ML18114A685
| ML18114A685 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 06/08/1979 |
| From: | Spencer W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 458, NUDOCS 7906130352 | |
| Download: ML18114A685 (18) | |
Text
',
VIRGINIA ELECTRIC AND POWER COMPANY R10HMOND, VIRGINIA 23261 June 8, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Serial No. 458 PSE&C/CMRjr:mac:wang Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37
Dear Mr. Denton:
ADDITIONAL INFORMATION REPORT ON REANALYSIS OF PIPING SYSTEMS SURRY POWER STATION UNIT 1 On June 5, 1979, a report was transmitted to you (Vepco Serial No. 453) regarding the reanalysis of safety related piping systems for Surry Power
. Station Unit 1. The report contained detailed information substantiating our request to restart the unit based on the analysis performed to date.
We have had the benefit of a meeting with your staff on June 5, 1979, to present the report and to answer questions related to its contents.
Several questions were asked that required responses that could not be given quickly and had to be researched further before adequate answers could be developed.
It is the purpose of this letter to transmit our responses to the remaining outstanding issues, as communicated to us by the staff at the June 5 meeting.
Base plate flexibility considerations will be incorporated in those supports for which the new loads exceed the original design allowable loads and for any new supports required by the reanalysis effort.
The guidelines to be used for flexibility considerations will meet the intent of IE Bulletin 79-02.
This procedure has been in effect for all completed support evaluations and will be in effect for all future support evaluations.
Hardware modifications were identified in our submittal of June 5 for Problems 743, 548A, 731A and 7318.
All modifications as delineated for these problems in the June S*report will.be completed prior to start up.
Table 3-1 of our June 5 submittal listed original and new pipe stresses for each of the problems affected by the Order to Show Cause of March 13, 1979.
The staff requested that, original total, original seismic and allowable stresses should be included in the Table.
The attached revised Table 3-1 (Attachment I) complies wit~ that request.
For those problems that are not yet complete as listed in Table 3-1 in June 5 submittal, we have identified a methodology for evaluating the significance of a potential modification as calculations continue during interim operation.
The methodology is delineated in Attachment II. This process amplifies and expands Section 5.2 of our June 5 submittal.
The reporting process to the NRC has been previously established and will be in compliance with the Surry Technical Specification Section T.S. 6.6.
our I
I
e e
VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton, Director Staff questions at the June 5 meeting on IE Bulletin 79-02 and its applicability to the pipe stress reanalysis effort required information in addition to that submitted in our letter of June 4, 1979 (Serial No.
146/030879A).
A sampling program has been established to meet the intent of the Bulletin.
The details of the sampling program are as follows:
- 1.
A statistical sampling program will be established in accordance with MIL STD 1050.
- 2.
A total of 200 anchor bolts will be sampled.
Of these, approximately 60% are in the containment building.
- 3.
The sampling program will incorporate such attribute$ as bolt size, backoff torque, thread engagement, visual inspection for cracked concrete (other than hair-line cracking normally associated with concrete), depth of set plug (for self-drilling anchors), and embedment depth of wedge type anchors (by UT inspection).
- 4.
The sampling program started on June 7, 1979, and will be completed on or about June 14, 1979.
Attachment III indicates the priority list and schedule for reanalysis of pipe stress problems.
All reactor coolant pressure boundary reanalyses have the highest priority. Attachment III also shows scheduled completion dates for pipe stress reanalysis and supports reanalysis for each problem.
The reactor coolant pressure boundary stress analysis problems which are part of this review effort have been reanalyzed.
As the problem list Attachment I shows, problem numbers 508, 708, and 630 have not yet been approved by Engineering Assurance so they are still not included in the 11complete 11 category. There are hardware modifications on problems 508 (residual heat removal) and 630 (pressurizer spray & relief) which are known in sufficient detail to insure that the hardware modifications will be incorporated prior to start up but are not known in sufficient detail to describe for this submittal.
The reactor coolant pressure boundary problems include much more pipe than is actually part of the reactor coolant pressure boundary and one of the modifications take place in areas other than the reactor coolant boundary areas.
The problems which include the reactor coolant boundary are identified in the comments column of Attachment III.
The Surry 1 and 2 FSAR was supplemented in 1973 with Appendix D entitled 11 Effects of Pipe Breaks Outside Containment.
11 This document presented the program for analysis of piping and equipment required to show protection from high energy lines.
Systems for reliable and redundant *detection of failures in the main steam system and feedwater system are incorporated in the FSAR in Chapter 7.2 Table 7.2-1.
2 I
e VIRGINIA ELECTRIC AND POWER CoMPA?.-Y TO Mr. Harold R. Denton, Director During the pipe stress reanalysis program for the Show Cau~e Order, only th~ main steam pipes outside the containment require seismic reanalysis because the other high energy lines are either smaller than computer-analysis size or were not seismically analyzed originally. The main steam line outside the containment is the only line in the original program affected by this effort. The break locations for the.main steam lines were selected on the basis of terminal points and highest stress intermediate points.
The terminal points remain as in the original program.
Only the intermediate* points which were listed in Table D.5-1 of Appendix D can change.
These nine intermediate points may change but since the stresses are usually highest in a pipe system at bends and elbows, the new locations are most likely to encompass most of the old points. Therefore, no appreciable number of location changes are
.expected and the criteria for selection of the points is still as stated in Appendix D.
The number of points should not change.
Continued applicability of Appendix D will be verified for the stress reanalysis.for the main steam line and the highest stress points will be reviewed by the same procedure used in the original break analysis.
Any break location changes will be available when main steam problem 346 is complete in late June, and the station will be notified if changes are required in the inspection program.
These responses conclude the outstanding issues on the Surry pipe stress reanalysis effort.
We once again reiterate the request of our June 5 letter.*
permitting us to start up Surry Unit 1 based on the satisfactory results of the analysis to date.
If other information is required beyond that presented in this letter, please* contact us immediately.
Attachments (3)
W. C. Spencer Vice President - Power Station Engineering and Construction Services 3
h1284622-lx 06.107/79
- 046 SURRY POWER STATION, UNIT 1 1.20
~.,;-*....
TABLE 3-1
~..
1.22 PIPE STRESS REEVAT;UATIO.N
SUMMARY
1.24
(.
- -- ~-*
NA - Not Available 1.27
- Table 4-1 of Seismic Design Review 1.28.
Equipment and Pi5inf Surry Power
- 1.29
'..;.~
Station, Sept. 1,
971 or 1.30
~reliminar:x Criteria
- Subsequent Reanalysis 1.31 Prob-Line PiHe Stress (Hsi>
, 1. 33
\\~:.::
lem System*
Isa{ 1.
'Sfze
- original Original New New Allow-
.1.34 l!Q_:_
Name NJh_
CNPS)
Total*
Seismic*. Total Seismic able 1.35 e.
SHOCK2 Problems 1.38 555 Low Head Safety 122 Dl
- 10 11 29290 NA 11180 5307 30882
- 1. 41 Injection 12" 1.42
<_j 1555 Low Head Safety 122 L1 12 11 25290 NA 1()392 3855 30882 1.44 Injection 1.45
~
706A Low Head Safety 122 Hl 6"
18451 10707 19830 8439 30769 1.47
- ...:_./
Injection 1.48 j;!
("")
707A Low Head Safety 122 Jl 6" 22436MH 19577MM 28980 1.50
- c Injection 1.51 3:
\\_)
fTI 2
708 Low Head Safety 122 Kl 6"
20552~0E NA 28980 1.53
--i Injection 1.54 H
-~
\\:.,!,)
Low Head Safety 127 El 2'z'671MM 731A 8"
NA 21503 13940 24750 1.56 Injection 1.57 t*,. \\
731B Low Head Safety 127 E2 8"
22671**
NA 21800.
16004 24750
- 2.1
j Injection 2.2 743 Low Head Safety 127 Fl 10" 24649**
NA*
16119 5496 33750 2.4
- .. _)
Injection 2.5
'-.J 727 Low Head Safety 127 Cl 6"
28909**
NA 30769 2.7 Injection 127 C2 10" 2.8 u
735 High Head Safety 127 Gl 411 6"
23011 13220 33750 2.10 Injection 127 G2 an:* 10" 2.11 525A.I Containment &
- 123 Al 8"
11999 10866 9409 3846 33750 2.13 u
1525A Recirculation Spray 10" 2.14 546/ Containment &
123 Dl 8"
2.17
()
560 Recirculation Spray 123 El 10" 28209
- 24753 31976 16024 32616 2.18 1 of 7
- u
~
..-1~. '
hl284622-lx 061'07/79 046 SURRY POWER STATION, UNIT l TABLE 3-1 (Cont>
PIPE STRESS REEVALUATION
SUMMARY
~reliminar:x Prob-Line Pi12e Stress (~s;U lem System Iso.
Size Original Original NeY Ney Alloy-
~ Name
~ CNPS)
Total*
Seismic* Total Seismic able 5461' Containment &
123 F3 8"
28209 24753 32616 2.21 5600 Recirculation Spray
- 1011 2.22 5461' Containment &
123 F2 8"
28209 24753 32616 2.25 e:*
5620 Recirculation Spray 10
2.26 548C Containment &
123 H2 10 11 15785**
11241**
29970 2.29 Recirculation Spray 2.30
~--~J 547 Containment &
123 Cl 8"
20953 5688 21960 19284 31482 2.33 Recirculation Spray 10 11 2.34
):,,
,.. '\\
7441' Containment &
123 Jl 8" 26721**
NA 33750 2.37
--l 754 Recirculation Spray 2.38 j;;!
n 548A Containment &
123 Bl 8"
11955 11256 31428 2.41
- c Recirculation Spray 1011 2.42 m
- z
-I 548B Containment &
123 Hl 1011 28660 26790 23251 18529 32616 2.45 i-,
Recirculation Spray
~.46 544 Containment &
123 Gl 10 11 13402 6986 6386
- 3766 28485 2.49 n
0 Recirculation Spray 123 G2 2.50 rl-_,,
544A Containment &
123 R2 10" 12853 11256 6814 3556 29970 2.53
/
C:
Recirculation Spray 2,54.
CD -
- a.
544B Containment &
123 Rl 10"
. 12853 11256 6628 4541 28485 2.57 Recirculation Spray 2.58 751 Containment &
123 Nl 1011 6010 5169 7085 5206 28485 3,3 Recirculation Spray 123 N2 3.4
\\,....)
- j 5621' Containment &
123 Fl 10" 12610**
NA 30000 3.7 i
546 Recirculation Spray 123 E2 3.8
)!
745 Containment &
123 Kl 8"
25702.
24579 33750 3.11 u ii Recirculation Spray
. 3.12 ii Main t
323A Steam 100 Dl 30 11 13824 6343 13064 354 27000 3.15 u
322A Main Steam 101 Dl 30" 13031 5548 11532 400 27000 3.17 2 of 7
<.)
w
___ lL
II hl284622-lx 061'071'79 046 SURRY POWER STATION, UNIT l TABLE 3-1 (Cont>
PIPE STRESS REEVALUATION
SUMMARY
PreliminarI Prob-Line Pi11e Stress C11si}
lem System Iso.
Size Original. Original New New Allow-
~ Name
~ CNPS)
Total*
Seismic* Total Seismic able 334A Main Steam 102 Dl 3011 18635 11082 15407 463 27000 3.19 346 Main Steam 103 Al 3011 19970
- 12563 33750 3.21 e.',
323B Feedwater 100 Gl 1411 15829 590 12923 8061 27000 3.23 322B Feedwater 101 Gl 1411 17927 13521 15965 1796 27000 3.25 334B Feedwater 102 Gl 14" 16025 12281 16145 9828 27000 3.27 417 Auxiliary Feedwater 118 Al 311 8568 NA 26769 14036 27000 3.30 118 A2 3,31 607 Auxiliary Feedwater 118 Gl 411 18681 NA 18331 5467 27000 3.34
):=,
-I 118 G2 6"
3.35 i!
n 636 Pressurizer Spray 125 Al 411 17671 9527 28800 3.38
- c
\\....J.
3:
& Relief
.3. 39 l'T'l z
630 Pressurizer Spray 124 Al 3"' 411 2Q500'1EM NA 33372
'3.42
-I
& Relief 124 A2 6 II t 12 11 3.43 1-1.
i.,..,
540 Residual Heat 117 Bl 311' 411 14746**
3374**
3.46 n
0 Removal 6"' 12 11 3,47
- s M- *-'
508 Residual Heat 117 Al 10 11
' 12 11, 16627 12375 3.50
- s Removal 117 A2 1411 3.51 C:
m a.*
465 Service Water 119 Al 2411 19101 18285 7778 5826 21600 3.54
- ...,,)
I 119 A2 3.55
- 1 119 A3 3.56 119 A4 3.57
()
488/ Component Cooling 112 C 18" 26723**
NA 27000 4.2 480 112 Al 4.3 507/ Component Cooling 112 Fl 8"
26695**
NA 27000 4.6 u
481 112 Bl 1811 4,7 614 Compone~t Cooling 112 AEl 12 II 254,48**
- NA 27000 4.10 u
lT2* AE'2 ra,i 4.11 3 of 7 u..
't
~
w I I
I h1284622-lx Prob-lem li.Q.:_
Preliminary System Name 512 Cq~ponent Cooling 603A Component Cooling 766 Component Cooling 605A Component Cooling 605B Component Cooling 509A Component Cooling 612 Component Cooling 1512 Component Cooling 2529 Component Cooling 2526 Component Cooling 2527 Component Cooling 527A Component Cooling 517 Component Cooling 603B Component Cooling 526A Component Cooling 06/07/79 046 SURRY POWER STATION, UNIT 1 TABLE 3-1 (Cont)
PIPE STRESS REEVALUATION
SUMMARY
Iso.
Line Size CNPS)
Original Total*
Pipe Stress (psi>
Original New New N..Q..:,_
Seismic* Total Seismic 112 ANl 18" 112 Sl 18" 112 AR 8"
112 T 26792**
NA 18710 13461 10449:KM NA 112 AAl 3", 611,
12420 112 AA2 18" 112 AAl 3", 611,
12420 112 AA2 18" 112 Gl 811, 12",
26566**
18", 24" 112 AKl 18" 1795810(
112 J 18" 26792**
112 AH 3", 6",
2 6 7 01 *
- 8 11, 10",
1411 D 18 11 112 AJ 2'2" 6" 1 NA 811 1 Io 11 112 AL 4 11 1 611,
811 I 10 11 112 Tl 112 Ml 112 M2 112 M3 4", 611,
811, 10",
14" 411, 6 11,
8 11, 10",
1411 D 18 11 112 Sl 18" 112. L3 61.1,
- 811, 12728**
10449**
17958**
18710 NA
- 4 of 7 6727 6727 NA NA NA NA NA NA NA NA 13461 NA Allow-able 27000 27000 27000 27000 27000 27000 27000 27000 27000 2700.0
- 27000 27000 27000 27000 4.14 4.16 4.19 4.20 4.23 4.24 4.27 4.28 4.31 4.32 4.35 4.37 4.40 4.41 4.42 4.45 4.46 4.49 4.50 4.53 4.54 4.55 4.58 5.1 5.2 5~5 5.7 n
0 :l rt
- l C:
fl>
- 0.
- r*-...,
- .._',.r*
,.:.J
'....).
UI
()
l) 0
II hl284622-lx 06/07/79 046 SURRY POWER STATION, UNIT 1
..,,..,..~
TABLE 3-1 (Cont)
PIPE STRESS REEVALUATION
SUMMARY
Preliminary Prob-Line PiRe Stress C:Qsi>
lem System Isa.
Size Original Original New New Ailow-
~ Name li2..:_
CNPS)
Total*
Seismic* Total Seismic able 526B Coml)onent Cooling 112 Ll 6", 8" NA NA 5.9 526C Component Cooling 112 Ll 6"' 8" NA NA 5.12 112 12 5.13 112 13 5.14 527B Component Cooling 112 T2 411' 6"'
12728**
NA 27000 5,17 8"' 10 11,
5.18 14" 5.19 527D Component Cooling 112 T3 411 I 6"'
12728~0E NA 27000 5.22 8"' 1011,
5.23 14" 5.24
):a
-t
-t 509B Component Cooling 112 G2 811, 12 11,
26566~0E NA 27000 5.27
):a n
18", 2411 5.28
- c 3:
509C Component Cooling 112 G3 8 II I 12 11,
26566~0E NA 27000 5.31 l'Tl z
18", 24" 5.32
-t 509D Component Cooling 112 G4 811,.12 11,
26566lElE NA 27000 5.35 18", 24."
5.36 n
0 137 Al 17554 1209 27000 5.39
- s CVl Containment Vacuum 8"
25750**
NA r+
. __ /
746 3" HP Steam 131 Al 311 26766**
27000 5.42
- s
.NA
' C: -
131 B2 411 5.43 Cl>
- 0.
131 C3 5.44 CFl Fire Protection 144 Al 2 II I 6"'
22255**
NA 27000 5.47 12" 5.48 CF2 Fire Protection 144 Bl.1 l/2 11,
10966**
3733**
27000 5.51
- ..J 2", 16 11.
5.52 1040 Diesel Muffler Exhaust 143 Al 24" 11857**
NA 22500 s.ss
(.)
Other Problems CHand Calculations and SHOCK0/1) 6.1 1000A Low Head Safety 127 Jl 2 II I 6" NA NA 6.4 u
Injection
~~ J C
6.5 5 of 7 u
0 II
hl284622-lx 06/07/79 046 SURRY POWER STATION, UNIT 1 TABLE 3-1 (Cont)
PIPE STRESS REEVALUATION
SUMMARY
Preliminary Prob-Line Pi'.Qe Stress C:Qsil lem System Isa.
Size Original Original New New Allow-lliw.._
Name fu1..:_
CNPS>
Total*
Seismic* Total Seismic able 1010A Low Head Safety 127 J2 2 II NA NA 24709 54*23 33750 6.8 Injection 6"
6.9 1020A Low Head Safety 127 J3 2"
NA NA 28401 6168 33750 6.12 Injection 6"
6.13 1020B Low Head Safety 127 J4 6"
NA NA 12305 5587 33750 6.16 Injection 6,i7 1020c Low Head Safety Injection 1n JS 6"
NA NA 22453 1270 33750 6.20 6.21 1030 Service Water 1119 Al 24 11 NA NA 3092 1421 21600
. 6.24
)>
-I 537 Low Head Safety 122 Al 4", 611 19247 13944 22928 17042 25789 6.27 n
Injection 122 A2 10", 12" 6.28
- c 3:
IT1 755 Containment &
123 Pl 12" 3950 2235 2400 867 33750 6.31 z
Recirculation Spray 6.32
-I 1-t 756 Containment &
123 Ql 12" 2077 1230 2638 1205 33750 6.35 Recirculation Spray 6.36 n
0
- s 611 Auxilary Feedwater 118 Ll 411' 6" 20347lOE NA 27000 6.39 rt
- s 32238 554 Residual Heat 117 Cl 6"
16627 12375 25955 21992 6.42 C: -
Removal 6.43 ro 0..
606 Component Cooling 112 ABl 12" 15033 13776
. 27000 6.46
.. J 613 Component Cooling 112 AD1 311 I 4"'
16729
- 15882 27000 6.49 8 II f 6" 6.50
'.. J 502 Component Cooling 112 Dl 18" 11172 7275 27000 6.53 506 Component Cooling 112 El l 811 8893 6807 27000 6.56
':..J 112 E2 6.57 112 E3 6.58 747 Spent Fuel Cooling 128 Al 12"
'6384 5604 28800 7.3 u
748 Spent Fuel Cooling*
128 Cl *12", 16" 18510 5777 28800 7.5 6 of 7 v
0
~~-*
11r!' iilft! ii
->l':a-H*iil*t'H*!'iAA FN*~~** -:F!" P
- i'IJRll*IJ"'"-
~~~r:n~*-l"!'i'l~,.,.m~****-*-...... T'...fi,..Y~'**v--~*n
j...
\\:..........
I..",:
\\_,.. --
i \\.'
- ~J I h1284622-lx Prob-lem li9..i_
749 NOTE:
Preliminary System Name Spent Fuel Cooling 0
06107/79...
046 SURRY POWER STATION; UNIT 1 TABLE 3-i (Cont)
PIPE STRESS REEVALUATION
SUMMARY
Line Pi~e Stress t~~il Iso.
Size Original Original New.
New Allow-
~ CNPS)
Total*
Seismic* *Total
£e i__5_m i_e_
able 128 Bl 12" 23333 19596 28800 Ihe original total and-original seismic stresses.shown in Table 3-1 were computed using the
~HOCK2, SHOCKl, or SHOCKO programs or hand calculated for the original design conditions.
Ihe new total and new seismic stresses were computed by the NUPIPE program using different mass models £nd, in some cases, different ARS's than the original calculations.
tlore importantly, the reanalyses were based on field-verified, as-built conditions in 1979, Rhich, in some cases, differ significantly from the original design conditions. for this reason, the new stresses and the original stresses in Table 3-1 are not comparable, as they go not necessarily represent the* same physical conditions.
Legend:
Allowable Stress* 1.8 Sh New Total Stress
- SLP + Sow + 1.s SoaEI + SoBEA (for SSI/ARS).
New Seismic* 1.5 SoaEI (for SSI/ARS)
. New Total Stress = Sow + SLP + SoBEI + SoaEA (original ARS)-
New Sefsmfc
- SoaEI (original ARS)
- 7 of 7 7.7 7.10 7.12 7.14 7.16 7.19 7.22 7.24 7.26 7.27 7.28 7.29
.)>
-f r,
- c 3:
IT1 z -f r,
0 :l
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- l C:
~ -
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- ~...
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~-~.*.)
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0 u
i Uj wl
e ATTACHMENT II 06/08/79 SURRY POWER STATION - UNIT 1
- 1.
Decision Matrix for Significant Items Found After Interim Startup The following is an elaboration of the evaluation.procedure, referenced in Section 5 of the Stone & Webster Engineering Corporation (S&W) Report 1 that shall be used in the continuing reanalysis program to determine the significance of and to define the reporting requirements for possible equipment modifications identified after start of interim plant operation.
This procedure addresses:
o Piping arid equipment nozzles for which stress levels exceed.
code allowable values.
o Pipe hangers for which loads exceed allowable.capacities.
1.1 Piping and Equipment Nozzles If calculated stress levels exceed code allowables after an analyst has
.reviewed all model parameters (i.e., the translating of information on piping isometric drawings into a computer program, input data), the supervisor will review.all computer input, program controls, data sources, and all output printouts.
The supervisor's _review will include an overview of all the input/output work and a review to determine whether a more detailed mathematical model will reduce the conservatism or whether any of the input assumptions used to simplify problem solutions
- should be revised to create a longer but more correct calculation.
If the supervisor's review does not result in a reduction of calculated stress to allowable levels, he will contact a system engineer familiar ::.
with the actual operating modes and conditions to determine if any of the input conservatisms can be reduced based on the way operating condi~
tions are actually combined as opposed to the way they are normally combined to produce worst possible case inputs to the problem and computer runs.
Th'is may require multiple computer reruns but produces less overconservatism than combining inputs for a worst case run.
If the stress level is still in excess of code allowable levels, a system engineer and an experienced stress engineer will immediately review the problem together to determine if the answer is comparable to:
what past experience would indicate is to be expected for this system
. arrangement.
If it appears unreasonable based on experience, they will immediately review with the supervisor all the conditions or input parameters that contribute to the largest component of stress.
They will continue their review for all inputs until they determine that an incorrect input 1Report on the Reanalysis of Safety-Related Piping Systems, Surry Power Station - Unit 1, June S, 1979.
Section S, page 2, line 2.12.
1: *of 4
~ ATTACHMENT II (Continued)~
06/08/79 SURRY POWER STATION - UNIT 1 exists or conclude that the stress is correctly calculated and is over allowable levels, If the stress is on an equipment nozzle, the vendor will be requested to perform an expeditious review to determine whether total nozzle capability is exceeded.
If total stress is over allowable levels, the prime causes of the overstress condition will be determined and the S&W stress review project engineer will be notified.
The S&W stress review project engineer will notify the Virginia Electric and Power Company (VEPCO) Station M~nager that an over-code allowable stress has been calculated, and will define the location in the pipe system and possible corrections.
He will then initiate a study of the overall impact of this stress level and provide additional information as it develops, to the Station Manager.
The reporting process to the NRC has been previously established and will be in compliance with the Surry Technical Specification Section T.S. 6.6.
2 of 4
e ATTACHMENT II (Coutinued) e 06/08/79 SURRY POWER STATION - UNIT 1 1.2 Pipe Hangers The following flow diagram represents the procedure used in making decisions in the evaluation.of pi1e hangers.
- 1.
Calculate actfa1 allowable hanger
- 2.
- 3.
- 4.
- 6.
capacity from sketches l,
Compare load from pipe stress run with actual allowable hanger capacity/
-~
Hanger overloaded Hanger OK
~
Would SSI~ARS reduce Complete loa~*
Yes No~
r!itiate SSI, Is new load greater than 90 percent of computer run.
yield strength or does an anchor bolt Go to 2.
have less than a 3-to-1 safety factor?
Yes~No Stress eng1.neer r~
- 5.
List htngers to
~
be reworked during
~
-~
steam generator If pipe stress Pipe requires repair outage.
- appears OK with active support support ineffec-to maintain allow-tive, go to 5.
able strjss levels.
f S&W Stress Review Project Engineer w~ll inform Station Manager that support exceeds "Notifica-tion of NRC Stress Criteria" and is in final review to determine any other mitigat-ing conditions.
3 of 4 7.-
l S&W Engineers Review.
- a. *Determine what portion of the load is due to self-limiting forces, e.g., anchor displace-ments which have only a finite travel possibil-ity, dead load which would be transferred to adjacent supports after dispiace-ment of this hanger, and reduce total by this dmount for temporary use.
i
. -**- *.. :. :..-~L*-***** ****-
- -t*----
... **.. _;-;. -.. _ -*~... ~;;\\*".,.,-----.;:--y-*,*- ~~~.---,*. ___ ::7,_._. ____._.... _____.:,.,:_.....~. --- *..,.J,. __. ________,_,. --*-*---~----------~-'-;-**---*-**' ~~----*.l;~
- ATTACHMENT II (Continued) *.,
06/08/79 SURRY POWER STATION - UNIT 1
+....*
- 8.
- b.
- c.
Calculate hanger deflection associated with load to determine if strain is less than 50 percent of ultimate stress elongation..
R
~
h.. 1*
eview system p ysica equipment arrangement to determine if stress would increase in equipment or valves or only in piping.if hanger was ineffective.
- d.
Review loads and hanger design to determine if m~deling and computer anal-ysis (STRUDL) will yield more accurate and lower stress *i..
Advise Station Manager of hanger evaluation.results and reasons why shutdown is or is not required.
- 9.
Station Manager determines impact on system operability and technical specifications and notifies NRC, I&E, Region II in accordance with technical specification procedures.
4 of 4
e ATTACHMENT II I PRIORITY LIST FOR REANALYSIS (SHOCK II AND HAND CALCULATIONS)
Scheduled Scheduled Problem/
Stress Supports Priority System Completion Completion Comments 1
636/PSR Comp
.June 25 RC Boundary 2
630/PSR June 15 July 2 RC Boundary 3
55S/LHSI
- Comp June 30 RC Boundary 4
1555/LHSI Comp July 2 RC *Boundary 5
508/RHR June 15 July 3 RC Boundary 6
1020A/LHSI Comp July 10 RC Boundary 7
707A/LHSI Comp July 10
.RC Boundary 8
706A/LHSI June 15 July 12 RC Boundary 9
lOOOA/LHSI Comp July 12 RC Boundary 10 708/LHSI June 12 July 16 RC Boundary 11 lOlOA/LHSI Comp July 16 RC Boundary 12 540/RHR June 15 July 16 13 743/LHSI Comp July 17 14 731A/LHSI Comp.
July 17 15 7318/LHSI Comp July 18 16 1020B/LHSI Comp July 18 17 1020C/LHSI Comp July 18 18 727/LHSI June 22 July 16 19 735/HHSI June 20 July 20 20 417/AFW Comp July 2.3 Page 1 of 4
e e
ATTACHMENT III (continued)
Scheduled
. Scheduled Problem/
Stress Supports Priority System Completion Completion Comments 21 607/AFW Comp July 24 22
- 746/HPS June 15 July 25 (525A/C&RS 23
( 1525A/C&RS June ll(Rev 3).
July 30 24 547/C&RS June 8 (Rev 3)
Aug. 2 25 548A/C&RS June 15
. Aug. 2 26 744, 754/C&RS June 15 Aug. 4 27 745/C&RS June 15 Aug. 6 28 544/C&RS June 15 Aug. 7 29 544A/C&RS Comp Aug. 8 30 5448/C&RS Comp Aug. 8 31 751/C&RS Comp Aug. 11 32 5488/C&RS Comp Alig. 1 33 548C/C&RS June 15 Aug. 13 34 546,560/C&RS Comp Aug. 13 35 546,5600/C&RS Comp Aug. 15 36 546,5620/C&RS
- Comp Aug. 17 37 562/C&RS June 15 Aug. 20 38 323A/MS
. Comp Aug. 21 39 322A/MS Comp Aug. 21 40 334A/MS Comp Aug. 22 41 346/MS June 22 Aug. 27 42 3238/FW Comp Aug. 29 43 3228/FW Comp Aug. 30 44 3348/FW Comp*
Sept.* 1 45 465/SW Comp Comp Page 2 of 4
~
9 ~*
e ATTACHMENT III (continued).
Scheduled*
Scheduled Problem/
Stress Supports Priority System Completion Completion Comments
. 46.
.488/480/CC
- June 30 Sept. 15 47 507/481/CC June 30 Sept. 15 48 614/CC June 30 Sept. 15 49 512/CC June 30 Sept. 15 50 603A/CC June 30
- Sept. 15 51
- 766/CC June 30 Sept. 15 52 605A/CC June 30 Sept. 15 53 6058/CC June 30 Sept. 15 54
- 509A/CC June 30 Sept. 15 55 612/CC June 30 Sept. 15 56 1512/CC June 30 Sept. 15 57 2529/CC June 30 Sept. 15 58 2526/CC June 30 Sept. 15 59 2527/CC June 30 Sept. 15
- 60 527A/CC June 30 Sept. 15 61.
517/CC June 30 Sept. 15 62 6038/CC June 30 Sept. 15 63 526A/CC June 30 Sept. 15 64 5268/CC June 30 Sept. 15 65 526C/CC June 30 Sept. 15 66 527B/CC June 30 Sept. 15 67 527D/CC June 30 Sept. 15 68 5098/CC June 30 Sept. 15 69 509C/CC June 30 Sept. 15 70 5090/CC June 30 Sept. 15
- page 3 of 4
ATTACHMENT III (continued)
- Scheduled Scheduled Problem/
Stress-Supports Pri oritt sistem Completion
- Completion Comments 71 CV-1/CV Comp Sept. 4
- 72 CF-1/FP June 30 Sept. 15 73 CF-2/FP June 30 Sept. 15 74 1040/DM June 15 Sept. 8 75 1030/SW Comp Sept. 8 PRIORITY LIST FOR REANALYSIS SHOCK 0 Problem/
Stress
. Supports Pri orit,Y sistem sistem Completion Comeletion 76 537 LHSI Comp Sept. 12 77 755*
cs Comp Sept. 13 78 756 cs Comp Sept. 13
_j 79
.611 AFW June 15 Sept. 15
.80 554 RHR Comp
- Sept. 16 81 606 ;
- cc June 30 Sept. 15 82 613 cc June 30 Sept. 15 83 502.:
cc June 30 Sept. 15 84 506.**
cc June 30 Sept. 15 85 747 s*
SFPC June 30 Sept. 15 86 748 SFPC June 30 Sept. 15 87 749
- SFPC June 30 Sept. 15 Page 4 of 4
. I