ML19241C119
| ML19241C119 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/23/1979 |
| From: | Utley E CAROLINA POWER & LIGHT CO. |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| GD-79-1844, NUDOCS 7907270230 | |
| Download: ML19241C119 (23) | |
Text
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- .e 2a :: w _ _.;- :: :;. y July 23, 1979 FILE:
NG-3514(B)
SERIAL:
GD-79-1844 Office of Nuclear Reactor Regulatica Attention:
Mr. T. A. Ip?olito, Chief Operating Reactors Branch Nc. 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:
In early April, Carolina Power & Light Company had discussions with the NRC staff concerning the use of the ADLPIPE comput.r code for analyzing pipe stresses for our P-rswick Steam Electric Flant (BSEP), Units 1 and 2.
This code e= ployed al aic su==atiens in the seismic analysis porrion.
In our April 24, 1979 response to IE Bulletin 79-07, Seismic Analysis of Safety-Related Piping, we cammitted to a total reanalysis of safety-related piping following NRC's acknowledgement that our new approach was satisfactory.
In our meeting on May 16, 1979, we further committed to have this analysis completed by July 21, 1979. We have completed the objectives of our co=mitment of May 16, 1979; however, subsequent commitments have resulted in additional reanalysis requirements which are in progress and which will be completed by August 3, 1979.
To assist in understanding the scope and the sequence of this reanalysis effort, a chronology of meetings and letters is presented in.
The safety-related piping for BSEP is identified on piping isometrics for stress analysis.
In the final aaalysii, 195 isometrics were reanalyzed.
These were arranged in six priority categories for reanalysis, where the larger piping in the pressure boundary and the emergency core cooling systems were analyzed first, as described in our letter of May 16, 1979. We have completed the pipe stress analysis for all six categories and have previously submitted to you the results of categories 1 through 5, and a " bump factor" category in our lette of April 24, May 21, May 29, June 4 and July 3,1979.
(Isometrics 300, 301 ana 302, service water lines in Category 4 have been deleted from further analysis as these lines are blanked off and are not in service at this time.) Attachment 2 to this letter presents the pipe stress analysis data for the final Category 6.
As shown, Isometries 213, 713 and 215 have calculated stresses which exceed allowable (stress greater than 1.8 3 ), but maintain structural integrity (stress less than 2.4 S ).
In ace rdance with our previous h
,,a 7 907270,q 17C L. (
49 004
.,., x.x
.m.
Mr. T. A. Ippolito July 23, 1979 commitments, we notified your staff on July 20, 1979 about these lines; and Lines 128 and 161 which were recently found to exceed allowable stresses, but maintained structural integrity, when analyzed for as-built differences and for valve eccentricity, respectively. A summary of the acceptability of the stress analysis for each isometric analyzed, and whether a modification is, or was, required is shown on Attachment 3.
As stated previously, commitments made subsequent to beginning our reanalysis effort have resulted in extending completion of the total reanalysis.
In the meeting with the NRC on bby 30, 1979, the modeling of valve operators in the pipe stress analyses was discussed.
Previous analyses considered valve operator mass lumped together with the valve mass at the centroid of the valve.
As a result of that meeting, it was decided to consider the effect due to the valve operator eccentricity in the reanalyses.
Recognizing that the effects of valve operator eccentricity are more significant in small pipe size systems (4" diameter and smaller), and that at the time the commitment to reanalyze was made approximately 112 lines had already been analyzed (without valve operator eccentricity), and that most of these lines were in the larger pipe size category, it was decided to first incorporate the valve operato1 eccentricity into the lines left to be reanalyzed. These lines, which were in Cr es 9, 4, 5 and 6, included the majority of the small size diameter piping system.
Several pipe stresses were calculated to exceed allowable (but maintain structural integrity) as a result of the valve operator eccentricity, and were reported to NRC by telephone on Jun e 22.
We are currently reanalyzing those lines previously reantlyzed where valve operator eccentricity should hase been considered. Approxima.ely 54 lines had to be reanalyzed, of which 29 are greater than 4" in diameter and 25 are 4" or less in diameter. Approximately 20 of these lines have been reanalyzed.
A preliminary investigation of the unreanalyzed lines indicates that pipe stresses for the larger diameter category isonetrics should not exceed the applicable allowable stresses.
In the case of the small diameter lines we expect that structural integrity will not be impaired, based on analyses to date. Effects on supports cannot be precisely predicted for all these lines wi'.hout a formal evaluation; however, va are confident that structural integrity of the supports and/or the associated piping will be maintained. Updated stress values will be cubmitted for these reanalyzed lines when they are available (about August 3, 1979.)
Our reanalysis effort on pipe support loads has progressed through the 195 isometrics. Currently, approximately nine supports, upon initial analysis, show structural integrity is not maintained and these are being reanalyzed taking out the conversatisms in accordance with our criteria given to you in our May 29, 1979 letter. These supports should be reanalyzed and identified as either a long-term fix or a short-term fix by July 30, 1979. During the supports reanalysis, it was determined that three supports would not maintain structural integrity and, because the affected lines would not maintain structural integrity, the supports required prompt fixes.
Supports at Data Points 50s and 80s on Isometric No. 4 (Unit 2) were reported by telephone to the NRC on July 2,1979.
At the time these support results became available, Unit 2 was in a shutdown mode and the supports were modified prior to startup.
On July 6, we notified 467 005
Mr. T. A. L to July 23, 1979 the NRC of a support at Data Point 80s, Isometric No. 4 (Unit 1) that did not meet structural integrity criteria.
The support was on a redundant 12" steam line from HPCI to the RHR steam condensing manifold (Loop B'.
The loop was declared out of service and the support was promptly modified. Approximately 2,000 supports have been evaluated on each unit, with approximately 200 per unit identified as long-term fixes (by the end of the next refueling outage).
A tabulation of the results of the evaluation of all pipe supports is not included with this submittal doe to the large amount of data. However, the number of supports which fall into the " acceptable", "long-term fix", or "short-term fix" categories are summarized on Attachment 3.
Containment penetrations and equipment nozzles were also analyzed for the revised loads. Allowable stresses for both upset (1.2 S ) and emergency h
conditions (1.8 S ) were not exceeded.
Safety class equipment, including the reactor pressure vessel, was reviewed to determine if the new load on nozzles exceeded the original loads previously accepted by the manufacturers. Where these original loads were exceeded due to increases from the new analyses, the manufacturers were contacted to determine the acceptability of the new loads.
Loads on all reactor vessel nozzles and five General Electric supplied equipment nozzles were forwarded to General Electric for verification of acceptability.
These have all been reviewed and loads are within code allowable. New loads on the RHR service water booster pumps are being reviewed by the manufacturer and results will be available about July 30.
Our commitment to compare as-built configuration verification with the as-designed (as analyzed) configurations has resulted in detecting approximately 64 differences. Many of these differences were small dimensional d4#'erences and had no impact on the analysis.
In accordance with our previou_Ay submitted criteria, these differences were and are being evaluated and, where required, reanalyzed. Over half of these differences have been evaluated.
Based on a computer reanalysis of some of the differences and the magnitude of the remaining, we have a high de Tree of confidence that the stresses and supports will be within the structural incegrity criteria.
This as-built verification and the evaluation and/or renalysis of the dif ferences, along with the criteria and data submitte=1 to you previously in this reanalys_ effort, satisfies the requirements of IE Bulletin 79-14, Seismic Analysis for As-Eitlt Safety-Related Piping Systems, dated July 2, 1979, for the Brunswick Steam Electric Plant, Units 1 and 2.
In summary, Carolina Power & Light Company has conducted a thorough reanalysis of the safety-related piping and supports using an acceptable computer code.
It should be noted here that no pipe stress exceeded allowable solely as a result of seismic analysis, but rather because of valve eccentricities or as-built /as-designed differences.
Various lines and pipe supports have been identified as requiring long-term fixes. Design modifications are in progress and the modifications will be made as time permits between now and the end of 467 cs36
Mr. T. A. Ippolito July 23, 1979 the next refueling outage for each unit.
Our commitment to totally review the as-built configuration to the as-analyzed configuration and make appropriate evaluations and/or reanalyses, has greatly increased our confiderae that the piping systems at BSEP are designed and installed in accordance with acceptable criteria, and that continued operation will be without undue risk to the heal *.h and safety of the public.
Yours very truly,
//
~
/ Mb/
y E. E. Utley Executive Vice President Power Supply & Customer Services MAC/j eb cc:
Mr. James P. O'Reilly U. S. Nuclear Regulatory Commission Region II 467 007
R&c/w,4 A'c 2 BRUNSWICK STEAM ELECTRIC PLANT SEISMIC ANALYSIS OF SAFETY RElATED PIPING CHRONOLOGY OF MAJOR EVENTS April 4 - 6, 1979 CP&L had several telephone conversations with the NRC Staff concerning the computer code that was used for the original Srunswick Steam Electric Plant Stress Analysis.
The results of a re-analysis of ten (10) lines completed by CE&C was discussed.
The results indicated that the stresses and the loads were within acceptablc values.
April 14, 1979 NRC IE Bulletin 79-07, Seismic Analysis of Safety Related Piping, was re-ceived which required a response f rom CP&L within ten days.
April 24 1979 CP&L filed its response to IE Bulletin 79-07, which discussed the computer code that was used in the original stress analysis, the results of the re-analysis of the ten lines, a commitment to re-analyce all safety related piping in accordance with a specified criteria, and a discussion of the new UE&C ADLPIPE - 2 Computer Code that will be used in the re-analysis.
The NFC was requested to acknowledge that this approach was acceptable prior to CP&L initiating the re-analysis effort.
Mav 15, 1979 CP&L met with UE&C and decided to proceed with the re-analysis even though the NRC had not responded to our request for their concurrence to our new approa.-h for the seismic re-analysis. A re-analysis priority criteria was established, and analysis was initiated.
May 16, 1079 A meeting with the NRC Staff was held in whict the quantitative results of the re-ar-lysis of the first ten lines (pipe stress only) was presented.
The details of the co=plete re-analysis effort was discussed with the NRC which included a discussion and quantitative results of t' e General Electric supplied safety related piping re-analysis.
During this meeting a commit-ment was made to complete the total re-analysis effort, in accordance with the criteria previously presented to the NRC on April 24, by July 21, 1979-A letter from CP&L to the NLC dated May 15, 1979 wa s given to the NRC docu-menting the material that was presented to them at this mee_ing.
May 21, 1979 A meeting was held with CP&L, UE&C, and the NRC to discuss additional requests for data that the SRC had made and to present some results trom the on going analysis. A letter f rom CP&L to the NRC, cated May 21, 1979, documented the material presented to th e NRC.
This letter identified the lines or isometrics that were to be re-analyzed, and placed each isometric into a re-analysis priority category.
In addition, schedule for the re-analysis was presented a
which shows completion by the July 21 date.
There was a discussion on the confidence that the as-ana lyzed was the as-built configuration.
The ru ess
& mn<,:Hw of up-dating engineering documents to as-built conditions was given to the NRC Also during this meeting, there was considerable discursion on the conservatisms that had been used previously in the relationship between the stresses calculated for the OBE case and those used for the DBE case.
Data was presented which shows that the relationship between DBE to OBE was a factor of 1.2 rather than the 2.0 that hac been used in the original a na lysis.
It was stated that this would be the basis for future seismic analysis.
May 22, 1979 A letter from CP&L to the NRC documented that computer listings for the UELC new seicnic stress analysis program and that used by Gencral Electric was given to the NRC during the May 21st meeting.
Fbv 25, 1979 in the course of this re-analysis, it was determined that several types of fabricated Lteel pipe supports were under designed for torsion loading.
In addition, it was discover i that several supports using concrete expan-sion bolts used the wrong allo able for specifying the required bolt size.
These problems were discovered on approximately 10 supports and six of these were shown not to maintain structural integrity under the identified support loads.
CP&L decided to shut the units down to make necessary modi-fications to the supports to insure their structural integrity and, at the same time, to have UE&C conduct a thorough review for all similar type supports to determine if these same problems xisted. During this evalua-tion, approximately 44 p.ne supports per unit were identified as requiring modification.
These pipe supports were modified prior to returning the units to operation on June 12 and June 15 for Units 1 and 2 respectively, May 30, 1979 A meeting was held with the NRC to discuss the additional information that had been requested, and the prvolems tha t had been encounterea 5 th the existing pipe support designs. A CPEL letter to the NRC dated May 29, 1979 discussed the pipe supports, the re-analysis of additional lines, the re-sults of an IE inspection of the as-built configurations at Brunswick Steam Electric Plant, discussed the validitv of the square root of the sum of the squares (SRSS) method versus the absolute sum method for the seismic analysis, identified the criteria that was to be used to report overstressed condi-tions to the NRC.
In addition, CP&L made the commitment to conduct an as-built verification for all safety related piping Inside the drywell trior to starting up the units. Also, at this meeting, the effect of the loca-tion of the valve operator in relation to the line stresses was discussed.
Although the NRC did not require a commitment from CP&L to analyze for valve operator eccentricity, the correct technical approach would be to consider this item and a decision was made that the stress analysis should account for the operator location.
Previous analysis lumped the mass of the operator at the valve location and not at some distance off of the line ac is the case.
It was decided to proceed with the remaining analysis in-corporating this change and to re-analyze those lines that had been completed 467 009
fhchmetf/L<
s where this additional loading was applicable following the initial analysis.
June 4, 1979 A meeting was held with the NRC to discuss the results cf the current pipe stress re-analysis effort. As of this date 112 lines had been re-analyzed for pipe stress.
These lines did not include the valve eccentricity loading.
In addition, the procedure used to determine if an as-built differ-ence required re-analysis was discussed. At this meeting, the NRC stated that the units could start up after completing the modifications to the supports that had bet previously identified, and after verifying as-built configuration in the.rywell.
June 22, 1979 UE&C notified CP&L that three lines had pipe stress levels exceeding their FSAR allowable limits, but maintained their structural integrity.
These high stresses resulted from valve operator eccentricity on small diameter
'ines (apprcximately 1" in diameter).
UE&C was continuing their stress analysis, support analysis, and analysis of approximately 64 deviations that were noted between as-built an as analyzed configurations.
The NRC was likewi se notified in accordance with our commitment.
July 2, 1979 Two pipe supports on Unit 2 were determined to be overstressed which would require a fix because structural integrity was not maintained under the identified loads. Unit 2 was _. ready in a shutdown mode and the modifica-tions were made before retu rning to service.
This was reported to the NRC in accordance with our previous co=mitments.
aulv 3, 1979 CPEL letter to the NRC transmi tted the results of the stress analysis for Categories 3, 4, and 5.
Julv 6.
1979 A pipe support on Unit 1 was identified as requiring a modification to main-
, tain s t ructural integrity.
This support was located on a redundant RHR loop and the modification was performed with the loop taken out of service, but the unit remained in service.
NRC was notified of these results on July 6.
July 20. 1979 Five lincs have been identified as being overstressed, but maintaining their structural integrity.
These lines were reported to the NRC.
Three lines were in Category 6 and two lines were a result of a re-analysis that was con-ducted for as-built differences and for valve eccentricities.
467 010
ATTACIDfENT 2
.ATECORY 6 EMERGE!!CY CONDITION STRESS
SUMMARY
(PSI)
ORIGINAL ORIGINAL (
NEW NEW(
RATIO SYSTEM ISO, NO.
LINE SIZE TOTAL SEISMIC TOTAL SCISMIC ALLOWA BLE ITr/A
- ontainment Atmospheric Control 148 20",24" 8101 1868 8054 1821 27000 0.29
'ontainment Atmospheric Control 149 18",24",30" 20687 18166 15567 13046 27000 0.57 nstrument Air 180 3/4",2" 9614 8364 7379 6129 27000 3.27
- ontainment Atmospheric Control 209,210 2",18",20",24 10870 8650 8031 5811 27000 0.29 tesidual lleat Removal 546,565 4"
20416 17624 11140 8348 27000 0.41 notrument Air 178 1",1b", 2" 11851 10710 3141 2000 27000 0.11 nstrument Air 679 3/4",2" 11940 8660 13193 9913 27000 0.48 iervice Water 216 15",2" 10090 8214 5413 3537 27000 0.20 nstrument Air 306 1"
12183 9043 7757 4616 27000 0.28 nstrument Air 682 1)",2" 12314 10975 6678 5338 00 0.24
.tandby Gas Treatment 150 18",24" 13333 12762 11392 10821 27000 0.42
.ns trument Air 177 3/4",2" 16406 15156 9824 8574 27000 0.56 nstrument Air 675 3/4",2" 11262 10112 8000 6850 27000 0.29 nstrument Piping 202 3/4",1" 11476 5764 9184 3460 27000 0.34
.ns trument Air 677 3/4",2" 21584 20244 7775 6435 27000 0.28 esidual IIcat Removal and 64 8"
17652 15652 8907 6907 27000 0.32
'uel Pool Cooling instrument Air 174 2"
2496 1246 1818 568 27000 0.06 467 0ii
ATTAC1DIENT 2 ORIGINAL ORIGINAL (I) MEW NF" ('
RATIO SYSTEM ISO NO.
LINE SIZE TOTAL SEIS'41C TOTA L SE T 'JII' A LIDWABLE UT/A Instrument Air 182 2"
20284 19034 10121 8871 27000 0.37 Fuel Pool Cooling & Residual 138,139 6",8" 21127 17094 13118 9358 27000 0.48 Heat Removal Instrument Air 186 2"
16026 14776 8919 7669 27000 0.33 Pesidual lleat Removal & Fuel 14 't,145 4",6",8",10",12" 15862 13712 10568 6526 27000 0.39 Pool Cooling Nitrogen System 231 3/4" 6436 5372 3101 2037 27000 0.11 Instrument Air 176 3/4",2" 2454 1204 2501 1251 27000 0.09 Nitrogen System 232 3/4" 8281 7204 4312 3235 27000 0.15 Service Water 218 1 ",2" 18318 15052 15356 12090 27000 0.56 Core Spray 166 3/4",1",2" 4898 2850 4300 1290 27000 0.15 Inctrunent Air 183 3/4",2" 14961 13836 7032 5782 27000
- 0. 9.6 Instrument Air 674 3 /4",1 \\",2" 6405 5065 4625 3284 27000 0.17 Service Water 214 3 / 4",1",1)", 2" 18556 17610 14714 12556 27000 0.54 Nitrogen System 233 3/4",1",2" 8901 7824 5765 4688 27000 0.21 Containment Atmospheric Control 213
",3/4",1",2" 11831 6408 26902 26750(3) 44800(4) 0.60 Containment Atmospheric Control 713
",3/4",1".2" 20574 14480 35604 29511(3) 44800(4) 0.79 Instrument Air 680 3/4",z" 8153 4812 9277 7936 27000 0.34 Instrument Air 681,691 3/4",?."
11483 10142 7211 5870 27000 0.26 Instrument Air 175 3/4",2" 14973 13812 18331 17081(3) 27000 0.67 Service Water 215
", 3 / 4",1",1 \\", 2" 19128 15778 35143 30892( )
44800I0) 0.78 (1) Original Seismic = 2 x OBE (3) New Seismic for this case calculated using actual DBE Response Spectra (2) Ucw Seismic = 1.2 x OBE (4) Structural integrity for stainless steel pm 467
ATTACllMEN E 3 For Notes see l' age 1.1 of t ii i. t 4ble NRC IE 79-07 RF-E"ALL'AT ION SMIMAR'(
PRIORITY CATEGORY I Pa80 I PIPE STRFSE 1
_5 ' TO R T IOADS Ic0 ACC'I' LTFO) 4fF(3)
ACC' LTF(2)
I!T (3)
NO.
SYSID1 Nl.
NO.
NO.
REIIARK I
Residual lica t Removal Yes None None 30 10 3
STF - Due to original support under design, fixes implemented May/ June 1979 2
Residual lleat Removal Yes None None 24 1
1 STF - Due to original support under design, fixes implemented May/ June 1979 3
Residual IIcat Removal Yes None None 34 5
2 STF - Due to original support under design, fixes implemented May/ June 1979 4
liigh Preasure Coolant Yes None None 16 1
2 STF - Due to sappc rt load carrying capability.
' Injection l
Repaired during caubber inspection outage in July - Unit 2.
Unit I repaired week of July 7.
5
'.csidual lleat Removal Yes None None 3
1 0
10 liigh Pressure Coolant Yes None None 15 3
0 Injection lHighPressureCoolant 11 Yes None None 6
2 0
l Injection 13 Residua 1 Ileat Remova1 Yes None None 18 0
0 14 Nuclear Steam System Yes None Nonc 13 0
0 15C I:uclear Steam System
.Ye s None None 8
0 1
STF - Due to original support under design.
fixes implemented May/ June 197')
21 Reactor Core Isolation Yes None None 9
0 0
Cooling 22 Reactor Water Cleanup Yes None None 3
0 0
25 Residual llea t Removal Yes None None 14 0
0 27 Control Rod Drive Yes None None 11 0
0 28 Residual lleat Removal Ye,
None None 24 2
3 1TF - Due to original support under design, I
fixes implemente.1 Itay/ Jour !ON 467 013
A rrAC11; tern 3 ror Notes see Page 13 o f t h i s table NRC IE 79-07 PE-EVALUATION St'M'!ARY Page 2 PIPF J1
.dfd J ')Pf0RT IOAlll 0) 150 ACC L1 Fl '
O)
ACC' I)
LTF(2) 377 31F No.
cYSIE 1 No.
NO.
NO.
REMARK 32 Nuclear Steam System Yes None None 21 0
0 36 Contr>l Rod Drive Yes None None 9
0 0
41 Residual lleat Removal Yes None Ilone 5
0 0
42 Residual lleat Removal Yes None None 2
1 1
STF s Due to original support under design, fixes implemented tiay/ Jane 1979 52 Residual lleat Removal Yes None None 5
0 2
STF - Due to original support under design, fixes implemented tiay/ June 1979 f
59 1:esid, t 1 Ilea t Remova1 Yes None None 5
0 0
60 Residual lleat Removal Yes None None 8
0 0
61 Re iidual llent Removal Yes None Ibne 11 3
0 87 Nuclear Steam System Yes None None 5
6 1
STF - Due to original support under 6esign, fixes implemented May/ June 1979 120 Nuclear Steam System Yes None None 6
0 1
STF - Due to orif,inal support under design, fixes implemented May/ June 1979 123 Nuclear Steam System Yes None None 8
1 2
STF -Due to original sapport under design, fixes implemented May/ June 1979 124 Nuclear Steam System Yes N
l' None 14 0
2 STF - Due to original support under design, fixes implemented May/ June 1979 126 Nuclear Steam System Yes None None 10 2
2 STF - Due to original support under design, fixes implemented May/ June 1979 128 Nuclear Steam System Yes None 15 0
0 LTF - Add 2 Supports - Formal evaluation of is-built vs as-enalyzed/ designed configuration discrepancies to be completed.
163 Residual llea t Removal Yes None None 25 1
0 173 Residual lleat Removal Yes None None 12 0
0 467 014
ATTAClif1Erir 3 For flotes see l' age 13 of this table NRC IE 79-07 RE-E'lALUATION SL?CIAR'1 Page 3' IPE STRF 4' L S_
S ' SRI I O ADS 150 ACC LTFU ilF(3)
ACC' )
LIF NI MT(3)
I NO.
C ':51 E31 NO.
NO.
NO.
REtliRK 187 Nuclea r S te am System Yes lione None 8
2 0
194 Reactor Core Isolation Yes None None 5
0 0
Coolant 196 Reactor Core Isolation Yes None Nonc 1
1 0
Coolant 199 Residua l lleat Removal Yes None None 4
1 0
1 524 l Core Spray Yes Nane None 8
0 0
605 Residual IIcat Removal Yes None None 31 3
0 606 l Se rv ice b'a t e r Yes None None 14 1
0 PRIOR FY CATEGORY II 7
Nuc lea r S team Sys tem Yes None None 9
0 0
9 liigh Pressure Coolant Yes None None 12 0
0 Injection 18 Core Spray Yes None None 7
0 0
19 Core Spray Yes None None 0
4 0
20 Core Spray Yes None None 5
1 0
23 Core Spray Yes None tione 7
1 0
26 Core Spray Yes None None 4
2 0
31 Residual llear I4emoval Yes None None 7
4 1
GTP - Due to original support under design, fixes implemented !!ay/ June 1979.
467 015
For Notes see Page 13 of this t ribl e ATTACliflENT 3 NRC IE 79-07 RE-EVALL'ATION Sli31?t-\\RY Page 4 I
S SE1 Sl'Ph1RT T/}A!H 150 ACC'
- l. [ Fl S'F(3, ACC'
)
I/I r i )
TIr(3)
NO, s'?STE:1 NO.
NO.
NO.
Rett \\RE 40 Core Spray Yes None None 1
1 0
43 Residual lleat Remova1 Yes None None 4
0 0
4't Residua 1 lie it Remova1 Yes None None 0
0 0
43 Residual lleat Removal Yes None None 7
1 0
46 Re sidua l llea t Removal Yes None None 9
3 0
47 Reaidua1 lient Remova1 Yes Mone None 6
1 0
48 Residua 1 licat Remova1 Yes None None 12 0
0 51 Residual lleat Removal Yes None None 0
0 0
53 liigh Pressure Coolant In-Yes None None 5
1 0
jection 54 Residual IIcat Removal Yes None None 6
1 0
55 Residua 1 lleat Remova1 Yes None None 11 0
0 56 Residual lleat Removal Yes None None 14 2
0 57 Residual lleat Removal Yes None None 6
0 0
58 Re.;idua1 Ileat Remova1 Yes None None 6
0 0
65 Residual treat Removal Yes None None 3
3 0
146 111gh Pressure Coolant Yes None None 12 1
0 609 Service Water Yes None None 9
4 0
i 467 016
ATTACllMENT 3 For Notes see Page 13 of this table NRC IE 79-07 RE-EVALl'ATION M"E tir;Y PRIORITY CATEGORY III Page 3 PIPE SIEFJEEE S!f PPORl LOADS.
ill 1I'F(2) 37p(3)
ACC ll) 1TF(2)
- 7 ( 3 )
ISO ACC NO.
SYSTEM NO.
NO, 30 Rett \\RK 12 Reactor Core Isolation Yes None None 14 5
1 S'lF - Due to original support under design, Coolant I1:u.s implemented ?!ay/ June 1979 33 Reactor Core Isolation Yes None Nonc 14 0
0 Coolant 34 Reactor Core Isolation Yes None None 3
0 0
Coolant 35 Reactor Core Isolation Yes None None 7
1 0
Coolr.n*
49 Reactor Core Isolation Yes None None 4
3 0
Coolant 50 Ecactor Core Isolation Yes None None 4
0 0
Coolant 63 Reactor Core Isolation Yes None None 5
2 0
Coo lant 66 Reactor Core Isola tion Yes None None 9
0 0
Coolant 67 Reactor Core Isolation Yes None None 6
1 0
Coolant 535 Reactor Core Isolation Yes None None 7
0 2
iTF - Due to original support under design, Coolant fixes implemented Ray / June 1979.
549 Reactot Core Isolation Yes None None 6
1 0
Coolant 563 Reactor Core Isolation Yes None None 8
0 0
Coolant 568 Reactor Core Isolation Yes None None 9
2 0
467 017 c~une
N I"'
P'.
I i of this table ATTAClit!Ctn 3 NRC TE 19-07 RE-EVAlfATION SMDtARY PRIORITY CATEGORY LV Page 6 PLPE Sif MES TfilFM 1
1:0
-\\CC ( l )
1.,1 F
'31F(3)
ACCI l 1,T FI STF l3)
'I0 SYTIEM NO.
I:0.
NO.
P.EMARK 29 Residual IIeat Removal Yes None None 6
1 0
30 Residual Heat Removal Yes None None 14 0
0 62 Service Water Yes None None 10 2
1 STF - Due to original support under design, fixes implemented Itay/.iune 1979 85 Residual IIeat Removal Yes None None 5
1 0
86 Residual IIcat Removal Yes None None 8
2 0
109 Service Watet Yes None None 9
3 1
STP - Due to original support under design, fixes i.mplemented May/ June 1979 110 Service Water Yes None None 27 0
0 142 Service Water Yes None Hone 12 0
0 162 Service Water Yes None None 10 0
0 163 Service Water Yes None None 0
0 0
217 Ser. ice Water Yes None None 15 0
0 662 Service Water Yes None None 5
0 0
PRIOR 1 TY CATEGORY V 8
Nuc1 car Steam System Yes None 16 4
0 LTF - Due to valve operator eccentricity 92 RCIC Misc. Lines @ Turbine Yes None None 2
0 0
13 2 Reactor Water Cleanup Sys.
Yes None None 15 2
0 151 High Pressure Coclant Yes None None 17 0
0 Injection 153 Hign Fressure Coolant Yes None Nore O
!2 0
467 018
ATTAC11 MENT 3 For Notes see Page 13 of this table NRC IE 79-07 RE-EVALUATION St?CtARi' l' age 7
'DRI I 01DS FIPE SIRESSE1 ISO ACC fl}
L1 Ff2) 'SIF(3)
ACC LTE(2)
STF (3 )
I NO.
SYSTEll NO.
NO.
- 0 REMARK 156 liigh Pressure Coolant Yes None None 13 1
0 Injection 157 Iligh Pressure Coolant Yes None None 8
1 0
Injection 159 liigh Pressure Coolant Yes None None 1
2 0
Injection 101 Reactor Core Isolation Yes None 11 1
0 LTF - Due to valve operator eccentricity Cooling 164 Reactor Core Isolation Yes None None 9
0 0
Cooling 170 Residual lleat Removal Yes tione None 2
0 0
171 Residua 1 Ileat Remova1 Yes None None 6
0 0
172 Residual lleat Removal Yes None None 6
1 0
195 Reactor Core Isolation Yes None Nene 1
0 0
Cooling 607 Service Water Yes None None 9
3 2
STF - Due to original support under design, fixes implemented May/ June 1979 608 Service Water Yes None None 17 3
0 657 Iligh Pressure Coolant Yes None 8
1 0
LTF - Due to valve operator eccentricity Injection 167 019
For Notes See Pane 13 of this table ATTACilifEttr 3 NRC IE 79-07 EE-EVALl'A !IO:. Sl3PLUE PRIORIFY CATEGORY VI P ve 8 PIPE STRNJiES Rgl'QR1 IDADS I
150 ACCfl)
L1F(23 STF(3)
ACC tII 1.T F ( 2 )
gp(3) a,
- E:1 NO.
NO.
NO.
REMARK 64 Residua 1 Ileat Remova1 Yes None None 9
0 0
138 ruel Pool Cooling Yes None None 14 1
0 139 fuel Pool Cooling Yes None None 10 1
0 144 Fuel Pool Cooling Yes None None 8
1 1
STF - Due to original support under design, fixes implemented May/ June 1979 145 Fuel Pool Cooling Yes None None 15 3
0 l
l '+ 8 l Fuel Pool Cooling Yes None None 3
1 0
I 14 9 Fuel Pool Cooling Yes None None 5
8 0
15n Standby Cas Treatment Yes None None 12 1
2 STF - Due to original support under design, fixes implemented May/ June 1979
! eedwater Yes None None 14 1
1 S!F - Due to original support under design, 160 F
l fixes implemented May/ June 1979 166 Core Spray Yes None None 4
3 0
174
, Instrument Air System Yes None None 11 0
0 175 Instrument Air System Yes None None 37 4
1 STF - Due to original support under design fixes implemented ;Iay/ June 1979 176 Instrument Air System Yes None None 10 0
0 177 Instrunent Air System Yes None None 36 3
0 f,TF - Due to supc? t load carrying capability 178 Instrument Air System Yes None None 16 0
0 180 lust rument Air System Yes None None 13 2
0 LTF - Due to support load carrying capability 182 Instrument Air System Yes Non2 None 6
0 0
183 Instrument Air System Yes None None 13 0
0 IE6 In s t rume n t Air System Yes None None 12 0
0 467 020
ATTACilMENT 3 ror tiotes see Page 13 of this table NRC IE 79-07 IE-EVALUATIO:1 StCIAih Page 9 P1PE S1RESSDi
_St!PPORT IOADS ISO ACC(I' lif F (2 )
STF(3)
ACC lI' LTF(2)
S IT ( 3 I NO.
SYSIDI NO, NO.
NO.
Im:1AltK 202 Inr.t rument Air Sys tem Yes None None 16 1
0 20)
Containment Atmospheric Yes None None 4
0 0
Control 210 Containment A t ospheric Yes None None 4
1 0
Control 213 Cont ainment Atmospheric Yes None 31 0
0 I;rF - Fornal evaluation of as-built vs. as-Control analyzed / designed configuration discrepancies to be completed.
214 Service Water Yes None done 5
2 0
215 Service Water Yes None 4
2 0
I!rF - Formal evaluation of as-built vs. as-
,naly: ed/ designed configuration discrepancies 216 Service Water Yes I;one None 15 0
0 to be completed.
218 Service Wa ter Yes None None 8
0
'l 231 Instrument Air System Yes None None 18 0
0 232 Instrument Air System Yes None None 18 0
0 233 Instrument Air System Yes None None 16 0
0 306 Ins trument Air System Yes None None 27 0
0 1
546 Residua l lleat Removal Yes None None 4
4 0
565 Residual IIca L Removal Yes
!!one None 5
0 0
674 Instrument Air System Yes None None 12 0
0 675 Instrument Air System Yes None None 31 12 2
lSTF - Due to original support under design, fifes implemented May/ June 1979 1
46/
DJ
A L TACll!!ErTr 3 For flotes see Page 13 of this tabic 3: E 19-07 RE-EVALUATION SU'pt\\RY Page 10
,llP4i;if1
'gi'O R1 IOADS I
ISO ACC-LIFU' 'STF(3)
ACCl )
LTF(2)
S IT (3 )
N0 SYSTE 1
!D,
NO.
NO.
REtuRK 677 Instrument Air System Yes None None 36 6
f STF-Due to original support under design, fixes implementedMay/. Tune 19)9 679 Instrument Air System Yes tione
!!one 6
16 0
lirF - Due to support load carryfng capability 680 Instrument Air System Yes tione lMone 14 1
0 LTF - Due to support load carrying capability 681 Instrument Air System Yes None Nonc 31 0
0 682 Instrument Air System Yes None
!!one 14 2
O I/IF - Due to support load carrying capability 691 Instrument Air System Yes None I None 3
0 0
/13 Containment Atmospheric Yes None 34 7
0 f,TF - Forma l eva lua tion of as-buil t vs. as-Control analyzed / designed configuration discrepancies to be complete?.
ORIGI:aL 10 6
Residua 1 Heat Remova1 Yes None None 3
0 0
15B Nuclear Steam Lyntem Yes None
'Jone 7
1 0
16 Feedwater Yes None None 14 2
0 17 liigh Pressure Coolant Yes None None 9
0 0
Injection 24 Core Spray Yes None None 1
1 0
121 Nuclear Steam System Yes None None 5
0 9 F - Due to original support under design, fixes implemented May/ June 1979 122 Nuclear Steam System Yes None None 4
0 3
STF - Due to original support under design, j
fixes implemented May/ June 1979 125 Nuclear Steam System Yes None None 11 0
1 STF - Due to original r, ort under design, fixes impicmented May c.a 1979 I
l 467 022
ATTACilMEttr 3 Fo r No t e s see Page 13 of titis table NRC IF_79-07 RE-EVAlfA TlON SL7 DIARY Page 11 MPE SLP]' SESSIF(3)
ACCil)
I.TF ( 2 )
3 TF (3 )
Sl' P PO RI LOADS 1
LIF ISO ACC NO-SYSTEB1 NO.
NO.
NO.
REB! ARK
'137 1;uclear Steam System Yes None None 4
3 2
STF - Due to original support under design, fixes implemented Flay / June 1979 310 liigh Pressure Coolant Yes None None 14 5
3 STF - Due to original support under design, Injection fixes implemented May/ June 1979 NRC PRIORITY 14A Nuclear Steam System Yes None None 6
0 0
39 Core Spray Yes None None 1
1 1
STF - Due to original support under design, fixes implemented ttay/ June 1979 82 Service Wa ter Yes None None 12 2
1 STF - Due to original support under design, fixes implemerted May/ June 1979 83 Service Water Yes None None 17 1
0 84 Service Water Yes None None 8
1 2
STF - Due to original support under design, 105 Core Spray Yes None None 35 3
0 106 Service Wa ter Yes None None 15 2
0 107 Service Water Yes None None 11 3
2 SIF - Due to original support under design, 108 Service Water Yes None None 20 2
0 119 Nuclear Steam System Yes None None 11 0
0 127 Nuclear Steam System Yes None None 3
2 1
iTF - Due to original support under design, fixes implemented tiay/ June 1979 152 Iligh Pressure Coolant Yes None None 3
4 0
Injection 154 High Pressure Coolant Yes None None 14 0
0 Injection 467 023
AT FACitMEttr 3 teor t:0: e: see Page 13 of thin table NRC IE 79-07 RE-EVALUATION SteDMRY Page 12
'IPE STRES:ll S
_ lit t P PORT TOAD 3
)
LIF(2) 1TF(3)
ACC(l)
LR(2)
HF f3)
ISO ACC NO, SYSICl NO.
NO.
';0.
REMARK 155 liigh Pressure Coolant Yes None None 6
2 0
Injection 158 liigh Pressure Coolant Yes None None 15 0
0 Injection 179 Instrument Air System Yes None None 17 2
0 181 Ins trument Ai r Sys tem Yes None None 29 1
0 184 Instrument Air System Yes None None 8
1 0
185 Ins trument Air System Yes None None 5
0 0
188 Instrument Air System Yes None None 7
0 0
189 Instrument Air System Yes None None 3
3 0
f.TF - Due to support lo.id carrying capability 190 Ins trument Air System Yes None None 0
28 0
I.TF - Due to support load carrying capability 191 Instrument Air System Yes None None 3
0 0
192 Instrument Air System Yes None None 4
1 0
193 Instrument Air System Yes None None 4
1 0
I.TF - Due to support load carrying capabilit-201 Instrument Air System Yes None "one 13 0
0 206 Instrument Air System Yes None None 7
3 0
LTF - Due to support Joad carrying capability 207 Instrument Air System Yes None None 12 2
0 LTF - Due to support load carrying capability 211 Containment Atmospheric Yes None None 2
0 0
Control 212 Containment Atmospheric Yes None None 3
3 o
Control 467 024
ATTACllMErft' 3 NRC IE 79-07 RE-EhLCATION SlW1ARY Page 13 PIPE S W liEl
' TI)ELT DADA E
SlF(3)
ACC' LTF ! )
(IF ( 3 )
U)
L1 F 150 ACC NO, S YSTE'1 M)
NO.
N0 R DL\\RK 230 Conta inment Venting Yes None None 4
2 0
545 Re idual IIca t Removal Yes None None 8
0 0
547 Pesidua1 Ileat Remova1 Yes None None 7
1 0
48 liesidua1 Ileat Remova1 Yes None None 8
1 0
690 Instrument Air System Yes None None 0
27 0
709 Containment Atmospheric Yes None None 3
0 0
Control 710 Containment Atmospheric Yes None None 12 4
1 STF - Due to original sa;' port under design, Control fixes impicmented May/ Jure 1979 716 Setvice Water Yes None None 15 0
0 NorES:
(1)
ACC - Acceptable - stresses / loads w Lthin FS/,R and/o origin il critet la alloaables (2)
LTF - Long Term Fix - indicates the fix may be made at the next scheduled refueling outage because it has been determined, that though overstressed by FSAR criteria, structural integrity can be maintained.
STF-ShortTermFix-indicatesthelfixsh(uldbemade (3) inwdiately because it has been determined that structural integrity cannot be n tintaine<1 in the overstressed pipe or pig e suppo rt.
f 467 023
ATTACHMENT NO. 4
SUMMARY
OF COMMITMENTS COMMITMENT STATUS Reanalyze all safety-? elated piping Completed with tii.s transmittal, within approximately 13 weeks of NRC approval of reanalyses criteria.
Provide interim reports of reanalysis.
Transmitted in CP&L letters of May 15, May 21, June 4 and July 3.
Verify as-built versus as analyzed Completed as reported in CP&L letter of conditions.
July 3.
Provide guidelines for uc.ermining Provided in CP6L letter of May 21.
if reanalysis is necessary due to as-buil-discrepancies.
Comply with IE Bulletin 79-02.
Completed as reported in CP&L letter of July 12, 1979.
Comply with IE Bulletin 79-04.
Completed as reported in CP&L letter of April 30, 1979.
Report any lines with calculated Three lir.as (total) reported to NRC - NRR stresses over allowable within and IE by telephone June 22.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> any systems CP&L telephone calls of July 2 ar.c Jul.y 6 that do not maintain structural reported three supports that had calculated integrity due to pipe support being overstresses. Affected supports were promptly stressed.
modified.
Reanalyze all pipe supports using Completed with this submi tal.
recalculated seismic loads.
Comply with IE Bulletin 79-07.
Completed with this submittal, prior submittals, and as-built verification.
Comply with IE Bulletin 79-14.
Completed with this and prior submittals,
and as-built verification.
Be prepared to discuss valve operator Categories 3, 4, 5 and 6 were reanalyzed eccentricity, with eccentricity incorporated. Categories 1 ar.d 2 are presently b ting re-evalua*ed on basis of eccentricity If any lines. e determined to be overstressed, NRC will be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and resultu will be submitted to NRC.
Provide ADLPIPE listings, benchwork Provided at May 21, 1979 meeting with NRC.
piping problems and PISYS Listing.
467 026