pipe stress reanalysis program. There are 98 pipe supports made up of snubbers, vendor catalog pipe supports, and fabricated supports. The recalculated loads compared to the original load and support structural capacity are tabulated on Table 1-1.
As can be seen on Table 1-1, the load did not increase appreciably due to the seismic stress reanalyses and recalculation of loads.
The load decreased for 30% of the supports and increased less than 25% for 60% of the supports. The load increased greater than 25%
for only nine supports, but the new loads were less than 75% of. capacity for these supports.
However, ten supports were found where the load exceeded the applicable allowable. Further investigation revealed that these ten supports were underdesigned initially. For these ten supports, the new loads were less than the old loads in five cases, increased less than 2.5%
in four cases, and in only one case, the increase was 19%.
These ten supports were analyzed in detail to determine if they would maintain their structural integrity under the specified loads even if they exceeded allowable. This is summarized on Table 1-2.
In four cases, including the one where the new load was 19% higher than the old load, the supports retained their structural integrity. Six supports would fail.
The six supports that would fail under the specified load (old or new) were redesigned to have-their stresses less than allowable. The new design loads for these pipe supports are shown on Table 1-2.
It has been concluded from the analysis of 98 pipe supports that the seismic stress reanalysis does not contribute to overloaded pipe supports.
However, it has been recognized that there is a potential for certain supports to be overloaded due to an error in the initial design.
These errors have been found to be with concrete expansion anchors and with torsion of the beam support. An investigation has begun to examine the pipe supports of the other safety-related piping for similar problems. The results will be reported at a later date.
2259 s50
4 TABLE 1-1
SUMMARY
OF PIPE SUPPORT IDADS UPSET IDAD CAPACITY OLD IDAD NEW IDAD RATIO RATIO LINE POINT LIMITING PART C
(OL)
(L)
L/0L L/C REMARKS 17 65S Strut.
13800 7037 7326 1.04
.53 Cat. P.S.
953 Cat. P.S.
15700 9129 9311 1.01
.59 110S Snubber 3920 1134 1430 1.26
.36 137S Cat. P.S.
4960 4192 4286 1.02
.86 172S S.S. Sup't.
1836 606 595 0.98
.32 175S Cat. P.S.
11630 6329 6527 1.03
.56 220S X Cat. P.S. Strut 3920 968 1176 1.21
.30 Y Cat. Sup't.
6230 5544 5L88 1.007
.90 Z Cat. P.S.
3920 1483 1634 1.10
.42 255S Cat. P.S.
8000 4088 4 60 1.01
.52 402S Snubber 3020 411 576 1.40
.15 24 136 XZ Snubber 3920 1339 2690 2.00 68 Y Snubber 3920 1352 1400 1.03
.36 154 X Snubber 3920
+1576
+1243 0.09 36 Z Snubber 3920
+1576
+1243 0.09
.36 ISB 61S Z Snubber 3920 956 1078#
1.12
.27
's 61S Y Snubber 3920 1381 1912 1.38
.49 105S Snubber 3920 1644 2133 1.29
.54 18S Wald 15300 11850 11466 0.96
.75 60S Snubber 22177 11837 11578 0.97
.52 71S Snubber 13666 7993 8392 1.04
.61 107S Snubber 13800 3187 3196 1.002
.23 73S Snubber 37600 25612 23214 0.90
.62 110S X Snubber 29090 5239 4700 0.89
.16 110s Z Snubber 20736 14316 13468 0.94
.65 2259 351
. tab 4E 1-1 (Cont'd)
UPSET IDAD CAPACITY OLD IDAD LEV IDAD RATIO BATIO IINE POINI LTMITING PART C
(OL)
(L)
L/0L L/C REMARKS 237 420 Cone. Anchors 678 4399 4014 0.91 (5.92) See Table 1-2 420 Y-RSSA-20 Strut 20000 11124 13311 0.11
.67 440 X-EX.W8x17 790 2784 2490
' O.89 (3.15)
(Torsion) 440
- -Strut RSSA-10 10000 5641 5792 1.02
.58
~
466 Snubber 3920 1884 1945 1.03
.50 472 Weld-Post Ex.St1 5374 4001 4551 1.13
.85 484 Snubber 13800 3546 3570 1.006
.26 503 Snubber 13800 3339 3539 1.05
.27 525 Weld-Post to Snub 6000 4615 4884 1.05
.81 122 2092 Snubber 3920 2117 2101 0.99
.54 2094 Snubber 23600 3818 3827 1.002
.16 2220 Snubber 13800 9008 9048 1.004
.66 2143 Snubber 13800 3674 5670 0.99
.41 2156 X -W6 x 15.5 660 2726 1775 0.65 (2.69) Sec Table 1-2 (Torsion) 2156 Z-Snubber 23600 1521 1504 0.98
.06 2230 Snubber 13800 4804 4182 0.87
.30 2240 Snubber 13800 6179 3685 0.59
,. 27 2174 W6 1 15.5(Torsion) 660 6714 6874 1.02 (10.42) See Table 1-2 2062 Snubber 13800 3044 3046 1.00
.22 121 3084 Conc. Anchors 840 2970 2991 1.007 (3.56) See Table 1-2 3083 Conc. Anchors 12960 3866 3207 0.82
.25 3067 Snubber 13800 6190 6244 1.008
.46 3066 Snubber 13800 8152 8010 0.98
.59 305S Clamp 11500 10798 9143 0.84
.79 (55) 2259 352
TABLE 1-1 (Cont'd)
UPSET IDAD CAPACITY OLD IDAD NEW IDAD RATIO RATIO LINE POINT LIMITING PART C
(OL)
(L)
L/0L L/C REMARKS 3048 Snubber 3920 2843 3214 1.13
.83 L/C =.95 for Emergency Condition Snubber 13800 3302 3405 1.03
.24 3200 Snubber 3920 712 1026 1.44
.26 s
6 13S Snubber 37600 31000 18000 0.58
.48 133S Snubber 23600 14499 13400 0.92
.57 232/
Fab. Sup't 7460 3341 5463 1.63
.73 233 16 27S Snubber 13800 10189 9377 0.93
.68 800S Snubber 13800 4628 5752 1.24
.42 806S Snubber 13800 6099 6863 1.12
.50 103S Snubber 13800 5106 8402 1.64
.61 719S Struct. Supt.
8800 11005 8600 0.78
.98 718S Snubber 13800 7383 5753 0.77
.42 24S St.ubber 13800 7651 6137 0.80
.44 725S Snubber 3920 1200 1249 1.04
.32 710S Snubber 3920 2114 2423 1.14
.62 4
724S Y Struct. Supt.
1000 3406 4112 1.20
.41 Z Snubber 3920 1601 1823 1.13
.47 901S Snubber 3920 2U4 2566 1.19
.63 900S
. Snubber 3920 1818 2160 1.18
.55 722S Struct. Supt.
12200 2826 3442 1.21
.28 108S Cat. Pipe Supt.
8900 7144 7693 1.07
.86 510 133 Struct. Steel 3563 1948 1961 1.006
.55 Supt.
125 Wald 3712 3361 3505 1.04
.94 132 Strue. Steel 1350 1546 1844 1.19 (1.36) See Table 1-2 Supt. Channel 2259 353
- TABLE 1-1 (Cont'd)
UPSET IDAD CAPACITY OLD IDAD NEW IDAD RATIO RATIO LINE POINT LIMITING PART C
(OL)
(L)
L/OL L/C REMARKS 195 Strue. Steel 27600 7049 7126 1.01
.26 Supt.
148 Strue. Steel 6040 2316 2276 0.98
.38 Supt.
120 Strue. Steel 9645 5510 5498 0.99
.57 Supt.
111 Bolts 1445 77 99 1.28
.02 137 Strue. Steel 3897 2265 2264 0.99
.59 Supt.
16 Bolts 5038 5601 5456 0.97 (1.08) See Table 102 230 Strue. Steel 3770 3615 3473 1.01
.93 Supt.
224 Strue. Steel 12992 11140 11311 1.01
.87 Supt.
225 Strue.
4960 3207 3211 1.001
.67 Supt.
40 Strue. Steel 5250 5271 5255 0.99 1.0 Supt.
116 Strue. Steel 3920 522 524 1.003
.13 Supt.
113 Strue. Steel 4960 3743 3744 1.00
.76 Supt.
e 30 Snubber 13000 8758 8881 1.01
.68 270 Strue. Steel 11800 18800 18900 1.005 (1.6)
- 136, Strue. Steel 3170 6556 6570 1.002 (2.07)
- 135,
.(Pipe Section) 803S 125 1072S Strue. Steel 4192 4122 4093 0.99
.98 1065S Conc. Andiors 12960 6302 7382 1.25
.61 1051S Snubber 11500 3477 3476 0.99
.30 1120S Snubber 13800 1452 1437 0.98
.10 1140s Cla=p 11500 10085 10488 1.03
.91 2259
.354
. TAB E 1-1 (Cont'd)
UPSET IDAD CAPACITY OLD IDAD NEW IDAD RATIO PATIO LINE POINT LIMITING PART C
(OL)
(L)
L/0L L/C REMARKS 1150S X Beam (Torsion) 790 1470 1301 0.88 (1.65)
Z Snubber.
3920 807 998 1.23
.25 1029S Snubber 13800 9111 9630 1.05
.70 e
2259 155
L:suLy JA
SUMMARY
OF PIPES RITfH L/C >1.0 OLD NEW CAPACITY CAPACITY TO RATIO IDAD IDAD RATIO REDESIGN LINE POINT LIMITING PART (C)
YIELD (Cy)
(NL/C or Cy)
(OL)
(NL)
OL/NL
_ LOAD ##x, 237 420 Conc. Anchors 678
> 1. 0 4399 4014 1.09 Y 13430 x 4192
- 440, 8W x 17 I-Bm 790
> 1. 0 2784 2490 1.11 6762 Torsion 122 2156 6W x 15.5 660
> 1. 0 2726 1775 1.53 2769 I-Bm Torsion 2174 6W x 15.5 660
> 1. 0 6714 6874 0.97 11342 1-Bm Torsion 121 3084 Conc. Anchors 840
> 1. 0 2970 2991 0.99 4582 510 132 SS (Channel) 1350 2194
.84 1546 1844 0.83 16 Bolts 5038 8184
.67 5601 5486 1.02 270 Stru. Steel 11800 19656
.96 18800 18900 0.99 PN) 136, 135 Stru. Steel 3170 6510
- 1.0 6556 6570 0.99 rs) 803S Pipe Section LT1 125 1150S I-Bm Torsion 790
> 1. 0 1470 1301 1.12 2243 t,a Ln Ch
- Cy does not apply
- Not redesigned for short term fix
- Redesi n Load consists of new calculated emergency load x 1.38 + transient o
to be less than AISC allowables (0.67 Sy).
ATTACHMENT 2 SEISMIC PIPE STRESS ANALYSIS CRITERIA As stated previously, the original seismic analysis for pipe stress '
used algebraic summation within each mode. A reanalysis effort was undertaken for all safety-related lines using the UE&C - ADLPIPE-2 Computer Code which employs the square root - sum-of-the-squares (SRSS) load ccabination within each mode.
The results of the reanalyses, given to the NRC Staff in our re-sponses to IE Bulletin 79-07, in letters dated April 24, May 15, and May 21, 1979, used the SRSS method. On May 24, 1979, the NRC Staff notified CP&L that the use of SRSS with a three-dimensional seismic analysis was acceptable, but for a two-dimensional seismic analysis the absolute sum method should be employed within each mode. The analysis for Brunswick uses a two-dimensional seismic input approach.
At the time BSEP was licensed, the two-dimensional SRSS analysis was the acceptable criteria. Therefore, the acceptability of stress levels should not be based on absolute sum. However, to use the most conservative case for comparison purposes only, the stresses calculated using UE&C - ADLPIPE-2 were multiplied by 1.38 (a number acceptable to the NRC Staff) to obtain stresses for the Operating Basis Earthquake (OBE).
As discussed in Attachment 7 of our letter to the NRC GD-79-1342, dated May 21, 1979, the previous seismic analysis used a most conversative approach of relating stresses for an OBE to that for a Design Basis Earthquake (DBE), known today as a Safe Shutdown Earthquake (SSE).
The stresses computed in the OBE were multiplied by 2 and used as the stresses for a DBE. As discussed on May 21, 1979 with the NRC Staff, our reevaluation of the OBE and DBE Amplified Response Spectra (ARS) indicates that the relationship between the two ARS in the frequency range that affects pipe stress is less than 1.2, and frequently less than 1.0.
However, a value of 1.2 has been selected for use to convert OBE stresses to DBE stresses.
For the thirty-nine lines already reanalyzed, the conversative stresses for comparison purposes for the DBE and total are shown onyble 2-1.
The DBE stresses in this table are calculated as follows: V DBE =
(7" OBE x 1.38 x 1.2, where CI~0BE is obtcined using the UE&C - ADLPIPE-2.
For the lines yet to be reanalyzed, the stress for a DBE was estimated as explained in Attachment 7 to our May 21, 1979 letter using a factor of 1.5 to account for the highest expected increase in stress due to the reanalysis for SRSS (within each mode) in lieu of algebraic sum (within each mode) and which is based on the data from the reanalyzed lines. For those lines identified on Attachment 3 to our May 21, 1979 letter, the stress for a DBE were estimated as follows:
(]DBE (I OBE x 1.38 x 1.2 x 1.5
=
Est.
- Orig, where C3 0BE was computed in the original analysis. Those lines whose Orig.
estimated stresses exceeded allowable are tabulated on Table 2-2.
2259 357
. EVALUATION As can be seen from Table 2-1, one line (RHR-60, Residual Heat Removal) exceeds the allowable (1.8 S ) by 1.7 percent.- However, this stress is h
less than the stress equal to 0.9 Sy (32,400). The BSEP FSAR allows the use of 0.9 Sy or 1.8 S, whichever is greater, as the allowable -
h stress during emergency condition (DBE). Therefore, the stresses are acceptable for all lines reanalyzed.
Table 2-2 shows that 39 of 411 lines yet to be reanalyzed exceed allow-able (1.8 S ).
These stress values are not necessarily based on coinci-h dent point maximums, but rather the summation of maximum stresses for each individual loading. It should be restated that these stresses are estimated and that they were derived using a conservative factor of 1.5 to cover the maximum increase expected for the reanlaysis (old algebraic to new SRSS combination within each mode). As discussed in Attachment 7 and shown on Attachment 8 of our May 21, 1979 letter, in over 58% of the lines already reanalyzed, the new seismic stress was,less than the original seismic stress. In over 87% of the cases, the new stresses were less than 1.25 of the original stresses.
As discussed previously in our letter, Carolina Power & Light Company commits to placing these lines in the highest reanalysis priority category, regardless of the priority category previously established on a function and size basis.
It should also be pointed out that of the 39 lines estimated to be overstressed, 27 are 2" or less in diameter.
2259 358
~
ATTAC11MENr 2 PI?E STRESS REEVAlllATION
SUMMARY
EMERGENCY CONDITION (PSI) 1 TOTAL LINE SIZE ORIGINAL ORIGINAL TOTAL SEISMIC TOTAL SEISHIC STRESS SYSTEM NAME ISO NO.
(NPS)
TOTAL SEISMIC 5/21/79 5/21/79 5/25/79 5/25/79 ALIMABLE ALIMABIE l
Hain Steam MS-ISB 24 10724 3942 10640 3858 9976 3194 27000 37 Safety / Relief SINL-121 10, 6 23012 12280 21910 11180 19987 9257 27000 74 Valve Safety / Relief SRVL-122 10, 6 19685 15800 24439 13352 22143 11056 27000 82 Valve Safety / Relief SRVL-237 10, 6 20432 12004 24588 16160 21809 13381 27000 81 Valve Safety / Relief SRVL-125 10, 6 24270 13347 24316 20270 20830 16784 27000 77 Valve Feedwater W-16 18, 12 18007 12420 20028 13296 17741 11009 27000 66 Residual IIcat RllR-6 20 19406 13582 12644 6820 11471 5647 27000 42 Removal Core Spray CS-24 10 16952 10076 14366 7490 13078 6202 27000 48 liigh Press Cool llPCIS-17 14 12200 6446 12502 6748 11341 5587 27000 42 Inice liigh Press Cool IIPCIS-510 14, 12, 10 12004 7994 12092 8082 10702 6692 27000 40 Inlet liigh Press Cool llPCIS-10 14, 12, 10 9733 3886 11152 5530 10201 4579 27000 38 i Inlet Residual lleat RllR-1 24, 20 24094 18584 17972 14366 15501 11895 27000 57 Removal Residual lleat RilR-2 20, 16, 12 13309 7654 11471 5948 10448 4925 27000 39 Removal Residual IIeat RllR-5 24 9848 3896 9514 2960 9005 2451 27000 33 Removal Residual lleat RllR-25 4, 6 18558 12904 18530 12876 16315 10661 27000 60 Removal Nuclear Steam NSS-14 24, 10 14745 8446 16335 10036 14609 8310 27000 54 Supply Safety / Relief S RVL-124 6, 10 25536 15928 25984 16376 23167 13559 27000 86 Thlve N
(.J1 4
u
( J1 D
.~
ATTACllMEttr 2 (CONr'D)
EMERGENCY CONDITION (PSI)
TOTAL LINE SIZE ORIGINAL ORIGINAL TOTAL SEISMIC TOTAL SEISMIC STRESS SYSTDI NAME ISO NO.
(NPS)
TOTAL SEISMIC 5/21/79 5/21/79 5/25/79 5/25/79 ALIMABLE ALIMABLE Safety / Relief S RVL-126 6, 10 23361 18000 22197 17422 19200 14425 27000 71 Valve Residual lleat RilR-52 14, 12 23271 17936 19539 14204 17096 11761 27000 63 Removal Reactor Core RCIC-21 3
7603 3588 7601 3586 6982 2967 27000 26 Isolat. Cool Resid IIcat Rem EllR-173-B 1h 3808 2186 3838 2216 3457 1835 27000 13 Drain Line lResidualllcat RilR-28 20, 16, 12 15298 8814 13626 7142 12398 5914 27000 46 Removal
! Nuclear Steam NSS-15' 24 10974 4420 10458 3904 9787 3233 27000 36 System
' Nuclear Steam NSS-120 10, 6 19443 8902 17899 8298 16472 6871 27000 61 System (ISC) i
! liigli Press Cool llPCIS-4 3, 6, 10, 12 23609 20876 25481 22748 21568 18835 27000 80 Inict l Nuclear Steam NSS-123 6, 10 21027 16098 18577 16782 15691 13896 27000 58 fSystem (15C)
- Huclear Steam NSS-187 10, 6 21856 11337 23596 l
16424 20771 13599 27000 77 l System (15C)
' Residual IIcat RilR-42 12, 14 18116 12976 17480 12340 15358 10218 27000 57 l Removal l Residual IIeat Rl!R-3 14, 16, 2Q, 25317 18328 23379 15590 20698 12909 27000 77 I Removal 24 Residual lleat RilR-13 4, 8, 14 12532 10018 12620 10105 10882 8368 27000 40 Removal Residual IIcat RllR-59 4, 6, 10 12664 9970 13344 10650 11512 8818 27000 43 Removal Residual IIcat RliR-60 4, 6, 10 34618 33658 32971 32012 27465 26506 27000 102 Removal Residual lleat RilR-168 1
23580 22658 26910 26198 22404 21692 27000 83 Removal hesidual lleat RilR-61 4, 6, 3/4 21117 17038 19393 15314 16759 12680 27000 62 Removal La CD
ATTAC!! MENT 2 (CONT'D)
I i
EMERGENCY CONDITION (PSI) l TOTAL LINE SIZE ORIGINAL ORIGINAL TOTAL SEISMIC TOTAL SEISMIC STRESS SYSTEM NAME ISO NO.
(NPS)
TOTAL SEISMIC 5/21/79 5/21/79 5/25/79 5/25/79 ALIMABLE ALIIXiABIE liigh Pressure IIPCI-11 16, 14, 6 21386 19790 17214 15618 14528 12932 27000 54 Coolant In-iunction
- Reactor Core Injunction RCIC-196 1, 3/4 22706 19504 22504 19302 19184 15982 27000 71 j cooling
' Residual lleat RIIR-41 3, 4 26802 20723 26824 20750 23255 17181 27000 86 Removal Residual lleat RllR-199 4, 1, li,
23410 20242 23398 20230 19918 16750 27000 74 f
Removal 3/4
, Reactor Core RCIC-194 2, 14, 1 24519 24118 24559 24156 20404 20001 27000 76 lIsol, cooling I
I I
I I
- Seismic stresses shown are obtained by multiplying the OBE Seismic Stresses by 2.
- Total stress (5/25/79) are based on:
i N
N (EBE x 1.2) 1.38 + (Total Stress - Seismic) 2 5/21/79 5/21/79 0
O e
0 e%
ATTAC1DIENr 2 - TABIE 2-2 IDCATION EMERGENCY CONDITION STRESS (PSI)
ISO /
INS. OR TOTAL SEISMIC TOTAL PROB.
SilEET OUTSIDE LINE STRESS (DBE)
TOTAL SEISHIC ALIDWABLE STRESS /
NO.
SYSTFM NO_
COPfr _
9T7F 5/?1/74 5/91/79 5/?5/74 5/95/74 (1_ R sg ALIDWABIE 2
Primary Steam Condensate 128 In 2
24225 20022 29070 24867 27000 108 Drain Inside Dry Well (East) and (West) 32 liigh Pressure Coolant 152 Out 1%
23242 22760 28750 28268 27000 106 Inj. (Main Pump to Barometer Cond.)
34 Iligh Pressure Coolant 154 Out 3/4 23191 20780 28220 25809 27000 105 Inj. (Misc. Vents &
Drains Booster Pump) t 35 liigh Pressure Coolant 155 out 2
23245 20990 28325 26070 27000 105 Inj. (Turbine Exh.)
Out 2
23346 21800 28622 27076 27000 106 Out 23762 23476 29443 29157 27000 109 38 liigh Pressure Coolant 158 Out 1
27538 25060 33602 31124 27000 124 Inj. (Misc. Vent, Test Out 3/4 24715 22760 30223 28268 27000 112
& Drains Lines)
Out 3/4 25923 22010 31249 27336 27000 116 l
46 Core Spray System (C.S.
39 Out 3
29154 26788 35636 33270 27000 132 Min. Flow By-Pass Pump 2A)
Out 3
25456 21860 30746 27150 27000 114 48 Core Spray System 105 out 2
25235 22438 30655 27858 27000 114 (RllR Conn. from C.S.1%np 2A)
Core Spray System 105 Out 2
25235 22438 30655 27858 27000 114' (RllR Conn, from C.S. Pump 2B)
N 54 Service Water Salt Water 82 Out.
20 29536 25810 35782 32056 27000 133 N
Supply to RilR, Service G
Water Pumps (South)
O O
O N
.~
TABIE 2-2 (Cont'd)
ATTACIIMENr 2 IDCATION EMERGENCf CONDITION STRESS (PSI)
ISO /
INS. OR TOTAL SEISMIC TOTAL PROB.
SIIEET OUTSIDE LINE STRESS (DBE)
TOTAL SEISMIC ALIDWABLE STRESS /
NO.
SYSTEM NO.
CONT.
SIZE 5/21/79 5/21/79 5/25/79 5/25/79 (1.8 S )
AllDWABLE h
55 Service Water System 106 Out 4
26139 25100 32213 31174 27000 119 6" Return Line from Pump Room Cooler 2A 56 Service Water System 107 Out 6
24965 17948 29308 22291 27000 109 6" Supply lleader (South)
Out 4
29907 28212 36734 35039 27000 136 57 Service Water System 108 Out 2
24300 21352 29467 26519 27000 109 6" Supply lleader (North) 70 Reactor Water Clean-up 22 In 6
24303 17134 28449 21280 25940 110 R.W.C.U. Pump Suction 80 cont. Atmospheric control 211 Out 2
26825 24396 32729 30300 27000 121 Valve By-Pass Piping 81 Cont. Atmospheric Control 212 Out 4
23948 21750 29212 27014 27000 108 Vent Purge Line From Drywell i
83 Containment Venting 230' Out 25268 14896 28873 18501 27000 107 89 Instrument Air System 179 Out 2
24598 23348 30248 28998 27000 112 Supply Line (North) West u g,n 93 Instrument Air System 184 In 2
30632 29382 37742 36492 27000 140 Supply lleader (North)
IN) 95 Instrument Air System 189 In 1%
25719 24512 31651 30444 27000 117 Pipe to Accum C) 96 Instrument Air System 190 Out 3/4 30507 29382 37617 36492 27000 139 SuPP y Lines to Filters l
C:3 D-0005 and D-0006 C:3 U
.~
A'rTACllMENT 2 - TABLE 2-2 (Cont'd)
IDCATION EMERGENCY CONDITION STRESS (PSI)
ISO /
INS. OR TOTAL SEISMIC TOTAL PROB.
SliEET OlffSIDE LINE STRESS (DBE)
TOTAL SEISMIC ALIDWABLE STRESS /
NO.
SYSTEM NO.
CONT.
SIZE 5/21/79 5/21/79 5/25/79 5/25/79 (1.8 S )
AlIDWABLE h
t 97 Instrument Air System 181 Out 2
28127 26922 34642 33437 27000 128 Outlet from RCVR at "18R" Supply West Col. "T" 98 Instrument Air Sys. Inner 192 In 2
28344 27094 34900 33650 27000 129 Air Supply 11eader Outer Air Supply lleader 99 Instrument Air System 201 In 3/4 25822 18666 30339 23183 25920 117 Recirc Pump 2B 101 Instrument Piping, Piping 206 In 3/4 24990 22410 30413 27833 27900 109 at Temp. Equalizing D003B 102 Instrument Piping 207 In 3/4 21655 19622 26403 24370 26028 101 Lines 2E21-701 & 702 108 Nitrogen & Off Cas Services 309 Out 3/4 27170 24018 32982 29830 27000 122 Bldg. Instr. Air Interrupt-able 110 RllR 545 out 4
23765 18128 28151 22514 27000 104 113 RllR Drain to lui 548 4
23806 21690 29055 26939 27000 108 116 RilR Pumps 1A & IB 605 2
24752 20878 29804 25930 27000 110 117 Service Water Sys.
606 6
'26321 23760 32071 29510 27000 119
'N125 Instrument Air 690 3/4 26655 25529 32833 31707 27000 121 N
,Cb 129 Cont. Atmos. Control 709 8
15579 11038 29288 13709-27000 108
!O Sys. Sup. Lines 710 C132 Service Water 716 1%
23143 18610 27646 23113 27000 102 O
4
.~
.R.
ATIACHMENT 3 AS-BUILT DRAWINGS As a result of the NRC-I&E walk-through of approximately 67 pipe supports on safety related lines, four discrepancies were identified:
1.
Isometric 17 High Pressure Coolant Injection main pump discharge line above el.18'-9" data point 45 does not agree with piping drawing. Actual location of support is 9'-2" from valve F006 in lieu of 7.0' as shown on the analysis isometric.
~
Coment: h analyzed location has been reviewed by stress analysist and confirmed that the actual placement of the support will have little or no effect on the results of analysis for the following reasons:
1.
The total maximum stress of the line is less than 507. of the code allowable stress. see attachment 2 2.
The placement of the support within approximately two pipe diameters of its analyzed position on this 14" Sch.120 pipe will not adversely effect the analysis.
2.
Isometric 20 Core Spray data point 101 is located approx 1=ately l'-3" closer to valve F015A chan shown on the isometric.
Comment: Review by stress analysist confirms that since the data point is a snubber placing it closer to the valve is better than the original placement. In addition the new placement will have no adverse effect on the stress analysis since the new placement is in the same plane as analyzed.
2260 005 __.-
,./
3.
Isometric 12 Reactor Core Isolation Cooling pump suction lines data point 272 vertical snubber is located on the opposite sice of an elbow than is shown on the analysis isometric.
Coment: Review by stress analysist confirm that placement has no effect on the stress analysis. The analysis program treats the elbow as a point in the model, therefore transfer from one side of an elbow to the other has no effect on the analysis re-suits as long as the snubber acts in the required direction.
Field check of the installation has verified that the snubber is acting in the correct (vertical) direction.
4.
Isometric 18 Core Spray Pump suction line 23, data point 236, is eleven inches closer to pump than shown on the analysis isometric.
Coment: Review by stress analysist confir=s that the location of the support within one pipe diameter will not adversely effect the stress analysis. In addition the support is a sliding dead' weight support which has no effect for seismic support.
It should be noted that in the above cases it has been determined by a stress analysist that there is no adverse impact on the pipe stresses. However, Carolina Power and Light has comitted to perform an as-built verification on all lines included in the reanalysis to increase the confidence that the as analyzed condition is consistent with the as-built condition. Thirty additional supports have been checked by field personnel and no additional problems'have been found.
e 2260 006 ge w e ee ema s-a
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