ML20138M139

Requests Approval to Eliminate Postulation of Intermediate Pipe Breaks for Main Steam Sys & RWCU Sys Inside Containment Unless Stress & Usage Factor Threshold Levels Exceeded or Located in Proximity of Welded Pipe Attachments

ML20138M139
Person / Time
Site:	Hope Creek
Issue date:	12/18/1985
From:	Corbin McNeil Public Service Enterprise Group
To:	Adensam E Office of Nuclear Reactor Regulation
References
NUDOCS 8512200260
Download: ML20138M139 (13)

Text

,

Pubhc Service Electric and Gas Ccmpany C:rbin A. McNeill, Jr. Public Service Electric and Gas Company P.O. Box 236. Hancocks Bridge, NJ 08038 609 339-4800 Vice President -

Nuclear December 18, 1985 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention: Ms. Elinor Adensam, Director Project Directorate 3 Division of BWR Licensing

Dear Ms. Adensam:

ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAKS HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Public Service Electric and Gas Company (PSE&G) requests approval for the Hope Creek Generating Station (HCGS) to eliminate the postulation of intermediate pipe breaks as specified by Standard Review Plan (SRP) 3.6.2 Sections II.1 and II.2 for the Main Steam system and Reactor Water Cleanup (RWCU) system inside containment unless such locations exceed the stress and usage factor threshold levels provided in Branch Technical Position (BTP) MEB 3-1 or are located in the proximity of welded pipe attachments.

This transmittal supersedes the PSE&G to NRC letter dated November 15, 1985 (R.L. Mittl, PSE&G to W.. Butler, NRC) as a result of discussions between members of the NRC Staff and PSE&G on December 11, 1985. 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 justifi-cation for elimination of arbitrary intermediate breaks.

\

EB (LIAW)

PSB (L. HULMAN) 2P*4 851220G260 851218 Eicss (salnivAsAx)

PDR ADOCK 05000354 RSB (ACTING)

A PDR FOR (VASSALLO)

AD - G. Lainas (Ltr only)

7-Director of Nuclear 2 12-18-85 Reactor Regulation 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, maintenance, and repair will be reduced over the life of the plant.

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 HCGS.

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 containment 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 accessi-bility.

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)

Director of Nuclear 3 12-18-85 Reactor Regulation 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 system is provided as follows.

Arb.

Nom. Interim. Pipe Whip Jet-Pipe Pipe Pipe Breaks Restraints Deflectors System Material Dia. Elimin. Dad Elimin.a Eliminated Inside Containment Main Steam CS 26" 8 0 0 RWCU CS/SS 4"/6" 3 0 0 NOTES: a. The quantities listed ere those restraints which have not yet been installed. Those restraints whigh 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 intermediate breaks and no such welded attachments are expected to be added in the future.

c. Deleted

d. All eliminated arbitrary intermediate breaks in the RWCU system are on carbon steel piping.

3. Provide a detailed discussion to justify why the systems identified in 2(A) are not susceptible to the following.

A. IGSCC B. Water / Steam hammer effects C. Thermal fatigue and mixing

Director of Nuclear 4 12-18-85 Reactor Regulation

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.

Although any stainless or carbon steel piping will e:hibit 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. Therefore, 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 affected 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 potential discussed in the Catawba position are specific to PWR plants and do not apply to the HCGS BWR design. Steam 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.

I

Director of Nuclear 5 12-18-85 Reactor Regulation C. As required by ASME B&PV Code Section III, a detailed 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 1 lines, conservatism is allowed for fatigue failure. The ASME Code limit for the Cumulative Usage Factor (CUP) 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 intermedia'te break locations involve CUPS 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 thermal fatigue due to mixing.

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 locations. The break locations for defining the worst case harsh environment conditions for all safety related equipment 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.

i Director of Nuclear 6 12-18-85 Reactor Regulation 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 l us.

Sincerely, Attachment C D.H. Wagner USNRC Licensing Project Manager R.W. Borchardt USNRC Senior Resident Inspector 1

3 3

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'i.b A : '. E JEAM SYSTE!' PIP:'.3 5. iS$ LEVELS AND PIFE EEE'.E DATA iPORIION I .',S I DE PE I M ARY CC? r A I N.'EN T )

Pipe Break Stress Stress Cumulative Limit Basis for Node Node By EO. 10 Uc => g e 2.4 Sm Break Break Point (2) Tvne(2) (ksi) Factor :ksi) TvoeC3) Selection (')

.. nes A & D Il 61 TTJ 30.63 0.0067 42.5 C fE I 200N EL 43.50 0.0158 12.5 C SFL 200F EL 41.18 0.0139 42.5 MSL 300N EL 7.57 0.0106 42.5 C M3L

@ iDCF 013 EL EL 34.;'

29.81 O.0089 0.0025 42.5 42 C MPL C TE M 400F EL 26.40 t . "' 0 3 7 .

.5 C TE

[b Lines B & C  !

192 TTJ 33.2 0.0071 42.5 C TE M 'lSBN EL '" 9 0.0169 42.5 C SFL 188F EL 42.54 0.0126 42.5 SFL 151N EL 40.46 0.0117 42.5 C MBL 151F - 42.05 0.0128 42.5 C M3L 116 EL 28.91 0.0050 42.5 C e

'cF EL 30.31 0.0053 42.5 C TE (2) Locations of the nodes are shown in Figure 3.6-2 l (2) Syrbols used to denote the node type are as follows:

ITJ -

Tapered transition joint EL Elbow

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o .. ,,

_ . .,-, , _. .m t .. . nw .

s .. s SWP - SK6GTAXGT (3) Break types are indicated as follows:

C -

Circumferential f - f.C 0 5(TU DI &

(4) Sy.Thols used to denote the basis for break selection are as follows:

TE -

Terminal end

-93L I,'... _jiat_ ir;.2 I c ;n i s..a o . l u t ml t o o t . ; f y -

thc :m  :, m

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..u~' r of L: 9^ [

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Strees aNd fatigue limits established in Section 3.6.2.1.1.3 are not met.

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, Line A'

-1 TTJ .29.94 0.010 42.5 C TE 45' EL . 62.07* 0.02* 42.5 C TE i

Line B 1 ' TTJ 24.95 -0.000 42.5 C TE

- 49 EL 53.61* '0.02* 42.5 C TE l- Line C i-f 1 TTJ 28.41 0.010 42.5 C TE 4 2 :- EL- 54.58* 0.010 42.5 C TE

[- Line D

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L l- ' TTJ: 27.9 0.010 42.5 C TE

- 39 EL 61.95* 0.020* . 42.5 C TE i.

• Based on Final GE Stress Report i

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HCGS FSAR 6/84 TABLE 3.6-10 SN&

Y J FRELihinARi RWCU SYSTEM PIPING STRESS LEVELS AND PIPE BREAK DATA (PORTION INSIDE PRIMARY CONTAINMENT)

Pipe Break Stress Stress Cumulative Limit Basis for Node Node By EO. 10 Usage 2.4 Sm Break Break Point (1) Type (2) (ksi) Factor (ksi) Tyoe(3) Selection (*)

95 BW 16.82 0.0003 42.86 C TE l g 100 TTJ 44.161 0.0179 42.86 C SFL l

/k() 230 TEE 43.99 0.0326 42.86 C SFL l ,

h hd (b 490 BW 16.69 0.0002 42.86 C TE WM l 518 BW 15.51 0.0002 42.86 C TE l

765 RED 47.64 0.0121 42.86 C SFL

(( l 800 TTJ 24.99 0.0029 42.86 C TE (2) Locations of the nodes are shown in Figure 3.6-15 l

(2) Symbols used to denote the node type are as follows:

TTJ -

Tapered transition joint l

J^ EL Elbow TEE T:

BW -

Butt veld

' RED -

Reducer 5#

(3) Break Srwr (Cad types are indicated as follows:

C -

Circumferential L- ~

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(*) Symbols used to denote the basis for break selection are as follows:

T -

Terminal end j r;E5L -

InLeunediacc uccak IvuaLivno ocicuicJ Lv ooLiefy the req"irerer.tc for 2 -ir.imur . urb Of brc;k SFL -

j]StressandfatiguelimitsestablishedinSection

!cr2 tier.c.

3.6.2.1.1.3 are not met.

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Amendment 6 i I

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INSERT B

- 90 BW 14.227 0.0001 43.60 C TE

- 101 BW 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 SPL 570 SW 64.676 0.6697 43.60 C SFL 575 SW 61.147 0.5535 43.60 C SFL

- 819 1SW 22.763 0.004 43.60 C TE c 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 TE

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