ML20117A126
| ML20117A126 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/30/1985 |
| From: | John Marshall COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 0016K, 16K, NUDOCS 8505080075 | |
| Download: ML20117A126 (19) | |
Text
-
l
^"
Commonwealth Edison
.y s
(
), one First Nttional Plaza, Chicago Ilhnois
(
~J Address R: ply to: Post Office Box 767
\\
,/ Chicago. lilinois 60690 April 30,1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
LaSalle County Station Units 1 and 2 Proposed Elimination of Pipe Whip Restraints Associated with Arbritrary Intermediate Pipe Breaks Facility Operating Licenses W F-11 and W F-18 NRC Docket Nos. 50-373 and 50-374 References (a): FSAR Section 3.6.2.1 (b): FSAR Appendix C (c): Byron /Braidwood letter to D. L. Farrar from B. J. Youngblood dated January 7, 1985.
Dear Mr. Denton:
Pursuant to 10 CFR 50.59, Commonwealth Edison proposes to amend the criteria used to define break and crack location and configuration as described in Reference (a) to permit removal of pipe whip restraints associated with postulated aribritary intermediate pipe breaks. Justifi-cation for this proposed change is provided in the attachment to this letter. A similar request on the Byron /Braidwood dockets was recently approved by your staff in Reference (c).
Commonwealth Edison has reviewed this proposed change in accordance with 10 CFR 50.59. Although no change to the Techncial Specifications is indicated by our review, we nonetheless request your prior approval and concurrence that this change does not constitute an unreviewed safety question.
Please direct any questions you may have concerning this matter to this' office. Pursuant to 10 CFR 170.12, a' fee remittance of $150.00 is enclosed.
f Ih 8505080075 8 h 73 l(Of ()
- DR ADOCK O sjy 4 po16
e_
e H. R. Denton April 30, 1985 Three (3) signed originals and thirty-seven (37) copies of this transmittal and its attachments are provided for your use.
Very truly yours, h !G H J. G. Marshall Nuclear Licensing Administrator 1m cc: Resident Inspector - LSCS A. Bournia - NRR G. Wright - State of Ill.
SUBSCRIBED AND SWORN to before me this SoMday of /144U;
, 1985 H lt b Afahru
'- Notary.Public 0016K i
m.
,i LASALLE COUNTY UNITS 1 & 2 ARBITRARY-INTERMEDIATE PIPE-BREAKS
'l.0 Introduction Commonwealth Edison Company (CECO) has followed closely the recent activities of the Nuclear Regulatory Commission (NRC) staf f and the nuclear industry relative to the treatment of design-basis pipe breaks in high energy piping sys tems.
In par ticular, it is noted that the NRC staff has expressed a positive interest in the industry's proposal to modify the historic pipe break criteria to eliminate f rom design consideration those intermediate pipe breaks generally referred to as arbitrary intermediate breaks, i.e.,
those breaks which, based on stress analysis, are below the piping stress limits and/or the cumulative usage factors specified in the initial NRC criteria, but which were inicially selected to provide a minimum of two breaks between te rminal ends.
NRC s taf f and industry discussions with the Advisory Committee on Reactor Safeguards. ( ACRS) on March 29 and June 2, 1983. have indicated general agreement with the elimination of the arbitrary intermediate breaks.
That elimination accrues considerable design benefit due to the deletion of the associated pipe whip restraints and related provisions which were to mitigate the ef fects of intermediate pipe breaks.
Additionally, operational advantages ensue from decreased numbers of pipe whip restraints to be inspected and maintained for 40 years.
~
t 2.0 Break Selection-Criteria The break selection criteria initially employed by CECO. for the LaSalle County Units 1 & 2 Stations was based upon NRC Branch Technical Position MEB 3-1.
That position required that pipe breaks be postulated at terminal ends and at intermediate locations where, depending on the pipe class, stresses or cumulative usage factors exceed specified limits.
If two intermediate locations could not be determined based on the above because the stresses and cumulative usage factors were below the specified limits, then the two highest stress locations were selected.
3.0 Industry Experience CECO concurs with other nuclear utilities in the belief that current knowledge and experience support the conclusion that designing for arbitrary intermediate breaks is not justified and that this requirement should be deleted.
This conclusion is supported by extensive p.
~
1 v so e
N; u k Yb$k.
. _ ~ -,
y.
,.m_..
n
,m-_,.
D operating experience in over 80 operating U.S.
plants and a number of similar plants overseas in which no piping failures have been known to occur that would suggest the need to design protective features to mitigate the dynamic ef fects of arbitrary intermediate breaks.
Arbitrary intermediate breaks are of ten postulated at locations where maximum pipe stresses are significantly less than the ASME Code allowables and within a few percent of the stress levels at other points in the same piping system.
Mitigating such breaks with pipe restraints results in complicated protective features at arbitrary specific break locations but does little to enhance overall plant safety.
. Benefits from-Removal 4.0 Elimination of the arbitrary intermediate break locations results in the elimination of the associated pipe whip restraints and other structural provisions to mitigate the consequences of these breaks.
Significant operational benefits are also realized over the 40-year life of each plant.
As identified in NUREG/CR-2136, these benefits accrue in the areas of plant reliability and reduced exposure of plant personnel to radiation during inspection of this excessive number of pipe whip restraints.
4.1 Access Access during plant operation for maintenance and inservice inspection is improved due to decreased congestion f rom these restraints and their supporting structural steel.
Also, fewer restraints must be removed to gain access for weld inspections.
In addition to the decrease in maintenance ef fort, a corresponding reduction in man-rem exposure can be realized from fewer manhours spent in radiation areas.
-4.2 Recovery from Unusual Conditions Recovery f rom unusual plant conditions can also be improved by elimination of conges tion due to excessive pipe restraints.
In the event of a radioactive release or spill inside the plant, decontamination is much simpler when complex shapes, represented by the structural f rameworks suppor ting the restraints, are eliminated.
This results in decreasing man-rem exposures associated with decontamination and restoration activities.
Similarly, access for control of fires within certain c.
areas of the plant would be improved, especially under low' visibility conditions.
Substantial overall benefits in these areas can be realized by redu%ing the number of whip. restraints required.
2
4.3 Decrease in Heat Loss By design, whip restraints fit closely around the high energy piping with gaps typically being on the order of half an inch.
These restraints and their supporting steel significantly increase the heat transfer to the surrounding environment.
Also, the insulation must be cut back in these areas.
This is done because ' thermal movement of the' piping system during start-up and shutdown can deform the piping insulation against the fixed whip restraint thus reducing insulation ef fectiveness.
This heat loss cortrDnites to the over-temoeratures 90 the LaSalle containments.
The elimination of whip
- restraints associated with arbi.trary interme.diate breaks would assist in reducing the normal environmental temperatures.
4.4 Removal Because the restraints are already in place at LaSalle, the. restraints determined not,to be required would be removed as access time and alara considerations allow during plant shut down, conditions.
5.0 Alternative Pipe Break Criteria-Based on the preceding information, CECO requests NRC approval of the alternative pipe break criteria given in 5.1 and 5.2 below which eliminate the original arbitrary in te rmediate pipe breaks, i.e.,
those breaks which, based on stress analysis, are below the stress limits and the cumulative usage f actors specified in the current NRC criteria, but which were initially postulated to provide a minimum of two breaks between terminal ends.
Application of the alternative pipe break criteria described below does. not alter CECO's commitment to quality that has been used in the design of. safety related structures, systems, and components.
The quality assurance program ensured that safety related structures, systems, and components have been designed, f abricated, erected, and tested -to the quality standards commensurate with their safety function.
5.1 ASME Section -III Piping Inside Csntainment o
Piping systems are designed to accommodate pipe breaks at terminal ends and locations where the
'#~~'
stress or usage f actor criterion of MEB 3-l' is
. exceeded.
No arbitrary intermediate pipe breaks are postulated where the stress and/or usage-factor criteria are not exceeded.
3
- u.
C 4,
... ~
N. --
o o
Elimination of arbitrary intermediate pipe breaks must not adversely af fect the remaining pipe breaks and their associated pipe whip restraints.
o For plant flooding evaluations, environmental qualification of equipment, and structural design of equipment in areas traversed by high energy piping sys tems, pipe breaks will continue to be postulated in accordance with the present project criteria, i.e.,
in each area traversed by the high energy piping system, non-mechanistic breaks are postulated at the locatio.n that results in the most severe e nvi'r o nme,n tal' 'co nseque'nc e s.
The re fo re,:
elimination of the arbitrary intermediate pipe breaks does not impact environmental qualifi-cation of equipment nor plant structural design.
PlpTniOTt'sl'dE' Cont'y Desi,g,ned -Non-ASME~
and Seismicall 5.2 ASME Section III aTn~5Fnt.
Section III
~
Piping systems are designed to accommodate pipe o
breaks at terminal ends and locations where the stress criterion of MEB 3-1 is exceeded.
No arbitrary intermediate pipe breaks are postulated unless the stress criterion is exceeded.
o Elimination of arbitrary intermediate pipe breaks must not af fect the remaining pipe breaks and their associated pipe whip restraints, o
For plant flooding evaluations, environmental qualification of equipment, and structural.
design of equipment in areas traversed by high energy piping systems, pipe breaks will continue to be postulated in accordance with the present project criteria, i. e., in eech area traversed by the high energy piping system, non-mechanistic breaks are pcstulated at the location that results in the most severe environmental consequences.
Therefore, elimination of the arbitrary intermediate pipe breaks does not impact the environmental qualification of equipment nor plant structural design.
6.0 Eliminated _ Pipe Breaks, c-Attachment A lists by subsystem those arbitrary intermediate pipa breaks and associated pipe whip restraints which can be eliminated f rom the design because the stress and usage factor limits are not exceeded.
The 4
~
FSAR will be revised after NRC approval of this submittal to show the physical location of the restraints within a given subsystem.
The application of the proposed alternative pipe break criteria results in the elimination of 10 break locations and 11 pipe whip restraints per unit.
6.1 Elimination of Breaks Not Yet Identified The existing guidelines in MEB 3-1 of the SRP (NUREG-0800) Revis' ion 1 will be met for those piping systems, or portions thereof, which are not included in this submital.
If other piping subsystems included within the systems identified' in Table D-1, but not specifically identified in this submittal, subsequently qualify for elimination under the alternative pipe break criteria of Section 5.0, they will be appropriately identified to the s taf f.
7.0 Additional Technical Justification In this submittal CECO is providing additional technical information to justify this request.
Specific NRC concerns are addressed in Attachments 3 through F as follows:
1.
Technical justification for elimination Attachment B of arbitrary intermediate pipe breaks 2.
Provisions for minimizing intergranular Attachment C stress corrosion cracking in high energy lines 3.
Provisions for minimizing the ef fects Attachment D of thermal and vibration induced piping fatigue 4.
Provisions for minimizing water / steam Attachment E hammer effects 5.
P.rovisions for minimizing local Attachment F stresses from welded attachraents
[
A
+
5
8.0 Conclusion CCCo has reviewed the basis for the postulated intermediate pipe breaks on designated high energy lines and has compared the design stresses and usage factors with the initial SRP MEB 3-1 Guidelines.
On the basis of ASME Code calculations, there is no technical justification for such postulated intermediate pipe breaks.
The probability of pipe rupture at the values of stress and usage assignable to these intermediate pipe breaks is extremely remote (10-9) to 10-11) whether in carbon steel or stainless steel pipes.
Elimination of these arbitrary intermediate pipe breaks affects the need for pipe whip restr.aints originally designed to mitigate these breaks.
The evaluation of these breaks under Alternative Criteria that fully conforms to the SRP MEB 3-1 guidance, but which omit these arbitrary intermediate pipe breaks on the high energy lines, results in a significant decrease in the number of pipe restraints required for high energy line breaks.
Removal of those pipe restraints associated with the arbitrary intermediate pipe breaks results in several benefits.
These benefits are identified in Attachment G.
Also avoided are the maintenance and surveillance of these surplus restraints over the 40 year life of the plant.
6
~
5 e
ATTACHMENT A Summary of Class 1 Piping Intermediate Break and Pi;ie Whip Restraint Reductions Pipe Whip Bre'aks/
Restraints /
In te rmedia te Break ID Restraint ID System Subsys tem Break Locations Eliminated Eliminated Main Steam MS02 (RCIC)' -Elbow welds between 2/Cl21 4/R46 horizontal leg following
/Cl23
/R47 valve /ver tical riser and
/R48 bottom of riser on Re-
/R50 actor Core Isolation Coolant line.
Residual RH01, RH04,
-Elbow welds at bottom of' 4/C69 2/R18 Heat RIIOS, RH06 12" riser containing valve
/C83
/R30 Removal E12-F090A.
/C87 System
-Elbow weld on riser between
/C91 RPV and valve for RHR lines 40BA, 400B, and 533.
Reactor RR01
-Elbow welds on horizontal 2/d95 2/R35 Water leg prior to riser on 4"
/C97
/R36 Cleanup branch line between the 6" line and closed valve.
Main Steam MS25
-Half couplings at 2" 2/C5 3/SR06 Miscellaneous branch lines 1MS14 AB and
/C9
/SR10 Piping 1MS14AC off 3" header.
/SR12 To tals i0 11 e
9 A,-
Attachment B TEC.HNICAL_JUSTIFICATIOJLF_OJ3 ELIMINATION OF ARBITRARY INTERMEDIATE BREAKS The following items provide generic technical justification for the elimination of arbitrary intermediate pipe breaks and their associated pipe whip restraints.
1.
Operating procedures and pipe and system designs minimize the possibility of intergranular stress corrosion cracking, thermal and vibration induced fatigue, and wa te r/s team hamme r in these lines in which arbitrary pipe breaks are currently postulated.
Detailed design provisions for these phenomena are provided.in Attachments C, D,
& E, respectively.
2.
. Welded attachments are not located in close proximity to the breaks to be eliminated.
Consequently, local bending stresses resulting f rom these attachments does not significantly af fect the stress levels at the break locations (refer to Attachment F).
3.
The remaining postulated pipe breaks and whip ~ restraints are not af fected by removal of the arbitrary intermediate breaks.
4.
Pipe breaks are postulated to occur at locations where, depending on the pipe class, stresses are only 80% of Code allowables or where the cumulative usage f actor is only 10%
of the allowable 1.0.
The arbitrary breaks to be eliminated all exhibit stresses and usage factors below these conservative thresholds.
5.
Pipe rupture is recognized in Branch Technical Position MEB 3-1 as being a " rare event which may only occur under unanticipated conditions".
6.
Arbitrary intermediate breaks are only postulated to provide additional conservatism in the design.
There is no technical justification for postulating these breaks.
7.
Elimination of pipe whip restraints associated with the arbitrary breaks can facilitate in-service inspection, reduce heat losses from the restrained piping, and reduce the potential for restraining pipe due to unanticipated, thermal growth and seismic motion.
B-1 9
e 9
l
,c.1... _.
_... ~. _.
8.
Pipe break related equipment qualification (EQ) requirements are not af fected by the elimination of the. arbitrary breaks.
Breaks are postualted non-mechanistically for EQ purposes.
It is concluded that the elimination of arbitrary intermediate breaks is technically jus tified, based on the preceding reasons.
g *-
n 0
M-O f b.
%-8 k
4%
4 mJs l
B-2 Y.
v
- -w;
- z c
.w L-
~s n
D
~
ATTACIIMENT C PROVISIONS FOR MINIMIZING STRESS CORROSION CRACKING IN HIGH ENERGY LINES y /~"
Industry experience has shown (NUREG-0691) that intergranular stress corrosion cracking (IGSCC) will not occur unless the following conditions exist simultaneously:
high residual tensile w:..
sdsceptible piping material, and a corrosive 1;,,N#' Stresses, environment.
Elimination of any one of these conditions will preclu,de'-the formation of IGSCC.
%1th6 ugh'any stal'nless' or: carbon steel piping will exhibit'some-7'.
' t
J de' gree of residual stress and material susceptibility to IGSCC, 4Gommonwealth Edison Company has taken positive steps to minimize the potenti al for IGSCC by choosing piping material with low
. q ';
susceptibility to stress corrosion, by conservative design margins and by preventing the existence of a corrosive J e nvi ro nme nt.
The material specifications consider. compatibility with. the system's ope rating environment (bo th in ternal and ex te r nal), as well as other materials in the system, applicable ASME code requirements, f racture toughness characteristics, and welding, processing, and fabrication techniques.
A summary of design changes made at LaSalle to minimize the potential for IGSCC. is included in response to FSAR Ques tion Q 121. 8.
Additionally, the IHSI treatment of stainless steel h6aders on the recirculation loops has been accomplished on Unit 21'aWd is scheduled for Unit'I at the first ref ueling outages 2
of suf ficient duration to accomplish this improvement against-IGSCC.
At LaSalle Units 1 and 2 only a portion of the RHR system is mate of stainless steel.
All other systems where arbitrary intermediate breaks have been postulated are made from ferritic type carbon s teel.
Replacement and removal of stainless steel f rom much of the primary pressure loop piping and equipment was accomplihhed during plant design and erection.
The likelihood of IGSCC in stainless steel increases with carbon content.. Consequently, only the lower carbon content stainless s teels (304, 316) were used for the stainless s teel por tion of the RHR system at the LaSalle Station.
The existence of a corrosive environment is prevented by strict criteria for
. q
' internal and external pipe cleaning, and by water chemistry control during start-up and normal operation.
y I
e
-(~,
e
~4 C-1 s
'.y t
For all other systems with postulated arbitrary intermediate breaks, ferritic type carbon steel was the choice for the piping, fittings, and valve bodies forming the pressure boundaries.
This ferritic material has been found satisf actory from the standpoint of non-susceptibility to IGSCC for the service conditions e nco un te red.
Because mos t BWR-5 sys tems are no t made of stainless steels, the ques tion of IGSCC in stainless steels does not arise for most systems at LaSalle.
In addition, all stainless steel piping has been inspected.
No IGSCC has occurred to date at LaSalle.
IGSCC is not a concern with whip restraints associated with strictly carbon steel piping.
g.
All piping involved in the elimination of arbitrary intermediate breaks has been cleaned and flushed as part of the pre-opera tional tes t program.
The piping has been flushed with demineralized water subject to written criteria for limits on total dissolved solids, conductivity, chlorides, fluorides and pH.
Flush water quality was monitored periodically.
The flushing was controlled by detailed procedures written for each sys tem.
Wa ter chemistry for preopera tional tes ting Vas controlled by written specifications.
During plant ooeration, water chemistry is monitored in the reactor planti.
The major water chemistry standards are included in the plant operating procedures for the lines in which arbitrary breaks were previously postulated.
The water chemistry requirements are provided in the Technical Specifications.
Table C-1 summarizes the sys tems in which currently postulated arbitrary intermediate breaks are to be eliminated.
e e
e 5
e C-2
....;r-O g
c.
n.3
-=-
ATTACHMENT C,
TABLE C-1 Elimination of Arbitrary Breaks and Restraints Systems Summary (Showing the. Preponderance of Carbon Steel. Piping vs. S tainless S teel Piping).
tiumber of Breaks /
Pipe Ope ra ting Restraints Piping System Material Temp. ( F)
Deleted Per Unit Main S team CS 550 2/4 Residual Heat 304 SS 550 1/1 (C69/R18)
Removal Sys tem CS 550 3/1 Reactor Water ~
CS 550' 2/2 Cleanup Main Steam Low 550 2'/3 Miscellaneous Carbon Piping Alloy S teel (5% Cr,Y$Mo)
To tals 10/11 SS - Stainless Steel CS - Carbon Steel C-3 e
h 3
e
~..-~.<,
- *,l,.
ATTACHMENT D PROVISIONS TO MINIMIZE THE EFFECTS OF T!!ERAML AND VIBRATION INDUCED PIPING FATIGUE I.
GENERAL FATIGUE DESIGN CONSIDERATIONS For ' Class 1. lines, fatigue considerations'are addressed by the cumulative usage factor (CUF).
To ensure that piping -
does not f ail due - to f atigue, the ASME Code has established.
^
the CUF limit at 1.0.
By definition, all arbitrary intermediate break locations have CUFs below 0.1.
i
'For C1'hss 2'and 3 lnes, f a'tigue i's considered 'in the
"'~5 allowable stress range check for thermal expansion stresses.
This stress is included in the to tal stress value used to determine postulated break locations.
All arbitrary break locations exhibit stresses less than 80% of the code Lallowables.
If-the number of thermal cycles is expected to be greater than 7,000, then the allowable stresses are.
f urther reduced by an amount dependent on the number of cycles.
II.
.TbERMAL DESIGN CONSIDERATIONS For Class 1 lines anticipated flow conditions that could result in piping thermal transient stresses have been defined.
Piping thermal transient stresses have been.
calculated for these conditions and the stresses have been
' included in cumulative. usage f actors and documented in y
stress reports for the piping.
III. VIBRATION DESIGN ~ CONSIDERATIONS Piping at LaSalle.is designed and supported to. minimize transient and s teady s tate vibra tions.
Preoperational and
~
start-up testing was performed 'to ensure that1 vibration of the piping systems was within allowable limits.
Preoperational and startup piping vibration programs have been completed at LaSalle Units 1&2.
The purpose of the program was to ensure that operational ~ piping vibration did not result in exceedances of allowable stress amplitudes nor result in-undesirable system responses.
The freedom from-restraint or snubber lock-up was also observed; and'a cold / hot walkdown was included.
A dynamic vibration monitoring system is permanently installed at LaSalle, however its usage is directed toward selective monitoring of.
rotating machinery.
D-1 D
I t
r-o ATTACHMENT E PROVISION FOR MINIMIZING STEAM / WATER HAMMER EFFECTS 1.
Water hammer is prevented in the ECCS discharge lines by maintaining the lines in a f ull condition.
The lines are kept full up to the discharge valves by fill pumps which replace any leakage from the lines.
Beyond the discharge valve the line is not drained when the system is not operating, so the discharge lines will remain full from the previous use.
The HPCS is a motor-operated system and has no steam supply line.
Water. hammer will be prevented in.the RCIC system during RCIC 1. -
. s tartup 'by' sequentially : opening the~ RCIC s team supply'
. g.-
isolation valves.
The RCIC steam supply line remains isolated until the reactor pressure reaches 50 psi.
At the 50 psi point the RCIC isolation valves are opened sequentially.
The outboard isolation valve is opened first and any condensation is drained f rom the line.
The bypass valve is then opened to allow condensation to drain from the line and allow the line to warmup to the reactor temperature corresponding to 50 psi; this allows the inboard isolation valve to be opened without water hammer occurring.
Additionally, the steamlines are sloped to allow any condensation in the lines to drain off to drain pots when the sys tem is no t operating.
Therefore, the steam supply line will be a dry steam line even af ter a cold shutdown.
- Hence, the steam supply line will be at the reactor temperature corresponding to 50 psi when the inboard isolation valve is opened, the reby preven ting wa te r hammer.
As the reactor pressure increases the tempera ture. of the line will increase, matching the corresponding saturation points.
2.
As indicated in part 1.
above, the ECCS discharge lines are maintained in a full condition by a jockey pump in each fill line to replace leakage from the lines to the suppression pool (the LPCS and RSR-A share a common pcmp). Pressure limit switches which alarm in the control room are provided to inform the operator of of f-normal conditions.
The pressure E-1 I
,4..
- switch setpoint valves and instrument accuracles are listed t
.below.
IIPCS LPCS RIIR range 0-100 psig 0-500 psig 0-600 psig contact-point 1 50 psig 475 psig 400 psig (Inc)
(Inc) contact point 2 40 psig (Dec) required accuracy
+ 1%
+ 10 psig
+ 2%
rated accuracy
+ 1%
+ 1%
+ 1%
~
3.
The main steam and feedwater systems are expected to experience steam and water hammer loadings, respec ti vely.
Analyses have been performed for. these 1.oadings and the, main steam and feedwater systems have been desidned to accommodate and minimize effects of these loadings.
Main steam piping has been analyzed and designed for the effects of isolation valves closures, turbine stop valve closures, and safety relief valve openings.
The feedwater piping has been analyzed and designed for the-effects of check valve closure caused.by flow reversal from the RPV after a feedwater pump trip.
The main s team and feedwater stress repor ts include the stresses and usage factors calculated from the analyses of these events.
{
o o
e mg E-2 0
9 L_
g>.
ATTACllMENT F PROVISION FOR MINIMIZING LOCAL STRESSES FROM WELDED ATTACIIMENTS CECO has reviewed those arbitrary intermediate break locations to be eliminated and has ~ de termined that in no case are welded attachments placed in close proximity' to postulated break loca tio ns.
As a result, local bending stresses induced by the attachment will not af fect the stresses.at the postulated break point.
To ensure that this is the case, the local stresses have been determined and added to the primary stress reports.
t' t
p o g 4
e i
0 4
6 i
D 9
e 8
e o
F
. w....s ATTACIIMENT G SdMMARY OF BENEFITS FOR THE ELIMINATION OF ARDITRARY INTEIU4EDE5Y5~5E55' fide 5NS'-
~ ~ ~ ~ ~A' TL_ _L E' ~C.F..dliT' ~$.Y_5YiO $
L_ S ~~
Changes Resulting fran Dreak Elimination Cost Savings
- Operational Benefits Elimination of 11 Pip 2 Unknwn o Potential improvement in
, c. s'iiij) Res.t.ra..ints pe..r..U.n..i.t.
., quality of inservice o Dose Reduction Costs o Dose reduction frcm
$ 55,000 improved parsonnel access during maintenance, ISI and recovery fran unusual plant conditions, e.g.,
radioactive spills, fires, etc.
Unknown o Improved capability to d
recover fron unusual plant conditions, e.g., decon-tanination, following radioactive spills, access for fire lighting, etc.
Material Replacznent o Reduced system heat loss res'ulting fran improved insulation design.
$ 50,000 o Dose reduction by eliminating the need to set and maintain restraint clearance gaps.
Elimination of Analys,cs o Load reduction and o Improved system layout arr3 Assbefated with tbe structural analyses design for future plant
~
~
~531c Effects and D
costs to accoTmodate modifications.
Lo_adfEclojfJTtio7.s~~
lower EMI load ratings S 50,000
'IUPAL SINiidS
$15$',~060 li3~ man-rem ~in dose
~ ~ ~ ~ ~ ~
~
~
~~~~
(UNITS 1 AND 2) reduction over the 40-year plant life.
- Estimate is Applicable to each Unit.
G-1 i._
..:..- '., gpij5ly L_
-