ML20247D156
| ML20247D156 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/31/1988 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20247D132 | List: |
| References | |
| PGE-1004-A-05, PGE-1004-A-5, NUDOCS 8907250068 | |
| Download: ML20247D156 (194) | |
Text
{{#Wiki_filter:^ ~~ - -- - q;rf ,' , ,
, [=
<;_ ,r:
a .3 5,E
~
'.PGE-1004 l,-
4 l- @. ( TROJAN NUCLEAR PLANT' ANALYSES.OF PIPE. SYSTEM BREAKS OUTSIDE CONTAINMENT i .It' qg December 1988 Amendraent 5 - Portland General Electric Company 121 SW Salmon Stree+.
. Portland, Oregon 97204
,.s % g -
8907250068 890i19 PDR ADDCK 05000344 p PDR
4
' CONTENTS
[ /~~'N Section Title' Page Y. k
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . 1.0-1 1.1 PLANT DESCRIPTION. . . . . . . . . . . . . . . . . . . 1.1-1 1.2 ANALYSIS CRITERIA. . . . . . . . . . . . . . . . . . . 1.2-1 9 2.0 PLANT DESIGN CRITERIA . . . . . . . . . . . . . . . . 2.0-1 2 .1 . SINGLE FAILURE CRITERIA. . . . . . . . . . . . . . . . 2.1-1 2.2 MATN STEAM SUPPLY SYSTEM DESIGN. 2.2-1 lh 2.3 CONDENSATE AND FEEDWATER AND EXTRACTION SYSTEMS DTSIGN 2.3-1 2.4 AUKILIARY FEEDWATER (AFW) PUMP TURBINE DRIVER STEAM SUPPLY DESIGN. . . . . .. . . . . . . . . . . . , . . . 2,4-1 Iff 2.5. STEAM CENERATOR BLOWDOWN SYSTEM DESIGN . . . . . . . . 2.5-1 2.6 PROCESS SAMPLING SYSTEM DESIGN . . , . . . . . . . . , 2,6-1 2.7 CHEMICAL AND VOLUME CONTROL _SY9 TEM LETDOWN AND CHARGING DESICW. . . . . . . . . . . . . . . . . . . . 2.7 1 2.8 RESIDUAL HEAT REMOVAL SYSTEM DESIGN. . . . . . . . . . 2.8-1 2.9 PROCESS STEAM SYSTEM DESIGN. . . . . . . , . . . . . . 2.9-1 2.10 AUKILIARY FERDWATER_(AFW) SYSTEM DESIGN. . . . . . . . 2 10-1 1 3.0 EQUIPMENT NEC2SSARY FOR SAFE SHUTDOWN OF THE REACTOR . 3.0-1 3.1 ,
NORMAL SHU?DOWN. . . . . . . . . . . . , . . . . . . . 3.1-1 3.2 EMERGEMCY_piUTCONN . . . . . . , . . . . . . . . .. . 3.2-1 3.2.1 Emstrency Shutdown with E Main Steam Line Rupture. . . 3.2-1 , l. l 3.2.2 h ge;ncy Shutdown with a Feedwater Line_Ru_pture . . . 3.2-2
- - 3.2.3 Emergency Shyt,fown witb,__ep_Juxi}J,ag Feedwater 9uuip, {
l Steam Suptly Lino Pupture. . . , . . . . . . . . . , , 3.2-3 " L 3.2.4 13ergency Shutdown with a Steam Generator Blowdown Line Rupture . . . . . . . . . . . . . . . . . . . . . 3.2-3 3.2.5 Emergency Shutdown with a Samplirm System Line Rupture. . . . . . . . . . . . . , . . . . . .. . . . 3.2-4 l 3.2.6 Emergency shutdown with a chemical and Volume ^' Control System (CVCS) Letdown Line Rupture . . . . . . 3.2-4 0
. i Amendment 5 !
(December 1988) a
(( ;% y l-l- .. CONTENTS (Contd) i e ( , Section Title Page 3.2.7- Rmergency Shutdown with a CVCS Cha;.Kinn-Line Rupture . 3.2-5 3.2.8. Emergency Shutdown with a Residual Heat Removal
' System Line Rupture. .. . . .......... . . . 3.2-5 3.2.9 Shutdown with a Process Stear Line Rupture . .. . . . 3.2-6 3.2.10- Emergency Shutdown with an AFW Line Rupture. 3.2-6 th 4.0 : HIGH ENERGY FKUID PIPING DESCRIPTION . ...... . . 4.0-1 4.1 PIPING SYLTEMS WITH TEMPERATURES HIGHER THAN 200*F' AfJD PRESSURE HIGHER THAN 275 POUNDS PER SQUARE _ g INCH GAGE (PSIC) . . . . . .. . . . ... ...... 4.1-1 v 4.1.1 Main Steam Supply Syst n Pipipg. _
...... . . . . . 4.1-1 4.1.2 Condensate and Feedwater and Extraction Syrtems Piping 4.1-3 h 4.1.3 Auxiliary Feedwater Pump Turbine Driver Steam Piping . 4,1-6 4.1.4 Steam Generator Blowdown System Piping .. .,... 4.1- 6 4.1.5 Process' Sampling System Piping 4.1-7 13 4.1.6 .,C_i}emical and Volume Control System Letdown and Charming Piping. . . . . . ...... ...... . . 4.1-8 j 4.1.7 Residual Heat Removal System (RHRS) Piping ...... 4.1- 8
.4.2 . PIPING CVSTEMS WITH TEMPERATURES HIGHER THAN 200'F OR E ESSURES MIGHER THAN 275 PSIG . ....... . . . . 4. 2- 1 4.2.1 Process Steau. System Piping. . ....... . . . . . 4.2-1 4.2.2 Auxiliary Feedwater System Piping. .. .. ... . . . 4 . 2- 2 ~i 5.0 ~ PIPE RUPTURE A4ALYSIS, , . . . ... .... . .... 5.0-1 l l
5.1 MAIN STEAM J.INF, RUPTURE. . .. . .. ... .. . . . 5.1-1 , c. 5.1.1 Areas Affected by a Steam Line Rupture . .. .. . . . 5.1-1 5.1.2 Pipe Whip. 5.1-2 13 5.1.3 Jet Tmpintement. 5.1-4 th 5.1.4 Compartment Pressurization . . .. . . ... .. . . . 5.1-4 I 11 Amendment 5 ! (December 1988) i 1 l
CONTF.NTS (Contd) Section Title Page 1 5.1.5 Flooding From Steam Line Break . ... .. . . . . . . 5.1-6 - l
- 5.1.6 Environmental Effects. . . ... . .. . . ... . . . 5.1-1 5.2 FEEDPATER LINE TUPTURE . . . ... .... . . . . 5.2-1 1
5.2.1 t_reas Affected by a Feodwater Line Rupture . . . . . . 5.2-1 5.2.2 Pipe Whip. . . . . . . . . . ..... . . . . . . . . 5.2-1 l 5.2.3 Jet Impingement. . . . . . . . .... . . . .. . . . 5.2-2 l 5.2.4 Compartment Pressurization . . . ... . . . . .. . . 5.2-3 5.2.5 Floodirs . . . . . . . . . . . . . . . . . . . . . . . 5.2-4 v 5.2.6 Enviro.imental Effects. . . . . . .. . .. ... . . . 5.2-4 5.3 CONDENSATE OR EXTRACTION LINE RUPTURE. . . . .. . . . 5.3-1 5.3.1 Areas Affected by a Condensate or Extraction Line Rupture. . . . . . . .. ... .. ... . . . .. . . 5.3-1
- 5.3.2 Pipe Whip. . . . . . . . . . . . ... . . . . . . . . 5.3-1 1
1 5.3.3 Jet Impingement. . . . . . . . .... .. ... . . 5.3-2 , 5.3.4 Compartment Pressurization . . .. ... . . . . . . . 5.3-2 l l 5.3.5 Floodin,g . . . . . . . . ... .... . . . . . . . . 5.3-3 5.3.6 Environmental Effects. . . . . . ..... . . . . . . 5 . 3--3 5.4 AUXILIARY FEEDWATER PUMP TURBINE DRIVER STEAM SUPPLY LINE RUPTURE . . . . . . . . ..... . . . .. . . . 5.4-1 5.5 STEAM CENTRATOR BLOWDOWN LINE RUPTURE. . . . . .. . . 5.5-1 5.5.1 Areas Affected by a Steam Generater Blowdown Line Rupture . . . . . . . .. . . .. . . . . . . . . 5.5-1 5.5.2 Pjpe Whip. . . . . . . . . . . .. ... . . . .. . . . 5.5-1 5.5.3 Jet Impinenme_nt.
. . . . . . . . . .. . . . . . . . . 5.5-2 O 5.5.4 Compart.._nt Pressurization . ..... . . ... . . . 5.5-2 5.5.5 Flooding . . . . . . . . . . .. . .. . . . . . . . 5.5-2 5.5.6 Environmental Effects. . . . . . .. . . . . . . . . . 5.5-2 111 Amend.nent 5 (December 1988)
I CONTENTS (Contd)-
;p' Section Title Page 5.6. CVCS LETDOWN LINE RUPTURE. . . . . . . . . . . . . . . 5.6 5.6.1 Areas Affected by a CVCS Letdown Line Ruptare. . . . . 5.6-1 5.6.2 Pipe Whip. . .. . . . . . . . . . . . . . . . . . . 5.6-1 , 5.6.3 Jet Imylngement. .. . .. . . . . . . . . . . .. . . . 5.6-1 5.6.4 Compartment Pressurization . . . . . . . . . . . . . . 5,6-2 lb .
l1 5.6.5 Flooding,. . . . . . . . . . . . . . . . . . . . ,, . . 5.6-2 l 5.6.6 Environmental Effects. . . . . . . . . .. . . . . . . 5.6 5.7 RESIDUAL HEAT REMOVAL LINE RUPTURE . . . . . . . . . . 5.7-1 5.8 'CVCS CHARCING LINE RUPTURE . . . . . . . . . . . . ... 5 "-1 1 5.8.1 ' Areas Affected by a CVCS Charging Line Rupture . . . . . 5.0 5.8.2 Charging Line Pipe Whip. . . . .. . . . . . . . . . . 5.8-1 5.8.3 Jet Impingement. . . . . . . . . . . . .. . . . . . . 5.8-1 5.8.4 Compartment Pressurization . . . . . . . . . . . . . . 5.8-2 13 5.8.5 Floodina . . . . . . . . . . . . . . . . . . . . . . . 5.8-2 5.8.6' Environmental Effects. . . . . . . . . . . . . . . . 5.8-2 l 5.9 PROCESS STEAM LINE RUPTURE . . . . . . . . . . . . . . 5.9-1 5.9.1 Areas Affec Md by~a Process steam Line Rupture . . . . 5.9-1 l 5.9.2 Pipe Whip. . . . . . . . . . . . . . . . . . . . . . . 5.9-1 5.9.3 Jet Tmpinnement. . . . . . . . . . . . . . . . . . . . 5.9-1 l l 5.9.4 Compartment Pressurization . . . . . . . . . . . . . . 5.9-2 l l L 5.9.5 Flooding . . . . . . . . . . . . . . . . . . . . . . . 5.9-2 5.9.6 Environmental Effects. . . . . . . . . . . . . . . . . 5.9-2 5.10 AUXTLIARY FEEDWATER LINE RUPTURE , . . . . . . . . . . 5.10-1 6.0 EMERGENCY SHUTDOWN PROCEDURE . . . . . . . . . . . . . 6.0-1 , tu v
1.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . 7.0-1 iv Amendment 5 (December 1988)
CONTENTS (Contd) Section Title APPENDIX A GENERAL INFORMATION REQUIRED FOR CONSIDERATION OF THE
- EFFECTS OF A PIPING SYSTEM BREAK OUTSIDE CONTAINMENT B EFFECTS OF A PIPING SYSTEM BREAK OUTSIDE THE CONTAINMENT C COMPARTMENT PRESSURE ANALYSIS FOR TURBINE BUILDING D STEAM LINE BREAK IN THE TURBINE BUILDING I E FEEDWATER LINE BREAKS IN THE TURBINE BUILDING AND l BLOWDOWN LINE BREAF.S OUTSIDE CONTAINMENT l
O O V Amendment 4 (August 1987)
I i i d TABLES I
-{/^) Number Title
%s - ;
2-1 Main Steam Line Safety VLives ) i 1 5-1 Main Steam Line Stress Summary / 5-2 Feedwater Line Stress Summary 1
)
v I I i i l 1 I l ( ' vi ! Amendment 4 (August 1987) ] q i 4 _ _ _ _ _ - _ _ _ _ . ]
l 1 I 1 FIGURES I ,N', Nunoer Title
\
]
1 1-1. Seismic Classification of Structures 1-2 Intentionally Deleted j 1-3 Intentionally Deleted 2-1 Variation of steam Pressure With Load 2-2 Intentionally Deleted 2-3 Intentionally Deleted - 2-4 Intentionally Deleted E\ 2-5 Intentionally Deleted 2-6 Intentionally Deleted 2-7 Intentionally Deleted 2-8 Intentionally Deleted 2-9 Intentionally Deleted 2-10 Intentionally Deleted 4-1 Isometric View of the Main Steam Piping in the Main Steam Support Structure and Turbine Building 4-la Intentionally Deleted h 4-2 Isometric View of the Feedwater Pipint in the Main Steam Support Structure and Turbine Buildind 4-2a Intentionally Deleted 4-2b Intentionally Deleted 4-2c Intentionally Deleted 4-3a intentionally Deleted O 4-3b Intentionally Deleted 4-4 Intentionally Deleted 4-5 Intentionally Deleted
-- 4-6 Intentionally Deleted vii Amendment 5 (December 1988)
p;
< FIGURES (Contd) l .,~f ' %/ Number Title 4-7 Intentionally Deleted 3
4-8 Intentionally Deleted 5-1 Deleted l 5-2 Deleted - l 8 5-2a Deleted 5-3 Deleted f% I viii Amendment 5 ( (December 1988)
l.IST OF EFFECTIVE PAGES yy PGE-1004 (
's._ ') Analyses of Pipe System Breaks outside containment Amendment 5 Amendment Page No. No. Date i thru iv 5 Dec. 1988 y and vi 4 Aug. 1987 vii thru ix 5 Dec. 1988 1.0-1 4 AtL. 1987 1.0-2 and 1.1-1 5 Dec. 1988 1.1-2 original June 1973 1.2-1 thru 1.2-3 5 Dec. 1988 I Figure 1-1 5 Dec. 1988 2.0-1 original June 1973 l 2.1-1 thru 2.10-1 5 Dec. 1988 )
Table 2-1 original June 1973 J Figure 2-1 5 Dec. 1988 j 3.0-1 thru 3.2-1 original June 1973 i 3.2-2 thru 3. 2-7 5 Dec. 1988 l f- s 4.0-1 original June 1973 "
/ \ 4.1-1 thru 4.2-2 5 Dec. 1988 k,_,/ Figure 4-1 and 4-2 original June 1973 5.0-1 original June 1973 5.1-1 thru 5.1-7 5 Dec. 1988 5.1-8 2 Aug. 1975
- 5. 2-1 thru 5.10-1 5 Dec. 1988 Table 5-1 2 Aug. 1975 ;
Table 5-2 1 Jan. 1974 6.0-1 and 6.1-1 5 Dec. 1988 7.0-1 4 Aug. 1987 A-1 thru A-13 original June 1973 A-13a thru A-13c 1 Jan. 1974 A-14 and A-15 original June 1973 A-16 thru A-36 2 Aug. 1975 B-1 original June 1973 B-2 1 Jan. 1974 B-3 original June 1973 B-4 1 Jan. 1974 B-5 and B-6 original June 1973 Table 2-1 (sheets 1 thru 3) original June 1973 B-7 thru B-39 original June 1973 C-1 thru C-5 4 Aug. 1987 D-1 thru D-6 4 Aug. 1987 l E-1 thru E-4 4 Aug. 1987 l (0) 1 ix Amendment 5 N/ (December 1988) i 1
e
1.0 INTRODUCTION
t ~ E V~ This r,eport was prepared in response to the Directorate of Licensing L letter of December 19,'1972 to Portland Ceneral Electric Company, in
' regard to the consequences of pipe. rupture during operation of the Trojan Nuclear Plant. This letter and its attachment titled " General Information Required for Consideration of the Effects of a Piping System Break Outside Containment" as modified by errata dated January 10, 1973, were the basis of this report. Discussions with the Atomic Energy Commission (AEC) Task l{
Force assigned to investigate the problems associated with high energy line pipe breaks outside containment took place in Bethesda, Maryland on January 22, 1973 and provided further interpretation of the criteria' outlined in the AEC letter referenced above. Pertinent correspondence regarding this subject is' enclosed herein within Appendix A. Revision 1 includes an updating of system descriptions and responses-
~
to the following AEC Directorate of Licensing questions and positions regarding the Trojan Nuclear Plant Final Safety Analysis Report: L
- 1) Question 3.4 forwarded in the AEC letter of June 29, 1973.
- 2) Question 10.2 forwarded in the AEC lett6r of August 10, 1973. ;
- 3) Question 3.24 forwarded in the AEC letter of November 30, 1973.
- 4) Question 10.6 forwarded in the AEC letter of January 18, 1974.
The revision also includes explanations and clarifications discussed in a i meeting with AEC representatives on November 27, 1973. l Revision 2 includes an updating of the pipe break analysis within the g main steam support structure. I l 4 Amendment 4 . (August 1987) l J = - _ _ - _ _ _ - _ - _ _ _ - . _ - -
f
'.j
.]
y.p- Revision 3 includes updates' to Figure 2-4 and Figure 4-2c to reflect the , Ms .- latest Auxiliar} 'eedwater System configuration.
)
Ameadment 4 includea updates of the analyses.of main steam and feedwater 1 line ruptures within the Turbine Building and the Main' Steam Support b Structure. The environmental effects from the main steam and feedwater j line breaks are discussed in Topical Report PGE-1025, " Trojan Nuclear
. Plant Environmental Qualification Program Manual".
~
Amendment 5 includes design changes to the steam generator blowdown
. system and makes numerous editorial changes to improve the accuracy and '
]w clarity of the' text. I i
I i i 1.0-2 Amendment 5 (December 1988)
s p 1.1- PLANT DESCRIPTION q
.The Trojan Nuclear Plant is a Westinghouse pressurized water reactor L (pWR) designed for a power output of 3423 megawatts thermal (MW)(t).. -
which is the license application rating. The equivalent warranted gross and approximate net electrical outputs of the plant are 1178 MW(e) and 1130 MW(e), respectively. The reactor is expected, ultimately, to be-capable of.an output of approximately 3570 MW(t), which would correspond to the valves-wide-open rating of the turbine-generator of 1219 MW(e) ' gross and 1170 MW(e) net. All plant safety systems, including Containment and Engineered Safety Features (ESP), are designed for operation at the higher power level. The power rating of 3570 MW(t) is also uced in the analyses of postulated accidents discussed in this report. The Nuclear Steam Supply System (NSSS) is similar to that at other nu- m. clear power plants (Diablo Canyon, Sequoyah). It uses a chemical shim and control rods for reactivity control and generates dry steam in
! vertical U-tube steam generators. The NSSS is located within a prestressed, post-tensioned, reinforced concrete Containment.
The steam produced drives an 1800-rpm. tandem-compound, six-flow turbine-generator. Major plant structures consist of the Containment which contains the reactor plant, the Turbine Building enclosing the turbine-generator, associated secondary plant pumps and hert exchangers, the emergency diesel generators and the auxiliary feedwater pumps, the Auxiliary Building which houses the reactor auxiliary and emergency i systems, the Control Building, and the Fuel Building. These structures and their relationship to each other are shown in Figure 1-1. h O w LO l 1.1-1 Amendment 5 (December 1988) e_______-__-
>,.- Portland General' Electric' Company is fully responsible for the complete k safety and adequacy.of the station. Aid in the design, construction, testing and startup of the plant has been and will he. supplied princi-pally by the~Bechtel Corporation and Westinghouse Electric Corporation.
Assistance has been and will also be rendered by other consultants and suppliers.as required, i
,Q v
i. f,s - 1.1-2 ! 1 i I l' 1
I l f 1 V l \ l [ n 1.2 ANALYSIS CRITERIA i' ( l , ' %.,} This report is based on the criteria outlined in the Atomic Energy Cornmission ( AEC) document "Ceneral Information Required for Consideration of the Effects of a piping System Break Outside Containment", which was _ l attached to a letter from the Deputy Director for Reactor projects dated U 4 j_ December 19, 1972. The letter cnd attachment are provided in Appendix A of this report. The analysis criteria as interpreted and implemented by l portland Geeval Electric are provided below. The systems analyzed were those piping systems whose operating teinpera-tures exceed 200'F or whose operating pressure exccods 275 psig. The effects of pipe whip were considered cnly for those piping systems where the operating pressure and temperature exceed 275 psig and .200*F. For piping with either a pressure of 275 psig or a temperature of 200'F, the effects of the critical break only were considered. A single pipe break only was considered. In Seismic Category I piping syst.emn or Seismic Category II piping systems Which were seismically ^ analyzed, the pipe breaks are sssumed to occur at:
- 1) Terminal ends.
- 2) Dranch connections.
- 3) Two intermediate points of highest combined stress lovels.
- 4) points where combined stress levels exceed 0.8 (Sh
- Sa )*
Where S is the basic material allowablo stress at design , h w v temperature and S, is the allowablo stress range for expansion stresses.
- 5) No intermediate breaks are assumed in short pipe runs (i.e., five pipe diameters or less).
1.2-1 Amendment 5 (December 1988)
The cumulative usage factor was not utilized in the analysis. The effect i} on the locations of postulated pipe breaks of including 1/2 Safe Shutdown - w Earthquake (SSE) in the plant conditions at the onset of the rupture has been assessed. Further, the effect on the predicted dynamic loads ex-erted upon the pipe restraints of including the particle motion assoc-isted with 1/2 SSE in the initial conditions as the pipe begins 10 whip following a postulated break has boon considered. At all locations the break was assumed to be a full area circumferentral or longitudinal. slot break. If the break was in Seismic Category II piping which has not been seismi-cally analyzed, it was assumed to occur at any location. (Full area cir-cumferential or longitudinal slot break.) The following types of breaks were postuinted at the locations identified above:
- 1) Longitudinal breaks in piping runs and branch runs 4-inch nominal pipe size and larger.
- 2) Circumferential breaks in piping runs and branch runs exceeding 1-inch nominal pipe size. Ih Critical crack breaks were assumed to occur at any location. They were located to maximize the consequences on required safe shutdown equipment or on structures. The crack length used was one-half the pipe insido diameter (ID), and the width used was one-half the wall thickness of the failed pipe.
plant conditions pri- to rupture were assumed to be normal steady state or hot standby, depending on which conditions provide the most limiting
^
effects. Worst-case temperature ef fects arc due to ruptures at power While worst-case pressure ef fects occur at hot standby. 1 0 1.2-2 Amendment 5 (December 1988)
q 1 .l
- g. No accident was assumed to occer concurrently with the pipe failure
,outside the Contairracnt. i l
I Concurrent loss'of both preferred and normal offsite power was assumed 1 for those accidents which cause protection system actuation effecting a 1 plant trip. Loss of function of onsite a-c power (diesel) and battories j i must be prevented in this case. A single failure of an active component ~was assumed within the combined { systems required to effect the cold shutdown condition (see Section 2.0), i i The effects considered included pipe whip, jet impingement, water ' )I
=!
flooding, pressure and temperature offects on structural integrity of , compartments and environmental effects of the pressure, temperature and 4 humidity on any equipment,. components, or structures important to safety. These effects must not compromise the capability to eventually-establish and maintain a cold shutdown of the reactor, l j
^ -
. i.,
Seismic Category I structural elements such as floors, interior walls,- exterior walls, building penetrations and the buildings as a whole havo j been analyzed for eventual reversal of loads due to the postulated accidents and are adequately designed for such loads. l i i i l i i I
)
t
)
. 1.2-3 Amendment 5 (December 1788) l
& rao t
s-o,e n.c-<,v.,
/, < mf sf. ..~ f ,
/
9z' 7,'/ 'a hib,// Ib!f g,.;sM% 7,7///,; f, Conin u '/ _,9,/-_
'/,/ l,// l' wE- m ...~.
a. O ,
/
/
/
r5 vy., , 7x: ~~~='~~~~v" ~ '
~ ~
' ~"?W *~ i
--};;. ~
~ ~ ~ ' '" * ^ # ' ~'
[' : ,' g., r: ji: r l' L 1 ~. p r i le/ f
/5/
/Es /
I r
//V
/*/ ,4 Onoams 8so<,
l_ /ff St I - LEGEN1D% TMg 1
'////l Catsoomv l Stauctums CARD Cetroomy D Stasc.vume If 4 /g .
- U erttsre '%
P // / murut r,< Y/ g - . (_. 3 6tthk Gcusacoa ' -.? w. uamy Puwe 2 ocw
.g- . C ocwo rowe, ,
- l
- ~ ' -
l. w ,._ n o o ,~ ,. . ..,,os
** " ~' I Amendment 5
'bf (December 1988)
TROJAN NUCLEAR PLANT
$tiSMIC CLASSMICATION OF STRUCTURES FICURE l-1
y
, i t ';
f [ i FIGURB 1-2 and FIGURE 1-3' , INTENTIONALLY DELETED 1 10
-)
l l . l O Amendment 5 (December 1988) u____________u-__. _ _ _ _
+<. 2.0 PLANT DESIGN CRITERIA
'q/
This'section describes the criteria' employed to establish the failure modes in accordance with accepted General- Design Criteria. It also enumerates and describes the systems considered in the evaluation.
/~ '
i i
.O 1 2.0-1 ,
l i i
(.-
., 2.1 SINGLE FAILURE CRITERIA An occurrence which reruits in'the. loss of capability of_a component to l perform its intended function is a. single failure. Multiple failures resulting from a single occurrence are considered to be a single
~
failure.-~ Fluid and electric systems are considered to be designed against an assumed single failure if neither a single failure of any active component (assuming passive components function properly), nor a single failure of a passive component.(assuming active components function properly) results in the loss of the. capability of a system to perform ito' safety function. An active failure is the failure of a component, such as a motorized valve or a pump, to operate, and a passive
~
failure is the failure of a static component such as a pressure boundary, i-The analysis in this report considered a single active failure within the l combined systems required'to effect the cold shutdown condition. The.
.following fluid systems were designed to provide their intended function 13 following the single failure:
i . Reactor Coolant System Containment Isolation System Emergency Core Cooling.Syntem Containment spray System Auxiliary Feedwater System Service Water System Component Cooling Water System Chemical and Volume Control System Diesel Fuel Oil System a lu j Auxiliary Steam System 1 O
\
I l i i 2.1-1 Amendment 5 , (December 1988).
-)
I l 2.2 MAIN STEAM SUPPLY SYSTEM DESIGN lh X 1' Q,i The Main Steam Supply System (MSSS) is designed to supply the stt 6 from th the four steam generators to the turbine-generator, moisture sep3rator reheaters and feed pump turbines, and steam jet air ejector during normal plant operation. The system also supplies steam to the auxiliary feed- Ih water pump turbine whenever the pump is required to operate. The system is also designed to bypass main steam directly to the condenser in case of load rejection or a trip of the turbino-generator and at plant startup or shutdown. Main steam line isolation valves are designed to isolate the main steam piping on a signal from the Engineered , Safety Feature Actuation System. The piping and cor.ponents from the steam generators up to and including the main steam line check valves downstream of the isolation valves outside the Containment are designed to meet Seismic Category I require-ments and in accordance with American National Standards Institute 10 ( n) (ANSI) B31.7, Class II. The remainder of the piping downstream of the main steam line check valves is Seismic Category II and was designed in accordance with ANSI B31.1.0. The design pressure rating of the MSSS lh piping is based on the maximum stesia pressure which occurs at no-load conditions, as shown in Figure 2-1. Ih The safety valves and power-operated relief valves provided in the MSSS are designed in accordance with the American Society of Mechanical _ Engineers (ASME) Boiler and pressure Vessel Code, Section III, Class 2. O The design capacities of the safety valves are shown in Table 2-1. The total capacity of the safety valves is at least sufficient to pass 100-percent of the steam flow generated with the turbine throttle valves ih in wide-open position. The setpoint of one safety valve on the main steam line from each steam generator is equal to the steam generator design pressure of 1200 psia, minus the pressure drop between the steam generator and the safety valve at full discharge conditions. Additional valves are set to open at higher pressures, as shown in Table 2-1. The gg l' 0 Amendment 5 2.2-1 (December 1988)
n -
'..),.,..
1' ,.> s ty.- /.,P. ~ , $ j,' i i
.., e ,
( t!. . n --
^
fy highest setpoint of the safety valves'does not exceed 110-percent of. iy jy-j . steam senerator ' design pressure, in accordance.with' the ASME Boiler -
., and Pressure Vessel Code. The power--operated relief valves have a total
'^
. design capacity of approximately 10-percent of the main' steam flow with< i.
the turbine thcottle valves in the wide-open position. f L. 7
- \. .
2.2-2 ' Amendment 5 (December 1988)
r~ - - - - = - ---
~ - - - - - - - - - - - - - -
- - - - - - - - - - ~ - = - - - - - - - ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - -
) i'.. [. . h
- n. .
r .'2.3 ~ CONDENSATE AND FEEDWATER. AND' EXTRACTION SYSTEMS DESIGN
~
The Condensate and Fee'dwater System is designed'to provide a continuous lh . feedwater supply to the four steam generators at required pressure and temperature under all anticipated steady state and transient conditions. The system.is designed with sufficient margin to maintain feedwater supply required for plant operation up to 70-percent of. full rated load Ih with only one of the two trains of pumps in service. l
; All components of the Extraction Steam and Feodwater Heater Drain and Vent Systems are also designed to handle the maximum flows that can occur .
under all anticipated operating conditions, including operation at 70-percent of rated load with one train of pumps out of service. Ih . 1 IL The Condensate and Feedwater System decign includes provisions for automatic isolation of the system from the steam generators when required to mitigate the consequences of a steam line break or malfunctions in the Condensate and Feedwater System. The f eedwater piping from the steam generator isolation check valves - outside the Containment to the four steam generators is designed and fabricated in accordance with American National Standards Institute lh (ANSI) B31.7 Class II, and to meet Seismic Category I requirements. All the remaining piping of the system is designed in accordance with ANSI B31.1.0. All pressure piping in the Extraction and Feedwater Heater Drain and Vent Systems is ANSI B31.1.0. Pressure vessels and pressure retaining parts of the components in the Extraction System are designed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Ih Pressure Vessel Code, Section VIII. All components and piping in the systems described above are Seismic Category II except where noted. ( 2.3-1 Amendment 5 (December 1988)
L. : ' 2.4 AUXILIARY FEEDWATER (APW)-PUMP TURBINE
' DRIV'R E STEAM SUPPLY DESIGN The steam-turbine-driven Pump P-102A is a single-stage; noncondensing ]
turbine equipped with an electrohydraulic speed control, overapeed trip mechanism,'and an integral trip throttic valve. The turbine can be supplied with steam from any of the four steam gene-rators. A connection is provided from each of the Seismic Category I ~C I sections'of the four main steam linas upstream of the main steam isola-tion valves. Each connection is provided with an air-operated supply . lC valve, a manually-operated isolation valve upstream, and,a nonroturn valve downstream for automatic isolation in the event'of a main steam ! line break. The air-operated supply valves are equipped with air i accumulators which ensure valve operation and auxiliary feedwater (AFW) C supply in the event'of a loss of all power. The four lines join into a
- single line and supply steam to the steam turbine through the combined trip and throttic valve. The 4-inch steam supply.-line enters an B inch 1 guard pipe as it passes.from the main steam support structure into the.
Turbine Building. The. steam exhaust from the turbino driver exits the Turbine Building through an 8-inch line encaced in a 12-inch guard pipo. This exhaust line vents to the atmosphere. These guard pipes protect the g 1. redundant AFW train from a rupturo in the steam supply line or exhaust line. The pressure piping from the main steam line to the air-operated supply valve is designed .in accordance with American National Standards Insti-tute ( ANSI) B31. 7,' Class I::. From the suppl.7 valves to the AFW pump turbine driver, the pressure piping was' designed in accordance with ANSI B31.1.0. The entire piping system is Seismic Category I. 4 i
'O .
2.4-1 Amendment S (December 1988) 1
)
2m m___ _ _ _ _.-__- _ _ _ _ _ _ _ _ _ _ _
2.5' STEAM GENERATOR BLOWDOWN SYSTEM DESIGN % ,/ The Steam Generator Blowdown System (SGBS) is designed to fulfill the following requirements:
- 1) To maintain the steam generator shell-side water chemistry within specifications. .
Oi ll
- 2) To accommodate a maximum blowdown flow rate of 100 gpm per steam j
^] ,
generator under normal plant operating conditions.
- 3) To isolate the blowdown system and prevent further blowdown if
^
radioactivity is actected in the steam generator shell side i water by the Process Sampling System.
- 4) To process a blowdown effluent of up to 218 spm in the event of primary-So-secondary steam generator tube leakage.
TN
) The part of the SGBS from the steam generators to and including Contain-ment isolation valves comprises an extension of the steam generator boundary. This portion of the system has been designed in accordance with Seismic Category I requirements. The remainder of the system is Seismic Category II/I. Ih The steam generator blowdown piping, frem the steam generators to the Containment isolation valves, has been designed in accordance with _
American National Standarde Institute (ANSI) B31.7, Class II, code for U Nuclear Piping. The remainder of the SGBS piping has been designed in
^'
accordance with ANSI B31.1, Code for Nuclear Piping, q O U 2.5-1 Amendment 5 (December 1988) l
f ~~S 2.6 PROCESS SAMPLING SYSTEM DESIGN . Y,, The Process Sampling System consists of two smaller systems, the Primary Sampling System and the Secondary Sampling System. The Primary Sampling System (FSS) is designed to collect samples of the fluids in the Reactor IC v , coolant System and auxiliary system process streams, for analysis by the plant operating staff. Chemical and radiochemical analyses are performed on these samples as appropriate to determine boron concentration, fission and corrosion product activity levels, dissolved gas concentration, chloride concentration, pH and conductivity levels, fission gas content, and gas compositions in various vessels. Analytical results are used to regulate boron control adjustments, monitor fuel rod integrity, evaluate ion exchanger and filter performance, specify chemical additions to the various systems, and maintain the proper hydrogen overpressure in the volume control tank. "he PSS is designed to permit the collection of samples during all modes of operation, from fell power to cold shutdown, without requiring access to the Containment. rm k,) The Secondary Sampling System (SSS) is designed to monitor water samples from the turbine cycle and the Circulating Water System. Water quality analyses are performed automatically on these samples as appropriate to determine pH and conductivity levels, dissolved oxygen, residual hydrazine, and sodium ion concentrations. Some of these measurements are used for the automatic control of water chemistry; tha remainder are recorded to permit appropriate corrective action by the operating staff. The PSS lines that penetrate the Containment can be isolated at the building boundary by automatic valves that close on receipt of a containment isolation signal. All sample lines penetrating the Containment are designed to meet Seismic Category I requirements from the 1 l first isolation valve inside the Containment to and including the first isolation valve outside the Containment. All other portions of sample l lines are designed to meet Seismic Category II requirements. 1 i
\
2.6-1 Amendment 5 (December 1988)
- 1. .
The PSS piping is' designed in.accordance'with thatl portion of.the
'k American National Standards Institute (ANSI). Code that also applies to lh the system.it serves. The PSS contains radioactive and potentially' radioactive fluids and is designed in accordance with-the American Society of Mechanical Engineers (ASME) Code, Section'III,' Nuclear Power i]
Plant Components,.for Class-III. components. Sampling system Containment. penetrations are designed to ANSI B31.7, Class II, from.the. isolation valve inside..to and including the isolation valve outside the Contain-ment. The SSS is designed to ANSI B31.1.0 requirements. LO: l O 2.6-2 Amendment 5 q (December 1988)
]
l
p W ' y hI
<, 2.7' CHEMICAL AND */OLUME CONTROL SYSTEM 1.TTDOWN AND CHARCING DESIGN ikg ) .
The charging and~1etdown functions of the Chemical'and Volume Control
~
~ System (CVCS) are employed to maintain a progrananed water level in the Reactor Coolant System (RCS)l pressurizer, thus maintaining proper reactor coolant inventory during all phases of plant operations. This is achieved by means.of.a' continuous feed and bleed process during which tha feed rate is automatically. controlled based'on' pressurizer water 1cvel.
The bleed rate can be chosen to suit various plant operational require-ments by selecting the proper combination of letdown orifices in the ietdown flou path. Reactor coolant is discharged to the CVCS letdown from the reactor coolant loop piping at a point upstream of the reactor coolant pump; it then flows through the shell side of the regenerat.ive heat exchanger L .Where its temperature is reduced by heat. transfer to the charging flow passing through the heat excht.nger tubes. The coolant then experiences a g
'large pressure reduction as it passes through a letoown orifice. The
'ietdown line then passes through the Containment and the Containment isolation valve. The coolar.t then flows through'the tube side of the F
letdown heat exchanger where its temperature is further reduced to the operating temperature of the mixed-bed domineralizers.
.The high-pressure charging flow normally comes from the centrifugal charging pumps. Charging nortnally flows from the pump through tho &
Containment to the regenerative hoat exchanger.
'The high-pressure piping outside the Containment in the CVCS is designed in accordance with the requirements of American National Standards 7 Institute (ANSI) B31.7, Class II, Code for Nuclear power piping.
O 2.7-1 Amendment 5 (December 1988) I _ _ _ _ _ _ _ _ _ _ _ l
l' l < l I fv 2.8 EESIDUAL HEAT hEMOVAL SYSTEM DESIGN
)
)
1 %J l The Residual Heat Removal System (RHRS) is designed to remove residual heat from the core and reduce the temperature of the Reactor Coolant C; { v I System (RCS) during the second phase .sf the plant cooldown. During the first phase of cooldown, the temperature of the RCS is reduced by l l transferring heat from the RCS to the Steam and power Conversion System J I through the use of the steam generators. i The RHRS is placed in operation approximately four hours af ter reactor
^f shutdown when the temperature and pressure of the RCS are less than 350*F and 425 psig, respectively. Assuming that two heat exchangers and two pumps are in service and that each heat exchanger is supplied with component cooling water at design flow and temperature, the RHRS is designed to reduce the temperature of the reactor cools.nt from 350*F to
^
140*F within 20 hours. The heat load handled by the RHRS during the l cooldown transient includes residual heat from the core and reactor coolant pump heat. The pressure piping in the RHRS outside the ()% Containment is designed in accordance with American National Standards Institute (ANSI) B31.7, Class II, Code for Nuclear power piping and is v Seismic Category I. 1 i l
. I /,g--- Amendment 5 %, 2.8-1 (December 1988)
v f-
. }l'.
V .; 2.9 PROCESS STEAM SYSTEM DESIGN-
.\
The Process Steam System supplies process steam to the following .1] l components: ,
- 1) Chemical and Volume Control System boric' acid evaporators and t] ,
feed preheaters.
- 2) ~ Boric acid batch tank.
- 3) Clean radioactive waste evaporater,
- 4) Cask wash pit.
The system supplies the required amount of process steam from the _
^
extraction line for Heaters SA and SB from the high-pressure turbino exhaust during normal operation and from a process steam boiler during plant startup or shutdown. All pressure vessels and pressure retaining parts of the process steam boiler are designed and fabricated in accordance with Ameiict.is Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section VIII. 7 The piping in the system is in accordance with American National Standards Institute (ANSI) B31.1.0. All components of the Process Steam System are designed to meet Seismic Category II requirements. i 2.9-1 Amendment 5 (December 1988) i __ - _ -- -- )
3 p- 2.10 AtrKILIARY FEEDWATER (AFW) SYSTEM DESIGN During emergency conditions, the auxiliary feedwater (AFW) system automatically maintains water level in the four steam generators. The m system includes titree pumps, two safety-related and one non-safety-related. The non-safety-related portion of the system supplies water to the steam generators during normal plant startup and shutdown. The safety-related AFW system is designed to provide two redundant means hi of supplying feedwater for removing heat from the reactor coolant. Each
^
pump is sized to supply 100 percent of the water flow required for a safe j3 cooldown of the RCS under any condition. Tht total dynamic head of the pumps is based on the most severo condition of pumping foedwater into the steam generators when the main steam safety valves.are discharging to atmosphere. Each pump is capabic of starting and delivering the rated flow of 960 spm within 60 seconds of receiving the start signal. 'Ih The steam turbine driving one of the two safety-related AFW pumps is designed to operate with steam produced in the steam generators. The turbine is rated for inlet steam pressure of 885 to 1085 psig at full load down to 110 ps's at plant startup and shutdown. The diesel engine driving the other safety-related pump is capabic of starting by batteries. The entire AFW system, with the exception of the non-safety-related pump; and its assnciated piping and components and the pump recirculation h .. t
'ines, is designed to meet Seismic Category I requirements. The pressure piping from the pump to the check valve is designed in accordance with l American National Standards Institute (ANSI) B31.7, Class II.
i O 2.10-1 Amendment 5 (December 1988)
.f_s . TABLE 2-1 MAIN STEAM SUPPLY SYSTEM MAIN STEAM LINE SAFETY VALVES Number of main steam lines. . . ....... 4 Number of valves per main steam line . .. . 5 Total number of safety valves . ......20 Design Data for Valves in Each Main Steam Line Set Pressure Flow Valve No. (psia) (1h/hr) 1 1170 671,083 2 1200 688,083 3 1210 911,779 4 1220 919,'226 5 1230 (126,674
%)
Total per line 4,116,84! Total capacity for four lines, Ib/hr 16,46',380 i
'i
)
- .w) ,.
- \
W I I .l l l
-a o
1100 Steam-Generator Outlet Prmure 1000
$: N r.
h Turbins. Generator inlet Pressure w w 800 700 I I I I 1 0 20 40 60 80 100 LOAD . PEft CENT , FIGURE 21 VARIATION OF STEAM PRESSURE WITH LOAD
< Amendment 5 (December 1988) 9
/
'1-4-
f, , O FIGURE 2-2
/
through FIGURE 2-10 i INTENTIONALLY DELETED l 0 O Amendment 5 (December 1988)
l k' i;I
. ,a-N 3.0- EQUTPMENT NECESSARY FOR SAFE STRITDOWN OF THE REACTOR N, .
This section describes the systems normally employed for safe shutdown Purposes. It goes on to enumerate the shutdown capability required for each event associated with the failures dealt with in this' evaluation.
't
\
3.0-1 - _ = _ - _ _ -- _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ _ _ _ _ _ _ -
1 l
-l l
-. 3.1 NORMAL SHUTDOWN
() 1 1 The systems necessary for normal, controlled safe shutdown of the reactor are those systems associated with the major functions in both the primary and secondary sides of the Nuclear Steam Supply System. There are no individually identifiable safe shutdr-3 systems per se. Ilowever, pre-scribed procedures for normally securing and maintaining the plant in a safe condition will be instituted by appropriate alignment of selected j fluid and control systems. The system functions required to be aligned for maintaining normal, controlled safe shutdown of the reactor are the minimum number that will:
- 1) Prevent the reactor from achieving criticality.
- 2) Provide an adequate heat sink such that design and safety limits are not exceeded.
The designation of systems that can be used for a safe normal shutdown h depends on identifying those systems which provide the following capa-bilities for maintaining a safe shutdown condition:
- 1) Reactor coolant boration.
- 2) Adequate supply of auxiliary feedwater.
- 3) Residual (decay) heat removal.
All systems required for a safe shutdown associated with these functions have been designed in accordance with the single failure criteria, as stated in Section 2.1. The following subsections identify systems which are required for safe shutdown of the reactor in the specific faulted conditions which this report is analyzing; these are conditions precip-itated by the rupture of high energy fluid lines outside the Containment. m 3.1-1 i
3.2 EMERCENCY SHUTDOWN LU ' An emergency shutdown may be caused by any natural or accidental event of infrequent occurrence. This includes related consequences, which affect the plant operations and require the use of other than preferred L systems or Engineered Safety Feature (ESF) Systems to bring the reactor to a safe shutdown condition.- In the context of this report an emer-gency shutdown is also considered to be precipitated by a pipe rupture that is of sufficient magnitude the: a plant shutdown is deemed necessary even though normal shutdown systems can be used to effect a reactor shut-down. All events are analyzed independently and are not assumed to occur
/
simultaneously. In this analysis, loss of preferred power is assumed to occur in those events which cause Protection System actuation effect-ing a plant trip. In this case, . loss of function of onsite a-c power (emergency diesels) and batteries mast be prevented. 3.2.1 EMERGENCY SHUTI)OWN WITH A MAIN STEAM LINE RUPTURE The equipment necessary for a safe shutdown of the reactor is the same for any pipe break location on the main steam line. The effect of any main steam break on plant shutdown is the loss of one steam generator for reactor decay heat removal immediately after reactor trip. For a large steam line break, the following must be available to accom-plish safety functions:
- 1) Safety injection to pump borated water into the core and, j thereby, limit the core power transient following the break.
)
J
- 2) Isolation of main feedwater to the steam generators to limit the Reactor Coolant System (RCS) cooldown. 1 l
i
- 3) Closure of the main steam isolation valves to limit ;
i RCS cooldown and reverse flow. l j
}
3.2-1 l
Ju f ;
'p ,
- 4) Auxiliary feedwater (AFW) is required to dissipate reactor decay k heat. In the event of a concurrent loss of offsite power, at least one of the two auxiliary feed pumps would be required within one minute to remove decay heat.
In order to cool the plant down.to the Residual Heat Removal System (RHRS); temperature end pressure, AFW must be available,.and the st'eam i 1._ generator power-operated relief valves must be operable. 3.2.2 EMERGENCY SHUTDOWN WITH A FEEDWATER LINE_RtrPTURE A feedwater rupture between the Containment and the feedwater check valve is considered to be~the worst-case feedwater rupture because of.the complete blowdown of one steam generator in addition to almost unrestricted flow from the feedwater pumps. For this rupture, the following must be available to accomplish their safety functions:
- 1) Safety injection to pump borated water into the core, and
( thereby limit the core power transient following the break.
- 2) Feedwater to the intact steam generators.
O w
- 3) Closure of main steam isolation valves.
In order to cool the plant down to the RHRS operating temperature and pressure AFW from at least one AFW pump must be available and the steam generator power-operated' relief valves must be operable. I For a large break between the feedwater pump and the main feedwater check valve, the feed line check valve will prevent water or steam release from any of the steam generators through the break. A large break at this point is thus essentially a loss of normal feedwater. In this case, the equipment that must be available to accomplish the safety function 1 I consists of the AFW System and the intact portlen of the Feedwater 'l 3.2-2 Amendment 5 (Dece:iber 1988) j l 1 a 1 _ _ _ _ _ . _ _ _ _ _ 1
e p, 4 43 s (System; safety injection is not required. This'caso covers.all lessor ((l' 'Feedwater and Condensate System high-energy line breaks. p 3!2.3 -EMERGENCY SHUTDOWN WITH AN AFW PUMP STEAM SUPPLY LINE RUPTURE A rupture of the AFW pump steam supply line between the main steam line and the normally closed air-operated valves is considered a less severe @ main steam line rupture, and its necessary emergency shutdown equipment is discussed in Section 3.2.1. After the rupture steam can be supplied-to the turbine-driven. auxiliary feed pump through the three.tenainin6 intact steam supply lines. The check valvos in each line limit the , blowdown from.this rupture to a single steam generator. O Based on the fact that.the steam is supplied to the turbine driver only during startup, shutdown, or test conditions, and upon the folloin,ng considerations, the probability of a rupture in this system between the air-operated valves and the AFW pump turbine driver is considered to be
~
lh extremely remote.
- 1) High leve1~of system quality control.
2). periodic inspection.
- 3) Low usage factor.
- 4) Short time the system exceeds 200*F and 275 psig annually.
- 5) Strict administrative controls on system operation during startup, shutdown and testing.
3.2.4 EMERCENCY SHUTDOWN WITH A STEAM CENERATOR BLOWDOWN LINE RUPTURE A steam generator blowdown line rupture outside the Containment would not cause a reactor or turbine trip. In this case, availability of offsite power is assumed and a " normal" plant shutdown can be effected. Break { 3.2-3 Amendment 5 (December 1988) l
- . 1
l'
-s locations have not been postulated between containment penetrations and L ( ,) the first outside isolation valve because the piping is conservatively reinforced and restrained beyond the valvo such that in the event of a 3 postulated pipe break, the transmitted pipe Irmds will neither impair the operability of the valve, nor the integrity of the piping or the Contain-ment penetration. The combined stresses in the piping between the Containment penetration and the first isolation valve are less than the h 0.8 (Sh + Sa ), all wable value as defined in Section 1.7 and a pipe whip restraint has been added at the Containment isolation valve, such that downstream pipe rupture loadings transmitted into the valve will be ;
so low as to not impair valve closure. For a steam generator blowdown line rupture dranstream of the isolution valve, the affected line could be isolated and en evaluation of the severity of the leak or rupture would determine the need to shut down the reactor. If needed, the shutdown procedure would use normal shutdown procedures. [\~- 3.2.5 EMERGENCY SHUTDOWN WITH A SAMPLING SYSTEM LINE RUPTURE Because of their small size, a rupture in any Sampling System line, even the RCS sample lines, would not cause a plant trip. In this case, availability of of fsite power is assumed, and a " normal" plant shutdown can be effected. 3.2.6 EMERCENCY SHUTDOWN WITH A CHEMICAL AND VOLUME CONTROL ,, SYSTEM (CVCS) LETDOWN LINE RUPURE QC , For a break in the letdown line between the Containment and the letdown heat exchanger, the following must be available:
- 1) Charging plus either normal makeup or makeup from the primary makeup water storage tank to maintain liquid inventory in the RCS until the break is identified and the letdown line isolated.
- 2) Letdown line Containment isolation valves.
3.2-4 Amendment 5 ( fm}
\-- (December 1988)
\
1 r j -( . ' 3')' APW to cool the plant down to the conditions necessary to s,- initiate operation of the RHRS. 4)' RHRS to bring the plant to cold shutdown conditions. C '
'3.2.7 EMERGENCY'8HUTDOWN WITH A CVCS CHARGING LINE RUPTURE For a-full area break in the charging line between the Containment and
~
the charging pumps, the following must be available:
- 1) Charging (from safety injection pipe) plus either normal makeup or makeup from the primary makeup water storage tank to maintain liquid inventory in the RCS.
2)~ AFW to cool the plant down to the conditions necessary to initiate operation of the RHRS.
- 3) RHRS.to bring the plant to cold shutdown conditions.
b V' 3.2.8 EMERGENCY SHUTDOWN WITH A RESIDUAL HEAT REMOVAL SYSTEM LINE RUPTURE The maximum temperature and pressure at which shutdown cooling can be. initiated is'350*F and 400 psig. At this point, the reactor is in a shutdown condition. The RHRS is designed and constructed to meet American National Standards Institute (ANSI) B31.7, Class II requiro- O ments. The RHRS is designed for Seismic Category I loadings. The rupture of a pipe elsewhere in the plant requiring the plant to be brought to the cold shutdown condition can be accomplished even when the postulated single active failure results in the loss of the ability to [h use the RHR system for normal plant cooldown. In this event, the reactor would be shut down and the plant maintained in a stable, steady-state N./ 3.2-5 Amendment 5 (December 1988)
J V: l l-- J a
,A condition.~ indefinitely'by dumping steam to dissipate reactor decay heat and maintaining steam generator water level by use of AFW until repairs : b to the RHR system could be effected. 1 3.2.9 SHUTDOWN WITH A PROCESS STEAM LINE RUPTURE The Process Steam System from the process steam boiler is not in operation when.the reactor is in the normal or hot standby conditions; therefore,'an' accident analysis is not required. A rupture in the Process Steam System from the fifth-stage extraction is covered in Section 3.2.2.
3.2.10 EMERGENCY'$HUTDOWN WITH AN AFW LINE RUPTURE The AFW System is an Engineered Safety Feature (ESP) System. It is not 'lh , only enquired to mitigate the consequence of accidents, but it is also required as a means of dissipating the energy from the RCS during periods when the main heat sink (main condenser dump and Main Feedwater System)
.is unavailable, for example, during a blackout situation.
A pipe rupture in the high-energy portion of the AFW System between the l}^ main feedwater line and the check valve in the common 3-inch line is i considered a less severe main feedwater line rupture. The necessary emergency shutdown equipment is discussed in Section 3.2.2. The criteria in Section 1.2 states that no accident is assumed to occur concurrently with a pipe failure outside Containment. Based on the
^
following considerations, the probability of a rupture in this system upstream of the first check valve is considered incredible.
- 1) High level of system quality control.
- 2) Periodic inspection.
b'. 3.2-6 Amendment 5 (December 1988) c__.___
1 l g .
- 3) Low usage factor.
! 4) Low temperature of the feedwater. OI
- 5) Strict administrative controls on system operation during startup, shutdown, and testing, d
i 1 1 i ( f Amendment 5 3.2-7 (December 1988)
t i.--____-_-__ 4.0 HIGH ENERGY FLUID PIPING DESCRIPTION
..d(% .
This section describes the~ classification and delineates specific piping, according to system and location in the plant which fall into the category of High Energy Piping Systems. O 4,0-1
,m N
\~/ 4.1 PIPING SYSTEMS WITH TEMPERATURES HIGHER THAN 200*F AND ,
PRESSURES HIGHER THAN 275 PSIG l$ High energy piping systems within the plant are classified into two groups in this report. The first group contains those systems whose operating temperature and pressure will exceed 200*F and 275 psig during normal power operation or while at hot standby. Piping systems of j interest falling within this classification are described in the
]
following subsections. The second group is discussed in Section 4.2.
^
4.1.1 MAIN STEAM SUPPLY SYSTEM PIPING l3 The Main Steam Supply System (MSSS) carries steam generated in the four lh steam generators through the Containment to the following components:
- 1) Turbine generator.
,. s
- 2) Moisture separator reheaters.
1 Q ,/
- 3) Steam jet air ejector.
4)' Turbine shaft gland seals.
- 5) Steam generator feedwater pump turbines.
- 6) Turbine-driven auxiliary feedwater (AFW) pump.
- 7) Turbine bypass.
- 8) Main steam blowdown and sample supply.
Figure 4-1 shows the isometric arrangement of the MSSS piping as it , passes from the Containment through the main steam support structure and U l l r% l f ) 4.1-1 Amendment 5
\d (December 1988) i
1 i (N into the Turbine Building. The main steam piping from the steam hji generators up to and including the check valves and the main steam piping to the AFW pump turbine have safety-related functions. Oa Saturated steam generated in the four steam generators flows out through Containment penetrations in four 28-inch main steam lines to the turbine 1 stop valves. A flow element is provided in each of the four main steam lines inside the Containment, which acts to limit the steam blowdown in the event of a main steam line break. The main steam line from each steam generator is provided with five spring-loaded safety valves and one power-operated relief valve. The safety and relief valves are 3.ocated in the main steam support structure between the Containment penetration and the corresponding main steam line isolation valve. The safuty valves are direct spring-loaded type with dual discharges. The power-operated relief valves are air-operated diaphragm type and are provided with necessary limit switches, stem c travel indicators, positioners, air sets, and interconnecting tubing to form a complete operating assembly. The power-operated relief valves are set to open before the first spring-loaded safety valve opens. The power-operated relief valves and the spring-loaded safety valves discharge to the atmosphere. The relief valves are mounted on a 28-inch m manifold above the main steam line at an elevation of 95 feet, 9 inches as shown in Figure 4-1. A quick-acting, pneumatic-cylinder-operated isolation valve and a check valve are installed in each main steam line outside the Containment and downstream of the safety valves. There is a 3-inch line for supplying main steam to the AFW pump turbine i on all four main steam lines outside the Containment and upstream of the isolation valves.
^
Each of the four main steam lines is connected to a 28-inch header through 14-inch lines upstream of the turbine stop valves, to equalize the steam im ! D} 4.1-2 Amendment 5 (December 1988) l l l l l _ _ _ . _ _ _ . _ _ _ - _ _ _ _ __ _ .l
l-l l n
' ,m
, pressure in the lines. The header also supplies steam to the moisture 13 separator reheaters, steam jet air ejector system, turbine shaf t gland seal system, steam generator feedwater pump turbines, and provides a connection to the Turbine Bypass System.
^
The MSSS piping centerline elevation is 81 feet through the main steam l3 support structure and into the Turbine Building to the stop valves. The
^
elevations of other major piping in the MSSS are shown in Figure 4-1. 3 4.1.2 CONDENSATE AND FEEDWATER AND EXTRACTION SYSTEMS PIPING The majority of the condensate piping and the extraction piping will not be described in detail because of its location in the Turbine Building in areas far removed from any engineered safety feature (ESF) equipment.
'^
The condensate piping of concern is located directly over the AFW area. i3 The analysis of the effects on ESF equipment in the Turbine Building is detailed further in Section 5.2. rm ( Tis section af the Feedwater System piping of interest for this report runs from the steam generator feedwater pump through the sixth- and seventh-stage feedwater heaters and out of the Turbine Building and through the main steam support structure to the Containment penetrations. The steam generator feedwater pumps take suction from the outlets of the fifth stage of feedwater heaters and the heater drain pumps and discharge
^
the feedwater into two cross-connected, parallel 24-inch lines as shown i ] in Figure 4-2. The two parallel streams pass through the sixth and seventh stages of feedwater heating and discharge into a 30-inch common ] header before entering the four 14-inch lines connecting to the steam 10 ) generators. Motor-operated isolation valves are located on the discharges of the
)
condensate pumps and steam generator feedwater pumps. Manually-operated valves are provided on the suction of all the pumps. Manually-operated 1 valves are provided in each feedwater heater train at the outlet of the 1 U 4.1-3 Amendment 5 l (December 1988) l l 1 l
l 3 fifth-stage feedwater heater, the inlet to the sixth-stage feedwater i 1 l ( ~/ heater and the outlet of the seventh-stage feedwater heater for isolation of the feedwater heater train when required. Air-cperated feedwater control va?ves with manually-operated isolation 7 valvos upstream and electrohydraulically-operated isolation valves downstream are provided on feedwater lines to the steam generators. 7 Seismic Category I isolation check valves are provided in each of thesc lines downstream of the electrohydraulically-operated valves. The 3-inch Ih auxiliary feedwater line for each steam generator joins the feedwater lines to the steam generators outside the Containment and downstream of the isolation check valves. The isolation check valves isolate the Seismic Category I AFW System piping from the Seismic Category II feedwater piping. The steam generator feedwater pumps are tripped automatically on any of the following signals: i s ) 1) Feedwater pump trip signal from the Engineered Safety Features Actuation System (ESFAS). This signal is generated on any of the following conditions: a) Steam generator high-high level. b) Any conditions which cause a safety injection signal. 3
- 2) High feedwater pump discharge pressure.
- 3) Low pump suction pressure.
- 4) Low pump / turbine lube oil pressure.
- 5) Pump turbine driver overspeed.
- 6) Pump suction valve not fully open, ry 4.1-4 Amendment 5 ,
(December 1988) .
l l
/~ 7) Turbine driver exhaust low vacuum.
l J
- 8) Turbine driver exhaust high temperature.
- 9) Condensate pumps tripped.
- 10) Thrust bearing wear.
The Condensate and Feedwater System is isolated from the steam generators . by closing the feedwater control valves and the feedwater control bypass C valves by closing the electrohydraulically-operated valves downstream of the control valves and the bypass valves on an isolation' signal from the ESFAS. An isolation signal is generated on any of the following condi-tions:
- 1) Two out of three high-high steam generator level signals from any one of the steam generators.
/x (v ) 2) Any of the conditions Which causco a safety injection signal.
I
- 3) Low T,y in coincidence with a reactor trip.
During the transient following a trip of one of the two steam generator feedwater pumps when the plant is operating at full load, the other steam generator feedwater pump can supply sufficient water (with two condensate pumps and two heater trains in operation) to prevent a reactor trip. With a failure of one of the two condensate pumps when the plant is operating at full load, a recctor trip would occur. The Seismic Category II portion of the Condensate and Feedwater and Extraction Systems, including the condensate pumps, low-pressure and high-pressure feedwater heaters, the steam generator feedwater pumps and the condensate and feedwater piping up to the isolation check valves is not essential for safe shutdown of the plant. In the event of failure of both steam generator feedwater pumps, the AFW pumps will start auto-f-~s (_,) 4.1-5 Amendment 5 ; (December 1988) 1 I l
, ~ g
. '~ , .7
-l matically and supply water for initial cooldown of the reactor. In tho' )
( . event of failure'of both condensate pumps, the steam generator feedwater pumps will automatically trip, and this in turn will start the AFW pumps. 4.1.3 AUKILIARY FREDWATER PUMP TURBINE DRIVER STEAM PIPING The steam turbine driving one of the AFW pumps is a single-stage noncon-densing turbine. Required steam supply for the turbine comes from any-one of the four steam generators. A 3-inch connection is provided from 'I each of Seismic Category I sections of the four 28-inch main steam lines ,
. upstream of the main steam isolation valves. Each 3-inch line is d provided with an air-operated supply valve, a manually-operated isolation valve upstream, and a check valve downstream for automatic isolation.
The normally closed air-operated valves isolate the high-energy main g
. steam from the remaining-AFW pump turbine steam supply piping and are .i located above the protective concrete floor at Elevation 69 feet in the lh i main steam support structure. The check valves in each line are located below the floor. The four 3-inch lines join into a single 4-inch line and then entors the Turbine Building through an 8-inch guard pipe at Elevation 57 feet. The guard pipe protects the redundant AFW train from a rupture in the steam supply line between the Turbine Building wall and g the turbine driver. The turbine exhausts low pressure steam through an 8-inch pipe to the atmosphere outside the Turbine Building. The B-inch exhaust piping is encased by a 12-inch guard pipe to protect the redundant AFW train from an exhaust line rupture. The system is Seismic Category I from the main steam line to the turbine exhaust outsido the Turbine Building.
4.1.4 STEAM CENERATOR BLOWDOWN SYSTEM PIPING Two 2-inch blowdown lines, one from each side of the bottom of the lh shell-side of a steam generator, are connected together, forwing one 2-inch blowdown line from each of the four steam generators. Each blowdown line penetrates the Containment and increases to a 3-inch line. 3 The lines are reduced back to 2-inches just prior to discharging into the I 1 steam generator blowdown tank. b 4.1-6 Amendment 5 (December 1988) ]
)
f i, l p Each blowdown line is provided with two Containment isolation valves ;
' (f located just outside the Containment. The valves are motor-operated globe valves which are normally open. The valves can be manually controlled from the control room or closed automatically on receipt of either an AFW pump start signal or Containment isolation signal. In addition, the second isolation valve also receives closure signals in the }'
event of AFW pump start, steam generator sample " alert" or "high" radiation signals or a steam generator blowdown tank high water level ; signal. A throttle valve is provided in each blowdown line just upstream of the steam generator blowdown tank connection. The throttle valve is provided with a calibrated handwheel which can be adjusted (with reference to a calibration curve for the valve) to modulate the blowdown flow rate. The Steam Generator Blowdown System (SGBS) steam piping outside the lh , Containment is located below the feedwater lines in the main steam support structure. The piping exits the south wall of the main steam 3 support structure and enters the Blowdown Building where it discharges to h the blowdown tank. The SGBS has no safety-related function. 4.1.5 PROCESS SAMPLING SYSTEM PIPING
'Ine Process Mmpling System includes all the equipment necessary to remotely collect representative samples of various process fluids in the Trojan plant. All Primary Sampling System lines are 3/4-inch or smaller ih and were not analyzed for pipe rupture in accordance with the critero ,
stated in Section 1.2. The following Secondary Sampling System (SSS) lines were analyzed as part of the systems they serve:
- 1) Condenser hotwell samples.
- 2) Feedwater-to-steam generator sample.
,% l !'~)
4.1-7 Amendment S
\
)
(December 1988) l
7
- 3) Main steam leader samples. i
..?
The above-listed lines.are all 1-inch lines or larger. The SSS lines _ lL smaller than one inch were not analyzed for pipe rupture in accordance O with the criteria stated in Section 1.2. l} l 4.1.6 CHEMICAL AND VOLUME CONTROL SYSTEM LICTDOWN 10* AND CHARGING PIPING t The 2-inch letdown line from the Containment to the letdown heat IG exchanger passes through the Containment penetration area e.nd into the Auxiliary Building at Elevation 70 feet. This line is encased by a 6-inch guard pipe to protect equipment from a rupture of the 2-inch letdown line. The 3-inch charging line starts at the positive displacement charging pump at Elevation 29 feet and passes through the Auxiliary Building as shown in Figure 4-6. The'line then traverses the Containment penetration area to its penetration located at Elevation 73 feet. 13 . The entire Letdown and Charging Line Systems are Seismic Category I. 4.1.7 RESIDUAL HEAT REMOVAL SYSTEM (RHRS) PIPING l} The RHRS piping outside the Containment consists of two trains of piping identical in function, but separate in location. The following descrip-tion of.one train applies to the other. The residual heat removal pumps both take normal suction from a single 14-inch line that passes through the Containment at Elevation 55 feet. The 8-inch pump discharge then passes through the residual heat exchanger adjacent to the RHR pumps at Elevation 8 feet, 6 inches. The 8-inch line then returns to the Containment at Elevation 55 feet. The entire RHRS is Seismic Category I. l t" 4.1-8 ~ Amendment 5 (December 1988) 3
t _h' 4.2 ; PIPING SYSTEMS WITH TEMPERATURES HICHER THAN 200*F OR' PRESSURES HIGHER THAN 275 pSIG The second group for classifying high-energy piping systems contains those systems whose operating temperature and pressure exceed 200*F or
. 275 psig, during riormal power operation or while at-hot standby. piping systems falling within this classification are described in the following subsection.
4.2.1 PROCESS STEAM SYSTEM PIPING During normal p1&nt operation process steam is supplied from the extraction line for Heaters SA and SB from high-pressure turbine exhaust. The 8-inch extraction line in the Turbine Building hau a pressure reducing valve before the line is reduced to foJr inches. This inch line then traverses the Control Building at Elevation 101 feet. [h enters the Auxiliary Building and connects to a control valve at Eleva-tion 95 feet. This control valve further reduces the steam pressure, p Individual control valves at the inlet to the components regulate the flow of steam. The steam is supplied to Chemical and Volume Control System (CVCS) boric th acid evaporators and feed preheaters, clean radioactive waste evaporator, boric acid batch tank and cask wash pit. The condensate from the CVCS boric acid evaporators, feed preheaters, and clean radioactive waste evaporators is returned to the condensate receiver and is pumped to the I main condenser. During plant startup or shutdown, When extraction steam is not available, the process steam boiler, located in the Fuel Building, is actuated to supply the process steam. The steam flows from the boiler through a pressure control valve and a moisture separator to the distribution system. The system flow paths are the same as described before except O 4.2-1 Amendment 5 (December 1988) i-
- - - - _ _ - _ _ - - - _ _ _ _ _ - - _ _ _ _ - - . _ _ _ . _ _ b
i: 1,: g.w that~ the steam is! supplied from the process steam boiler and the 1 condensate returning from the components served is now pumped'to the boiler instead of the main condenser. 4.2,2 AUKILTARY FEEDWATER SYSTEM pTPING IC v The auxiliary feedwater (APW) pumps are normally supplied with condensate g l
- from the condensate storage tank. The pumps can also be. supplied from the Seismic Category I Service Water System (SWS) in the event of loss of the supply from the Seismic Category II condensate storage tank.
Each pump can feea any or.all of the four steam generators. A 6-inch th
. discharge line from each feed pump traverses the AFW area at Eleva- 13 tion 57 feet. Each pump discharge line is provided with a nonreturn th valve and an isolating valve to permit maintenance of the pump and the 13 nonecturn valve. Each discharge line then branches into four 3-inch th 1- . lines to supply the four steam generators. Downstream of the branches, each of.these eight lines is provided with a manually-operated isolation
. valve, a flow element, a motor-operated control valve, and another ,
manually-operated isolation valve. During AFW pump operation, the flow C. elements will detect a high flow condition resulting from a downstream AFW line rupture. The high feedwater flow signal will shut the normally open motor-operated control valve, which is located below the protective concrete flooring at Elevation 69 feet to isolate AFW to the ruptured th pipe. A 3-inch AFW line from one pump joins with a corresponding line from the second pump into a single 3-inch line. Above the protective concrete h flooring at Elevation 69 feet, this common line is provided with a flow indicator for remote and local indication and a check valve to separate the high-energy main feedwater from the AFW System. The single AFW line ;
~
then joins with the main feedwater line in the Seis:nic Category I section between the feedwater line isolation check valve and the containment in the main st'eam support structure at Elevation 71 feet. _I
$i 4.2-2 Amendment 5 (December 1988) l
____i___.____._ _ _ . _
ms.-.~,-..-._,-., . _ _ , . . ,__ ., F% %
/
.. g -
<?
</
s/ 6,/ 9 9
/
s ko. , . xN
\ N
,N
., \ '
+'%
g?.. i J ( g' ( s 7 \!, (c .s.27 . ..
/ -
, )6 yf ?
.ef$ ' '
J,;4'i,Y5
$y ,
c .% n y-fC' 3
- /N$. e 4/
/./ , oj7
.f
.y q
,/
/ /
I. ' I
,,/ (* fl,$*/
.,' y .
m
, l
. _/ 1
/
- i Gl?' l i
- 1 i
. / s .ii g%'
l
-- - .)
.gjj;]5'^
.. I i
4. s s ' ss .e e 1 g 6
/j -g'- sy
/ g . s c-@,9' e g s, /'s ~ k) 3 [F'e s X L: q g i s _ f - t#EkTN'
- CARD
(@s $jj:fd 6 )' MS [*fe 2
%, lA*. W f
~
no' l \ ~ ,p
- Ab % f {7 #
/,'
pq y Oh. qLy y%cN + -{ I s W
?I
~
fk,.a Yp
~
' N" ' $
AO ( 3 1.c a... igQ;gg3'%s s# Of'x L4 4 Ds J' !h* gkN'I'df l 4 j' N'k e y-en p l
, s ,- -%\ ,-
9 SRD196DOSt-Og ' TADJAN NJCLEAR PL ANT
'p , ISOMETRIC VIEW Or THE MAIN 5TE AM PIPING N THE MAIN STE AM SUFPORT I STRUCTURE AND TURBINT BJtLDING f!GUPE 4-l
FIGURE 4-la INTENTIONALLY DELETED l i O Amendment 5 (December 1988)
' f. " , t ,, k k'N % j
/
,g!>'\
y,e3*S ,v. .
,es '
p W,re/ ;
, G
/;7~
f Y ; d s
-/ j
- ^
j xx
+, ;
r ,, a
^
/ '
/ps "+
{!?
*p.
,y Ml)// s . R/
.x ~
s?g ' s-
,/
( l Am. _ __
g a gnz.- v @ J' L @ . p W ' v .- cs r > p -
-1@ ...,,,'
, ./'- 'l.,
- S
/
Ng',' u x
,# g
" 'gl 9 , ljR
,X 'N ,
E 5 s >,t$, k y,' ,p,,./
- / #,, ,6 sR -
t ds/ '/x '
~
*[/ s
,#N e4, >'f r' hy \p_,; ,+ ' N y' % yM . . ,,,, . 'N s 19 w ;
s e. ep 6 s k 3 ; N
< ,, ;.j N'.. .g gj$,, j.e N .
GIS'p4 s' x s 1 , y
<e ,# K
# ,, a/; a g' p,:> -
, s f .. > 4 ..- p
',,' % + . N g.,
j'S mbYY N& .sNEfS hh,,h xy 'y h . [ N.y:q s;:(sg2 ; .y flt x .m ...,a.,
/ / ,, % N[g..'g ,.A N / f.pp..
c .t,3s' -
,i p' y s ,~ ~ * ,
x /y
.y b <..
x-(y,/_ ALLSt,%gr
,o .-
. /.A,gn ,, ,-
n / .
-- :Y !' .
1 l f"#,f p? n /, . . '. J' i l I [ ,/
... / %
ero ,, g B .i ./ h' I. I , / / d
.PERTUlW.
',/ - /
/ ' l; -
CARD
/ , ,
,/
Also Available m
's[3c s I( Aperture Card 9,9A1R500(J- 03_
TROJAN NUCLEAR PLANT ! OMETRIC VIEW OF THE FEEDWA1ER PIPING lN T HE MAIN LTEAM $UPPOilT LTPUCTUTE AND TURB!NE BU;LDING m FIGURE 4 -2 Gf
i f:._ I 1 I 1 O 1 1
'I j
FIGURE 4-2a through
.I FIGURE 4-8 j
I I INTENTION , DELETED t I i o l i 1 l O Amendment 5 (December 1988) t I
- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ l
m t. f ' u s, 5 l. s 1 ,_, . . , 5.O PIPE RUPTURE ANALYSIS-7 l l In the.following sections, each of-the high energy fluid system pipe rup-l;
; ture analyses is' sumarized. The sumary includes a discussion of the .
effects of the postulated pipe breaks and the plant'modif1 cations required to mitigate the rupture effects. I y . s.o-1 u_i,____;______ _____:A---------- i
i
-[ i i
.. 5.1 MAIN STEAM LINE RUPTURE (v)_~
'"he 28-inch main steam lines carry saturated steam between 557"F and 1107 psia for.no-load hot standby, and 530*F and 888 psia for 100 percent-load operation. The pipe break criteria, as described in Section 1.2, i were used for the analysis of the main steam lines, except for the main ;
~
steam safety valve header inside the Main Steam Support Structure (MSSS). This portion of piping, which is between the Containment penetration and the second main steam isolation valve, has been analyzed in accordance with_the Branch Technical Position - Mechanical Engineering _ Branch (MEB) Number 1, " Postulated Failure and Leakago Locations in' Fluid $ System piping Outside Containment" dated July 1, 1974 (sec Appendix A). The piping downstream of the second main steam isolation valve is Seismic h Category II. The Seismic Category II portion of the main steam line was analyzed to determine the high stress locations. Using this analysis, which is summarized. in Table 5-1, the pipe break locations, shown by circled numbers in Figure 4-1, were postulated. Both a full area longitudinal and a double-ended rupture were considered at these break locatiens. 5.1.1 AREAS AFFECTED BY A STEAM LINE RUPTURE O v The consequences of a steam line rupture would be limited to the Turbine
~
Building and the MSSS. Equipment, components, and systems located in the lh Control Building, Auxiliary Building, or the Fuel Building would not be exposed to the effects of a ste 4a line rupture. Equipmer.t. structures, and components important to safety located in arean possibly affected by a steam line rupture are as follows: l l C
~
- 1) MSSS.
l 2) Auxiliary Feedwater System (AFW) and compartments.
- 3) Emergency diesel generators (EDGs) and compartments.
1. l l 5.1-1 Amendment 5 (December 1988) L_- _ _ _ _ _ _ -
.g .
*lb >
4)- Engineered safety features (ESP) Class 1E switchgear. 1]
.V h
- 5) Main'StearcSupply System upstream of the isolation valves. .i]
- 6) Containment' adjacent'to the MSSS. Ib
- 7) ' Auxiliary feedwater remote control / remote shutdown pans.i (C-160).
- 8) Main feedwater isolation valves.
2
- 9) Main steam and turbine first-stage pressure transmitters.
- 10) Train A safety-related cabling.
5.1.2 PIPE WHIP The methods outlined in Reference I were used to analyze the full area pipe breaks postulated above for pipe whips. Because of the large b resultant jet thrust forces, plastic hinge formation was assumed for'all unrestrained break locations'on the steam lines. The methods used for t.he analysis of the effects of the pipe whip forces on structures are outlined in Appendix B. Existing pipe whip restraints were reanalyzed and were found adequate to prevent damage to ESF electrical and fluid systems for pipe rupt.ures at'
-Location 1 as shown in Figure 4-1. These ruptures are located in the
't]
Seismic Category I portion of the main steam lines and the restraints
-provided prevent damage to the Containment structure and to the electrical and mechanical portions of the main steam isolation valves. The analysis of these break locations was identical for all four steam lines. It I
should be noted that tha stress levels at these break locations are below \ the limits defined in the American Society of Mechanical Engineers (ASME) g
~
Section III Class 2, portion of Section IB of MEB Number 1. Addition-ally, pipe whip restraints designed in accordance with Section 2C(1), l-o 5.1-2 Amendment 5 (December 1988)
Appendix A of MEB Number 1, are located immediately downstream of the
/ ,. } second main steam isolation valve which are capable of resisting bending
/
and torsional moments. A pipe rupture is not postulated for the branch connection on the main steam safety valve header designated Locations 2 and 3 on Figure 4-1. This portion of piping, which is between the containment penetration and the second main steam isolation valve, has been analyzed in accordance _ with MEB Humber 1. In applying these criteria, no pipe breaks nave ' peen U postulated on the main steam safety valve header piping inside the main steam support structure because: 1) the piping stress levels are below the limits defined in the ASME Gection III, Class 2, portion of Section IB of MEB Number 1, 2) pipe whip restraints designed per the requirements of MEB Number 1, Appendix A, Section 2C(1) are located immediately downstream of the second isolation valves, and 3) insievice inspection will be conducted on 100 percent of all Class 2 welds in this region defined as inspectable in accordance with ASME Section XI, Winter 1972 Addenda.
/~s The pipe whip analysis for Break Location 4 was performed to investigate the possible damage to the AFW area located at Elevation 45 feet, which I is partially under the main steam lines. The analysis indicated the need for a pipe restraint to prevent a longitudinal rupture at Location 4 from ]
whipping the pipe down to the floor and collapsing the AFW area root. l l The restraint was found to be necessary only on the steam line from the "A" Steam Generator. This is because this line is located farthest north 4 l of the four steam lines and the impact on the floor of any of the other ; three lines would not affect the AFW area. ^k Dj pipe restraints for breaks at Locations 5, 6, 7, and 8 and the rest of the Main Steam System were not provided, because any unrestrained pipe motion resulting from a rupture in these areas on all four lines would not joopardize any equipment, components or systems important to safety. 4h , 1 1 l 5.1-3 Amendment 5 (x (December 1988) , l I l I
= _ - _ _ - _ _ - _ _ _ _
+
?' 'The jet forces from a critical crack are not significant enough t'o creato a pipe whip affecting the safety-related items. Ib 5.1.3 JET IMPINGEMENT-The jet impingement force caused by the momentum change of fluid flowing.
through the break is a function of the upstream fluid conditions, fluid. enthalpy, source pressure. pipe flow restrictional friction and dimen-sions. The jet forces acting upon the pipe are computed using the method out-lined in Reference'1. The jet forces are assumad to be instantaneous th (with zero rise ~ time). The forcing function is assumed to be a straight' line which changes so slowly that the variations up to the time of maximum response are negligible. The methods used for the analysis of the effects of jet impingement forces on structures are outlined in Appendix B. The main steam support structure was analyzed to determine the effects of
%) '
jet impingement forces on the safety-related equipment located in the ih area. The support structure was found adequate to withstand the jet impingement loads applied by full area ruptures at Locations 1 and 2. Therefore, the main steam isolation valves are protected from the effects U of jet impingement, th-It was determined that any jet impingement loading resulting from a critical crack anywhere in the main steam lines or any postulated full 13 area longitudinal or circumferential pipe break in the remainder of the
' main steam lines would not affect any structures, systems or components necessary for a safe shutdown due to a steam line break.
5.1.4 COMPARTMENT PRESSURIZATION The postulated main steam line ruptures in the Turbine Building were l[ ' analyzed to determine the effects of the resulting compartment pressuri-zation. The various postulated breaks in these piping runs were analyzed @( N/ l 5.1-4 Amendment 5 (December 1988)
= _ ___- _ _ _ - - _ _ _ - - - - -
.l l I
g ' to determine the worst-case break in the Turbine Building. Compartment
.h pressurization analysis for the main. steam support structure will be addressed in a subsequent revision to this report.
The double-ended guillotine main steam line break at Location 4 (see 1 Figure 4-1) was found to be the worst-case steam pipe break in the Turbine Building. 7 The methods outlined in Appendix D were used to calculate steam mass and - 3 energy blowdown rates for a full area main steam pipe rupture. 1 Usin6 the methodology described in Appendix C, an analysis was performed to predict the peak pressures expected in the various compartments of the Turbine Building following the postulated worst-case steam line rupture. j The results of this analysis indicated that in order to ensure acceptabic environmental conditions within the safety-related equipment compart-ments, as well as to preserve the structural integrity of safety-related portions of the building, the following modifications were required: IC
~
O 1) Modification of a portion of the east and south Turbine Building exterior metal siding between Elevations 63 feet and 93 feet such that it will blow out at a Turbine Building internal pressure of approximately 0.2 psig. Use of blowout-type panels in lieu of permanent removal of the existing metal siding provides the necessary pressure relief, _ while maintaining the environmental protection normally 0' J afforded by an enclosed structure. l' - 1
- 2) Modification of the following equipment maintenance / access doors as required to prevent their structural failure:
i l Door 102 to the diesel-driven AFW pump room, Door 106 in the access corridor between the Turbine Building and the railroad bay, Loor 109 to the " Train B" EDG compartment, Door 110 at the east end of the railroad bay, and Door 132 to the " Train A" ESF switchgear room. O 5.1-5 Amendment 5 (December 1988) ( l C ___ _
- 3) Installation of backdraft-type dampers in the ventilation f -~ 7 exhausts for the following rooms to limit steam intrusion'into the safety-related equipment compartments: diesel-driven'AFW pump room, turbine-driven AFW pump room, and A-train AFW- {
control panel room.
- 4) Structural reinforcement of the smoke exhaust - knum for the th
" Train A" ESF switchgear room as required to prevent' collapse l}
due to the steam line. break pressure transient. O w
- 5) Installation of a flapper-type damper in the ventilation intake opening of the A-train AFW pump control panel room to limit Ih steam intrusion into the room.
3
- 6) structural reinforcement of the ventilation intake' duct for the turbine-driven AFW pump room as required to prevent collapse 'th due to the steam line break pressure transient. 'lh These modifications have been completed. Ih i(
5.1.5 FLOODING FROM STEAM LINE BREAK The Turbine Building contains the following safety-related equipment:
~
l l
- 1) EDGs located et Elevation 45 feet in the diesel generator ,
compartments. *lj l l- 2) AFW pumps, piping and/or controls located at Elevation 45 feet 10 I x in the AFW compartments. IC
- 3) Class 1E switchgear in the Turbine Building switchgear room located at Elevation 63 feet.
Ol v i Each of the rooms containing this equipment is isolated from the others
.and from the other portions of the Turbine Building by protective l}
O 5.1-6 Amendment 5 (December 1988)
I l i
,_ features, including flood dikes. ThereCore, thie equipment will not be l[h
/ i .) s.- affected by failures outside the rooms except where flooding would be allowed to continue to a point above the 48-foot elevation of the dikes. i{ j The design basis internal flooding rate for which the Turbino Building is lh l designed is 500,000 gpm, which is assumed to occur following a circula-ting water system rupture. Sinea this design basis flooding rate far exceeds the maximum blowdown rate calculated for the worst-case steam ; line rupture, there would be no flooding of ESF equipment due to a main steam line break in the Turbine Building. There is no safety-related equipment that would be affected by flooding ih in the main steam support structure as the result of a main steam line 1[h break. 5.1.6 ENVIRONMENTAL EFFECTS The environmental ef fects of a main steam line rupture in either the Turbine Building or the main steam support structure are discussed in (L. j i Reference 2. ()
,m 5.1-7 Amendment 5 (December 1988)
, q. . . There are no electrical or mechanical components or systems necessary 3 for a reactor safe shutdown that would be affected by the environmental
- conditions resulting from a main steam line rupture in the Turbine Building.
,A
/m 5.1-8 Revision 2 August 1975 j
s i q i I-
- f. 5.2- FEEDWATER LIVE RUPTURE
%/
The four 14-inch feedwater lines carry water at 300*F and 1140 psig at
~
no-load condition and 440'F and 920 psig at 100 percent load. The pipe lh_ break criteria as described in Section 1.2 were used for the analysis. The feedwater lines from the feedwater pumps to the check valve outside the containment penetration'are classified as Seismic Category II. The L Seismic Category II portion of the Feedwater System from the discharge of the feedwater pumps to the inlet nozzle of the sixth-stage feedwater heaters and from the discharge nozzle of the neventh-state feedwater. heaters to the Containment penet. ration, was analyzed to determine the high stress location. Using this analysis, which is summarized in Table 5-2, the pipe break locations identified by circled numbers in Figure 4-2 were postulated. A full area longitudinni and a double-ended circumferential rupture were both considered at these break locations. 5.2.1 AREAS AFFECTED BY A FEEDWATER LINE RUPTJRE lh The consequences of a feedwater line rupture would be limited to the Turbine Building and the main steam suppert structure. Equipment, 13 components, and systems located in the. Control Building, Auxiliary Building, or the Fuel Building would not be exposed to the effects of a , feedwater line rupture. Safety-related equipment located in the Turbine U Building and the main steam support structure is listed'in Section 5.1.1. L2. 2 PIPE WHIP The methods outlined in Reference,I were used to analyze postulated full area feedwater pipe breaks for pipe whip. Because of the large resultant ih jet thrust force, plastic hinge formation was assumed for all break loca-- tions in the feedwater lines. The methods used for the analysis of the effects of pipe whip forces on structures are outlined in Appendix B. Pipe ruptures at Locations 9, 10, 11, 12, 13, and 14 were analyzed th becaus_e of their proximity to the auxiliary feedwater pump compartment. 5.2-1 Amendment 5 (December 1988)
p ' Because of the pipe orientation, pipe whip motion would be away from I
\ _
auxiliary' feedwater (AFW) area for Break Locations 9,10,11,' and 12. , pipe Break Locations 13, 14, 15, 16, and 17 are at Elevation 83 feet and above. At these elevations there are no components, equipment or systems important to ' safety which could be affected by feedwater pipe whip. Therefore, no pipe whip restraints are necessary in these areas. The feedwater piping from Break Locations IS and'16 through the sixth and Ih seventh stage feedwater heaters and to the flow elements located on the 14-inch feedwater lines to the containment at Elevation 73 feet passes throu6h areas where no safety-related equipment is located. Because of g its physical isolation from any safety-related components, equipment or systems, pipe whip restraints are not necessary for any of the piping described above. i Analysis of pipe whip due to the postulated ruptures at Break Loca-tions 18 and 19 revealed that no componento, equipment or systems impor- . tant to safety could be affected by a feedwater pipe Whip. The analysis l for Break Locations 18 and 19 in identical t'o the break analysis for the j other three 14-inch feedwater lines in the main steam support structure. ! The jet forces from a critical cract are not significant enough to create pipe whip which could affect sr.f ety-rejsted structurts or equipment. h 5.2.3 JET IMPINCES42WT The jet Empinsement force etused by the momentum changc, of fluid flowins through the bre&k is 9 function of the vpstru m fluid condit14ns, fluid < enthalpy, pour,e pressure, pipe flow restric.tlenal friet. ion, and dimensions. I The jet forces actin,s upon the pipe are computed u1 Lng the method ottt-lined in Ecfcrence 1. rhe jet forces are assumed to be instantaneous th , (with icero ris,e time) . The fedcin5 function is assumed to be a straight l O 5.2=2 Amendment $ (December 1988) ] q l l
f- line which changes so slowly that the variation, up to the time of the maximum response, is negligible. The methods used for the analysis of the effects of jet impingement forces on structures are outlined in Appendix B. The feedwater piping from the feedwater pump discharges through the sixth- and seventh-stage feedwater heaters and to the flow elements th located on the four 14-int.h lines and is routed through areas without any th l safety-related equipment, components or systems. Because of its physical th isolation jet impingement effects from any full area or critical crack break in the piping described above could not damage equipment or systens l_- v necessary for a safe reactor shutdown. The main steam support structure is designed to withstand the jet impinge-ment loads arising from a full area pipe rupture at Break Locations 18 g and 19 without affecting any components, equipment or systems important to safety.
.j
.{m, .
The APW compartment was analyzed for the effects of a critical crack in either of the two main feedwater pump discharge lines which pass on both sides of the compartment. The AFW compartment was found to be able to th withstand loadings from a critical crack in a feedwater line anywhere in the vicinity. There were no postulated full area breaks in the feedwater system that could cause an additional load on the AFW pump compart:nent, j L2. 4 COMPARTMENT PRESSURIZATION l
'the postulcted main feedwater line ruptures in the Turbine Building were l]
analyzed to determine the effects of the resulting compartment pressuri- lj zation. Becauze of the lower energy release rate associated with a g
~
feedwater line break, the Turbine Building compartment pressurization
~!
l would be less than that already presented in the pressure analysis 18 , i n l discussed in Section 5.1.4. 13 l t 5.2-3 Amendment 5 (December 1988) L L- -- --_ - - - - -
- - - - - - - - - - - - - - - -- . - - - - - . - _- -- -- .-_ --_------_ _----- _ _--_---_ _ o
- _ _ -_ - - -.- - - - --_- - - - . - - - - - - - - - _ -- ._~
I x .l Compartment pressurization analysis: for the main steam support structure ,l
- will be addressed in a subsequent revision to this report. I
-5.2.5-' FLdODING.
As with the flooding associated with'a main steam line break discussed in Section 5.1.5, flooding from a feedwater pipe break in' the Turbine g-Building is enveloped by.the design basis flooding resulting from a
. circulating water system rupture. Therefore, no engineered safety .
feature equipment would be affected by flooding as'a result of a feed- 0 water pipe rupture in the Turbine Building. 'Furthermore, there is no. safety-related equipment that would be affected by flooding in the main- g=
~~
steam support structure from a main feedwater line break. 5.2.6' ENVIR'NMENTAL O EFFECTS The environmental effects of a main feedwater pipe break in either the-
. Turbine Building or the main steam support structure are discursed in g-Reference 2.
O. I' 5.2-4 Amendment 5 (December 1988)
l I J n 5.3 CONDENSATE OR EXTRACTION LINE RUPTURE !
, \ s D f The condensate and extraction piping carries water at temperatures up to 450*F and pressure up to 475 psig. The break criteria described in 13, Section 1.2 were used for analysis. All condensato and extraction piping is Seismic Catescry II; therefore, for this analysis, full area j l
longitudinal and double-ended circumferential ruptures were assumed to occur anywhere in the system piping. j
]
5.3.1 AREAS AFFECTED BY A CONDENSATE OR EXTRACTION LINE RUPTURE I The consequences of a condensate or extraction line rupture would be j limited to the Turbine Building. Equipment, components and systems th located in the Control Building, Auxiliary Building, Containment or the j Fuel Building would not be exposed to the effects of a condensate or - extraction line rupture. Safety-related equipment located in the Turbine Cj Building is listed in Section 5.1.5. ( 5.3.2 PTPE WHIP ,
'% f i
The methods outlined in Reference 1 were used to analyze postulated full area condensate or extraction pipe breaks for pipe whip. The methods ' used for the analysis of the effects of pipe whip forces on structures are outlined in Appendix B. J I With only one exception, all condensate and extraction piping is located far from any safety-related equipment, components or systems in the lh Turbine Building, such that any unrestrained pipe motion would not affect ] the ability to bring about a safe shutdown of the reactor. This assumes a pipe break located anywhere in each pipe, which is consistent with the criteria outlined in Section 1.2 for Seismic Category II piping. One of the two feedwater pump suction lines from the condensate pump passes over the AFW pump compartment roof. The analysis of piping in ' this area and this pipe in particular found that a full area pipe break C ()\ 5.3-1 Amendment 5 (December 1988) l l
E ' i L k ( , . - . and subsequent pipe motion would not affect the AFW area, except for a
>A.r.
j break in the vertical run near.the intersection of Columns o and 64. It was fc,und necessary to install a pipe restraint to prevent unrestrained pipe motion down to the AFW area roof.' I [
, Jet fore.es'from a critical crack are not significant enough to create a pipe whip affecting safety-related structures or equipment.
(( J 5.3.3 JET IMPINGEMENT 1 l: 1 The jet impingement force caused by the momentum change of fluid flowing i i through the break is a function of the upstream fluid conditions,' fluid I enthalpy, source pressure, pipe flow restrictional' friction and dimen- 'l s sions, d I The jet forces acting upon the pipe are computed using the method out-- lined in Reference 1. The jet forces are assumed to be instantaneous i{ i (with zero rise time). The forcing function is assumed to be a straight line which changes so slowly that the variation up to the time of maximum
~
response is negligible. The methods used for the analysis of the effects of jet impingement forces on structures are outlined in Appendix B. - With the exception of the pipe described in Section 5.3.2, there is no condensate or extraction pipe located where jet impingement forces resulting from a pipe break could affect any safety-related equipment, components or~ systems. l{ l q i i 4 The feedwater pump suction pipe which passes over the AFW area was ) analyzed for jet impingement effects. It was found that jet impingement forces from a full area pipe break would not damage the structures )i containing the AFW pumps, j i
)t 5.3.4 COMPARTMENT PRESSURIZATION Because of lower energy releases associated with a condensate or o 5.3-2 Amendment 5 1
(December 1988) l l 1 I _ = _ _ _ _ - _ - - _ _ - _ _ _ _ _ _ _ _ - _ a
l 1 extraction pipe break, the compartment pressurization in the Turbine (a} - Building is bounded by the main steam compartment pressure analysis as 10 v i i discussed in Section 5.1.4. 5.3.5 FLOODING Flooding from a condensate or extraction line pipe break in the Turbine g Building is bounded by the analysis in Section 5.1.5. 5.3.6 ENVIRONMENTAL EFFECTS 1 The environmental effects from a condensate or extraction pipe break in the Turbine Building are bounded by the analysis described in 13 Section 5.1.6. l
'n /
J l O i 5.3-3 Amendment 5 l (December 1988) l l l i l l
f3
) 5.4 AUX EIARY FEEDWATER PUMP TURBINE DRIVER STEAM SUPPLY LINE RUPTURE The 3-inch steam supply lines to the turbine-driven auxiliary feedwater (AFW) pump upstream of the normally closed air-operated steam supply 7l valves carry saturated steam between 557'F and 1107 psia for no-load hot th standby and 530*F and 888 psia for 100-percent load operation. The pipe lh break criteria as described in Section 1.2, were used for analysis. This piping is classified as Seismic Category I. ;
A rupture of this steam supply line between the main stcam line and the air-operated steam supply valves is considered a less severe steam rup- , ture with regard to consequences than the main steam line break discussed
- n in Section 5.1. 1 0
The steam supply line for the turbine-driven AFW pump downstream of the - normally cloried air--operated valves was not analyzed for the effects of a pipe rupture on a reactor shutdown for the reasons outlined in ( Section 3.2.3. C i l I i l l 1 i I i l 1 i l lj i 5.4-1 Amendment 5 ' (December 1988) i ! l L________________. _ _ _ _ _j
5.5 STEAM CENERATOR BLOWDOWN LINE RUPTURE n The four steam generator blowdown lines carry water at a maximum D tem, rature and pressure of 557'F and 1107 psia. The piping and valves from the containment penetration to the second Containment isolation - valve (MO-2808, -2810, -2812, and -2813) are seismic Category I. The l remaining piping from the second isolation valve to the steam generator blowdown tank is Seismic Category II/I. The pipe break criteria as b described in Section.1.2 were used for the analysis. No pipe rupturcs were postulated in the blowdown piping between the Containment b penetration and the first isolation valve as discussed in Section 3.2.4, or within the confines of the main steam support structure based on the pipe break criteria of Section 1.2. *j A steam generator blowdown line rupture downstream *of the first Contain-ment isolation valve enn be isolated by closing this valve with no O adverse effect on the capability to eventually establish and maintain a cold shutdown of the reactos if required.
\'~'/ 5.5.1 AREAS AFFECTED BY A STEAM GENERATOR BLOWDOWN LINE RUPTURE The consequences of a steam generator blowdown line rupture would be limited to the area outside the south wall of the main steam support structure and the blowdown tank buildinf,, No enfety=rcicted equipment would be affected by a rupture in these areas, b i 5.5.2 PIPE WHIP Unrestrained pipe motion resulting from a postulated rupture of the steam generator blowdown piping would not damage equipment necessary for safe shutdown of the plant.
5.5-1 Amendment 5 L (December 1988) [ l
I 5.5.3 JET IMPINGEMENT t i 1._/ There is ao equipment necessary to shut down the reactor located in the area that would be subject to jet impingement forces. 5.5.4 COMPARTMENT PRESSURIZATION There is no equipment located in the steam generator blowdown building that is necessary for safe shutdown of the reactor. Therefore, compart-
^
mental pressurization effects are inconsequential for the postulated ruptures in the steam generator blowdown line. 5.5.5 FLOODING Postulated ruptures of the steam generator blowdown piping would not result in flooding of safety-related equipment. 5.5.6 ENVIRONMENTAL EFFECTS I
,r x
'~~
The environmental effects of a blowdown line rupture would be limited to the imnediate area of the rupture whien contains no safety-related equipment. 1 i-m t )
\- / 5.5-2 Amendment 5 (December 1988)
i
.I 5.6 CHEMICAL AND VOLUME CONTROL SYSTEM LETDOWN LINE RUPTURE lu-The 2-inch letdown line carries water at 350'F and 304 psig from the- 10 Containment penetration'through the letdown heat exchanger to a control valve where the temperature and pressure are reduced to 100*F ind l] .
107 psis. The pipe break criteria described in Section 1.2 wero used for i the analysis. The entire Letdown System is Seismic Category I. Full-area pipe ruptures and critical cracks were postulated at any point on the letdown line between the Containment penetration and the pressure reducing control valve. 5.6.1 AREAS AFFECTED BY A CHEMICAL AND VOLUME CONTROL SYSTEM h LETDOWN LINE RUPTURE The consequences of a letdown line rupture would be limited to the Containment piping penetration area and the Auxiliary Building at G Elevation 63 feet. Equipment, components, and systems located in the Control Building, Turbine Building or the Fuel Building would not be exposed to the effects of a Chemical and Volume Control System (CVCS) IC letdown line rupture. 5.6.2 PIPE WHIP The methods outlined in Reference I were used to analyze pipe breaks for pipe whip. Because of the relatively small forces and the structural isolation of the CVCS letdown line, any unrestrained pipe motion resulting from a pipo break would not damage equipment necessary to bring about a 13 safe reactor shutdown. 5.6.3 JET TMPINCEMENT There is no equipment necessary to shut down the reactor located in the 13 letdown line area that would be damaged by jet impingement forces. 5.6-1 Amendment 5 (December 1988)
5.6.4 COMPARTMENT PRESSURIZATION Because of'the low calculated mass and energy release rates associated-with a letdown line rupture.. compartment pressurization is insignificant
- because the normal. ventilation in the affected area would remove the
.' released steam.
. 5.6.5 FLOODING Because of the relatively low mass release rate, area f'loor drains would handle any potential flooding problem in the Containment penetration area or the Auxiliary Building until operator action would terminate the blowdown by closing the letdown line isolation valve.
5.6.6 ENVIRONMENTAL EFFECTS . The environmental effects of a letdown line rupture would be limited to
- the immediate area around the letdown equipment which contains no . l{
equipment neceseery for a safe reactor shutdown.
'l I
i k i 1 l [ Amendment 5 5.6-2 (December 1988) r O .
l l l1 f. L-f f .- 5,7 RESIDUAL HEAT REMOVAL LINE RUPTURE l / l ( The Residual Heat Removal System was not analyzed for the effects of pipe th l ruptures on a reactor shutdown for the reasons outlined in Section 3.2,8. i3 1 1 R
,r-t-) ' 5.7-1 Amendment 5 (December 1988)
n 5.8 ' CHEMICAL AND VOLUME CONTROL SYSTEM CHARCING LINE RUPTURE lu O G. ' The 3-inch Chemical and Volume Control System (CVCS) charging line is 4 inches in diameter at the discharge of the centrifugal charging pumps (CCPs) and reduces to 3 inches in diameter prior to entering - Containment. The positive displacement charging pump also discharges to this 3-inch line. The CVCS charging line carries water at a maximum of 130*F and 2350 psig; therefore, critical crack breaks only were considered. The pipe break criteria as described in Section 1.2 were used for the analysis. The entire CVCS charging system is Seismic Category I. The CVCS charging system itself with the exception of the positive displacement pump is an Engineered Safety Feature (ESF) system. Its safety function would be performed by the high pressure safety irdection pumps (CCPs) in the event of a pipe rupture. The Charging System and High Pressure Injection System are designed in accordance with the single failure criteria to ensure containment isolation and Reactor Coolant System (RCS) coolant supply, respectively. " f 5.8.1 AREAS AFFECTED BY A CVCS CHARCING LINE RUPTURE The consequences of a charging line rupture would be limited to the Containment fluid penetration area and the Auxiliary Building at Elevation 25 feet. Equipment, components and systems located in the lh j Control Buildin6, Turbine Building or the Fuel Building would not be exposed to the effects of a critical crack break. H 5.8.2 CHARCING LINE PIPE WHIP The CVCS charging line was not analyzed for pipe whip because the maximum fluid temperature is below 200'F (see Section 1.2). 1 5.8.3 JET IMPINGEMENT 1 There is no equipment necessary to shut down the reactor located in the @ O 5.8-1 Amendment 5 (December 1988)
charging line area that would be damaged by jet impingement forces ( ,) resulting from a critical crack break. 5.8 4 COMPARTMENT PRESSURIZATION There would be no compartment pressurization from a charging line break because the charging fluid temperature will never exceed 130*F. 5.8.5 FLOODING l Because of the relatively low ness release rate, area floor drains would ! handle any potential flooding problem in the Auxiliary Building or Containment penetration area unti,1 operator action would terminate the charging pump flow. 5.8.6 ENVIRONMENTAL EFFECTS There would be no adverse environmental effects in the areas around a [' charging line critical crack break that would affect any safety-related l{3
~'
equipment, components or systems. i 1 l l t I
,A) \_- 5.8-2 Amendinent 5 (December 1988) t I
i; . h f'f? Q _ ::l f l 5.9 PROCESS STEAM 1.TNE RUPTURE During normal plant operation, process steam is supplied to the various n
' equipment listed in Section 2.9 from the fifth-stage extraction. The '$'
remainder of the' system is not in operation during normal or hot standby th conditions. The O l Pe rupture analysis'for the extraction line in.the I Turbine Building is covered in Section 5.3. Tne portion of the fifth-stage extraction in the control Building and the Auxiliary Building will be discussed in this section. The extraction line in the Control Building carries steam at 375*F and th r 170,pais; therefore, only the offects of the critical crack break were considered. After it enters the Auxiliary Building, it passes.through a control valve, which reduces the pressure to 75 psig. The process steam piping is all' Seismic Category.II. The pipe break criteria described in Section 1.2 were used for.the analysis.
. 5.9.1 AREAS AFFECTED BY A PROCES? STEAM LINE RUPTURE The consequences of a process steam line rupture would be limited to the Control Building and Auxiliary Building. The Turbine Building or the Fuel' Building would not be exposed to the effects of a process steam line critical crack break.
5.9.2 PTPE WHIP
- The process steam line was not analyzed for pipe whip, because the maximum fluid pressure is below 275 psig (see Section 1.2). Ib 5.9.3 JET IMPINGEMENT An 8-inch guard pipe around the process steam line in the Control th Building protects any equipment located noar it from the effects of jet
[ 5.9-1 Amendment 5
'I (December 1988)
i l 7_ impingement forces resulting from a process steam line critical crack ( ,jb break. The guard pipe is the same schedule pipe as the 4-inch pipe it l l] - contains. The process steam lines in the Auxiliary Building contain fluid at a pressure of 75 psig, which is too low to produce any significant jet 1 impingement forces. ; 1 5.9.4 COMPARTMENT PRESSURIZATION There would be no compartment pressurization from a process steam line critical crack in the Control Building because ot the guard pipe, or in the Auxiliary Building because of the comparatively low energy release rate. 5.9.5 FLOODING There would be no flooding in the Control Building because of the guard ( rm)
\/
pipe, which is continuous throughout the control Building, rhere would be no flooding in the Auxiliary Building because the floor drains would contain any fluid dischtrge from a process steam line critical crack break until operator action would terminate the blowdown. 5.9.6 ENVIRONMENTAL EFFECTS There would be no environmental effects from a process steam line break in the Cont tol Building because all fluid would be contained in the guard pipe. There would be no adverse environmental effects in the areas around a process steam line critical crack in the Auxiliary Building because of the low mass and energy release associated with such a break, f l r~'T
'\il 5.9-2 Amendment 5 (Deceuber 1988)
, 5.10 AUXILIARY FEEDWATER LINE RUPTURE V
The four 3-inch high-energy auxiliary feedwater lines from the main th feedwater line to the first leck valve contain water at 300*F and l l 1140 psig at no-loed condition and 440*F and 920 psig at 100 percent , l- load. The pipe break criteria of Section 1,2 were used for the ,f l analysis. The piping is classified as Seismic Category I. O 1 i A rupture of this auxiliary feedwater line between the main feedwater line and the first 3-inch check valve is considered a less sever: nain j l
^
feedwater line rupture with consequences bounded by those discussed in Section S.2 for the main feedwater line break. The auxiliary feedwater lines upstream of these check valves were not ; analyzed for the effects of a pipe rupture on a reactor shutdown for the ^- 1] reasons outlined in Section 3.2.10. ( l O i l 1 1 j i l I
'(
( 5.10-1 Amendment 5 (December 1988)
h TABLE 5-1
' . \.
)
MAIN STEAM LINE STRESS SL991ARY (FROM CONTAINMENT TO STOP VALVES) Break Point Thermal Seismic Weight Pressure Total I"3 (See Figure 4-1) psi psi psi psi psi ! 1 Terminal End - - - 2 Branch Connection - - - - 3 Branch Connection . 10,041 20,900 4 384 1500 8975 i . 5 9,212 472 1500 8975 20,159 ! 6 3,510 264 1500 8975 14,249 l 7 7,756 649 1500 8975 18,880- q 8 Terminal End - - -
)
l [a] 0.8 (Sa + Sh) = 35,000 psi
.k [b] No break postulated as discussed in Sectit: 5.1.2. 3 1
l i
- i V
Revision 2 August 1975 i _ _ _ - _ . . - . _ . . - - . _ _ - . - - _ - . _ - - -.-__. ._ . - - _ . . _ - _ - - _ - - - 'E
l
.9 TABLE 5-2
(. >3
- FEEDWATER LINE STRESS
SUMMARY
(FROM FEEDWATER PUMP DISCHARGE TO 6TH-STAGE HEATER AND FROM 7TH-l"i STAGE HEATER TO CONTAINMENT PENETRATION) Break Point Thermal Seismic Weight Pressure Total ") (See Figure 4-2) psi psi psi psi p_si 9 Terminal End - - - d 10 Terminal End - - - 11 4239 468 1500 6660 12,867 12 7036 1314 1500 6600 16,460 13 4043 461 1500 6600 12,664 14 6233 872 1500 6660 15,265 l 15 Branch Connectior - - - ! 16 Branch Connection - - - 3 I 17 Branch Connection - - - 18 8620 1733 1500 4202 16,055 19 7930 659 1500 4202 14,291
,.- s
% [a] 0.8 (Sa + Sh) = 46,875 psi il l
l l p_ k Revision 1 January 1974 1
Y? I 6.0 EMERGENCY SHUTDOWW PROCEDURE
-W . .
3 The Trojan Plant Operating Manual Emergency. Instructions provide for bringing the reactor to a safe shutdown and maintaining it in that condition in the event of a main steam line break or main feedwater line break. These' types of breaks provide the greatest challenge to reactor protective functions and thus are considered limiting case for high energy line break (HELBs) outside Containment with regard to safe plant and reactor shutdown. The reactor automatic. protection equipment is h designed to shut the reactor down cafely in the event of the above
'HELBs. 'The plant Engineered Safety Features systems operate from offsite electrical power or from onsite emergency. diesel-electric power, should offsite power not be available. Detailed descriptions of the above accident including assumed plant conditions and responses, are provided in the Trojan accident analysis, Chapter 15, of the Final Safety' Analysis Report, i
G 6.0-1 Amendment 5 , (December 1988) ! l 4
.-g N
jI SECTION'6.1 through SECTION 6.3 INTENTIONALLY DELETED
'l
-1 1
I l 1 i 6.1-1 Amendment 5 (December 1988) 1 i j
t.
7.0 REFERENCES
'~
- 1. BN-TOP-2 (Revision 1), Desir.n foe Pipe Break Effects, September 1973, @*
Bechtel Power Corporation.
- 2. PGE-1025, Trojan Nuclear Plant Environmental Qualification Program Manual. 7 l
1 j
\- 7.0-1 Amendment 4 (August 1987) i w - _ - _ _ - _ . . - - _ . . _ . . . _ _ _ .
_7.___, _ _ _ , _ _ , _ _ _ _ _ _ _ _ _ , ___ ____
, r ,_,__ . ,_ _ _ __ , , ._ __,.,, __,-_. __ y ._,..,_, ..- , , , . . _ , . _ . _ - - -,_.___7_ ,._,__,
}' l V '!s '.
{'.
'T
' APPENDIX A
-l i- i PIPE BREAK ANALYSIS CORRESPONDENCE 1
'P I I' i
l'
f:2
'l ~ x 4
.! "% UNITED STATES .
- ATOMIG, ENERGY COMMISSION W
- k*h U ' '
wAsm NETON. D.C. 20545 Q sj,. DEC10 1972
, Docket No. 50-344 m
Portland General Electric Company ATTN: Mr.' Joe L. Williams Vice President, Engineering ] and Construction 621: Southwest Alder Street. Portland, Oregou. 97205' Gentlemen: The Regulatory staff's continuing review of reactor power plant safety ; indicates that the consequences of postulated pipe failures outside of ' the containment structure, including the rupture of a main steam or feedwater-line, need'to be adequately documented and analyzed by. licensees and applicants, and evaluated by the staff as soon as possible. Criterion No. 4 of the Commission's General Design Criteria, listed in l LAppendix A of 10 CFR Part 50 requires that: g
" Structures, systetu , and components important.to safety shall-be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal opera-tion, maintenance, testing and postulated accidents, including loss-of-coolant accidents. These structures, systems, and com-ponents shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit."
The. previous version of the' Commission's General Design Criteria also reficcts the above requirements. l- Thus, a nuclear plant should.be' designed so that the reactor can be shuc down and maintained in a safe shutdown condition in the event of a postu-lated rupture, outside containment, of a pipe containing a high energy fluid, including the double ended rupture of the inrgest pipe in the main steam and feedwater systens. Plant structures, systems, and con 4ponenes . important'to safety should be designed ,and located in the facility to accommodate the effects of such a postulated pipe failure to the extent necessary to assure that a safe shutdown condition of the reactor can be accomplished and maintained. '
. i A-1
=l_-__-_=__- _ __ _ !
( I 1 i Portland General Electric Company 7g
.g~
V Based on the information we presently have available to us on the Trojan facility there is a potential for flooding of the auxiliary feedwater pumps should this postulated accident situation occur. However, sufficient information is not available for us to conclude ' that the design of your facility in other respects is in accordance with our criteria. We request that you provide us with analyses and other relevant infor-J mation needed to determine the consequences of such an event, using the guidance provided in the enclosed general information request. The enclosure represents our basic information requirements for plants now being constructed or operating. You should determine the appli-cability, for the Trojan, facility, of the items listed in the enclosure. If the results of your analyses indicate that changes in the design of structures, systems, or components are necessary to assure safe reactor shutdown in the event this postulated accident situation should occur, please provide information on your plans to revise the design of your facility to accot:nodate the postulated failures described above. Any design modifications proposed sbould include appropriate consideration of the guidelines and requests for information in the enclosure.
, Please inform us within 7 days after receipt of this letter when we (V,,) may expect to receive an amcndment with your analysis of this postu-lated accident situation fdr the Trojan facility, and a-description of any proposed modifications. Sixty copies of the amendment should be provided.
- A copy of the Commission's press announcement on this matter is also enclosed for your-information.
Sincerely, D3 A. Giambusso, Deputy Director l for Reactor Projects 1 Directorate of Licensing
Enclosures:
- 1. General Information Required
- 2. Press Release P-429 cc: See page 3 I l
O V . A-2
i Portland General Electric Company DEc19 1972 cc: Mr. H. H. Phillips 621 Southwest Alder Street Portland, Oregon 97205 lI l s O f I
.l A-3
\
j
/") .
k 4 (_ / Concral Information Required for Consideration ! of the Effects of a Piping System Breck Outside Containment The following is a general list of information required for AEC review of the effects of a piping nystem break autside containment, including the douhic ended rupture of the largest pipe in the main steam and feed-I I water systerm, and for AEC review of any proposed design changes that may be found necessary. Since piping layouts are substantially dif ferent from plant to plant, applicants and licensees should determine on 'an individual plant basis the applicability of each of the following items for inclusion in their submittals:
- 1. The systens (or portions of systems) for which protection against pipe whip is required should be identified. Protection from pipe whip need ;
s i i 1 not be provided if any of the following condicions will exist: V (a) Both of the following piping system conditions are met: (1) the service temperature is less than 200* F; and (2) the design pressure is 275 psig or less; or (h) The piping is physically separated (or isolated) from structures, systems, or components important to safety by protective barriers, , or res trained f rom whipping by plant design features , such as k l concrete encasement; or l j i j (c) Following a single break, the unrestrained pipe movement of either end of the ruptured pipe in any possible direction about a plastic , i hinge formed at the nearest pipe whip restraint cannot impact any structure, system, or component important to safety; or es l
\ ,,
A-4 _.m_
L
. Q (d) The internal energy level associated with the whipping pipe p
can be demonstrated to be insufficient to impair the' safety function of any structure, system, er component to r.n ! unacceptable level.
- 2. The criteria used to determine the design basis piping break locations in the piping systems should be equivalent to the following:
(a) ASME Section III Code Clags I pipir,g breaks should be postulated to occur at the following locations in each piping run or branch run:
/
/ (1) the terminal ends; (2) anyintermediatelocationsbekweenterminalendswhere the primary plus secondary stress intensities S . (circum-a ferential or longitudinal) derived on an elastica 11y 1
The internal fluid energy level associated with the pipe break reaction may take into account any line restrictions (e.g., flow limiter) between the pressure source and break location, and the effects of either single-ended or double-ended flow conditions, as applicable. The energy level in a whipping pipe may be considered as insufficient to rupture an impacted pipe of equal or greater nominal pipe size and equal or heavier wall thickness. - 2 Piping is a pressure retaining component consisting of straight or curved pipe and pipe fittings (e.g., elbows, tecs, and reducers). A piping run interconnects components such as pressure vessels, pumps, and rigidly fixed valves that may act to restrain pipe movement beyond that required for design thermal displacement. A branch run differs from a piping run only in that it originates at a piping intersection, as a branch of the main pipe run. O . A-5 1
- - _ _ _ _ _ . . _ . . __ ._. . -- . _ = _ -
O t l i
' calculated basis under the loadings associated with one -
i half safe shutdown earthquake and operational plant for ferritic steel,.and conditions exceedu 2.0 S
- i "or austenitic steel; 1 (3) 'lliate locations between terminal ends where the (S2 dative usage factor (U) derived from the piping fatigue analysis and based on all normal, upset, and testing plant conditions exceeds 0.1; and .
(4) at intertaediate locations in addition to those determined by (1) and (2) above,. selected on a reasonable basis as necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run or branch run. (b) ASME Section III Code Class 2 and 3 piping breaks should be postulated to occur at the following locations in each piping run or branch run: ,
-(1) the terminal ends; Operational plant conditions include normal reactor operation, upset j conditions (e.g., anticipated operational occurrences) and testing j
conditions. i S S ,is the design stress intensity as specified in Section III of the ASE Boiler and Pressure Vessel Code, " Nuclear Plant Components." i U is the cumulative usage factor as specified in Section III of the ASE Boiler and Pressure Vessel Code, "Huclear Power Plant Components." l 1
& A-6 l
I
u q ,
'~
(2) any intermediate locations between terminal' ends where either the circumferential or, longitudinal stresses' derived on an elastica 11y calculated basis under the loadings associated with seismic events and operational plant conditions exceed 0.9 (Sh+I) A I the expansion stresses exceed 0.8 Sg ; and (3) intermediate locations in addition to these determined by (2) above, selected on reasonable basi 3 as necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run or branch run.
- 3. The criteria used to determine the pipe break orientation at the break locations as specified under 2 above should be equivalent to the V) following:
(a) Longitudinal breaks in piping runs and branch runs, 4 inches nominal pipe size; and larger, and/or I Sh is the stress calculated by the rules of NC-3600 and ND-3600 for Class 2 and 3 components, respectively, of the ASME Code Section III Winter 1972 Addenda. S A is the allowable strees range for expansion stress calculated by the rules of NC-3600 of the ASME Code, Section III, or the USA Standard Code for Pressure Piping, ANSI B31.1.0-1967. 0 Longitudinal breaks are pirallel to the pipe axis and oriented at any l point around the pipe circumference. The break area is equal to the effectiva cross-sectional flow area upstream of the break location. Dynamic forces resulting from such breaks are assumed to cause lateral pipe movements in the di:ection normal to the pipe axis. O O A-7 I
- n. _
- n. y (b) Circumferential breaks in piping runs and branch runs' exceeding 1 inch nominal pipe size.
- 4. A summary should be provided of the dynamic analyses applicable to the design of Category I piping and associated supports which determine the resulting loadings as a result of a postulated pipe break including:
(a) The locations and number of design basis breaks on which the dynamic analyses are based. (b) Thepostulatedbaptureorientation,suchasacircumferential and/or longitudinal break (s), for each postulated design basis break location.
.(c) A description of the forcing functions used for the pipe whip dynamic analyses including the direction, rise time, magnitude, p
d duration and initial conditions that adequately represent the-jet stream dynamics and the system pressure difference. (d) Diagrams of mathematical models used for the dynamic cualysis. (e) A summary of the analynes which demonstrates that unrestrained . 1 motion of ruptured liues will not damage to an unacceptable degree, structure, systems, or components important to safety, such as the control room. l
- I 9Circumferential breaks are perpendicular to the pipe axis, and the break area, is equivalent to the internal cross-sectional area of the ruptured pipe. Dynamic forces resulting from such breaks are assumed to esparate the piping axially, and cause whipping in any direction normal te the pipe axis.
A-8
n, _ _ f'
- 5. A description should be provided of the measures, as applicable, to
. protect against pipe whip, blowdown jet and reactive forces including:
(a) Pipe restraint design to prevent pipe whip impact; (b) Protective provisions for structures, systems, and components l- required for safety against pipe whip and blowdown jet and reactive forces; (c) Separation of redundant features; (d) Provisions to separate physically piping and other components of redundant features; and 1 (e) A description of the typical pipe whip restraints and a summary ' of number and location of all restraints in each system.
- 6. The procedures that will be used to evaluate the structural adequacy p]
of Category I structures and to design new seismic Category I structures should be provided including: (a) The method of evaluating stresses, e.g., the working stress method and/or the ultimate strength method that will be used; (b) The aliowable design stresses and/or strains; and (c) The load factors and the load combinations.
- 7. The ' design loads, including the pressure and temperature transients, the dead, live and equipnent loads; and the pipe and equipment static, thermal, and dynamic reactions should be provided.
i i I O v 1
- 1 A-9 !
)
l
i 1
,~
%)
'8. Seismic Category I structural elements such as floors, interior i
valls, exterior walls, building penetrations and the buildings l as a whole should be analyzed for eventual reversal of loads due f I l to the postulated accident. f
- 9. If new openings are to be provided in existing structures, the capabilities of the modified structures to carry the design loads should be demonstrated.
- 10. Verification that failure of any structure, including nonseismic Category I structures, caused by the accident, will not cause failure of any other structure in a manner to adversely affect:
(a) Kitigation of the consequences of the accidents; and 13 (._ l (b) Capability to bring the unit (s) to a cold shutdown condition.
- 11. Verification that rupture of a pipe carrying high energy fluid will not directly or indirectly result in:
(a) Loss of redundancy in any portion of the protection system (as defined in IEEE-279), Class IE electric system (as defined in IEEE-308), engineered safety feature equipment, cable pene-trations, or their interconnecting cables required to mitigate the consequences of the steam line break accident and place the reactor (s) in a cold shutdown condition; or ['i 1._/ A-10
J I i
\~ / l (b) Loss of the ability to cope with accidents due to ruptures of pipes other than a steam line, such as the rupture of pipes 3 I
causing a steam or water leak too small to cause a reactor i accident but large enough to cause electrical failure.
- 12. Assurance should be provided that the control room will be habitable and its equipment functional after a steam line or feedwater line break or that the cap
- ability for shutdown and cooldown of the unit (s) will be available in another habitable area.
- 13. Environmental qualification should be demonstrated by test for that .
electrical equipment required to function in the steam-air environ- j ment resulting from a steam line or feedwater line creak. The in-ym i
) formation required for our review should include the following:
(a) Identification of all electrical equipment necessary to meet i requirements of 11 above. The time after the accident in which they are required to operate should be given. (b) The test conditions and the results of test data showing that the systems will perform their intended function in the environ-ment resulting from the postulated accident and time interval of the accident. Environmental conditions used for the tests should be selected from a conservative evaluation of accident conditions. ( I (c) The results of a study of steam.syste=s identifying locations where barriers will be required to prevent steam jet impingment from dis-abling a protection system. The design criteria for the barriers should be stated and the capability of the equipment to survive (/} within the pr otected environment should be described. A-ll
i 1 l %.
\J~
(d) An evaluation of the capability for safety related electrical 1 equipment in the control room to function in the environment l 1 that may exist following a pipe break accident should be ! provided, Environmental conditions used for the evaluation j i should be selected from conservative calculations of accident j conditions. 1 (e) An evaluation to assure that the onsite power distribution system and onsite r,ources (diesels and batteries) will remain l operable throughout the event.
- 14. Design diagrams and drawings of the steam and feedwater lines including branch lines showing the routing from containment to the turbine , building should be provided. The drawings should show
(,) m elevations and include the location relative to the piping runs of q safety related equipment including ventilation equipment, intakes, j i and ducts.
- 15. A discussion should be provided of the potential for flooding of safety related equipment in the event of failgrr of a feedwater line or anf 1
other line carrying high energy fluid. I i l
- 16. A description should be provided of the quality control and inspection j
\
programs that will be required or have been utilized for piping systems I outside containment. j
; 7. If leak detection equipment is to be used in the proposed modifications, a discussion of its capabilities should be provided. l1 l
I (~3 j U l A-12 ) i l l l I l
'V' 18.- A summary should be provided of the emergency procedures that would be followed siter a pipe break accident, including the automatic and manual operations required to place the reactor unit (s) in a
. cold shutdown condition. The estimated times following the accident for all equipment and personnel operational actions shonid be included in the procedure summary.
- 19. A description should be provided of the seismic and quality classi-fication of the high energy fluid piping systems including the steam and feedwater piping that run near structures, systems, or components important to safety.
i
- 20. A description should be provided of the assumptions, methods, and '
/% results of analyses, including steam generator blowdown, used to
\
calculate the pressure and temperature transients in compartments, pipe tunnels, intermediate buildings, and the turbine building following a pipe rupture in these areas. The equipment assumed to function in the analyses should be identified and the capability of systems required to function to meet a single active component failure should be described.
- 21. A description should be provided of the methods or analyses performed to demonstrate that there will be no adverse effects on the primary and/or secondary containment structures due to a pipe rupture outside these structures.
Q 1 A-13
4+ , *'< UNITCD STATES pc .n ..
.t #, Q # <*
ATOMIC ENERGY COMMISSION . s' I(W x); L 'l *
% ,Ei g e , 'f
'.' wmuma e on. n:c. w.c.
1:
,y- }o73 L
I' '.To Those' Listed Below ERRATA. SHEET FOR "CENERAL 10FORMATTON REQUIRED FOR CONSIDERATION OF' I. . Ti!E EFFECTS OF' A PIPING SYSTEM BREAX OUTSIDE CONTAINMENT" I. . Enclosed is the crrata'aheet for the modifications to the
" General Information Cuide For Consideration Of The Effects Of A Piping System Bronk Outside Containment" which wa's transmitted to-all applicants and licensecs.
} ;> t') W ..~
Robert L..Tedesco, Assistant Director' m for Containment Safety O Directorate of Licensing'
- /
Enclosure:
As stated above' 4 cc: 'E. C. Case-J. M. !!cndrie A. Ciambusso
'A/D's P/TR 8/C's P/TR j .-
A-13a Revision 1
. January 1974
i. 4 i l (
. - ERRATA SHEET FOR "CF.NERA1. INFORMATION REQUIRED F0j_t CONSIDERATION OF TiiE I CFFECTS OF A PIPING SYSTEM HREAX GUTSIDE COKYAINMENT" \
1 The following lists the changes that have evolved on our initial information l request: 1
- 1. Page 2, Item 2--Insert the following'in 2. to precede the existing first sentence:
i
" Design basis break locations should be selected in accordance with the following pipe whip protection criteria; however, where pipes carrying high energy fluid are routed in the vicinity of structures and systems necessary for safe shutdown of the nuclear plant, supplemental protection of those structures and systems shall be provided to cope with the environmental effects (including the effects of jet impingement) of a singic postulated open crack at the most adverse location (s) with re ard to those ;
enscncial structures and systems, the length of the crack being chosen not to exceed the critical crack size. The critical crack size is taken to be 1/2 the pipe diameter-in length and 1/2 the wall thickness in width." [ 2. Face 2, item 2(a)(2)--change nomenclature to read "any interr.ediate ( locations between terminal ends where the primary plus secondary stress intensities S,... r
=
- 3. Face 4, Item 2. (b)(2)--change 0.9 (Sh+8)t 0.8 (Sh+8)*
A A
- 4. Page 6, Item 7--Add " structural" to read "The structural design loads..."
- 5. Page 7,
- Item 11. (a)--Add " required" so as to read, " Loss of reouired redundancy..."
- 6. Pano 7,_ Item ll. (a)--Delete "the steam line break" and replace with "that" to read "...the consequences of that accident..."
- 7. P::nt;,8[ item 11. (h)--Repince (b) with the following: (b) "Envf ronmentally Jnduced '.*ailurcs caused by a le:4k or rupture of the pipe which would not of itself result in protective action but does disable protection functions. In this regard, a loss of redundancy is permitted but :
' loss of function is not permitted. For such situations plant shutdown is required."
O A-13b Revision 1 January 1974
e 1 u i F.rroto Slicet I*or "Cencral Information Itenuired For Consideration Of The
' ~~~
4 F.ffects Of Piping System _ Break Outside Containment"
. s-
.w
~
P . 8. Page A, item 13--Change wording in the first sentence to read
" Environmental qualification should be demonstrated by test for that electrical equipment required to function in the steam-air environment resulting from a high energy fluid line break."
- /
,'b O A-13 e Revision 1 January 1974 _x _-- _ _ _ . _ ._ _ _ . _ _ _ _ _ _ _ _ _ _ __- .____ ________ __ _ _ __.. _ _ ___ _ ____ _ _ ___
l 1 PonTLAxn GExsnAL ELEcTnic CoupAxy Ettf;tmc fBuscano
.' [ .
Jo EPM L WlLUAMS
' wer earseewy January 2,1973 Docket No. 50-344 U. S. Atomic Energy Commissian
! Directorate of Licensing l ATTN: Mr. A. Giambumo Deputy Director for
' Reactor Projects Wcshington, D.C. 20545
Dear Str:
We ere in receipt of your letter dekel Geeember 19,1772 which discusses the need for analysis of consequences of postulated pipe failures outside the containment and requests documentation of the results*o'f~these cnolyses.
/ We wish to inform you that these onelyses are currently being performed and thet we expect to provide the fermetion requested by you in the form of en cmendment to the Trojan FSAR by Mey 1,1973. l Sincerely,
. l/? .
J. L. Williams Vice President Engineering-Construction JLW:mo
)
(v^'\ A-14
m Pon1FLvND GENERAL ELucTaxc. COMPANY ELECTMic BUILDINO
, , , . PORTLAND.OncooN o7205' a \cSEPN L. WILLIAMS
, vsce omtement a
May 1,1973 Docket No. 50-344 11., S. Atomic Energy Commits!on - Dhectorate of Licensing ATTN: Mr. A. Giambusso Deputy Director for Reactor Projects Wo:hington, D. C. 20545
Dear Sir:
With reference to your letter dated December 19,1972 end our response dated Jcnucry 2,1973 related to cur.cnclysis of cer.:cquences of pc:tulcted pipe . (], failures outside the confeinment; these enolyses have preven to be more involved then originctly contemplated and we find our: elves unable to comply with our originally estimated completion date of May 1,1973 for submittel of this information. We feel reasonchly confident that we con provide the information requested in the form of on cmendment to the Trojan FSAR by June 15, 1973. and request your concurrence in the acceptability of this revised submitral
^
date. Sincerely,
, [M' ( E _
Joseph L. Williams Vice President Engineering-Construction JLW mo O A-15 =- ___ _ _ _ _ _ _
;~
t
\ ),--
BRANCH TECHNICAL POSITION - MEB NO. 1 . 1 POSTULATED FAILURE AND LEAKAGE LOCATIONS IN j FLUID SYSTEM PIPING OUTSIDE CONTAINMENT 4 i i i j l i i i 1
)
Revision 2 A-16 August 1975
o , 7/1/7.4' ' BRANCH TECitNICAL PCiLTION-NEB NO 'l' l-!ECl!ANICAL L: GINEERING BRANCil
- _DIREC10MIE OF LICENSING CRITERIA FOR.
_{ , POSTULATED FAILURE AND LEAKAGE LOCATIONS IN-FLUID SYSTC! PIPING OUTSIDE CONTAI::IE sf-
.The following criteria are,within the review responsibility.of the Mechanical Engineering' Branch with'the exception of I.A., II.A., II..D., 1 II.E and 1.a., 1.b., 1.c., 2.a and 2; .(3) of Appendix A. t
'. I. ' High-Energy Fluid SysicNPipinc A. Fluid Systems Separated from Essential Structurcs, Systems &
Components For the purpose of satisfying the separetion provisions of 1.a. of Appendix A a review of the piping layout and plant arrangement drawings should clearly show that the eff ects of pc:,tulated piping breaks at any location are isolated or physically remote from
^
essential structures, syster:s, and components. At the designer's option, break 1r,cacions as deterreined from I.C., I.D. , and I.E below may be selected for this purpose. a 1 i' 1
. B. Fluid System Piping Between Containment: Isolation Valves Breake need not be postulated-in those portions of piping identified in 2.C. (1) and 2.C. (2) of Appendix A provided they meet the requirements of AS:!E Code, Sect. ion III - Subarticle j' NE-ll10 and are designed to meet the following additional requirements:
See Glossary for definitions of italicized phrases. A-17 Revision 2 August 1975
.-_._x_- _ .-. - - _ _ . - . . - - - -
p
*1. The following design stress and fatigue limits should not be exceeded; n) 5 For AS!!E Code, Section III, Class 1 Piping (a) Maximum stress ranges should not exceed the following limits:
Ferritic steel 1 2.0S Austenitic steel 5,2.4Sm. (b) The maximum stress range between any two load sets (including the zero load set) sh..;uld be calculated by Eq. (10) in Par. NB-3653, A' 'E Code, Section lil, for upset plant conditions and an OBE event transient. If the calculated maximum stress range of Eq. (10) exceeds the limits of I.B.1(a) but is not greater than 3S , the limit of 1.B.1(c) should be met. () ^
,if the calculated maximum stress range of Eq. (10) exceeds 35,, the stress rar.ges calculated by both Eq. (12) and Eq. (13) should meet the limits .of I.B.1(a) and the limit of I.B.1(c).
(c') Cumulative usage factor 10.1, as required by I.B.1(b). For ASME Code, Section III, Class 2 Piping Maximum stress range as calculated by Eq. (9) and (10) dn Far NC-3652, ASME Code, Section III, considering upset plant ; l ccnditions (i.e., sustained loads, occasion'al loads, and thermal I j em - U Revision 2 1 A-18 ! August 1975 - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . . - . n.
l cxpansion) and an OBE cvent should not exceed (S +Sp2/ .
. s
- 2. "Ucided attachments, for pipe supports or other purposes, to l%
%)
these portions of piping should be avoided c cpt uhcr'c detailed
' stress analyses, or tests, are performed to demonstrate compliance vich.the limits of I.B.1. .
s
- 3. The number of piping circumferential and longitudinal velds .
and branch connections should be minimized. - i
- 4. The icngth of the piping run should be reduced to the minir.um . . ,
Icngth practical.
$. The design at points of pipe fixity (e.g. , pipe anchors or .
wclded connections at containment penetrations) shou'ld not require uciding directly to .hc outer surface of thu pipin;
,m ,
s (e.g., fluid integial forged pipe fittings may be used) c:: cept khcre detail'ed stress nalyscs are performed to demonstrate s.. 8 e ' compliance with the limits of I.D.1.. -
- 6. Geometric . discontinuities, such as at pipe-to-valve scetion transitions, at branch connections, and at changes in pipe vall thickness should be designed to minimize the discontinuity -
- r. tresses. -
. . e
- c. Fluid Sysicens Enclosed uithin Protective structures
- 1. Breaks in AS:!E Code, Section III, Class 2 and 3 piping should O -
+S
~
2/The limit of 0.S(1.2 S + S ) nay be used in lieu of (Sc
- h*
A-19 - Revision 2 i
' August 1975 ,
l I l 1 E-_____
.ba postulated at the following locations in each piping and
~
branch run (except those portions of fluid systen piping
,r 3 (j. ideritified in I.B.) within a protective structure containing essential systems and components and designed to satisiy the provisions of 1.b. or 1.c. of Appendix A:
- a. At terminal ends of the pressurized portions of the run if located within the protective structure.
- b. At intermediate locations selected by either of the following criteria:
(i) At each pipe fitting (e.g. , elbow, t ee, cross, and non-standard fitting) or, if the run contains no fittings, at one location at each extreme of the run (a terminal end, if located within the protective [%_ structure may substitute for one intermediate break). 3 (ii) At each location where the stresses / exceed (53 + S,)2/ but at not less than two separated locations' chosen on the basis of highest stress4 / . In the case of a straight i l pipe run without any pipe fittings or welded attach- ! ments and stresses below (Sh c + S )' " i"i "* f "" location chosen on the basis of highest stress. l
-3/Stresses associated with normal and upset plcnt conditions, and an OBE event as calculated by Eq. (9) and (10), Par. NC-3652 of the AS:1E Code,
- Section III, for Class 2 and 3 piping
-4/Two highest stress points; select second point at least 10% below the l
. highest stress.
f' ()) A-20 Revision 2 ' August 1975 1 g I
,3
, 2 2. . Breaks in non-nuclear t. tass piping should bo postulated at the following locations in each piping or branch run::
(
- a. At 'tcrminal ends of the pressurized po rtions -of the run if'
'l located with'in the protective structur1.
.b.- At each intermediate pipe fitting and wolded attachment.
D. ' Flu.id Systems Not Enclosed Within Protective Structures -
- f. Breaks in ASME Code, Section III,' Class 2 and 3 piping, should f
be postulated at the following locations in each piping and l branch run (except those portions of fluid system piping
- identified in I.B) outside but. routed alongside, above, or below a protective structure containing essential systers and components and designed to satisfy .the provisions of 1.b, or
- s. 1.c of Appendix A. .
Q.
- a. At terminal ends of pressurized portions of the run'if located adjacent to the protective structure.
- b. At intermediate locations selected by either of the .
'following criteria:
(1) At each pipe fitting (e.g., cibow, tee, cross, and non-standard fitting). l . (ii) At dach location where the stresses 1 exceed (Sh*S)~ c' but at not .tess than two separated locations chosen on the basis of highest stress4 / . In the case of a i d. A-21 Revision 2 August 1975 - m __n _u_ _.
lNy$ ' .,
-2 ,
- l31 A. g
. ;straidhtlpiperunwithoutanypipe.fittingsor-l Q. welded attachments'and stresses below (S3 + Sc ), a '
minimum of one location chosen on the basis of
-l
, highest stress.
O
- 2. . Breaks in non-nuclear class piping should be postulated at the'following locations.in each piping orl branch run:
- a. At terminal ends of pressurized portions of. thel run if located adjacent to the protective structure.
f
- b. At each intermediate pipe fitting and welded attachment.
- 11. Moderate-Ener;p Fluid Sustcm piping A. Fluid Systems Saparated irom Essential Struct :res, Systems' &
Components
-G For the purpose of satisfying the separation provisions of 1.a. of b
Appendix'A, a review of the' piping. layout'and plant arrangement drawings should clearly show that the' effects of through-wall . leakage cracks at any location are isolated or. physically ret:ote from essential structures, systems,. and components. B. Fluid System Piping Between Containment Isolation valves Breaks need not be postulated in those portions of piping identified in 2.c. of Appendix A provided they, meet the requirements of ASME Code, Section III - Subarticle NE-1110, and are designed such that l y the stresses do not exceed 0.5(S # S )5/ f r ASME Code, Section III, h c Class 2 piping.
'F
% -5/The limit 0.4(1.2 Sh * #A) may be used in lieu of 0.5(S, + S )*
h A-22 Revision 2 August 1975 L _ _ _ _ ____ _----- -_---.--__- -
C. Fluid Systems Within or Outside and Adjacent to Protective Structures j (,) v Through-wall Icakage cracks should be postulated in fluid 1 ( I system piping located within or outside and adjacent to i protective structures containing essential systcms and components and designed to satisfy the provisions of 1.b'. or 1.c. of Appendix A, except where exempted by II.B. II.D. 1 or in those portions of ASME Code, Section III, Class 2 or 3 j piping or non-nuclear piping where the stresses are less than j 0.5(S3 + Sy )b . The cracks should be postulated to occur individually at locations that result in the maximum effects from fluid spraying and flooding, and the consequent hazards or environmental conditions developed. D, hoderate-Energy Fluid Systems in Proxtetty to High-Energy Fluid f3 Systems b ! Cracks need not be postulated in moderate-energy fluid system piping located in an area in which a break in high-energy fluid system piping is postulated, provided such cracks would not result in more limiting environmental conditions than the high-energy piping break. Where a postulated leakage crack in the moderate-energy fluid system piping results in more limiting environmental conditions than the break in proximate high-energN fluid systcm piping, the provisions of II.C should be applied. C. Fluid Sysrcmc Qualifying as High-Energy or l oderate-Energy Syntemc Through-vall leakage cracks instead of breaks may be postulated g V A-23 Revision 2 August 1975
[ 's, ., { ['
.q
' ' .g
,i in the piping of thosa fluid systems that qualify as high energy 3 7
fluid systems for only'short operational periods 6/' but qualify; t 4 as moderate-energy fluid systems for the major operational l period. III. Tvpe of' Breaks and Leakage Cracks in Fluid Syste-2 Pioing. [ A. Circumfcrential Pipe Breaks x-The'following circumferential breaks should be. postulated in high-energy fluid system piping at the locations specified in , Section 1 above:
- 1. Circumferential breaks should be postulated in fluid system piping and branch runs, exceeding a nominal pipe size of 1 inch, except that, if the maximum stress range in'the circumferential direction is at least twice that in the axial direction, only a longitudinal break need be postulated. Instrument lines, one L inch and-less nominal pipe size for tubing should meet the provisions of Regulatory Guide 1.11.
- 2. Where break locations are selected at pipe fittings without the benefit of stress calculations, breaks should be postulated at each pipe-to-fitting veld. If detailed stress analyses i
-6/An operational period is considered "short" if the fraction of time that the system operates within the pressure-temperature conditions specified for higII-c':ergy 'f;uid sys:ces is less than 2 percent of the time that ~
the system operates as a calcrc:.: energy ; aid eya:cm (e.g. , systems such as the reactor decay heat removal systems qualif y as isd.' rate-energy fluid sysrc s; however, systems such as auxiliary feedwater systems operated during Pt?R reactor startup, hot standby, or shutdown
. qualify as higli-energy f;uid sys: cms).
O A-24 Revision 2 : i
. August 1975
. a eque a e eggyp
_m_--.u__m_m ..___,._________,_____,_m__,___,,____y__ _ _ _ _ _ _ _
(' l
- 1. Longitudinal breaN in fluid system piping and branch rund chould be postulated in nominal *p'ipe sizes 4-inch and larger, l.n) ,
except that, if the maximum stress ranc.c in the axial direction is at Icast twice that in the circumferential direction, only a circumferential break need be postulated.
~
- 2. ' Longitudinal breaks need not be postulated at terminal ends if the piping at the terminal ends c*ontains no longitudinal pipe '
velds and major geometric discontinuities at the circumferential veld joints' of the tenninal ends are designed to minimize dis-
., continuity stresses.- -
, U 3. Longitudinsi breaks should be assumed to result in an axial
.. oplit without' pipe severance. Splits should bc located (but not concurrently) 5t two ulametrically uppvoed puir.ts on the
,-s) ,
\x / -
, piping circumf erence such tha t a jet reaction causing out-of-
- . , , . planc bending of the piping configuration results.
- 4. The dynamic force of the fluid . jct discharge should be based on a ci,rcular or elliptical (20 x 1/2D)'- break arca equal to l
- u. . ;
.the ef f ective cross-sectional ficw area of the pipe at the
. , breah location and on a calculated fluid pressure r.iodified by 5 . .
} ,
. an analytically or experimentally determined thrust coefficient )
as deterr. tined for a circumf erential break at the same location. flou limiters, positive pump-co'ntrolled flow,
~
Line restrictions, i and 'the absence of energy reservoirs voy be taken into account,
,,-_ as applicabic, in the reduction of jet discharge. .
(~- . A-25 Revision.2
--. .bPSVGt..].91L ~
- - - - - = - - - - - - - - - - - - - - - - - - -
< < g-y a
(e.g. ,: finite element analyses): or tests are performed; ths : maximum stressed location in the fitting may be'sclected . i+
.instead of th'e pipe-to-fitting weld. .
1 I
- 3. Circumferential breaks should be assumed to result'in pipe severance and separation amounting to a one-diameter lateral- ,[
displacement of: the ruptured piping sections unless physically j limited by piping-restraints, structural menbers, or piping. :{
, stiffness as may be demonstrated by inelastic limit analysis (e.g., a' plastic, hinge in the piping'is not developed under
-loading).
1 f
- 4. The dynamic force of the jet discharge at the break location should be based on the effective cross-sectional flow area
' of' the pipe and on a calculated fluid pressure as modified
[% ' by an analytically or experimentally determined' thrust coefficient. Limited pipe displacement at the break location -
'line restrictions, flow limiters, positive pump-controlled .
l flow, and the absence of energy reservoirs may be taken into q LL account, as applicable .in the reduction of jet discharge. -l
]
1
- 5. Pipe whipping should be assumed to occur in the plane defined by the piping geometry and configuration, and to cause pipe I
movement in the direction of the jet reaction. ! l: B. Longitudinal Pipe Breaks The following longitudinal breaks should be postulated in high-energy fluid system piping at the locations of each circumferential
. break specified in III.A.:
A-26 Revision 2 August 1975 1
.l, j ' - l
~'
, , (
g !
~,5. Piping movenant should be assumed to occur in the direction of th'e jet-reaction unless limited by structural members, piping -
i! D)J restraints, or piping stiffness as demonstrated by inelastic
' limit analysis. I C. Through-Wall Leakage Cracks s
The following through-wall' leakage cracks should be postulated in j moderate-energy fluid' system piping at the' locations specified in Section 11 above:
- 1. Cracks,should be postulated in moderate-energy fluid system !
piping and branch runs exceeding a nominal pipe size of 1 inch. .l
- 2. Fluid flow from a crack should be based on a circular opening of area equal to that of a rectangle one-half pipe-diameter in. l
( length and one-half-pipe wall thickness in width.
- 3. The flow from the crack'should be assumed to result in an environment that wets all unprotected components within the compartment, with consequent flooding in the compartment and communicating compartments. Flooding effects may be determined on the basis of a conservatively-estimated time period required to effect corrective actions, g
't . l b -
A-27 Revision 2 August 1975 %- -----------_.______-_m.o__ - - - - . _ - _ -
I APPENDIX A
,m PLANT ARRANGDil3T CRITERIA AND SELECTED PIPING DESIGN FEATURES
).
d
- 1. Plant Arrangement ,
I Protection of essential structures, systems, and components against postulated piping failures in high or moderate energy fluid systems that operate during normal plant conditions and that are located out- l side of containment should be provided by one of the following plant arrangement considerations:
- a. . Plant arrangements should separate fluid system pfping from essential structures, systems, and components. Separation should be achieved by plant physical layorts that provide sufficient distances, betueen essential structures, systems, and components and fluid system piping such that-the effects of any postulated
'N' piping failure therein (i.e., pipe whip, jet impingement, and the environmental conditions resulting from the escape of contained fluids as appropriate to high or modcrate-energy fluid system piping) cannot impair the integrity or operability of essential structures, systems, and components.
- b. Fluid system piping or portions thereof not satisfying the provisions of 1.a. above should e enclosed within structures or compartments designed to protectfncarby essential structures, systems, and coxsoncnts. Alternatively, essential systems and 1
O A-28 Revision 2 August 1975 l l L
_ - _ - - _ - - --- -- -= .--- q
-i components may~be enclossd withis, structures or compartments In (
kk J j% designed to withstand the effects of postulated piping' failures j j
\
l Jin nearby fluid systems.
,1 j
- c. Plant arranger.ents or system features that do _not satisfy the provisions oi either 1.a. or 1.b. above should be limited to is
-those for which the above provisions are impractical. Such
,i
- cases may arise, for example, (1) at interconnections between d fluid systems and essential systems and components, or (2) in i
fluid' systems having dual functions (i.e.,-required to operate during nomal plant conditions as well as to shut down the reactor). 1
;In such cases, redundant design features, ' separated or otherwisc ,
protected from effects of postulated piping failures, or additional protection.should be provided so that reactor shutdown is assured Ib in the event of'a failure in the interconnecting piping <of1(1), l j V i or in the dual fun'etion piping of .(2). Additional protection may be provided by restraints and barriers or by designing or y testing essential systems and components to withstand the effects associated uith postulated piping failures. 1
.2. Design Features 4
- a. - Ecsential systems and cor:ponents sbould be designed to meet the 1 seismic design requirements od Regulatory Guide 1.29.
- b. Protective structures or cc part:ents, fluid system piping restraints, and other protective ceasures should'be designed in '
accordance with the following: f. P A-29 Revision 2 f August 1975 ( l a _i_ _ -_ l
r V'< x ,
-(1) Protective structures or compartments needed.to implement ,
1.b. or 1.c. above shoul'd be ' designed to Scismic Category I eO: . requirements.. The ef fect's of a postulated piping failure
,{j (i.e. , pipn whip, jet impingement, pressurization of compart-ment, water-spray, and flooding, as appropriate) in combination with loadings associated with the Saf e Shutdown Earthquake and normal' operation should be used for the design of required protective structures. Piping restraints, if used, may be taken into account to limit effects of the postulated piping failure.
~
-(2) Eigh-energy fluid system piping restraints and protective measures.should be designed such that the effects of a.
postulated break- in.one pipe cannot, in turn, rupture other nearby pipes or components which could result in 3 ' unacceptable offsite consequences or in loss of capability of essential systems and components to initiate, actuate, and comp 1'ete actions required for reactor shutdown.
. c. Fluid system piping between containment isolation valves should meet the following design provisions:
-1/
- In the design of piping restraint, an unrestrained whipping pipe should be considered capable of (a) rupturing impacted pipes of smaller nominal pipe sizes and (b) developing a through-wall leakage crack in larger nominal pipe sizes with thinner wall thicknesses except where experimental or analytical data for specific impact energies demonstrate the' capability to withstand the impact without failure.
I "O i 1 i A-30 Revision 2 ] August 1975 l i
)h;,
J ' mu (1) Portions of fluid'systch piping between isolation valves
~
. .of single. barrier containment structures (including any
,,y n-
!M rigid connection to the containment penetration)Lthat connect, on a. continuous or intermittentJbasis to the reactor. coolant pressure boundary.or the steam and feedwater systems of..PWR
. plants should be designed to the stress limits specified.in I.B. or II.B. of this document.
These' portions of high-energy fluid. system' piping should be provided with pipe whip restraints - (i.e., capable of resisting
, bending and torsional moments) located reasoncbly close to the containment isolation valves.I The restraints should be designed to withstand the loadings resulting from a postu!ated piping.
~
failure beyond these portions of piping so.that .ncither isolation:
# valve operability nor the leaktight integrity of'the containment
) .
~ '
will be impaired. Termina!. ends of'the piping runs outside containment should be considered to originate at'the pipe whip restraint locations' v outside containment. Where containment isolation valves are not required inside containment, those portions of the fluid system piping extending from the outside isolation valve to either the rigid pipe connection to the containment penetration or the first pipe ; 1 l . i A-31 Revision 2 August 1975 m _-_m__. __m__.____...-_m-_. .. __
=_ _ - - _ . - - - - --- - --
l structure or guard. pipe, a i full flow arca break should be assumed-in that portion of piping.within the enclosing- .1 structure or guard pipe, f 1 l (3) For those- portions of_ fluid sys cm piping identified in' 2.c.'(1) and 2.c. (2). above, the extent of inservice examination N conducted as specified in Division 1 of Section XI of the ASME
' Code during each inspection interval should be increased to , l provide volumetric examination of 100 percent of the circum- -
4
. j-.. .
ferential;and longitudinal weld' joints in piping identified in Section III.A.1. and Section III.B.1. of this document.. The
. j' areas subject to examination should comply with the require-ments of.the following categories as specified'in Section XI of the~ASME Code:
(a) ASME Class 1 piping welds, Examination Category B-J in
-Table IWB-2500.
(b) ASME Class 2 piping welds, Examination Category C-F and, C-G in Table IWC-2500. t c l J A-33 Revision 2 August 1975 _n_________ _ _ __ _ _ _ _ _ _ - . _ _ _ _ .a
I CLOr.S ARY_ p ' Structures, systems,. Esscntini Str r turcs, S:sstenc, av.d Consoncnts.
' and' components required for reactor shutdown without off-site power.or
.to mitigate the consequences of a postulated piping failure in fluid system piping that results in trip of the turbine-generator or the-reactor. protection' system.
Fluid Suctcms. High and moderate energy fluid systems that are subject to the postulation of piping failures against which protection of essential structures, systems, and components is needed. Riah-Enercu fluid S:.estens. Fluid systems that, during normal plant-conditions, are either'in operation or maintained. pressurized under-conditions where either or both of the following are met: c.
- a. maximum operating temperature exceeds 200'F, or
- b. maximum operating pressure exceeds 275 psig.
Noderaie-Enercu Fluid Esstens. Fluid systems that, during nornal plant conditions; are either in operation or maintained pressurized (above 2 atmospheric pressure) under conditions where both of the following are met:
- a. maximum operating temperature is 200*F or less, and
- b. maximum operating pressure is 275 psig or less.
- p k
A-34 Revision 2 August 1975
---- _ __ -_ _ = __-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
#om:I Plant ConRtions. Plant operating. conditions during reactor
.startup, operation at power, hot standby, or reactor cooldown to cold
' shutdown condition.
Uosat Plant Conditions. Plant ' operating conditions during system I transients that'may occur with moderate frequency during plant. service life and arc anticipated. operational occurrences, but not during system testing. l l' Postulated Piring Failto'es. Longitudinal.and circumferential breaks L'- ..in high-energy fluid system piping and through-wall leakage cracks in moderate-energy fluid systen. piping postulated according to the provisions of this document. Sh ' #c' #"d #A. Allowable stresses at maximum (hot) temperature, at minimum (cold) temperature, and allowable stress range for thermal expansion respectively, as defined in Article NC-3600 of the ASME Code, Section III. Sg Design stress intensity as defined in Article NB-3600 of the ASME Code, Section III. Sinale Active Ccconent Failure. Malfunction or loss of function of a component of electrical or fluid systems. The failure of an active component of a fluid system is considered to be a loss of component function as a result of mechanical, hydraulic, pneumatic, or electrical malfunction, but not the loss'of component structural integrity. The direct consequences of a single active cer:ponent failure are considered ] to be part of the single failure. O A-35 Revision 2 l August 197S l
r-- - -
- Temics Tds. Extremetics of piphug runs that connect to structures, components (e.g. , vessels, pumps, ' valves), or pipe anchors that act as
!'v) - - rigid constraints to piping thermal expansion. A branch connection to a main piping run is a terminal end of the branch run. .~%( )\
I m ( ) LJ A-36 Revision 2 August 1975
J
.i O
i i l j APPENDIX B EFFECTS OF A PIPING SYSTEM BREAK OUTSIDE THE CONTAINMENT O 1 5 1 I i 1 i O l _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ )
7 j - c: ,
.' r f
xi TABLE OF' CONTENTS 1.0 GENERAL B 2.0 LOADING B-4 3.0- ANALYTICAL TECHNIQUES B-7 3.1 General. B-7 3 3.2 Impulsive Loading ~ B-7 "i ' 3 '. 3 Impact.with Impulsive Load .B-9 3.4 Resistance-Displacement Response B-13 4.0. CRITERIA B-18' 4.1 Resistance B-18 4.2- Ductility and Maximum Displacement B-19
- 5.0 STRUCTURAL ADEQUACY' B-28 5.1 General B-28 5.2 Examples B-29 l
B-1
a .
'~
1.0. GENERAL ' ! The consequences of a high energy piping system rupture out-side of the containment is the object of the following docu-- ment. The pipe rupture-can be classified as guillotine, slot, This rupture may cause-a jet impingement or a pipe
~
or crack. whip on structures whose integrity is important to a safe'- shutdown of the plant. The possible points of pipe rupture have been selected in accordance with the AEC criteria' set forth in Appendix.A. . A. simplified analysis and design approach is-chosen based on-conservative assumptions in material and structural behavior. ( The basic reference in determining the structural repsonse from the effects of impact and impulse loadings resulting from < pipe failure is BN-TOP-2, Revision 1, " Design for Pipe Break Effects", C September, 1973. The material presented herein supplements BN-TOP-2 in the following areas: l
- 1) Expands the material given in BN-TOP-2 and demonstrates l'
the analytical techniques to cover plastic impact.
- 2) Gives the equations necessary to evaluate resistance and yield displacement for several typical systems. 4 L
- 3) Lists ductility ratios for various materials and structural systems.
- 4) Gives examples illustrating the analytical techniques. I i
l O E~2 Revision 1 January 1974
- --___-__ ___--_-______x__--____ -__-_- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _
p. i i(3'L f In general, any whipping pipe is a low velocity blunt mis- l i sile,'and penetration and spalling.are not appreciable for j the wall thicknesses of interest. I
)
l 1 4 1 A b l
.f l
1 i B-3 l I L- - --__ - l
2.0 LOADING l ; The type of loading is a function of the mode of failure, direction v of flow and the location of the rupture. Some examples are shown in Figure 2-1. For a jet, a 10' half angle is used to simulate dispersion. The loading combinations to be considered due to a pipe rupture are given in Table 2-1, and correspond to those in m O AEC Document B, " Structural Design Criteria For Evaluating The Effects Of liigh Energy Pipe Breaks On Cateogry I Structures Outside The Containment". When pipes under pressure fail, a fluid jet is created and a fcrce is exerted on the piping. The force exerted on the piping system has a rapid buildup and decreases to zero as l the system blows down. This force-time relationship is illus-trated in Figure 2-2. The dynamic solution for a system sub- l O jected to this rupture force may be simplified if the force is f I conservatively assumed to build up instantaneously and then l remain constant as shown in Figure 2-2. The assumption that the force remains constant with time also means that it l remains a constant with respect to displacement,and simpli-fied analytical techniques can be used which are shown in 1 Section 3.0 l l l i I f~ l
\
B-4 Revision 1 January 1974
t \_/ . T .
/
/ F /
hF
\
\
\
\ '1 ht~ (
'i
/j/ 3
') I /
F F , _ p
/
'/
/ } / y i/
/ / [/
/
o C(.'
/
' /[
/
/ '
m Q' i Slot Failure Guillotine Failure Pipe Whip
/
10 [/
- /
Pipe % (( Wall
/
x f/ r Jet Impingement Figure 2-1 Pipe Rupture Examples B-5 k W ..a-------.----.--w_.---- - - . - _ - _ _ - - - - - - - - _ -------_-_---_--_-_.--._--__..--.--_---._-_-._.-__..--__--._____--.___.___-__-__._-_-_-,-_____.__---__.-_--_--_-__.___.-_.-a
,< ~(. i i I N,_ .)
1 Rupture Force Time i Actual Force Time History p Rupture Force Time Simplified Conservative Force Time History ) I i l Figure 2-2 Rupture Force vs Time Relationship i l i 1
?
( B-6 l _ _ ~ _ _
i Sheet 1 of 3- ! TABLE 2-1 j []
\./
LOAD COMBINATIONS' The following load combinations should be satisfied: 1 l A) For concrete structures:
- 1) U = D + L + T, + R + 1.5 P, j
- 2) U = D + L + T, + R, + 1. 25 P, + 1. 0 l
(Y r +Y j + Y m) + 1.25 Fego
- 3) U = D + L + T, + R, + 1. 0 P, + 1. 0 (Y r +Y j + Y m) + 1.0 Teqs p B) For steel structures:
4 If elastic working stress design methods are used:
- 1) 1.5 S = D + L + T, + R, + P,
- 2) 1.4 S = D + L + T, + R, + P, + 1.0 (Y +Y +Y + Fego r m
- 3) 1.5 S = D + L + T, + R, + P (Y) + Yr + Y m) + Feqs If plastic design methods are used:
- 1) .90 Y = D + L + T, + R, + 1.5 P,
v%f
..3 s ;
.c; , t .. ,
h- 1
.. t j.-
i.I
- f. jtl:1 i '
Sheet 2 of 3 _ TABLE 2-1 %K,
.i M
fi - '
- 2) . .90 Y = D + L + T + R'+ 1.25 P + 1.0
.a a- a
, a
~ (Y ' + Y : + Y ) + 1.25 Feqo yf.
j r- m
- 3) .90 Y = D + La +T a + R- a
+ 1.0 P + 1.0 a
(Y .'+ Y + Y-)
-- j ' m
+ 1.0 Fegs c .
r Where:
'D'---- De'ad loads and their related monents and forces including any permanent equipment, loads, and hydrostatic pressures, if any.
L'-- . Live -loads, present during the pipe rupture event, and their:related monents and forces. lD ( .. P, --- Design pressure load within .or across a compartment and/or
' building, generated by a postulated _ break, and including
.the dynamic. effects due to the pressure time history.
T, ---- Thermal effects due to' thermal conditicns ge ersted by a postulated break and ~ including T,. R, --- Pipe reactions under thermal conditions generated by a postulated break and including R . Y --- L ad on the structure generated by the reaction on a r ruptured high-energy pipe during the postulated event. y) ---- Load on a structure generated by the jet impingement from a ruptured high-energy pipe during the postulated event. ls .
.\
m en- .
l3 L
,'_ ,o Sheet 3 of 3 TABLE 2-1 Yg - The energy resulting from the impact of ruptured high-
. energy pipe on a structure or pipe restraint during the' postulated event.
l Fego -- Loads generated.by thel Operating Basis Earthquake or, if i an OBE is not specified, loads. Fegs -- Loads generated by the Safe Shutdown Earthquake. ,
, ' U ---- For concrete structuus, U is the section strength .
required ta resist design loads and based on methods
' described in ACI 318-71.
S'----- For structural steel,-S'is the required section strength
~
based on the elastic design methods and the allowable
' stresses defined in Part 1 of the AISC " Specification
. for the Design, Fabrication and Erection of Structural l
Steel for Buildings," February 12, 1969. Y ----- For structural steel, Y is section strength required to resist design loads and based on plastic design methods
- d. scribed in Part 2 of AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Build-ings," February 12, 1969.
i m I
A
.t ,
, .N O 3.0' ANALYTICAL TECHNIQUE .l 3.1 General To obtain a solution for the actual complex system, the si.ructural system is converted to an equivalent one 'or two t degrees-of-freedom system. Of particular interest are the the responses to an impulse by a fluid jet and an impulse-impact from a whipping pipe. These structural responses will be treated in subsequent sections. L 3.2 Impulsive Loading To illustrate the concept of impulsive' loading, con-sider the simple system shown in Figure 3-1. Tne work
- done by the dynamic load is Fx and the nergy absorbed by the equivalent system is equal to the area under the resistance-displacement curve given by x
R(x-h (B-2) where x y is the displacement at yield. Equating the work done to the energy absorbed leads to the following
~
general solution'for a constant magnitude impulsive load acting on an elasto-plastic system: and Fx=Rfxx (B-3) Fx R= x (B-4) x-y O B-7 ,
p , i s ': 1 1 Defining.the ductility ratio by p = (B-5) y the solution can be expressed as
, t
= "
R y .(B-6) u7 As an example, if a system is allowed to displace ~to three times yield, i.e., p = 3, then R = 1.2F This requires that the resistance or capacity must be at least 20% higher than the dynamic load F to stop the
~&. .
system within a displacement of three times yield. For a brittle failure within the elastic range, the ductility ratio is taken to be p=1 and ' R = 2F Therefore, if the response is to remain elastic, the resistance must be twice che applied load since the solution assumed the load was suddenly applied. The preceding solution for an impulsive load is appli-cable to situations where the fluid jet is impinging on a structural element'or the rupture force is acting O B-8
Nl i i, ! r oon a piping system which is in contact with its { supports or restraints. 3.3 Impact Combined with Impulsive. Load An impact problem is here defined to be the collision of two bodies with an energy transfer during impact. A common situat' ion associated with this. type of loading occurs when the ruptured. piping system travels across a gap to strike a structural element. An equivalent simplified system is shown in Figure'3-2.
~
If the-effective masses are not easily' evaluated or a-very conservative solution is desired, then the following
) equation may be written which assumes a full transfer of energy'to the restraint.after the pipe system crosses the gap.
x 7 x -x 1 Fx = Rp x- f+RrI X ~X g 2 (B"7) W w , i w k Work Energy Energy Absorbed Done Absorbed by Restraint by Pipe The only unknown is x and its solution will yield the pipe displacement and the restraint displacement is given by (x - x g) . I
, - I B-9 i
;/R :-
( 4
; [ The preceding solution is valid when x >x 9 P x >x P
x >x consider next-the case where the pipe remains' elastic'in the previous problem, then the pipe resistance; curve in Figure 3 2 is changed tot e g x, _ _ _ _ _ _ _ _ K m , w I a ; l- i j . i x x x ("% Displacement
.Q) and the solution is:
2 x -x - Fx = KpT+R (x-xg) - 2 ( -8) for the limits - (x > x , x < x , x > x )
- g p r Since there are many different solutions depending on the displacement relationships, it is important to understand the analysis technique and handle each problem on a case-by-L case' basis, if necessary-ll B-10 I
m . , + < t '/ I. 1 i' "'I 3
' ! - Q,, ;.:.
[ .. The previous solutions assumed'a full energy transfer. upon impact which is'very conservative. .A more realistic-solution involves'a fully plastic. impact which assumes that a portion of the mass.of the pipe remains'inLcon-
..,, tact with the restraint mass afterfimpact., This' fully l plastic impact. concept is valid.since large. forces-4 remain on the piping system after impact and since the-s' pipe. deforms locally.
The concept of plastic impact utilizing the conservation of momentum can be. illustrated-by considering'the two masses shown'below:.
.M m .M 2
T o y 2 Before Impact After Impact The conservation of momentum gives: mv = (m + M)V and V= 3),v The initial kinetic energy before impact is: 1 2 E=ym 1 and after impact is: B-ll i _________1___-_._____._ .__ __
' , , o
'~h !
_ya s
.k 2
Y
. KE ' = h(M + mMm )
4 KE'
, = f mv Mm ." ' Mm j Therefore, the kinetic energy after impact is
[ times the initial kinetic energy. The energy which-appears lost has been absorbed in the deformation of the masses during~ impact. Applying the concept of plastic impact to the system shown in Figure 3-2, gives
~ ~
m - x- ( 3 m m 9 P( 9
~
+( ~
9} ~ P I* ~ 7 ) p r . d ~, J W M External Energy Work Done Energy Absorbed Transfer During after^ By Pipe After Plastic Impact Impact Impact x =x
-+ R *~* ~
r g g 2 M mn A Energy Absorbed By Restraint Again x is the only unknown and its solution represents L. , the pipe displacement while the restraint displacement is given by x - x g. This solution is valid when: i X # X p
'b x>X x >x 9 P B-12
.i; I ,
IL g: In summary, the force-time relationship was conserva-tively simplified-so that the force is a constant with-respect to displacement. This enables a solution by
' equating work done to energy absorbed and eliminates the necessity of solving an equation of motion for the' system which would yield the same results.
3.4 Resistance-Displacement Relation The previous section showed the necessity.of having a resistance versps displacement curve for each resisting element in the system. As an illustration of the deter-mination of a resistance-displacement curve, consider the
' (* ' (
fully fixed beam shown in Figure 3-3.. For the load at 4 the center of the span, applying the virtual work princi-ple to the lower bound limit analysis mechanism gives the following relationship between the applied load P and the yield moment M u P6 = 4M e since 6= eh then p, "u L Since the failure mechanism is associated with a fully hinged condition, the load P will be equivalent to the B-13 - _ - - _ _ - - - _ _ - - - b
y {
. ; \
o .. i L
~-
w ' .; q the? resistance or capacity at yield: R' = P = -8M"
]
L j For a= steel beam, Mu = 0.9f ypS where f .is the yield stress, S is the plastic section' y p modulus, and O'.9 is the capacity reduction factor. ~ For !
- a. reinforced concrete member, M w ill be based on the u
strength principle of ACI-318-71. This will be covered in more detail in'Section 4. Since.R is the load at yielding, its application to the fully fixed elastic beam equation will give a miriimum estimate of the' displacement at yield: RL y " 172 EI where'E is the modulus of elasticity and I is the section moment of inertia. With R and xy determined, the elasto-plastic curve can be constructed as shown'in Figure'3-3. o I B-14 i _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ __ i
y r .: t
...),
27 . (..?
~
4 l
. q/ -
O / m l, -- K l f. , F ---+ " I WA%v ' Elasto- / Plastic' , Spring. p. fi 1 V e R "-- I
=
g l .;
.N l l-T I I j E l l. I 1 X x /
y Displacement R = Resistance Force xy = Yield Displacement j F- = Jet Force Figure 3-1 Impulsive System Example i f O B-15 , l 1 _ _ _ _ _ = - _ - _ _ _ _ - _ - _ _ _ _ - _ - _ _ - - - _ _ - -_-_ _ ___- _ _ _ _ __ _ ____-._-________-___--_______-_L
r.: i O R = Resistance
'K- = Stiffness %F "p m m = Mass
$p- ,
xg . r Y r x-= Displacement Piping - - F = ' Jet Force X Restraint System'. System l NOTE: All' springs are elasto-plastic. ! l
'e o PIPE c
4
- y- R --
p
'( m E.
i I x g *
*P 4, 1
Displacement ! I 4 RESTRAINT eR o e r , e I e .t to -
=rd tD / g e
* . _.._t i .
X X X g r i Displacement t-Figure 3-2 Impact + Impulsive System Example B-16 i
t l l
' ~.s l n
/
'J ~ P f
4 -
/
$ 'I _.
s
;= = = ... 21 1 1 7 7 i P
"u a .e 6 e7.-
u 20 J 4 M P u P ! I 7 iv) l, 1 i e
>U
$ R .-- 7 -- ,
a i i m w i , m i i g
+
CG t i . i _ . . . . _ _ . . _ . ..e__. . ._ ___ xy xu " "*y Displacement 1 Figure 3-3 Resistance-Displacement Fxample ] 1 l f~
\ f N)
B-17 i l
p pw 1o l 4.0 CRIT 2RIA 4.1 Resistance , The resistance of any structural component is based on its minimum' strength. As an example, for a 1' concrete . slab subjected to an impulse loading, yield line theory'is applied, and slab strength is evaluated based on its flexural, shear and punchout capacity. The lowest of the capacities determines its resistance. The capacity of the restraint elements is based on the dynamic material strength which is defined by: allowable dynamic stress = allowable static stress () times DIF, where DIF is the dynamic increase factor given in i Table'4-1. The following codes have been used where applicable to I determine the cross-sectional strengths of various members: a) Reinforced and Prestressed Concrete
'" Building Code Requirements for Reinforced Concrete (ACI-318-71) American concrete Institute" b) Structural Steel l "AISC 7th Edition 1970 - Manual of Steel Construction" 1
O B-18 l 1 i
- - _ _ - _ - - _ _ _ - _ _ _ - _ - _ _ 1
t - 4 l y $. 1,.M c) Steel Pressure Vessels
~
"ASME Boiler and Pressure Vessel Code Section III -
- 1971, Nuclear Power Plant Components" d) Piping Systems "USAS-B31.7 - Nuclear Power Piping, 1969" s',
a
, Tables 4-2 and 4-3 are provided to determine the resistance capacity based on cross-sectional moment capacity and member dimensions.
i I 4.2 Ductility, Yield and Maximum Displacement Tables 4-2 and 4-3 provide yield displacement approxi-mations and the resistance and displacement values.are i 1 only applicable to systems whose flexural strength defines minimum capacity. In evaluating the yield displacement with the usual elastic analysis, an adjustment must be made to the moment of inertia to account for cracking of the. concrete. The empirical relation used follows Section 5-8 of " Structures to Resist the Effects of Accidental Explosions", TM5-1300, June, 1969. The average moment of ine'.tia I, used in calculating deflection is: I I +I U I 3
=
2 (B-10) l' l B-19 l c _r _ __ _ _ _ _ _ _ _ _ ____ _ _
f"
!U where !
I -= Moment of inertia of gross concrete cross-section g of thickness t about its centroid (neglecting steel
, areas) =ht !
I c-
= Moment of inertia of the cracked concrete section 3
of width'b = Fd . F is given in Figure 4-1. This value of I, has been used in the displacement equations given in Tables 4-2 and 4-3 for all reinforced ; concrete members. In" situations where the energy input is essentially unlimited, such as a jet (impingement) loading aeting on a wall, the allowable ductility value is limited to 3. This provides a resistance of at least 1.2F based on R= where p = 3. This limit is only g applied.to the primary resisting component.such as the i
~
I wall. Any secondary resisting components such as the cracked pipe may exceed the limit of 3. In a situation where pipe restraints are provided to limit pipe motion, the restraints are considered as l the primary load resisting components,and the piping is the secondary component. l . .i B-20 l w ._- _--_--
p5a-l 1, A.;
~
( ) When the loading consists of both impact and impulsive loading, such as a pipe whip situation, the maximum
' allowable displacement is based on the ductility 1,
ratio given in Table 4-4. In this case, the minimum resis-tance is also limited.to 1.2F for the primary load o resist.ng component. 7
'l For the case of plastic impact, the mass ratio in Equation (B-9):
mp
-s m +m r
%J can be written in terms of.the weight ration p
Np + "r - W is taken to be equal to one-half of the whipping P pipe weight bounded by the nearest plastic hinge (s). W r for a slab or wall is the weight.of.the volume: Vy = (d+t) 2 (t) (B-11) where d is the striking pipe diameter and t is the slab or wall thickness. W for a beam is the weight of the r volumes vb = (d+2t) 2 (t) (B-12) where d is the striking pipe diameter and t is the depth () of the beam. B-21
- l. .._&,'.
l
/ .:
l '\
. For. restraints subjected to compressive' loadings, the resistance-displacement curves include any reduction due to either elastic or inelastic instability. Flexural loading.on restraints and piping may cause lateral instabi-lity in the beam compression flange or local buckling in the pipe, 'and these effects are considered in devel-oping-the resistance-displacement. curves.
L f o
.B22
i J' ' t TABLE 4-1
\ :
DYNAMIC INCREASE FACTOR (DIF) I. Reinforced or Prestressed Concrete Concrete DIF Compression 1.25 Diagonal. Tension & Direct Shear (Punch Out) 1.00 Bond 1.00 Reinforcing Steel Tension. 1.20 Compression 1.20 Diagonal Tension & Direct Shear 1.00 7 (Stirrups) II. . Structural-Steel _ Flexure & Tension 1.20
-Compression 1.20 Shear 1.00 III.. Piping Flexure & Tension 1.20 Compression 1.20 Shear 1.00 I
Reference:
- " Structures to Resist the Effects of l Accidental Explosions", TM5-1300, June 1969. CE) B-23 a :__ ___-_ - - _
\- .Q..' .
TABLE 2 RESISTANCE-YIELD DISPLACEMENT VALUES FOR BEAMS Yield Description- Resistance Displacement
.(1) Cantilever R
M 3 [ u R=3 x
=h
/ L (2) Simply Supported
- h ..
R 3 4M u x RL p R= =g h L/2 _ - L/2 _ mn 9 (3) Fixed Supports R '
/
e l 8M u RL
# ~
R=
/ j L y 192EI L/2 , l; L /2 ;_ l l
(4) Multi-Spen 3 8M RL R= u xy e) J pg y y 3
=
92EI
' L L/f'L/ " L B-24 L_______________. _._
1 i i I TABLE 4-3 l RESISTANCE YIELD DISPLACEMENT -
- 1 VALUES FOR SLABS- {
Yield' Description Resistance Displacement (1) Sirply Supported on all i 4 sides with load at ] center i 1 b e 2 a +R R = 2wMu y
- II~" I a
.1 b -
b/a 1.0 1.1 1.2' 1.4 1.6 ~~1.8 1 2.0 ^3IO*
- a .1390 .1518 .1624 .1781 .1884 .1944 .1981 .2029 2031 i
(2) Fixed Supports on all v = Poisson's ratio 4' sides with load at t = Thickness center. E = Modulus of Elasticity T a R = 4wM u ""
.R *y "1 1 a
y t b/a 1.0 1.2 1.4 1.6 1.8 2.0 F b. a .0671 0776 .0830 .0854 .0864 .0866
*The resistance functions are taken from:
- 1. K. W. Johansen, "Pladeformler", Polyteknish Forening, j Copenhagen, 2nd Edition, 1949. i L
- 2. K. W. Johansen, "Pladeformler Formelsamling", Polyteknish Forening, Copenhagen, 2nd Edition, 1954.
The displacement equations are taken from: S. Timoshenko & S. Woinowsky-Krieger, " Theory of Plates and i Shells", McGraw Hill, 1959 3-25 l
.. _, 3 ,
i, '
'o g
3,,' j-TABLE 4 i \f f 3< DUCTILITY RATIOS p.
- n .
Reinforced Concrete Max. Value of u E . Slabs & Beams- (Tension steel only) 1S f 30
.P Slabs & Beams (tension and compr. 10
< 30 '
steel) p_p ' - Walls & Columns Compression- 1.3-Composite T Beam 8 p is the. percent of tensile reinforcement lp' is the percent of compressive reinforcement. A p = gh (100) and. p6 1.5% Steel Components Steel Beams. (lateral load) 26 (Note: To develope this ductility, the flanges must be thick enough to prevent local plastic buckling.)- Steel Beams (lateral and axial ' load) 8 LWelded Portal Frames (vert. load) 6 - 16 Piping 26
"( Referenc'ei Air Force Design Manual AFSWC-TDR-62-138, Dec.1962 B-26
= - _ _ - _ _ - _
F~ I-l ! TM 51300iMAVFAC P 397/AFM 88 22 ) ,"
- g I \. . .. .i
....- 4
... ..e.
r . g. .- l >(
.. .. 4. .
e 3
.4 . ... .e ..
. . , ...I,
.... ... ei. .
sg / .. [ . . . .. . ... .... ... . i j .
.... ... . I. . t,..
..l.... . . ..i.
. . J. l.. g.. .$. 1
. i l . . .g.e .i.,. %g [. .... . .. ^.
... . e
. . .g .
g 0 070
.muus.u. lu e , p .
p g..
....l.. ... ...
. ...r. , ., . ...
, l. .e ..
c.s
. , .s. .
.. . . l..! . .
., ..l
.. ... ,.. ... ... .... . .. ...l l.I .....,I ..l .!.
4
.i.). .!....
e
.i ....
U.. .. ) . g..g. . .. . , . ..l.... ...)
.. .g.
. g 4 .. . .... .l.
.s
. . .. . ..l . . . g..
- c. . ..... .... - .g. ...
..l.. . ..
=.......
..g.
.s
.i.'. .. ..... ... .... . mi
..s .
g 3g. ...
..,... 4 ....'
t.u D'he e., e.-- ei .. i
.u. .m. ... ....
uu. .
.u.m.. .
1
. l .. . .
..g
. g
. .g.
gg g
. p
. l" . ,
.u ,. . .
.v.0U.V ...l... ....r...y ... fN . [.....
....3 ..
. ... .. I.. ....
. .. e. .... .. .... i... n.... . .
.... .... s.
.s 4 ..e .
,.,. .. m.. e.. . .. - ..
. . .. ...m
.. l..
. .... .g.
l
... ... .....j _
s . . ... ..g . g
.. ..g*.
./ ....* . . u
.g. .. . . . ... . . .
m.,
..... I. ,. . gy
.g..
, ..- .)....
r l i t ,,
. .. 3. .......l ... . ... ...... .I ....
.....] ....... .... . . .... .... ...I.
p 4 .
,.t...
. .. ..s...... .... . .... ....
..l.. .
...,.ls.
.l..
- u. . . .
.T..
. f...
j
.. .I
.... ..i..
.l.. ..
(). C)S O .. .... ... .. ...... ... ..... I t. 'I.. 1. . . . . .. l.... ,1 .
..l.,. .. ..
... ....... _i .l. l...,....
. ii.. . l, .
.. . I. i. ..1.... . l._ o )..1
-..., .IL. I 1, l...... .. i. ..ti. .l... . .. .ll. -
.1. .. . .. ,1,. .. ,.. ..l. . . . . . . O
. ,.... ...l. . . ,...
- 1. ,.. .... ... . ... ! ..
.1
... ,I l l. .
... l.l. .... .. .
.. . . .1..... . . . .. .. . ..
- l. .
...I
...l.. . l. .. i . . l. . ..1. ,. .
. . .1.
],.il . l. l. . ./.. . ... e..
I... ........ ..l. . ., .. ..... .. l r g... . ..
. .I ..../.... 1 i,
.... ......i l1. ..
.I ..
. . I. ,) . . l. .
. 11.
,. l
... .... .... ... .. .. .l..
- i. s.. A.. I.. . . L.. . . ....
u, -.
.... .I .. ... ..,
l! . . . . . .
- l. l .l.
., ..l.
.i.....I .l...i. ... . .
- l. , . ..
. .. .l.., .
.l. . . !. ../
.!!. l. .... .
1.
.. t.... . I...l.
1I l.1. .. ... ... . .. . F, 0. 0 4 0 ... . . . . .
, . , ,/ e. ,. . . I I ., ... .,
.l.. . . . . . ..... .. ..1 .... ,,..l.. ..
- i. l . ..e.
, l. ..
..... ...1...
. ,l, .
, . , . ..... . !. . ... .l...l.. , .... .. ,
go .l..
. . . l.....
...l. .
() .,..
.11 l..., .. . ....
..I.
. . I...
- ..,... 1.!! . I.1. .l l .i 1 .
.a.... .s e. .
.l e
I el . ..... h.
- u. ... .. ... .. .., . ..
.I ... ,,... ,..
... 11 ss - .... ...... . .
... . ...l. . . ... .. ... .g....,... . ...
.. , .l..
... . .,1, .....
. 11 l-e. .1 . ..... ..
t,p
.. .I... l .,.
- a. . ...
1..ll. ...... ,f.... . .i.... ...... .1. . ... ._..
.f o . _
.. I.
. ... , . . .....1....
.i g
x,- () gig.. .. i
.. .., .... l.... i..g..g i
.g I
. .;g I. . .;i. 1.I. I..l...
.. ....I.g.g .
g O.O3O ... . . . .. . ., .
.....,l... .!.
i..... .. .....
... ..I. l ..,.!.,.. ,. .I... l...,1 . .... ....
). /..l..
.... .. . ,.,, 1. !. 1. .. ... .l.
. l .- ,.sl l. i. ...
;e . .. .s..
.t ... . ... . .
. l. .. . . . .. . .. ...
. .l..
.... J..
11.. ...
...l., .. {
. . ..... . l. l.l. .
. ... ..ll
.s. i.i.
.. . .. 1 ..l.
. .. .gl.. . .. . 1.. . , .. ..
... I.. .
..i . . . . . . l.. .
.. l.1 i
.e.. l. 1
- l. . . . . 6 .
. . . ..g. ... .. ,. ,. _....
].
..... ... .t...
.s.. .. . .. ....
. .. .. .... .."1
, .l.. . ..l... ..... .
6, . n E f"[c .... . g.... . . . ..l-
.. .. .. .1.
.. ..... *j'
.,g.. . . ..i. .
..s. .. . ..,
i ....... .... . ,.
..g.
.. ..l.*. *. l *. ..
.l.. 3 .t'. '
l.p.
... ..'../ . . .
O.O20 j. *.r, LVU .4 .... . ..
.l. . .I- .
..,, .... ,. C .
. l
.. . ,...il.
..l.
.-l... .
. .a.
e
.. -l p.
.e..'..e. ... . 1. s. _ .. e...
l....
.. g g.,
.g g..,
g
.g ... l. ...
.gl . .
...4
.4g..
)..
.I... .gl i .. l 9.
. . .. s. ... . 5. .. . . g ... .. .
. , . ... . .. . e .,
..' I I ....
.... ..eI .
.6. . . .. l..g.g p. ..... l .
. . ... .... I .l.. . I.
... . . t l .,
..g g..g. I....
s
...,.e . .. .
l..... 6, g .... p . g s . I-. . l e. . ... .. p . I. t
,. .e...
. l.
M.I
. ... ... , .., .., , . .. ... .. ...,. .4.
_5.j.. ... ...
. 4 .I,.
... .. .l....,.. . ... .. . ..
- 1.
. . .....l e1 .e . 41. .. i...... . .. . .
. .. 1. .!. ..
. l . . .i !.. ! . 6 1i, . .. -6 Oa ()l() .. ... .
. i,
,,.1
. . 1., 8 l.. .e1.. l I... ,
.--1. .......... I. l .. ...
.I.. .! ..l..I .!.. .I. .fl. ..... . ...
1.
. . . .I1
....l i l .. . . . .
..... .l... . . ...
.,1 .
..l,. 11.
. ... . . . ,,l I. ...!
.lll t.
. .. I. ,.
+. .-
.e. . .
..-..l.. l,..
. . . ..l..
. ..... .... ... . I. .
. . ..l.
..I 1. .
..l.,. 1.l . i..i..
. . . . I. .. . .f. ... . ! ! .... .,.
i.
. ... .... . .,.1 ..
....... ...... .... .>.. Lt. ..... l. .l.
I
...... I. .
/..... . .11.. I.. .l. 1 ... . -11 l..
... _. l~. .
. .l. ... .. . .... .. ..l .... .... .... . . . . .
t.
... - = . . . ..... .,,. .., ...
.. . . ...... .1.,.. . .
. . . . , .l...,
..t.
..1. .
l.. l. .... .. 1. ..
. . . ... . __. ..t,
. l. .l..1 . --.
... l. . . .
... l. . ....
. l. . .
.. .. , l ._.. . ..t..
..l
.._ ..t..
.. .. .I. . i. t. .l. .
c) _ O O.004 0.008 0.0 12 0.016 0.010 REINFORCEMENT R ATIO p: A3 /bd Figure 4-1. Coefficients for moment of inertia of cracked I,'.,*/~) sections with tension reinforcement only. B-27
( f%
- Q 5.0- STRUCTURAL ADEQUACY 5.1--General The following general steps are necessary to determine-the adequacy of the. structural system in resisting the applied loading:
- 1) Determine the magnitude and location of the peak dynamic force which is usually taken to be:
L F=C D PA f where C D is a fluid dynamic factor, p is the fluid pressure, and A f is the cross-sectional flow area of the' fluid jet.
- 2) Evaluate the resistance versus displacement curve to
() the maximum allowable displacement which is~given by px for all resisting elements. y
- 3) Using the applicable impulse or impulse and impact solution technique,, evaluate the maximum displacement of the structural element under consideration.
<4 ) If the displacement found in Item 3'is less than the maximum allowable displacement of Item 2, then the system has successfully resisted the load; otherwise, failure is predicted.
-~
s)- . B-28 i
D-5.2 Examples of Impulsive Loading and Impact Combined with Impulsive Loading. The following example will illustrate a situation where a pipe located 2 feet from a concrete wall may rupture. The pipe is cupported at 30 foot intervals and it is postulated that a slot failure occurs at mid-span. If the rupture occurs on the side of the pipe nearest that wall, then the case becomes one of jet impingement on the wall. pn'the other hand, the rupture may be.on the side of the pipe away from the wall, in which case the pipe will be subjected to a whipping action, and the pipe will subse-quently' impact against the wall.
' /"L
~V 5.2.1 Example of Impulsive Loading - Jet impinges at center of one face of a concrete tunnel.
Tunnel '6 "
* [ 45 0 12" e.w.
e.f. Wall "Q o " Pipe h.f ~o a c 2 d; .,
" o V.'s n c.
4_ 2" _ 22" ; f.
,k B-29 I
+
33 9 N.
' n,
' L h't '
5.2.la . Structural Sizes and Material Constants k Concrete Wall , Reinforcement ~ ~ Concrete A,: = 0. 31 'in /f t. .f = 3 ksi-
#5 0 12 in. b = 12 in.
y = 1.2 (40) =.48 kai~
,, f .d = 22 in.
^
where 1.2 is Dynamic 1 Increase Factor (See Table 4-1) 6 E, = 30.x.10 psi -n = 10' A o=5[=0.00117 E c
= 3000'ksi Piping System
= Mean' Pipe Radius = 6 in.
) .r p
t= Wall Thickness = 0.5 in. I.D. = Inside Diameter = 11.75 in. Af = Fluid Area = 108.4 sq. in. y p = 1.2 (YieldfStrength). = 1.2 (30) = 36 ksi where'l.2 is the Dynamic Increase Factor p = Fluid Pressure = 1000 psi C D = Dynamic Fluid Factor = 1.25 t n ( .k O- . 8 B-30 Cr______.____________. _ _ _ _
5.2.lb Evaluation of Wall Resistance Function. Section Strength - M, determination.: Mu is to be based on the strength principle of ACI-318-71* (see Section 3.4 of this document) . Check that p 6 0.75 ob, where
.p b" O.85K(![87000 f
1 where k y = 0.85
- y. (87000 + fy
, (0.85) 2 (3) 87000 = 0.029 48 [135000
( j
. Checking 0.00117_ < 0.75 (0.029) . OK Af sy 0.31(48) a= , 0. 8 5 (3) (12) = 0.486 0.85f b M = 4A,f (d-0.5a) = 0.9 0.31(48)'(22 0.42 I
= 291.4 k-in/ft = 24.28 k-ft/ft From Table 4-3, Case (2)
Rr = 4wMu = 4w(24.28) = 305k Also From Table 4-2, assuming a 10' x 10' clamped wall panel, 2 X =" (1-v2) where a = 0.0671 y a = 10 x 12 = 120 I v = 0.2 l t = 24 in. y , 0.0671(256) (120)2(.96) ,7.859 y 12(3000)I, I,
*The notations used herein conform to Appendix B of ACI-318-71.
B-31 [: L - _ _ . - _ _ _ _ _ _ - _ _
D,- This form of the equation is used so that the value t /12 (moment of inertia of gross concrete section) may be averaged with the cracked section-moment of inertia. I = Gross.I = I2 = 1152 in.4/in. of width Determine cracked section moment of inertia I ' c For n = 10 and p = 0.00117 .... F = 0.0085 l (See Figure 4-1.) l 4 I c = Fd = 0.0085(24)3 = 118 in /in. of width 50 4 I, = = 118 = 634 in./in. of width Substituting I, into the previous equation gives.... X = 6.596 = 0.0104" y 634 This establishes the following wall resistance. function: 4 R r
= 305k l
l
,Xy .0104" This completes the analysis of the wall resisting l medium.
O . O B-32 ,
H > s , , i
\
I J L f
)x J 5.2.lc Evaluation of. Pipe Resistance Function The following is a simplification which neglects stresses in the pipe due to pressure and temperature, also , dead load, etc.
Assume the pipe is supported at 30 ft. intervals. Slot Failure is assumed to occur at midspan. L = 30 ft ; l .e p- P l -Slot Failure j .
. 1
. .\ . : S = Plastic Section Modulus p
= 4r t = 4 (6) 2 (0.5) = 72.in 3 4 I = wr t = 340 in Find Mp to yield the pipe:
Use 0.9 as Capacity Reduction Factor. j i Mp = (0.9)y pp S g p
= (0.9) (36) (72) = 2333 k-in. ,
)
From the assumption that plastic hinges form at the support points and at mid-span (see Table 4-2, j Case (3)). ! B-33
]
- l) 3
- h. ,
-,q;7
- . j -
. _( 1 ce :8M-w- --
, J , 8 (2333) = 52K~
- p. L 12(30).
< Find'the' maximum displacement at mid-span for
. yielded condition.
3
~' 1 R, L
. x, = y >y.
K for simply support,ed span = 48 '
, . Average K for fixed supports = 192 k = 120 s
, y ,,52(30)3(12)3 = 1. 9 8 in .
': p 120(30,000)(340) p = Maximum ductility ratio for pipe = 30 p
... Maximum Allowable Pipe Deflection X = 1.98 (30) = 59 in.-
f This establishes the'following pipe resistance function. Rp .= 52k
- u. c i
x j, g __ 4 ) X
- 1.98 in.'
l P j i i 5.2.1d Jet' Impirigement on Wall j considar a 12 in, diameter pipe located 2 ft. from t.he center of the wall span. Total Force of jet load ... F B-34 i . _ _ - - - - - -------.----.----s-
g '
- ,o -
hjf y
.{s "Kj F= CDPA f 1
(see ' Section . 5.1)
= 1.25 (1) (108.4) = 136k' l
Use a 10 dispersion angle. j i
'l
'V .\;
~ 45 -l Pipe 10 L .
0 - Wall n j h g -,, 1
" h
.J- - -- 2
\t
_ d = 22" h y = 12 + 2(24 Wan.10 = 20.5 in. ! h2=hy + d = 42.5 in.
- Shear area = wh d = n (42.5) (22) = 2937 sq. in.
2 Shear capacity = 2940(44) f /1000
= 2940 (4) (0.85) 3000/1000 = 548K i
., *In this case, the slot bresk was assumed to be a circular area since this has a minimum perimeter
. and is conservative for punch out shear evaluation. ,
1 . B-35 l e -- - . _
=}t , a
, i I\' Since shear capacity (548k) is greater than flexural ' capacity (305k) , then flexural capacity is the governing condi. tion, and the previously calculated load displacement (see Section-b above) is valid. For impulsive loading and elacto-plastic resistance, an allowable ductility ratio u must
'be considered.
p = 3 = Allowable ductility ratio value'... Sec. 4.2. Rr (required) = _" = 1. 2F = 1. 2 -(13 6) = 163k j Since R r = 305 > 163,. then the minimum requirement for R.has been met. tO
%./
2 Pipe Whip Effect --Full Energy Trant.fer. Here the pipe traverses the 24-in. gap (X ) - g before striking the wall (see Section 5,2.lc), and the jet reaction (136k ) continues to act after the pipe str kes the wall. The resulting flexural behavior of the wall is evaluated and- i the wLil displacement is limited by a maximum 1 allowable ductility ratio of 30. ff9/1//21-as Pipe . Wall
' i F
l .'.1
.O - -
- x, _
= 24"
- B-36 L_
- - - .)
_ - _ . . _ _ - = - _ - _ _ - _ _ _ _ _ _ _ _ _ j :n; , S r , {(] 5.2.2a First assume full energy. transfer'for conscrvativs solution (no plastic impact credit) . l l F = 13GK-Work Done By Force b., E
- p = 52K
[f Energy Absorbed By Piye x d *- 1. 9 8 in . P r
= 305K 1
Energy Absorbed O. By Wall X = 2" ' g 24.0104" 136X = 52 (X- ) + 305(X-24 0.0104). w.
- w < n ,
Work Done Energy-Absorbed Energy Absorbed By Pipe By Wall t Solving ... X = 33.4 in. 1 The solution is valid since it satisfies the original assumptions. i.e., X>X g + Wall Yield .... 33.4 > 24+0.0104 X > Pipe Yield ......... 33.4 > 1.98 (" Wall Displacement = 33.4 - 24.0 = 9.4 in. B-37
Max. Allowable' Displacement = VXy.= 30(0.0104) = 0.312 in.
.:. .. Failure is indicated.
d 'j- ~ , 5.2.2b. Pipe Whip Effect with Credit for Plastic Impact. This cciculation'is made since failure was indicated in a) above withoutlthis credit. 1 Pipe weight per ft = (1) (3.14) (1) (20) = 62.8 lbs. Water weight'perTft = 0.785(1) 2 (62.5) = 49.0 lbs. 111.8 lbs. Half of.30 ft, pipe span.shall be considered to be~ effective in plastic impact, ' i.e., W = 15 W = 1680 lbs. P
-w The sketch indicates the.
4 . volume of wcil effectively t/2'
~ ~ ~ ~
.e C,T. - - - -y~ Pd impact.
a ./2
+
Vol .= (dp +t) 2t = - (1+2) '(2 r = 18 cu. f t. WW = 18 (150) = 2700 lbs. k9 w A e t -. Plastic Impact Factor =
"p
- 1680 W+W 1680 + 2700 = 0.384.
p w The appropriate equation when considering impact energy is Equation 3-5 with M and p M rep l aced by W and W . r p B-38 _ = _ _ _ - - _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ __
i: m
.f 1 -W'- X-N 9s ,,[ E + . F (X-Xg ) =-
FX -Ry(X g- y )- p w- ~ y - . . a.s '- < Reduced. Energy Work Done-Transfer From- + After Equals Plastic Impact Impact
" . X# -X "
Rp (X-Xg) + ~R r X-X 2 a s . Energy Absorbed ' Energy Absorbed By Pipe'After + By Wall
' Impact Restraint WP '
y ,, = 0.384 R p
= 52K X ~X r g
=.X y = 0.0104 i F = 136K
~
X = 1.98 in. P X = 24 in. R r
=.305K r
0.384 .136(24) - 52 (24-0.99)
+ 136(X-24)
= 52 (X-24) + 305 (7-24-0.0052) solving, 1 X = 27.6 in.
The' solution is valid since it satisfies the original assumptions. i.e., X > X g + Wall Yield ....' 27 . 6 ?> 24 . 0 + 0. 010 4 X > Pipe' Yield . . . . . . . . . 2 7 . 6 ;>> 1. 98 Wall Displacement = X - X g = 27.6 - 24 = 3.6 in. Maximum Allowable Displacement = pX y = 30 (0.0104) = 0.312 in.
- l. Failure is still indicated sven with impact ,
'[D x /.
credit. 1 B-39 L' i
1 l 1 [h I k,_) 4 i i APPENDIX C l' ( COMPARTMENT TEES 3URE ANALYSIS FOR - TURBr.NE BUILDING $ G .V Ame:2dment 4 (August 1987)
Compartment Pressure Analysis for Turbine Buildinr.
,y
.t ! N' Introduction This appendix describes the modeling technique used in evaluating a main 1 steam line rupture in the Turbine Building. A feedwater line break in the Turbine Building was evaluated using the RELAP 5/ MOD 1 Code. This code is considered commonly known, an. a description is therefore not included here. Modeling techniques used in evaluating high energy pipe rupiures in the main steam support structure are discursed in Portland General Electrical Topical Report PGE-1025< Hodel Description A main steam line rupttere in the Turbine Building was evaluated using _ Bechtel's FLUD (NE017) Version 7 Code, Thermofluid Dynamics for a System S of Interconnected Compartments. []v FLUD is a computer code used to calculate pressure and temperature transients in a network of interconnected compartments which is subjected to postulated pipe break accidents. Compartments are regions that are characterized by relatively complete mixing and low fluid velocities; flow paths see re6 1ons that connect compartments and have relatively higher fluid velocities. FLUD is similar to the existing COPDA computer code in that it performs the same kino of calculations and has similar options. It is different from CDPDA in that it allows modeling of con-vective heat transfer and steam condensation using the Uchida h6st transfer coefficient (or a constant heat transfer coeff 4 n?ent selected by the user): it can be used to model heating and ventilating system f lows Where f ans are present; it can model time-dependent variatians in atmospheric presrure, temperature, and relative humidity; and it can model variable flow path areas.
,g s-C-1 Amendment 4 (August 1987)
A discussion of the assumptions made in FLUD and of'their limitations is
' presented.below.
d, r%j The FLUD' Validation Report describes features and options of the code
. and the various beneLaark problems. run to verify FLUD. Detailed instruc-tions on the execu'lon and use of the FLUD program on Bechtel's Univac computet. system rce presented in the FLUD User's Manual . j Summary of A lumptions and Solution Approach
- 1. Assumptions Made by FLUD
.The meaning, implications, and rationale for the assumptions made in i the analytical development are discussed in detail in this section.
The significant assumptions used in the FLUD calculations are: l
- a. Thermodynamic equilibrium exists at all times in t.e compartment volume. 2!
- b. The compartment volume contains a homogeneous mixture of steam, I
air, and water.
- c. , Air is treated as en ideal gas, and steam is treated to second order in the virial equation of state.
- d. Flows between compartments are calculated using quasi-steady-state or an implicit technique.
- e. Condensation is treated by using either a Uchida or a constant 4
heat transfer coefficient as selected by the uner. Assumptions I and 2 l 1 Thermodynamic equilibr(uia means that the air, steam, and water compo-nents of each compartment atmosphere are at the same temperature and are homogeneously mixed. Thus, there are no temperature, pressure, or density gradients in the compartment during a given time step. This assumption can be satisfied by choosing compartment volumes small enough or integration time steps large enough so that I . 1 C-2 Amendment 4 (August 1987)
7m the relanation time for the compartment is less than the calculation The assumption of equilibrium greatly reduces the [l time step,. l enmplexity of the governing thermodynamic equations,'thus avoiding .j.
' consideration of nonequilibrium thermodynamics. This assumption can !
be further justified emp',rically, because its adequacy has been ! I demonstrated by comparing existing subcompartment analysis codes with many containment experiments.S The assumpti(n of homogeneous mixing of the air, steam, and water in ! a compartment implies that there is no water dropout, and thus, no consideration of liquid sump formation. For a large class of prob-lems where high enthalpy steam blowdown is the driving force for j compartment pressurization or where rapid flashing is present, this assumption has been observed to be well founded. The steam or j
'j flashing water forms a mist of very fine water droplets with insignificant dropout during the t.imes of interest. '
i n 1 For cases where substantially subcooled fluid fs being released or $j L for cases where long-term pressure and temperature transients are desired, FLUD gives the user the option of dropping out the liquid ; 8 (non-flashing) portion of the blowdown fluid rather than mixing it into the compartment atmosphere. The liquid portion of the blowdown is removed from the compartment. This has the effect of maintaining the compartment atmosphere at the saturation point. This option yields slightly lower compartment peak pressures than if homogeneous j mixing were assumed, but it gives vtuch more reasonable estimates of the compartment temperature, particularly for long-term transients such as for equipment qualification problems. I 1 Assumption 3 The assumption that air is treated as an ideal gas and steam is treated to second order in the virial equation of state agrees well with the respective air and steam tablec. This assumption simplifies the calculation immensely, because time-consuming steam table look-ups are avoided. l C-3 Amendment 4 (August 1987) 1 O_-____-_____
Assumption'A q
, O The use of an implicit flow calculation scheme is in keeping with the practice of other existing state-of-the-art thermohydraulic computer
-programs used in nuclear safety analysis. This method is based on i t
the governing equations of mass momentum and energy conservation which describe the flow. This method and the quasi-steady state flow method also used by FLUD adequately describes the range of flow regimes for which FLUD is used as demonstrated by the favorable comparison with available test data and with the results'of other l computer programs.
'Assumptfo_rj Condensins heat transfer is modeled'using a heat transfer correlation
~
developed by Uchida from measureraents of relatively quiescent
. steam condensing on a vertical flat plate in the presence of variable amounts of air. The correlation is applicable to the extent that the actual situation is similar to the original experiment. Effects on condensation such as compartment turbulence or steam impingement are not considered by this correlation, and thus, its application to
'these riituations is approximate only. The effects of turbulence increase the rate of condensing haat transfer. .Thus, the use'of the Uchida correlation is concidered conservative during the blowdown portion of the transient.
- 2. FLUD Solution Method 1
Since FLUD is concerned with calculating transient pressures and temperatures in a system of interconnected compartments, the tracking of mass cnd energy within the system is of great importance. FLUD uses the mass and energy in a given compartment to calculate the state point (pressure and temperature) of that compartment'. The stats point in turn is used to calculate the subsequent mass and l A, C-4 Amendment 4 (August 1987) ( , l; 1 L - _ -__ . _ - _ - _ _ _ _ _ l
m q-- - .] 4 ; Lj i s . i 1 ' energy, flow rates except.for the implicit flow method.Where a relationship between pressure and er.ergy is used. To advance the V calculation through a given transient, the flows of mass and energy-are-advanced in a discrete, finite difference manner through a given- ,
. time step. Thus, the main FLUD calculation loop proceeds as follows:-'
~
i t/ 1 a4 -
- a. Calculate the new system state point from the existing mass and.
energy, 4 b.s " Calculate mass and energy derivatives (flows and heat transfer rates).
< c. Integrate the mass and energy derivatives over one time step to i determine new masses and energy.
q References
- 1. Braddy, R. W., and J. W. Thiesing, Subcompartment pressure and -
Temperature Transient Analysis BN-TOP-4, Bec.htel Power Corporation, 6, 1976. l 9 a-
~
O 2. Uchida, H., et al, Evaluation of Post-Incident Coolint Syste'ns of 'J 7 ' Light Water Power ReactoJps, Third International Conference on the-Peaceful Uses of~ Atomic Energy, Geneva, 1964.
- 5. Nt017, FLUD', Thermofitiid Dynamics for a System of Interconnected ;
t
. Compartments, Validation Report, Revision 1, Bechtel Eastern Power (
Corporation, June 1986. J
)
4.- NE017, FLUD, Thernofluid Dynamics for a System of Interconnected I Compartments, Users Manual, Revision 2. Bechtel Eastern Power Corporation, June 1386.
- 5. COFDA, Bechtel Power Corporation; DDIF-1, Combustion Engineers; WARLOC, Sargent & Lundy; THREED, Stone & Webster; COMPRESS, United 3 L
l' Engineers & Constructors; TMD, Westinghouse; MNODE, Gilbe.rt r Commonwealth. t
.a C-S Amendment 4 (August 1987) .
l 1 _z._i___..i_____.___i__ '
I.e#
-N i
i l 3 APPENDIX D
^ '
(' STEAM LINE BREAX IN THE TURBINE BUILDING l Amendment 4 (Auguat 1987)
L Steam Line Break In the Turbine Building l f) V Introbuetion 1 The following outlines a method for obtaining mass flow rates which may , be used for determining local pressures resulting from a steam line break l in the Turbine Building. The flow rates obtained using this procedure are defensible upper bound values for any break location in the Turbine a Building. Note that they are intended to be used for local short-term 4 l compartment pressure calculations and are not applicable for calculating j long-term mass and energy releases, thrust loads, or jet impingement l forces. Also note that in the method that follows, back flow from the intact steam lines (and steam generators) is included. The temperature j effects of a steam line break are discussed in Portland General Electric Topical Report PGE-1025. I Basis f,or the Calculations _ , g- Since detailed flow rate calculations for steam line bre!.ks in the k Turbine Building are a ?.metion of p) ant piping layout t.nd break location, the effort required for tratisient blowdown analyses for all I cases involving assessment of consequential damage resulting from the break becomes quite prohibitive. For this reason, it is desirable to obtain defensible upper bound flow rates which are independent of break l location.
.The limiting plant condition in terms of both steam generator mass inventory and initial secondary system pressure are obtained when the plant is at hot standby. Because of the high flow rates associated with a steamline break, frothing in the steam generator causes a rspid increase in water level, resulting in a large decrease in the qua.itf of l
fluid expelled from the steam generator. Although the enthalpy of tW.s l low quality fluid is less than that of dry steam, the critical mass f2ow rate is higher due to the higher pressure, resulting in a net increase in the energy release rate from the break. Current evaluations show this to , I l G D-1 Amendment 4 (August 1987) ; ! _ _ _ - . _ __ -. - --_______ _____ O
be the'limitirag case for determining maximum pressure in vonted compart-g ments. - The blowdown can be broken up into time required f or both forward j and backward flow from a double-ended guillotine break. !!aximum pres-suces are reached very quickly'followins the rupture, and it is found that the piping inventory determines the magnitude of thu maximum compartment pressures. The mass flow production of the steam generators only plays a role in determining peak temperature values as discussed in PGE-1025. Computational Method The piping, both upstream and downstream of the break, may act as a reservoir during the decompression following the break. The flow is calculated using a method of characteristics computer program, Bechtel's PATHFINDER code, as discussed in the following. In the analysis, the blowdown flow rate after reaching its peak value was conservatively assumed to remain constant until the piping inventory was depleted. The peak value was verified with Moody's chart.
- Description of PATHFINDER Code 1
PATHFINDER SIMULATION OF THE FLOW OF AN IDEAL GAS j BETWEEN TWO RESERVOIRS USING '?HE l METHOD OF CHARACTERISTICS It has long been known that the method of characteristics (MOC) when spplied to transient fluid flow problems offers many advantages: it is the most accurate of any of the finite difference methods for simulating fluid flow, criteria for stable solutions are firmly established, transi-ent two-phase flow with heat transfer is easily handled, and the method l 1 ends itself to the simulation of flow in very complex piping systems. The PATHFINDER computer program uses the method of characteristics to simulate fluid flow in order to accurately simulate the flow of an ideal l gas between two reservoirs.
/*
k
~
Amendment 4 (August 1987)
+
The basic equations governing fluid motion are essentially nonlinear (more precisely, quasi-linear).. hyperbolic, partial differential equa-ly~)s.ik--J tions. _Because of this basic nonlinearlity, exact solutions are rare and are usually self-similar solutions. This class of solutions. This class of solutions result from the reduction of the governing partial dif-forential equations to ordinary differential equations by virtue of a high degree of symmetry. Therefore, recourse is made to numerical techniques for the solution of the fluid equations of motion. We explicitly assume the flow is one-dimensional and adiabatic. The equations governing fluid flow are the conservation equation for mass, SE " l- (pu) = 0 Ill 8t Bz for momentum SE + u lu + 1 la + g sine + F = 0 [2] at 8z p bz n and for energy $ G I- [p(e+u2)] + _ [pu(h-u ,)] + pug sine = 0 I31 at 2 8z 2 i where u is the fluid velocity, p the pressure, h the specific enthalpy, i d p the density, e the specific energy, and the remaining variables have standard meanings. The variable F represents the flow resistance per unit mass and includes the frictional resistance at the pipe-fluid boundary, as'well-as the localized losses: i F=1 Af + K ulUl [4] 2 D L where f is the Fanning friction factor, D the pipe diameter, and K/L the local losses per unit pipe length. The umss, momentum, and energy conservation equations are quasi-linear, hyperbolic, partial differential equations, and a closed form solution, is in general, not possible. Solution of these partial differential
]
i D-3 Amendment 4 (August 1987)
f equations (PDEs) via the method of characteristics consists of their p,. transformation into ordinary differential equations (ODES). A conse- _quence of such a transformation is that the resulting ODES are valid only along specific space-time trajectories known as characteristics. ! i This transformation rer;ults in six equations:' two propagation equatione, one transport' equation, and three equations describing the trajectories of propagation and transport. The propagation of pressure and velocity in the direction of positive flow is: d_2 + pa pu = -a 2 uF - pa [F + g sine] = K,' [5] , dt dz Bh p While the equation describing the trajectory of this propagation, the C, characteristic, is:
$1 = u + a. [6]
dt g The equation describing the propagation of pressure and velocity in the b direction of negative fluid flow is: gg - pa gu,= -a 2 dt dz
~[B ah p uF + pa [F + g sin 9] = L, [7]
and the equation describing the trajectory of this propagation, the C-characteristic, is:
$1 = u - a. [8]
dt Energy is transported rather than propagated; the equation describing the transport of-thermal energy (expressed in terms of enthalpy) is: dh - 1 gP = uF = M, I93 dt p dt D-4 Amendment 4 (August 1987) ! l i
and the equation describing the trajectory of energy transport, the C D P- characteristic, is:
- l ,% l- y
[10] l
!!E = u .
, dt I In the above equations describing pressure propagation and energy transport, the parameters K, L, and M represent losses due to flow resistance and heat transfer. (Far this simulation study, heat transfer was neglected.) Since the three conservation equations (Eqs[1] - [3]) contain four unknowns ~(p, u,~h, p), an additional equation must be specified. For l this additional equation, we choose the equation of state for an ideal i gas which expressed the fluid density p in terms of fluid pressure p and specific enthalpy h: p + p.(p,h) = k P [11] [k - 1) h 3 -(O/- where k is the ratio of specific heats. The junction between the two reservoirs and the flow path constitute two separate boundary conditions for the flow path. Gas is either flowing from or to, out from or into, a reservoir. For the case where the flow is from the reservoir into the flow path, the pressure at the flow path-boundary is assumed to undergo an isentropic expansion: l
= 1+k-1 -l [12) j where P, is the bulk average reservoir pressure. The variables p, u, and a are the pressure, fluid velocity, and sonic speed at the entrance !
I to the flow path.
~l O :
D-5 Amendment 4 (August 1987)
For flow from the flow path into the reservoir, the junction pressure is taken to be the bulk average reservoir pressure for O subsonic flow [(u/a) < 11 For sonic flow [(u/a) = 1], the junction b pressure is greater than the reservoir pressure. I l O O D-6 Amendment 4 (August 1987)
c. e y; A 1 y s h, - 1 i'
., , /
c lf [ l APPENDIX E m-FEEDWATER LINE BREAKS,IN THE TURBINF 41ILDING AND _
~
BLOWDOWN LINE-BREAKS OUTS *iDE CONTAINMENT Amendment 4 (August 1987)
.- --m _A__ _ _ . . _ _ ._a_____ ___, _ _ _ , ,__ ,,, ___
q, 7. y e
~
Feedwa+er Line-Breaks'in the Turbine Buildina and-i
- g Blowdown Line Breaks Outside Containment
'lf
, - Ihtroduetion
. The following outlines methods for obtaining' mass-and energy discharge.
. rates following a feedwater line in the "urbine Building or a. blowdown line rupture outside the Containment. -The methods for calculating mass
~
.and energy discharge rates result in conservative upr-r-bound values which should be used for the calculation of short-term compartment pressures resulting from piping rupture. The discharge rates calculated' .. !
by- these methods should not be used for calculating long-tem and energy releases, the,sst loads, or jet impingement forces. The temperature j effects of a feedwater line break are discussed in Portland General-
. Electric Topical Report PCg-1025.
- Basis for the Coleulations 7
E 1 1. Feedwater f,ystem l
'l Under notaal operating conditions, the feedwater system will contain pressu;1 red, subcooled water. The assumption of.a double-ended-guillotine break under these cer.ditions. results in a decompression-wave propstating through the system at sonic velocity with the pres-sure behind the wave corresponding to saturation pressure of the liquid. Because of the very low compressibility of'subcooled water, subcooled blowdown cannot be sustained Tor more than a few milli-seconds, and the total mass release under subcooled blowdown condi- )
'bns is q, tite small. Following this extremely short term initial i
. Ph ass , the pressure will correspond to saturation pressure of the feedwater.
l The' net ma ss flow rate throu5h a feedwater line break was developed using the RELAP 5/ MOD 1 code. The limiting conditions were obtained l asruming the highest feedwater temperature to be expected at that location under normal conditions. E-1 Amendment 4 3 (August 1987) l
y
- 2. ' Blowdown System
&^N t-.
c S.J . Sin'ce the blowdown system piping consists of much smaller lines than the feedwater system, as well as less interconnecting piping, the effects of friction losses in the piping are included in the calcula-tion. The blowdown flow rate is calculated using the results of-Moody's evaluation of blowdown o'f a reservoir through connected piping. I.se steam generator pressure at no load conditions is used . I to determine.the mass flow rate for saturated liquid. 1 Computational Method l l The zero loss maximum blowdown flow rate for saturated liquid based on ] 1 the Moody correlation has been fit to a simple function:of pressure to obtain the following relation: 1 C = 250 p 300 < P < 1200 pounds per (E-1) square inch absolute (psia) $j where P = Saturation pressure of the liquid psia G=Massvejocity,poundsmasspersqttrefootsecond (1bm/ft -see) 1 1 This function may be used to calculate G for the blowdown line rupture. Definitions: APf w = cross-sectional area of pipe at break location, square feet. l l Ps at = saturation pressure cceresponding to feedwater temperature ar, i full load. NL = no load secondary system pressure (this is consistent with a plant PSG trip from full load near the time of a feedwater line rupture). C = sonic velocity of compressed water - 4500 feet per second. 10 l O E-2 Amendment 4 (August 1987) O_ _-_ _
i i L8G = length of pipe between steam s'enerator and break, feet. I Lp = length of pipe between nearest main feedwater pump and break, (~} feet. .Af Wy = forward flow, from break, pounds mass per second. l WB = back flow from break, pounds mass per second. j k Since the blowdown system consists of smaller piping than the feedwater system with less interconnected piping which may act as reservoirs during the blowdown, the effect of line resistance may be included in the j procedure. ] I 1 Feglecting line resistance can lead to overpredicting the blowdcwn rate i i by a factor of two to five. Allowance for resistance, however, results in a best-estimate blowdown rate with a possible uncertainty of i 40 per-cent. For an upper-bound prediction of the blowdown rate, the appropri- l ste relation is, therefore, $- Gg = 1.4.x 250 x P x Gg /G = 350 P Gg /G,, i where G /G is a function of f L/D and is shown on Figure 1. g 4 The flow rate is determined by: i NL W=Apx 350'.(P 3g)1/2, 0'< t < ty < t seconds starting new loads where ty is the time required to discharge the piping volume between the break and a steam generator and leg. NL W=A x 350 (P gg)1/2 x G 3/G,), u d seconds where G /G is taken from Figure 1 as a function of piping resistance f L/D. i i r"% E-3 Amendment 4 (August 1987)
',f _f* f s' l Figure 1 is a condensed verrion of.the results given in ReferenceL2,- '
. ,.m .
which is-applicable over the parameter range of interest.- If credit-is 7 .. k taken'for a flow area in the line that is smaller'than the. flow' area of the pipe, such as a partially closed valve, the mass velocity should be determined assuming zero resistance, is: j 1 W = A,x 350 (P SG)1/2 ,
-where A is the flow area of the restriction.
'l Forward and reverse flow out the break should be determined separately and summed.
7] The energy release is determined by manitiplying the mass flow rate by the enthalpy of the fluid'being expelled. References n
- 1. F.'J. Moody, " Maximum Flow Rate of a Single Component, Two-Phase
)!
' Mixture",' Journal of Heat Transfer Trans. ASME,' Series C, Volume 86 February 1965, P. 134.
- 2. F. J. Moody, " Maximum Two-Phase Vessel Blowdown from Pipes", Journal of Heat Transfer, August 1966, P. 285. W
.j l
I l I b E-4 Amendment 4 5983k (August 1987) L___.-__n-_L____-___. l}}