ML20138R949
| ML20138R949 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/15/1985 |
| From: | Douglas R, Mittl R Public Service Enterprise Group |
| To: | Butler W Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8511190296 | |
| Download: ML20138R949 (13) | |
Text
$
o O PS G Cornpany Pubhc Servce Ekx:tnc and Gas 80 Park Plaza. Newark, NJ 07101/ 201430 8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation November 15, 1985 Director of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention:
Mr. Walter Butler, Chief Licensing Branch 2 Division of Licensing Gentlemen:
ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAKS HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Public Service Electric and Gas Company requests approval for the Hope Creek Generating Station to eliminate the postulation of intermediate pipe breaks as specified by SRP 3.6.2 Sections II.1 and II.2 for the main steam and RWCU systems inside containment unless such locations exceed the stress and usage factor threshold levels provided in BTP MEB 3-1 or are located in the proximity of welded pipe attachments.
In support of this request, the following information is provided in accordance with your letter of September 20, 1985 (W. Butler, NRC, to R.L. Mittl, PSE&G):
1.
Provide a short discussion of the technical justifica-tion for elimination of arbitrary intermediate breaks.
RESPONSE
The technical justification for elimination of arbitrary intermediate breaks is as follows:
a.
Deletion of whip restraints will improve access for operation, inservice inspection, and maintenance.
b.
Occupational radiation exposure during inspection, (h
maintenance, and repair will be reduced over the life of t
I I
the plant, B511190296 851115 PDR ADOCK 05000354 The Eneray People A
PDR 95 4912 (4M) 7 83
l-Director of Nuclear Reactor Regulation 2
11/15/85 c.
The additional accessibility to the piping systems may improve the efficiency of inservice inspections.
d.
Postulating arbitrary intermediate breaks provides only additional conservatism with no physical basis, e.
Deletion of arbitrary intermediate break locations will not impact the environmental qualification of safety related equipment and components since the harsh environment conditions have already been defined and will not be revised.
f.
The NRC has accepted a similar position for other piping systems on the Hope Creek Generating Station.
g.
For pipe whip restraints which are currently installed, but not required based on the elimination of arbitrary intermediate breaks, the whip restraints may be retained.
However, substantial cost savings will occur since notching of insulation around shimpacs is not
. required, resulting in reduced heat loss to the contain-ment and ease of insulation, installation, and removal.
h.
The option exists to remove unnecessary existing pipe whip restraints if maintenance / inspection operations could be simplified by enhanced accessibility.
2.
Provide a table or summary which includes the following information :
a) identification of all affected piping systems.
b) pipe diameter and material of each system in (a).
c) estimated number of breaks eliminated in each system in (a).
d) estimated number of rupture restraints and jet deflectors eliminated in each system in (a).
RESPONSE
A summary table of the affected systems is provided as follows:
Director of Nuclear Reactor Regulation 3
11/15/85 Arb.
Nom.
Interm.
Pipe Whip Jet Pipe Pipe Pipe Breaks Restraints Deflectors System Material Dia.
Elimin.D,c Elimin.a Eliminated Inside Containment Main Steam CS 26" 7
0 0
RWCU CS/SS 2"/4"/6" 3
0 0
NOTES:
a.
The quantities listed are those restraints which have not yet been installed.
Those restraints which have been installed may remain, however, several restraint shimpacs may not be required.
b.
Welded piping attachments are not located in the proximity of any eliminated arbitrary inter-mediate breaks and no such welded attachments are expected to be added in the future.
c.
The actual number of arbitrary intermediate breaks to be eliminated will be based on final stress analyses.
3.
Provide a detailed discussion to justify why the systems identified in 2(a) are not susceptible to the following:
a.
b.
Water / Steam hammer effects.
c.
Thermal fatigue and mixing.
RESPONSE
The above systems are not susceptible to intergranular stress corrosion cracking (IGSCC), steam / water hammer effects, or thermal fatigue and mixing due to the following:
a.
Industry experience has shown per NUREG-1061 that IGSCC can occur when the following conditions exist simultaneously:
high tensile stresses, piping material susceptible to cracking, and a corrosive environment.
i Director of Nuclear Reactor Regulation 4
11/15/85 Although any stainless or carbon steel piping will exhibit some degree of residual stresses and be exposed to tensile stresses, the potential of IGSCC is minimized by choosing piping material with low susceptibility to stress corrosion and by ensuring that a corrosive environment does not exist.
The likelihood of IGSCC in stainless steel increases with carbon content.
There-fore, only a low carbon content stainless steel has been used (304L) in the portion of the 6-inch transition piece connecting the RWCU system to the recirculation system.
The remainder of the af fected system piping is ferritic carbon steel which has been found not to be susceptible to IGSCC.
The existence of a corrosive environment is minimized by specifying stringent criteria for internal and external cleaning and by following water chemistry guidelines during power ascension and normal operation, b.
The steam / water hammer effects discussed in the Catawba position are specific to PWR plants and do not apply to the HCGS BWR design.
Steara hammer loads are anticipated for the Main Steam system and are included in the design as discussed in FSAR Section 3.9.1.
Analyses have been performed for these loadings and the Main Steam system has been designed to accommodate and minimize effects of these loadings.
The RWCU system is continuously in operation to purify the reactor water, and the lines will be filled, thus minimizing the potential for water hammer.
As required by ASME B&PV Code Section III, a detailed c.
fatigue analysis is performed on all Class 1 piping systems.
Such analyses have been performed for the Main Steam and RWCU systems.
For ASME B&PV Class I lines, conservatism is allowed for fatigue failure.
The ASME Code limit for the Cumulative Usage Factor (CUF) is 1.0 to assure that pipe fatigue failure will not occur.
The pipe break postulation limit is 10 percent of this number, and all of the Class 1 arbitrary intermediate break locatiens involve CUFs below this limit.
Based on the system design and layout which minimizes thermal stratification and cyclical stresses, and the analyses performed to verify the piping will experience no fatigue failure, the Main Steam and RWCU systems are not susceptible to ther..aal f atigue due to mixing.
=
Director of Nuclear Reactor Regulation 5
11/15/85 4.
Provide a commitment that all systems in 2(a) will be included in the preoperational piping testing program.
RESPONSE
The Main Steam and RWCU systems are within the scope of the piping startup testing program.
Each system will be tested to verify that steady state vibratory levels are within acceptable limits for operating conditions anticipated during service.
5.
Provide a commitment that all safety related equipment in the vicinity of the eliminated breaks has been environmentally qualified to withstand the effects of a non-mechanistic break.
RESPONSE
Elimination of arbitrary intermediate breaks will not affect the environmental qualification of safety related equipment in the vicinity of the arbitrary intermediate break loca-tions.
The break locations for defining the worst case harsh environment conditions for all safety related equip-ment have been evaluated, which include the arbitrary intermediate break locations, and the results documented in the FSAR.
These worst case conditions will not be revised based upon elimination of the arbitrary intermediate break locations.
In addition to the above information, attached for your review are proposed FSAR changes to Section 3.6 eliminating the postulation of arbitrary pipe break locations.
These changes will be incorporated upon approval of the above request.
Should you have any questions in this regard, please contact us.
Very truly yours,
Director-of Nuclear.
' Reactor ~ Regulation 6
11/15/85 C
D. H. Wagner USNRC Licensing Project Manager A. R.
Blough USNRC Senior Resident Inspector OH 18 01/06A
~
i HCGS FSAR 6/84 TABLE 3.6-2 Pn: LIMINA"Y MAIN STEAM SYSTEM PIPING STRESS LEVELS AND PIPE BREAK DATA (PORTION INSIDE PRIMARY CONTAINMENT)
Pipe Break Stress Stress Cumulative Limit Basis for Node Node By EQ. 10 Usage 2.4 Sm Break
. Break Point (1) Type (2)
(ksi)
' Factor (ksi)
Type (3)
Selection (*)
tines A & v
~
l 6 1 TTJ 30.63 0.0067 42.5 C
TE 2 30N EL 43.50 0.0158 42.5 SFL 2 )0F EL 41.18 0.0139 42.5 C
MBL 3 )0N EL 37.57 0.0106 42.5 C
MBL 300F EL 34.93 0.0089 C
MBL 4
0 3
EL 28.81 0.0025 42.5 C
TE 400F EL 26.40 0.0 42.5 C
TE i
L.nes B & C 1 12 TTJ
.28 0.0071 42.5 C
TE 1 18N EL 48.19 0.0169 42.5 C
SFL 1 18F EL 42.54 0.0126 42.5 C
SFL I S1N 40.46 0.0117 42.5 C
MBL{
I S1F EL 42.05 0.0128 42.5 C
MBL 1
EL 28.91 0.0050 42.5 C
TE Insert A T'2T EL 30 3; 0 0053
'2 5 C
T2 i -
3 (2) Locations of the nodes are shown in Figure 3.6-2 l
(2) Symbols used to denote the node type are as follows:
TTJ Tapered transition joint EL Elbow TEE Tec O'4 Butt veld REO Reducer SwP-5 wee. pole 1 (3) Break types are indicated as follows:
C Circumferential Long M u-din &l L
(*) Symbols used to denote the basis for break selection are as follows:
TE Terminal end MBL inLecmedicta hreak lecctions selected tc ccticfy-C-the requirements for a minimu- " umber Of break 4L, lccetienc + -
Stress and fatigue limits established in Section SFL 3.6.2.1.1.3 are not met.
Amendment 6
INSERT A Line A 1
TTJ 29.94 0.010 42.5 C
TE 45 EL 58.27 0.010 42.5 C
TE Line B 1
TTJ 24.95 0.000 42.5 C
TE 49 EL 64.86 0.030 42.5 C
TE Line C 1.-
TTJ 28.41 0.010 42.5 C
TE 4
EL 51.83 0.010 42.5 C&L SFL 42 EL 58.71 0.010 42.5 C
TE Line D 1
TTJ 27.9 0.010 42.5 C
TE 300 SWP 48.85 0.182 42.5 C&L SFL 304 TTJ 18.86 0.187 42.5 C&L SFL 39 EL 58.17 0.010 42.5 C
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HCGS FSAR 6/84 TABLE 3.6-10 Ja-rn:LIMIIiARi' RWCU SYSTEM PIPING STRESS LEVELS AND PIPE BREAK DATA (PORTION INSIDE PRIMARY CONTAINMENT)
Pipe Break Stress Stress Cumuistive Limit Basis for Node Node By EQ. 10 Usage 2.4 Sm Break Break Point (1) Typeca)
(ksi)
Factor (ksi)
Type <3)
Select'icn(*)
^5 BW 16.32 0.0003 A2 RA r
- E l
1 00 TTJ 44.161 0.01/9 42.86 SI L l
210 TEE 43.99 0.0326 42.
C S1'L l
42.86 C
T E l
4 30 BW 16.69 0.
t 5 18 BW 1
0.0002 42.86 C
T E l
7 55 47.64 0.0121 42.86 C
SFL l
6 TTJ 20.99 0.0029
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(*) Locations of the nodes are shown in Figure 3.6-15
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(2) Symbols used to denote the node type are as follows:
TTJ Tapered transition joint EL Elbow "-
TEE Tcc ^
BW Butt weld Reducer RED SocXet weld s it)
(3) Break. types are indicated as follows:
C Circumferential L
l-o ng M uci; nal
(*) Symbols used to denote the basis for break selection are as follows:
TE Terminal end MSu~ -
Intcrnedi=Fa hraak locaticn; selected tc estisf' fur, the requircm:nt for a ninimu.? number of break _-(__,
locatienc Stress and fatigue limits established in Section SFL 3.6.2.1.1.3 are not met.
Amendment 6
' ?.
I INSERT ~B 90 BW 14.227 0.0001 43.60 C
TE 101
'Inf 68.251 0.8853 43.60 C&L SFL 480
.BW 10.344 0.0000 43.60 C
TE 518 BW 10.332 0.0000 43.60 C
TE 760 RED 65.771 0.125 43.60 C
SFL 800 BW 10.357 0.0002 43.60 C
TE 108' TTJ-76.297 0.5401 43.60 C&L SFL 109 DSW 52.044 0.1346 43.60 C&L SFL 570 SW 64.676 0.6697 43.60 C
SFL 575 SW 61.147 0.5535 43.60 C
SFL 819 SW 22.763 0.004 43.60 C
TE 705 TTJ 49.52 0.0139 34.64 C
TE 710 TTJ 75.508 0.862 34.64 C&L SFL 910 RED 48.52 0.0152 43.60 C
SFL 920 SW 8.93 0.0003 43.60 C
TE 855 BW 12.67 0.0000 43.60 C
TE 902 TTJ 45.5 0.0085 34.64 C
TE 905 TTJ 78.52 0.921 34.64 C&L SFL 984-RED 48.52 0.0152 43.60 C
SFL 988 SW 9.01 0.0003 43.60 C
TE 968 BW 10.10 0.0000 43.60 C
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