ML20136G018: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
 
Line 1: Line 1:
{{Adams
#REDIRECT [[NSD-NRC-97-5018, Submits Responses to Four NRC Open Items Re Piping Identified as Action W in Encl 7 of NRC .Responses & Associated Ssar Revs Permit Staff to Close Out Open Items in Section 3.12 & Provide FSER Input]]
| number = ML20136G018
| issue date = 03/13/1997
| title = Submits Responses to Four NRC Open Items Re Piping Identified as Action W in Encl 7 of NRC 970207 Ltr.Responses & Associated Ssar Revs Permit Staff to Close Out Open Items in Section 3.12 & Provide FSER Input
| author name = Mcintyre B
| author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
| addressee name = Quay T
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05200003
| license number =
| contact person =
| document report number = NSD-NRC-97-5018, NUDOCS 9703170163
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 19
}}
 
=Text=
{{#Wiki_filter:._          _ . _ . _ _ .      _. . _ _ . _ . _ _.. . _ .                        _ .          .- . . _ _
14      ,
4 Westinghouse              Energy Systems                                              Ba 355 Pittsburgh PennsyNda 15230-0355 Electric Corporation i'
j                                                                                              NSD-NRC-97-5018 i                                                                                              DCP/NRC0767
)'                                                                                            Docket No.: STN-52-003 March 13,1997 Document Control Desk
,            U. S. Nuclear Regulatory Commission
;            Washington, DC 20555 ATTENTION: T. R. QUAY
 
==SUBJECT:==
SER SECTION 3.12 OPEN ITEMS
 
==Dear Mr. Quay:==
)            Attached are the responses for four NRC open items related to piping (DSER Section 3.12) that were
;'          identified as Action W in Enclosure 7 of the NRC leuer dated February 7,1997. These responses                                        :
j            along with the associated SSAR revisions will permit the staff to close out the open items in Section
;            3.12 and provide the FSER input. The Westinghouse status in the open item tracking system will be                                    ;
            - changed to Confirm-W. Draft SSAR markups are also attached.
1 4            The items included in the attachment are as follows:                                                                                  +
3                                                  DSER 3.12.3-1 (OITS #822) i DSER 3.12.5.3-1 (OITS #832) i                                                DSER 3.12.5.12-1 (OITS #838)
DSER 3.12.5.9-1 (OITS #836)
SSAR changes for the two items in Enclosure 7 (DSER 3.9.3.3-1 and DSER .3.12.6-1) statused as W-Confirm were included in SSAR Revision 11.
Please contact D. A. Lindgren at (412) 374-4856 is you have any questions.
sy//
Brian A. cInty e, Manager Advanced Plant Safety and Licensing                                                                                                  .
1I          l jml                                                                                                                            j Attachment cc:    D. Jackson, NRC (w/ Attachment)
N. J. Liparulo, Westinghouse (w/o Attachment) 9703170163.970313            i EDR    ADOCK.05200003 i                                  5 E.
M..E'P    ""1.ME''p'E P F-m                      PDR    \
 
e n
Attachment to NSD-NRC-97-5018 I
In the February 7,1997 NRC letter updating the status of many items in the ECGB review area the following requests were included on the AP600 approach for functional capability, modeling uncertainties, and thet al stratification for piping. The Westinghouse response is provided for each of the items below.
DSER 3.12.5.3-1 (OITS 832)                                                                                l Item (E) - Westinghouse response in 10/28/96 letter was discussed in the 12/5/96 meeting.
1 Table 3.9-11 of SSAR, Revision 10 should be further revised to make Footnotes 3 and 4 also applicable to Class I piping. At the 12/5/96 meeting, the staff took the action to internally discuss the use of ASME Level D versus Level C regarding allowable stresses for the evaluation of the functional capability of piping systems. The staff finds that the Westinghouse proposal is inconsistent with NUREG-1367 and therefore, is unacceptable.
Subsequent to receipt of the position provided above, a letter dated February 20,1997 updated the staff position on piping functional capabi'ity as follows:                                                ,
l The AP600 design for piping functional capability should implement the criteria in Section 9,
                    " Conclusions" of NUREG-1367, with the following clarification:                                    )
When slug-flow loads are combined with other design basis loads, e.g., safe shutdown              i earthquake, pipe break loads etc., then ASME Level D allowable stresses are acceptable for        I
;                  functional capability. When slug-flow loads are only combined with pressure, weight, and other    l sustained mechanical loads, then ASME Level C allowable stresses should be used, and the          :
l 0.25 Sy limit for steady-state loads does not apply. The 0.25 Sy limit applies for steady-stata loads, which consist of pressure and dead weight loads only.                                      l l
Wes.#-house Resoonse                                                                                      !
l The latest statf position is acceptable to Westinghouse. The attached draft SSAR markup shows how the revised criteria will be implemented.
l DSER 3.12.5.12-1 (OITS 838)
Th s item is pending resolution of OITS 832.(E) above.
y/rmnghpuse Resoonse l
DSER 3.12.5.12-1 should be closed.
w                                                    I
 
    '                                                                                                                      i 3            .
l
        ,    s 8
i e                                                                                                                    l 4
Attachment to NSD-NRC-97-5018        l 4
4 l
* Modeling uncertainties:
DSER 3.12.3-1 (OITS 822)                                                                                  {
l              Item 2.b -- Westinghouse letter, dated 11/11/96, was discussed in 12/5/% meeting. The changes in Subsection 3.7.3.17 of SSAR Revision 10 are acceptable. However, as discussed at the 12/5/96                ,
meeting, the mixed use of time history analysis and response spectra analysis for the four soil cases is    I unacceptable. The SSAR needs to be revised to exclude this option.
Westinchouse Resnonse 1
i
~
The SSAR will be revised to delete the option of mixing time history and response spectra analysis.
The third paragraph of subsection 3.7.3.17 will be revised as follows.
Four separate soil cases, as described in subsection 3.7.1.4 are considered in the seismic I      analysis. One app:c=h i: :c p=fc= When time history analysis is used, a time history I      analysis for each soil case is performed. ^ =$= app c=h i: :c p=ferm :%.: his:c y ^.=!ysi
,                      for de h=d rcck ;ci! :=: =d : :!:;;!: :=pc=: p=:= :=!ysi fc: i: ==:i." ;; 6::: ;ci!
eases-For time history analysis of piping system models that include a dynamic model of the          ,
i                    supporting concrete building either the building stiffness is varied by + or - 30 percent, or the    l l      time scale is shifted by + or - 15 percent, to account for uncertainties. Alternately, when
  .                  uniform enveloping time history analysis is performed, modelling uncertainties are accounted 4
for by the spreading that is included in the broadened ;esponse spectra. In this case, the four soils are accounted for in the broadened response spe.:tra, i
i h
i
                    % 9 4
i
  )
l m                                        .
2
 
l.
l                                                                                                  Attachment to NSD-NRC-97-5018        ;
+                                                                                                                                        i l
* Thermal stratification:
l                      DSER 3.12.5.9-1 (OITS 836) l                      In the 12/5/% meeting, the application of the EPRI report to the AP600 was discussed, and i                      Westinghouse calculation in GW-PLC-001 was audited. Westinghouse agreed to delete the SSAR reference to the EPRI report. The results of the Computation Fluid Dynamics (CFD) plan for the i                      normal RHR line, PRHR return line, and ADS Stage 4 line, as Westinghouse proposed in its j                      12/16/% letter for addressing uncertainty of the temperature profiles, should be submitted.
i
!.                      Westinohouse Resnonse
                                                                                                                                          )
In the December 5,1996 meeting at the NRC offices in Rockville, the AP600 thermal stratification                  1 j                      calculation GW-PLC-001, Revision 0 was reviewed by the NRC. Three open items which relate to
: j.                      uncertainties in the temperature distribution of the normal residual heat removal (NRHR) suction line,
!                      the passive residual heat removal (PRHR) return line, and the automatic depressurization system (ADS) stage 4 lines were discussed. The NRC requested that these open items be addressed in the design certification phase. To determine the normal power operation temperature distributions (axially i'                      and diametrically) for the lines, and to determine the effects of these distributions on the structural          )
integrity of the piping, Westinghouse has completed detailed computational fluid dynamics (CFD)                  I analyses and pipe stress analyses, as discussed below.
Normal Residual Heat Removal Line The NRHR line was investigated for adverse stresses since there is a long section of piping that is sloped slightly downward from the hot leg. This layout may have the potential for stratification in the line due to turbulence and convection from the hot leg.
CFD analyses show that thermal stratification does not occur in the piping.                                      )
Stress analysis was performed in parallel with the CFD analysis using a stratification temperature differential of 71*F, conservatively assuming a step change temperature distribution across the pipe diameter, with the step occurring at the midpoint of the pipe. The results of this analysis show that the pipe stresses satisfy the ASME Code stress limits and leak-before-break bounding analysis curves.
The CFD analyses show that the use of the 71*F stratification temperature differential is conservative.
Passiv Residual Heat Removal Line The combined effects of turbulence from flow within the steam generator, the purification line flow, and pipe metal thermal conduction were investigated for stratification due to heating in the horizontal piping directly below the steam generator.
The CFD analyses show that the' forced flow perturbations from the purification branch line and steam generator outlet plenum flow / turbulence flow fields results in thermal penetration into the line.
Forced flow is the dominant mechanism near the steam generator, and free convection is the primary mechanism for spreading higher temperature weer to the more distant parts of the line. The more severe thermal penetration is caused by the operation of the purification line. Therefore, stratification is possible witHa the horizontal piping directly below the steam generator outlet plenum.
m                                                            .
3
                                                                            .- . ~ .  , _ _ _ . __          __  __.  .__      _ _ _
 
      '                  4 l                                                                                                      Attachment to NSD-NRC-97-5018 l
Based on the CFD results, a conservative stratification temperature differential of 100*F was used for        !
the piping stress analysis. For the horental riping sections between the steam generator and the first        !
:                      closed valves, it was conservatively assu.. ed that the stratification profile is a step change in j                      temperature, with the step occurring at the midpoint of the pipe. The vertically inclined piping from i                      the steam generator nozzle to the horizontal piping, as well as the vertical piping upstream of the horizontal section of piping below the steam generator, are at cold leg temperature, with no
;                    stratification. This is based on the fact that vertically inclined piping cannot be stratified, and the cold leg temperature is the source temperature for the hot water in the horizontal piping. The results of i                    this analysis show that the pipe stresses satisfy the ASME Code stress limits and leak-before-break j                    bounding analysis curves.
l                      Since the PRHR piping was determined to be hot, the line will be insulated from the steam generator i                    connection to the first closed valves. This will result in somewhat higher temperatures in the line but, i                    because all of the piping will be insulated, the fluid / piping temperature gradients will be less. The stress analysis conducted with 100'F stratification differential results in additional conservatism. The system design documents and calculations will be revised to incorporate the higher temperature fluid in the piping between the steam generator and the first closed valves, and also the addition of piping insulation.
Automatic Depressurization System (ADS) Stage 4 Lines The vertically sloped piping that connects the high point of the line with the horizontal piping containing the ADS valves may not be sufficiently long to provide a cold trap. Thermal stratification was investigated in the horizontal piping which contains the valves.
CFD analyses show that thermal stratification is possible in the horizontal piping which contains the valves. This horizontal section of piping receives a sufficient enthalpy influx from the hotter water at the highest piping level. Heat transfer in the fluid was due strictly to free convection driven by the        l temperature difference between the piping and fluid closest to the hot leg connection and the ambient          I l
temperature. Most of the heat transfer to the external air occurs where the piping is uninsulated at the highest elevation adjacent to the closed valves. The thermal driving force for the free convection consists of the hot fluid at the lower elevations and the cold fluid at the higher elevations. This            l situation results in the development of stable free convection, with hotter fluid rising and colder fluid falling. The source of cold fluid is continuously replenished by heat transfer to ambient in the uninsulated, horizontal section of piping. The steepest piping temperature gradients occur in the 45 degree vertically inclined section between the high point and the horizontal piping containing the values / The largest such gradients are below 120*F.
A conservative value of 120*F is used for the piping stress analysis, with a step change in temperature occurring at the midpoint of the pipe. The short vertically inclined section of piping which joins the horizontal piping that contains the valves to the horizontal piping at the high point of the line is also assumed to be stratified. The piping from the hot legs to this point is at the hot leg temperature, since this piping extends upward and horizontally from the hot legs. . The results of the stress analysis show that the pipe stresses satisfy the ASME Code stress limits and leak-before-break bounding analysis curves.
Since the ADS Stage 4 piping was determined by the CFD analysis to be hot, the line will be insulated from the hot leg to the first closed valves. This will result in somewhat higher temperatures m                                          .
4
 
      .. . . . .- .. .-.. - - -.. .                              - . -.. .- .. . ~ . . . . - . _ - _ . - - . - .                - . - . . - - _ . - - _
9              \
Attachment to NSD-NRC-97 5018 in the line but, because all of the piping will be insulated, the fluid / piping temperature gradients associated with the interface between insulated and uninsulated piping will be less. The stress analysis '
conducted with 120*F stratification differential results in additional conservatism. The system design documents and calculations will be revised to incorporate the higher temperature fluid in the piping between the hot legs and the first closed valves, and also the addition of piping insulation.
Due to the results in the analyses outline above, NRC review comments, and updates to the feedwater line design, the SSAR in subsection 3.9.3.1.2 will be revised. The proposed SSAR changes are shown in the attached markup.
1 1
4 5
i
                                                                                                                                                              ~
lll4 A                                      ,              5
 
l l
: 3. Design of Structures, Components, EquipmeEt, cnd Systsms 1
3.9.3.1.5      ASME Classes 1,2, and 3 Piping The loads for ASME Code Classes 1,2, and 3 piping are included in the loads listed in Tables 3.9-3. Table 3.9-6 lists additional load combinations and stress limits for Class I piping. Table 3.9-7 lists the additional loading combinations and stress limits for Class 2 and 3 piping.
l 1
Piping systems are designed and analyzed for Levels A, B, and C service conditions, and corresponding service level requirements to the rules of the ASME Code, Section Ill. The        l analysis or test methods and associated stress or load allowable limits that are used in        l evaluation of Level D service conditions are those that are defined in Appendix F of the        l ASME Code, Section III. Inelastic analysis methods are not used.                                ;
1 Subsection 3.7.3 summarizes seismic analysis methods and criteria. Subsection 3.6.2 summarizes pipe break analysis methods.
l l
The supports are represented by stiffness matrices in the system model for the dynamic analysis. Alternate methods for support stiffnesses representation is provided in            l subsection 3.9.3.4. Shock suppressors that resist rapid motions and limit stop supports with    I gaps are also included in the analysis. The solution for the seismic disturbance uses the      l response spectra method. This method uses the lumped mass technique, linear elastic            !
properties, and the principle of modal superposition. Alternatively, the time-history method    l may be used for the solution of the seismic disturbance.
The total response obtained from the seismic analysis consists of two parts: the inertia response of the piping system and the response from differential anchor motions (see subsection 3.7.3). The stresses resulting from the anchor motions are considered to be secondary and are evaluated to the limits in Table 3.9-6 and 3.9-7.
l The mathematical models used in the seismic analyses of the Class I, 2, and 3 piping          j systems lines are also used for pipe rupture effect analysis. To obtain the dynamic solution for auxiliary lines with active valves, the time-history deflections from the analysis of the reactor coolant loop are applied at nozzle connections. For other lines that must maintain 1
structural integrity or that have no active valves, the motion of the reactor coolant loop is applied statically.
The functional capability requirements for ASME piping systems that rnust maintam an          ;
adequate fluid flow path to mitigate a Level C or Level D plant event are shown in            !
l                Table 3.9-11      Ib    requirements are based on References 10 =d 20.                        I r l                                                                                                      l i
i l
o wnswa moum          Revision: 12 3.9 5c;                                      Draft,1997 3 W85tingh0US8
 
  ,a ur n
3
: 3. Design of Structures, Components, Equipment, and Systems 1
DRAFT                      .
I I
: 14. NRC BULLETIN NO. 88-11: Pressurizer Surge Line Thermal Stratification, December 20,1988.                                                                    ,
: 15. "AP600 Implementation of the Regulatory Treatment of Nonsafety-Related Systems Process," WCAP-13856 September 1993.
: 16. NRC IE Bulletin 79-13, " Cracking in Feedwater System Piping," June 25,1979 and Revisions 1 and 2, dated August 30,1979 and November 16,1979.
: 17. " Investigation of Feedwater Line Cracking in Pressurized Water Reactor Plants,"      ;
(Proprietary) WCAP-9693 June 1980,
: 18. "AP600 Reactor Internals Flow-Induced tVibration Assessment Program,"                !
WCAP-14761, March,1996.                                                              l 1
l              19. " Functional Capability of Piping Systems," NUREG-1367, Nuclear Regulatory            )
l                    Commision, November 1992. "F= :ic=1 Capabili:y C:iteria for E=n::a! Mrk Il          l Piping," Centra! E! ::rie Ccmp=y, NEDO 21985, 78NEDl'", E. C. Rotbagh-              l Sep:mber,1978.                                                                      l I              20. Deleted "F= :ic=! Capabili:y of ASME Cl= 2/3 S:ain!= S:: ! B+ w.; =d E!bc=,"
ASME 83 P"P 66, T H Liu, E. R. khn:ca, K C Ch=g.                                    ,
: 21. " Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8236 (Proprietary), WCAP-8238 (non-proprietary)              l l
l s ;
Revision: 12 a    imm ai2.o3:297 Draft,1997                                      3.9-l%                            W W95tingh0US8
 
      ,      e        . . . . .
: 3. Design of Structures, Components, Equipment, end SystIms DRAFT l                                                    Table 3.9-11 (Shx: ' e' 3)
PIPING FUNC7IONAL CLASS 1,2, AND      CAPABI{)1TY 3(              - ASME Wall Thickness:                                            Do/t 5 50, where Do. t are per ASME 111 i      Service Level D Conditions                                Equation 9 s smaller of 2.0 S and 3.0 S (2,4,5) l                                                                Equation 9 5 smaller of 2.0 Sf and 3.0 Sh3,4,6)
External Pressure:                                        P external sP nremal TE + SCVE                                                  C2*M*D0/215 6.0 Smf )(NB-3650)
Equation 10s 'NC3653.2) s 3.0 S c W TF + SCVF                                                  C2*M*D0 /215 6.0 S (2) (NB-3650g)
Equation 10a (NC 36I3.2) $ 3.0 Sc(
Notes:
: 1.        Applicable to Level C or Level D plant events for which the piping system must maintain an adequate fluid flow path
: 2.        Applicable to ASME Code Class ! piping for all loading conditions and analysis methods
: 3.        Applicable to ASME Code Class 2 and 3 piping -- h= 'h: fc!!cwing F-i:dc= x: me::
l 3'    Dyn= 6:.i ze rna:!ag (1:g Sc            . := h;m s :'= & c: non mve d ag)                            l 3.2 Dynanic mcm=:: re coub::d m:n; = !=:!: rnpenx :p=: rem =;!y= c41 !5 9-- p d brcein n;                  l
:nd =: x; :h;- 5 N- imping l              13 S:=dy =:: bending ::nn dca n;; =:=d:
B2
* M 5; 0.25 Sy Z
Fa ^.S"E C'r: 2 =d 3 piping S:: dem ac: n=fy S: F-::=c= : =:: 3 2:=, f=c: =;! =pe"!:y i:
m;r:d -- h= i: :q=.:!;= cn 2:= 2 =d 3              : r.::.
l 4.          Applicable to ASME Code Class I,2 and 3 piping when the following limitations are met:
s        ,
[
I              4.1 Dynamic loads are reversing (slug flow water hammer Isids are non-reversing)
I l              4.2 Slug flow water-hammer loads are combined with other design basis loads ( for example: SSE: pipe break I                  loads).
1 l              4.3 Steady-state bending stress from deadweight loads does not exceed:
Revision: 12 a swnirmo9a monm Draft,1997                                                3.9 122                              W W85tiligt100S8
:'      3. o.si o o, S.,oe,.,es. Com,ooezzts. E,.,,me=,. aod S,,,      .,
r-g; DRAF7 50.25 S y l
I l  5. For Class I piping, when slug-dow water hammer loads are only combined with pressure, weight and other i        sustained mechanical loads, the Equation 9 stress does not exceed the smaller of 1.8 Sy and 2.25 Sm.
I I  6. For Class 2 and 3 piping, when slug-Gow water hammer loads are only combined with pressure, weight and I        other sustained mechanical loads, the Equation 9 stress does not exceed the smaller of 1.8 Sy and 2.25 Sh.
l l
l I
1 i
s ,'
o wwnizexn. ni243:291    Revision: 12 3 W85tingh00S8                                    3.9-123                                        Draft,1997
 
                                                -                                                                  =                    ,                      -              . .
4 s
2                  w
                                    +
s        '
: 3. Design of Structures, Compon:nts, Equipmint, and Systems m
f        q. 3 ss . -
                                                                                                                                                                                                                                - -}
h              k
                                                                                                                        *r'                      1i fet
                                                                                                                          . L. t. 1 n.
                                                                                                                                          .            p......9 ,.      c. i s PlD.IM.P.
                                                                                                . .          E'I. .f M,
                                                                                                                      ~      PT. .inM..A. _I P.&. D. .A.D.~I f I.T. V.                          .&._U. %. f. t''
P
                                                                                                                              ~~ I~ .i.c e ,. , _ , .i. u. . n 2 P_.......a...
                          . . , . _                                                                                        _ca_ .. . _(*    ~ _ t.._ . . t . f _..
                                                                                                                                                        . . .                                                                                          ca._ .      _.
Mmit
                .e.....L.... ...          n..v h                                                                          n. n _ , . . i n..n~
t, . .e c- ,. .
h                                                                                  D. 1, he n..~rh,r9. ..              DS.LhA1EL.nL,
                                                                                                                                                                    .                ~.
D. 9.pke .t s..,n._  .~                              9. . A c ,. .
: t. ... . ~
                .c . .. !.                  .c.. . . I. e t. L.~_ . . . h
                                                                      ..                                                  D.1.h&D.fT.r9.          ...        D 9.he t..s n~                                                                            1. .. o  .c. ,. .
                      ........t......
                .k.? . _      ..'..    .                .. . ~.1..
                                                        .c.a        E t L.~_ . . . h                                      D t.h3D. rh,,f9...                D. 9.hu 1.A n..
                                                                                                                                                                          .                                                                              i. . f .c,,. .
n.-2. . _ h
                .            ...                                                                                          n. rL i 4. .          r.~.eMa k.in~  .                                                                                      t . .c  .c. ,. .
Ben          11
                                    .?._t..l'__ . . . , T.
h                                                        D.1.ha n. rL,i9.    .~
D 9.Lh k%..f L,n.~L
                                                                                                                                                                                      -                D. 9. .ht h. A.
                                                                                                                                                                                                        . .                  ~. ._.n.
c
: 9. . A. - , . .
t.' .. L. :. ... ... ..  - .a T. . h                                                                      n.-t% f 4.            re.
h.e t. .E.,n~                                                                                    1. . .C,c    ..
i 1
kee*
                .f.                  f_,.t....J...-.
L....
                                                          . . .        ... . t. J ,. I. ,! ._.!. ._ ..i n,. J
                                                                                    .                          ,_ .-._.._J.r.....__,..I,._
                                                                                                                                                    . . . ~ . . , c11. ..
                                                                                                                                                                              ..                      .....1.,.,
: t. f . J ,I ,m,I.. _~                    ,..1. 1. . I n...,.__.. ....!.e _.
                                                                                                                                                                                                                                                                . e.. f.
i 1
: t. 2 m1      1. :,..
                                                            .        _..n..._.,..J
                                                                                ........            L .-.._.. J ,!_:_.
r D.
1 _
                                                - n ~c . ._ _.,    m
                                                                          .,. I .f..    *i. l. . . D.
                                                                                                    . S. L ,,. D. 9..
                                                                                                                          .    ! ..(19~, .L._. . D.1
_ 9 19 S'S ='.5(R ,Kg                              M fe 24 )M                                                            b;!;;; k" 0;" l'3 fTbVFdf'm *p I D 9. _ -_ n. f,i n. ,
                                                                                          ,g i ,y L . . ...~.1.      _.
                                                                                                                                    ....L..._.. 4.f3 g,4, _ , ,---Tb* ,,          '
2, c;p                gap - - L . .. . J __ AciXC FM UD 1 f /M
                                                                                                                                  - -                    '- ~~''
s ,.
Revision: 12 os-nimo9n ni24nm Draft,1997                                                                                                                    49420                                                                        W    WBStingh00S8 e
: 3. Design of Structures, Components, Equipmint, and Syst:ms e
d    . Qq uvL faw "Y.'..      ,. s.-1A        1I
                                                                                                                                                                          /gui C L.w..w.m .
1 tf 11 m
sf g }g [ %'44  $,
j DIDf MP t't fMPTi(\M A f P A D & DII ITt? - 4m u s&vC a4 EasI' s aa a a i ss a ss a i w a a si a i s a as u s a a a a as a as a a s PI ACC 1                        'l    A MT) 1
                                                                                                                                              % ass a uss ay an y a a a i as e7 Wen
        'T        D1 _ A1 u            -
v.
A.IL v - i L...
vu.            ..v.
: m.        t.._
                                                                                                          .ws
                                                                                                                        .Lm e ius e AA _ . _
u.v      ..v.
                                                                                                                                                          ._m..
g i + =.w          . L. u.is Af
: v.        41..,
i ,iv f.m .
                                                                                                                                                                                                          .v.      Dva 9 _- t , #1, 3  v  D u t. _- A.s vf D.,9        -- 1 1/L
                                                .: i                          _l_L
                                                                          ,u s y            .u u AA J - _ _ _ _ ..
e      -v      wwgswwas
                                                                              *m d v_i usysiu            I_Lm _- ev        AAww      . i . gs_ _wmwm1, f_ _
1, ,vA, ivi u s yi nu -
_t_L_            _ n-e swwg.ww.      .. _ _ .,.
1Lm          L-,m.I            - . .
L.      u,,s _ w. J u ysiu vusv u vi .                                A v i . f*C 1 i.s Y.TT, k?Duvuvv                1 L IM
                    !_._        _ _t.u.,_                      _.._1..          f..
                                                                                                  ..1...,,              ..f                          AA f ..____ D9 _                          im...
                      . .w i y v m
                                                  . 1: s_iv ua y                s              -suv.                v: s u    1. y_. L,-  u. iv uwgews .                  u.      sm.v. sw as  .L..,._i..
is t
                                                                                                                                                                                                                        ..Av 0
s.
D1 _
u a -                  A.
v    t.
A
: v. 4L,          L...
vu.              .v.
Im_.,.
sw s.
                                                                                                                        .L._
                                                                                                                          .iu.=      AA_-..v'
                                                                                                                                      .i . v
                                                                                                                                                            -._.-.L-O'*"'**            ' ' " ' Af' " ' *A1m
                                                                                                                                                                                                  * * ' ' f-' ' " D' '9' *- ' ''"'
_1A  "' Dt  _ Af
                  ,D. S.        - t .1,rL.
                                                        .N        f__                                m      __
                                                                                                                    -_. nn          s ...ws _ . .. ,
                        .                . -                      svi u trL _i su -v.                                v t 17 /L . , r_                  iv.          .. t y_ L. ._. _-.t. v uw g ww a f .a...._..
_ As___
1.
                                    - v. n,      f,. _ _u f. y.. L. . .- ir w=g wws L , u._. J  w 1 L .. L_m.1
                                                                        .sy n iu v u.sw w s
                                                                                                                  -- AC11C evai            fff, s .
MD ifIM iu    suvv
                    ".                      fm..      t:                      f,. _
vi,f.I_L=                s AA J _ _ _, .                  D9        m. t-a-                1 A
                      ..-m      . yva u.w                  a  .m u_f y. . ave                  ..1.....,uwe v
syi          , ev uw gi vw w.                  u.        s  sws    .L..._.
                                                                                                                                                                                                          . .            e i
l A
_A7g!,L...us. f.                  1._.,. .L__
d t
: e.        f.?rv f.                u                                . , .      u.us.                A,
                    ,. _ Af u.w                A v.v.A 1 I.._yL.uIf%        g sg                              ,L...wu.
: s. gsvu.wa
                                                                                                                                                                  .L_
                                                                                                                                                                    . u. 9.A v s
d
                                              ._a            1            L.m.f                                                              M/9 1IIM                                                                                                          ll g,      u. iu usyiL=.                v usw w                  -_ A Cava-vi,            1 f.f.s* ff. a f ,    n i i.,    avvv i
I D1 _                                                                                                                                                    D1 _ 9 11
: 1. Av. u            -
A..f,w., .. y. " f. w_ . '. .L.. _. D.
v i
                                                                                                          . 9. L.v. __-D9.            . _    .,    4. .f9.i .L.w__.
                                                                                                                                                                  . s. u      -
u-
_ -()                                                L...          _m.        1___                      .4 11 D 9.Lusv                  v . ,4 g.s
                                                      /D                                ,    vu.          i.v.        ivas        .Lu_..
                                                                                                                                    ..            r ,
h l
L...          _m.                      .L m          4 11                                                                                                            j D9
: u.          _ -Av.vg    f tD.s                                  ,    vu.          ..v.        f. w. _asusu      . <
L_ _.sJ.                          A. C. R.u E.f*..f. .f fuM.        . D i_f /M
                , ,                                    ,                            v..                .
11
: f. *F T. _ A    -
: v. ','f-if ! , Lv. ...    .  .v.
: m.          t...m. . .      .L..,._. . . t. ~A
                    =-v.,./L..
_ AA
                                                          .      L. L,._...!
vu              .
vi.          .A C 1 f.f*.s ...
fit      M, /,.9.    .i vvv f /M                                                                                                    J l 1
il J
l l
l I
d l
Revision: 12 o w,neoma anau m Draft,1997                                                                                                                                            3.A '203                                                                  W W85tingt10US8
 
a        V" Y
2                E                                  3. Design of Structures, Components, Equipment, and Syst:ms 1
* g        ',  J
                                                                                            ?
Thermal Strati 0 cation, Cycling, and Striping Thermal stratification, cycling and striping (TASCS) are phenomena that have resulted in pipe cracking at nuclear power plants. As a result of these incidents, the United States Nuclear Regulatory Commission has issued several bulletins, which are discussed below.
Thermal stratification may occur in piping when Dow rates are low and adequate mixing of hot 3d cold fluid layers does not occur. Thermal cycling due to stratification inay occur because of leaking valves or plant operation. Thermal striping is a cyclic mechanism caused by instabilities in the hot-cold fluid interface in stratified fluid during relatively steady flow conditions.
The design of piping and component nozzles in the AP600 includes provisions to minimize the potential for and the effects of thermal stratification and cycling. Piping and component supports are designed and evaluated for the thermal expansion of the piping resulting from potential stratification mcdes. The evaluation includes consideration of the information on thermal cycling and thermal stratification included in NRC Bulletins 79-13, 88-08, and l              88-11, EPRI =per: " 103581 (Ref =ne: 13), and other applicable design standards.
NRC Bulletin 79-13 Bulletin 79-13 (Reference 16) was issued as a result of a feedwater line cracking incident at Donald C. Cook Unit 2. This bulletin required that inspections of operating plant feedwater lines be performed. This resulted in the discovery of cracks in the feedwater lines of several plants. To provide a uniform approach to address this issuet a Feedwater Line Cracking Owners Group was established. The specific tasks of the Owners Group Program were to evaluate the thermal, hydraulic, structural and environmental conditions which could individually or collectively contribute to feedwater line crack initiation and growth. The Feedwater Line Cracking Owners Group was disbanded in 1981, after the original investigations were completed. The results of this program indicated that the primary cause of the cracking was thermal fatigue loading induced by thermal stratification and high-cycle thermal striping during low How auxiliary feedwater injection. The mode of failure was concluded to be corrosion fatigue. This information is documented in WCAP-9693 (Reference 17).
I            The ^,P600 steam generators are equipped with separate nozzles for the main feedwater and I              startup feedwater lines. Analyses of the AP600 main feedwater nozzles are performed to demonstrate that the applicable requirements of the ASME Section III Code are met. M
                    =dyi!=!uis =dimi, of p!=: cpca::=,:hc=d am:iS=:!ca, hc=d cyding =d
                . th:=d adping, u:ing ::mp;m::= dindbu:i=: =d :r=in:: -Sich am &vdoped frcm th: synem & sign, piping kycu: ad a p; & ne Y the f::d=::- lin: emeking ;; sue.
l            Thermal stratification is prevented in the main feedwater line based on the flow rate I              limitations within the main feedwater line and the flow control stability for feedwater I            control. Low feedwater flow duty is provided by the startup feedwater line while higher i              feedwater flow rates are provided and controlled via the main feedwater line. The I              switchover from the startup to the main feedwater lines occurs above a minimum flow rate I              to prevent thermal stratification for limiting temperature deviations. Main feedwater control I              valve positioning during normal operation is the function of the plant control system. The Revision: 12 owmenai2 o3i297 Draft,1997                                          3.9-50                              W W65tiligh0llS8
 
.                                                                                                                        l l
: 3. Design of Structures, Components, Eq:lpment, and Systems
.                                                                                                                      l DRAFT l                control scheme enhances steam generator level stability and thus reduces potential feedwater i                thermal stratification resulting from temporary low flow transients. A men.:cing prcgmm
                      .;i!! be impicmen:cd by he Combined Licen;c Sc!d= :: :he Sn: AP60^ :c :=crd                      ;
:cmpem:= di;;ribu::cn; =d th= mal disp!=cmen:: cf the feedwatc liac piping, a; wc!! a:            l per:i=n: p!=: pmame:cm :=h a; ;; cam genem:ct :cmpem:= =d !=c!, =d de=m:ee tempcm =c. ."Oni:c-ing .'!!! be p=f=med dring het f=ctice.;! c;;ing =d dring :he Sn:
f=! cycle. He resu!:ing moni:cring data 'i!! Se eva!=:cd c -hov tha: .: i; i:hin :he              l bo=d; cf :he =aly::ca! :cmpaatum distribu:i=; =d disp!=cmen: .
NRC Bulletin 88-08                                                                                j Bulletin 88-08, Supplement 1, Supplement 2, and Supplement 3 (Reference 12) were issued following the discovery of cracks in unisolable piping at several nuclear power plants.          l These cracks were attributed to unanalyzed thermal stresses resulting from isolation valve        '
I leakage. This bulletin required that utilities: 1) review systems connected to the reactor coolant system to determine whether unisolable sections of piping connected to the reactor coolant system can be subjected to stresses from temperature stratification or temperature oscillations that could be induced by leaking valves and that were not evaluated in the          ,
design analysis of the piping,2) nondestructively examine the welds, heat-affected zones        )
and high stress locations, including geometric discontinuities and base metal, as appropriate,  j to provide assurance that there are no existing flaws, and 3) plan and implement a program to provide continuing assurance of piping integrity. This assurance may be provided by designing the system to withstand the stresses from valve leakage, instrumenting the piping      I to detect adverse temperature distributions and establishing appropriate limits on these          i temperature distributions, or providing a means that pressure upstream from isolation valves that might leak into the reactor coolant system is monitored and does not exceed reactor coolant system pressure. In addition to leakage into the reactor coolant system, leakage out    ;
of the reactor coolant system may also result in adverse thermal stresses as discussed in        I Supplement 3 of the bulletin.
For adverse stresses from leakage to occur in unisolable piping, three conditions are necessary:
: 1) A component with the potential for leakage must exist. In most cases, this will be a valve.
: 2) A pressure differential capable of forcing leakage through the pressure-retaining            i component must exist. Leakage in unisolable piping sections may be directed toward the reactor coolant system ("inleakage"), or from the reactor coolant system
            ,j                ("outleakage").
: 3) A temperature differential between the unisolable piping section and the leakage source sufficient to produce significant stresses in the event of leakage must exist. For cases involving inleakage, this could result from a cold leakage entering hot sections of unisolable piping. For cases involving outleakage, this could result from hot leakage from the reactor coolant system entering cold sections of unisolable piping.
o \ssarrvl2Wim Rl2Gl297 Revision: 12 W  W8stinghouse                                  3.9-51                                        Draft,1997
 
a
: 3. Design of Structures, Components, Equipment, and Systems u.
P)-      -*a 3 h The criteria used in the evaluation of the AP600 systems design for susceptibility to adverse stresses from valve leakage are summarized below:
Single isolation valves can leak, regardless of design except for explosively actuated valves.
                  . It is generally assumed that two or more closed valves in series are sufficient to limit the amount of leakage to a magnitude which would have a negligible effect on piping integrity.
                  . Valves which have external operators may leak through the valve seat and packing.
In the case of leaking through the packing, additional in-series closed valves may not be beneficial.
                  =    A positive pressure difference should be considered as a possible leak source.
* Cross-leakage is possible between interconnected lines that are attached to different reactor coolant loop pipes and are isolated by single check valves.
* Sections of piping systems which have a slope of greater than 45 degrees from the horizontal plane are no: subject to thermal stratification, cycling and striping thermal loadings.
* Pipe lines, or sections oflines less than or equal to 1-inch nominal size do not require a thermal stratification, cycling and striping evaluation.
The unisolable portions of the following lines connected to the reactor coolant system have been reviewed and are not susceptible to thermal stratification, cycling or striping:
l
* Direct vessel injection lines from the reactor vessel nozzle up to the accumulator injection valves, core make up injection valves, in-containment refueling water storage tank injection valves, and normal residual heat removal injection valves.
* Core make up lines from the cold legs to the core make up tanks.
* Passive residual heat removal lines from the hot leg to the passive residual heat removal heat exchanger, =d fecm :he ha: :=h=ger :c :h: ::=m gemm:ct ch==!
I                    head.
                    =    Auxiliary pressurizer spray from the pressurizer spray line to the auxiliary spray check valve.
Au:cma:i &pra=is:ica S::g: ' lix fecm h h : !:g::c :he S: g: t &p:==i m:ica cdca.
I
* Chemical and volume control purification line from the pressurizer spray line to the letdown valve.
Revision: 12 owniuno% niz.o3::97 Draft,1997                                          3.9-52                          W  W85tingt100S8
 
1 1
: 3. Design of Structures, Components, Eq:1pmelt, cnd Systems mm Chemical and volume control purification line from the passive residual heat removal line to the charging valve.
* Pressurizer safety valve lines from the pressurizer to the safety valve.
l
                        . Pressurizer spray lines from the cold legs to the pressurizer.
l
                        . Automatic depressurization Stage 1, 2, and 3 lines from the pressurizer to the                I depressurization valves.
* Normal residual heat removal suction lines from the hot legs to the isolation valves.
I                The unisolable ponions of the following lines connected to the reactor coolant system have          l l                been reviewed and are determined to be susceptible to thermal stratification, cycling or            I l                striping-                                                                                          1 I                                                                                                                    l l
* Passive residual heat removal line from the passive residual heat removal heat                l 1                      exchanger to the steam generator channel head.                                                I I
l l
* Automatic depressurization Stage 4 lines from the hot legs to the Stage 4 depressuri-        l zation valves.
l l                                                                                                                    l l                Analyses of the passive residual heat removal line and the automatic depressurization Stage        j l                4 lines are performed to demonstrate that the applicable requirements of the ASME Section          '
I                III Code are met. This analysis includes consideration of plant operation and thermal I                stratification using temperature distributions which are developed from finite element fluid I                flow and heat transfer analysis.
1 NRC Bulletin 88-11                                                                                  l Bulletin 88-11 (Reference 14) was issued after Portland General Electric Company                  i experienced difficulties in setting whip restraint gap sizes on the pressurizer surge line at      l Trojan plant. The cold gaps were adjusted to design settings several times and were found to be out of specification after each operating cycle. The gap changes were caused by              l plastic deformation in the surge line piping resulting from excessive thermal loadings. The thermal loadings were determined to be caused by thermal stratification based on monitoring and analysis. Several similar incidents were subsequently discovered in other surge lines,        i and an industry-wide program to evaluate this phenomena was undertaken by the various PWR owners groups.
I              The purpose of Bulletin 88-1I was a request to addresses, establish, and implement a i                program to confirm pressurizer surge line integrity in view of the occurrence of thermal I                stratification, and to require addressees to inform the NRC staff of the actions taken to I                resolve this issue.
1 I                The actions requested in the bulletin are discussed below, and the manner in which AP600 l                addresses the actions, if required, for surge line stratification:
o \surrv12V)3Mn Rl2 031297 Revision: 12 3 W85tingt100S8                                  3.9-53                                            Draft,1997
 
s      IF "!i
: 3. Desigm of Structures, Components, Equipment, end Systems
  *  ,                                                                                                                I w <lwj l              For all licensees of operating PWRs:
I I              Request 1.                                                                    ,
I l
1                  The actions included under this heading are not applicable to the AP600.                    l l                                                                                                              l i              For all applicants for PWR Operating Licenses:                                                  l l
l              Request 2. a)
I                                                                                                              .
I                  Before issuance of the low power license, applicants are requested to demonstrate that      i I                  the pressurizer surge line meets the applicable design codes and other FSAR and            '
I                  regulatory commitments for the licensed life of the plant. This may be accomplished I                  by performing a plant specific or generic bounding analysis. The analysis should I                  include consideration of thermal stratification and thermal striping to ensure that I                  fatigue and stresses are in compliance with applicable code limits. The analysis and I                  hot functional testing should verify that piping thermal deflections result in no adverse I                  consequences, such as contacting the pipe whip restraints. If analysis or test results i                  show Code noncompliance, conduct of all actions specified below is requested.
I I                  AP600 Conformance I
1                  Analysis of the AP600 surge line considers thermal stratification and thermal striping, I                  and demonstrates that the surge line meets applicable code requirements for the I                  licensed life of the plant. Hot functional testing requirements for the AP600 ensure        l 1                  that piping thermal deflections result in no adverse consequences.                          1 I
I            Request 2. b)                                                                                    ,
I l                  Applicants are requested to evaluate operational altematives or piping modifications I                  needed to reduce fatigue and stresses to acceptable levels.
I I                  AP600 Conformance I
l                  Analysis of the AP600 surge line ensures that stress and fatigue requirements are I                  satisfied, therefore the evaluation of operational alternatives or piping modifications is I                  not required.
I            Request 2. c) l                  Applicants are requested to either monitor the surge line for the effects of thermal I                  stratification, beginning with hot functional testing, or obtain data through collective I                  efforts to assess the extent of thermal stratification, thermal striping and piping i                  displacements.
Revision: 12 ownimmmonm Draft,1997                                        3.9-54                            3 W95tiflgh0US8
: 3. Desig of Stnictures, Compone:ts, Equipment, and Syst:ms 3
P T3J,& P "*
c
* a "g L\[      d  a l                        AP600 Conformance l
l                        As part of the Westinghouse Owners Group program on surge line thermal stratifica-I                        tion, Westinghouse collected surge line physical design and plant operational data for I                        all domestic Westinghouse PWRs. In addition, Westinghouse collected surge line I                        monitoring data from approximately 30 plants. This experience was used in the I                        development of the AP600 thermal stratification loadings. Monitoring of the AP600 I                        surge line is therefore not required.
I I                  Request 2. d) l I                        Applicants are requested to update stress and fatigue analyses, as necessary, to ensure l                        Code compliance. The analyses should be completed no later than one year after I                        issuance of the low power license.
i l                        AP600 Conformance I
l                        Revision of the stress and fatigue analyses is not required for the AP600 surge line, I                        since the design analysis considers thermal stratification and thermal striping.
l l                  Request 3)
I i                        Addressees are ' ' iested to generate records to document the development and I                        implementation ot the program requested by Items 1 or 2, as well as any subsequent I                        corrective actions, and maintain these records in accordance with 10 CFR Part 50, 1                        Appendix B and plant procedures.
I                                                                                                                i I                        AP600 Conformance I
l                      AP600 procedures require documentation and maintenance of records in accordance I                      with 10 CFR Part 50, Appendix B.
A=!ysi: cf :he p= mud =: :=ge !!= c p dcmed ic &me=:m:0 th: tb appli=b!:
mquimmau cf :b ASME S=:!ca !!! Cc& =: m:: Hi =dy:i indud= cc=i&m:ien cf p!=: cp;m:i=, :b=d ::m:!5=:i= =d :b=d ::dping, u:ing ::mpem:um db:d6:ic=
                          =d :==ient -hich m: &vdcped fmm = ped =ce = =l::ing p!=: m=i:cnn; p= gam:.
A ==l:cnn; p= gam ci!! k imp!==n::d by 'h: Cembi=d L;==: hc!&r a: :he Sm:
AM :c m:=d tmp;m:um di::db:! = =d :k=d di;p!= mat of :b surg: !!n:
piping, = d! = p=:ina: p!=: p===== :=h = p m::d=: ::=p;m:um =d !:vd, he:
                ,j        ::; tmp;m:um, =d ==::::cc!=: pu=p : :=. M=i: dag u!'! bc imd=med dudag b:
f=::!c=! :=:ing =d idng :he fm: fud :y:!:. n: :=u!:ing monHdng 6:: !!! b cvd=td :c :hcv 'b: .: b : ;$in $ k=6 cf $ =dy:i=1 ==p;m::= di::dbtic= =d di:p!= ; m a n.
own 2mo9.a:2m:297      Revision: 12 W  Westinghouse                                  3.9-55                                      Draft,1997 e
 
                                                                                                                        )
    ,  w l
q                                                                                                          .
_a      -
: 3. Design of Structures, Components, Equipment, and Systems e  ,
D D 8 TN 3.9.9          References LJhi-u 1                          !
: 1. ANS/ ANSI N51.1-83," Nuclear Safety Criteria for the Design of Stationary Pressurized    I Water Reactor Plants."
: 2. ANSI /ASME OM Code-1990, " Code for Operation and Maintenance of Nuclear Power            l Plants."
: 3. Kuenzel, A. J., " Westinghouse PWR Internals Vibrations Summary Three-Loop Internals Assurance," WCAP-7765 AR, November 1973.
t
: 4. Bloyd, C. N., Ciaramitaro, W., and Singleton, N. R., " Verification of Neutron Pad and    l 17 x 17 Guide Tube Designs by Preoperational Tests on the Trojan i Power Plant,"          '
WCAP-8766 (Proprietary) and WCAP-8780, (Nonproprietary), May 1976.                      a j
: 5. Bloyd, C. N., and Singleton, N. R., "UHI Plant Internals Vibrations Measurement          .
Program and Pre- and Post-Hot Functional Examinations,"                  WCAP-8516-P    !
(Proprietary) and WCAP-8517 (Nonproprietary), March 1975.                                l
: 6. Abou Jaude, K. F. and Nitkiewicz, J. S., "Doel 4 Reactor Internals Flow Induced Vibration Measurement Program," WCAP-10846 (Proprietary), March 1985.
: 7. Bhandari, D. R. and Yu, C., " South Texas Plant (TGX) Reactor Internals Flow-Induced Vibration Assessment," WCAP 10865 (Proprietary) and WCAP-10866                          .
(Nonproprietary), February 1985                                                            I
: 8. Takeuch;, K., et al., "Multiflex-A Fortran-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708-P-A, Volumes 1 and 2 (Proprietary) and WCAP-8709-A Volumes I and 2. (Nonproprietary), February 1976.          j j
: 9. Cooper, F. W., Jr., "17 x 17 Drive Line Components Tests - Phase IB i1 I11 D-Loop        j Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Nonproprietary),            j December 1974.                                                                            !
: 10. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, Nuclear Regulatory            I Commission, July 1980.                                                                    i I
: 11. "Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More," ANSI N14.6.
I s l
: 12. NRC BULLETIN NO. 88-08: Thermal Stresses in Piping Connected to Reactor                  !
Coolant Systems, June 22,1988, including Supplements 1,2, and 3, dated: June 24, 1988; August 4,1988; and April 11,1989.                                                  l i
l                13. Deleted E =d Pcwn R===h in::::u:: (EPRI) Repen S 103581, "H=d                            I S: :i'~.=::ca, Cy '! ; =d S:dping (TASCS)," R===h Scja: 343 02, "=ch 199 '.
i o swn12e3o9n ai2 03:291  Revisien: 12 W Westirighouse                                3.9-105                                      Draf2, 1997
                                                      .}}

Latest revision as of 00:22, 14 December 2021

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=ML20136G018&oldid=3374669"