ML20070H275
| ML20070H275 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 12/17/1982 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| JNF-53-82, MFN-193-82, NUDOCS 8212230301 | |
| Download: ML20070H275 (144) | |
Text
. _ _ _ _ _ _ _ _ _ _ _ _ _
GENERAL h ELECTRIC uuctean
,0 wen SYSTEMS DIVISION GENE N 08[92b-C NF December 17, 1982 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regelation Washington, DC 20555 Attention:
Mr. D. G. Eisenhut, Director Division of Licer. sing Gentlemen :
SUBJECT:
IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAD _TY ANALYSIS REPORT (GESSAR II);
DOCKET NO. STN 50-447
)
MEB DRAFT SIR QUESTIONS / RESPONSES Attached please fir.d General Electric's responses to the Mechanical Engineering Branch Draft SER Questions.
These responses address both the " formal" questions and additional questions brought up at the meeting requiring follow up action.
l Essentially all questions are addressed in this transmittal.
The remaining responses will be completed by January 14, 1983. An amendment is scheduled for January 1983 to formalize those responses requiring changes to GESSAR II.
Sincerely, Glenn G Sherwood, Manager Nuclear Safety & Licensing Operation GGS :td Attachments
)
cc:
..J. Miraglia (w/o attachments)
C.0. Thomas (w/o attachments)
D.C. Scaletti L.S. Gifford (w/o attachments)
H.L. Brammer (w/o at*3chments)
Ob
- c..W r; 4 8212230301 821217 PDR ADOCK 05000447 E
GENERAL ELECTRIC'S RESPONSES TO NRC'S MEB QUESTION ON CHAPTER 3 0F GESSAR II
m,,
a Question _1
~
Table 3.2-1
Response
Mc response required, (),,7h g,.gf O
O O
l l
O 2
Qu2stian 2 Clarification and justification of several of tM classifications
'Q if this table are requcsted.
Response
No response required.
M4.arbi4 41/Me/rgq, O
I O
e O
1 i
lO 3
Gauk 9
3.
Ohy are Nuclear Boiler System air accumulator vessels Safety Class 3 and Guality Group C instead of Safety Class 2 and Quali ty Grous a?
O
RESPONSE
The reasons for classifying the Nuclear Boiler System aic accumulatoc(for ADS valve actuation) safety class s and quality group c are as follows; 1)
The air accumulator is not a part of the reactor coolant pressure boundary and it will not create a safety ha7ard as defined for safety class 2 equipment if it doesfail.
2)
The air accumulator function is similar to cessential service water system which provides the service function for the essential equipment.
The service water system is classified safety class 3, and by the same reasoning the air accumulator is classified safety class 3 and quality group c t
+
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y-w----
-y,ww-,-r-gv,r w--
o---ww
=c-----ee mw-+------e~----w
,----y-y---=w-e+y-+--yw-w+*
' -, - -_,,,,*y~-ye 9-e-
-,e
-~-e-*
--a-,*w
Qunstien 4 Several components are listed with two possible k:lassifications -
p for Sa#ety Class and/or Quality Group.
For example, fo.' the RHR System the piping within outermost isolation valves has a Safc ty Class of "1, 2" and Quality Group of */./B".
In these instances where multiple classifications are listed, the table should be made more specific to indicate which component subset is associated with which Safety Class and
' Quality Group.
In most instances where multiple classifications are listed.
the second classification is lower than that previously presented by GE in other FSAR's.
Justify these reductions in Safety.Ciasses and Quality Groups.
,/
O a saoas-When mulitple classifications are used, refer to ttee applicablefoote in margin (a through x). Also note reference to note (a) is a typographical error, it should be note (g).
~FeevacQs M o unte L C C L A SSJf tC A P 'J:
Aec
- Aus, coRa t cToz I r4 TH c f o L L<> w u c lt CVis CD WA.
- 3. 2 - l pgcg5 o
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1 O
O
O O
O O
O Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Safet Classi-Assurance Seismic Principal Component
- Class Location fication Requirement
- Category Comments MEB II (Continued)
.g 4
(/)
12.
Valves, other - isolation valves 1/2 C,D A/B B
I and within outermost isolation I
valves l
13.
Valves - instrumentation beyond 2/Other A
B/D B/N/A I/N/A g(/)
[
outermost isolation valves ao i
z I
14.
Mechanical modules -
2 C
N/A B
I f
l Instrumentation with safety y
7 function y
4 W
e 15.
Electrical modules wi'.') safety 2
C N/A B
T (i)
H function z
16.
Cable with safet;* function 2
C,D,A,X N/A B
I O
i III Reactor Recirculation System l
1.
Piping 1
D A/B B
.I (g) i i
2.
Pipe suspension - recirculation 1
D A
B I
l l ine.
3.
Pipe restraints - recirculation 2
D N/A B
I 1ine
,e i
4.
Pumps 1
D A
B I
~
]
o l
i
)
)
i
.:q.y@.MOL 2.
Tabl 2-1 O,'
O q,3 EQUIPMENT CLASSI ATION (Continutd)
Quality
)
Group Quality l*
nasurance Seismic Safetg Incation" fication Requirement
- Category' Comuments Classi d Principal Component" Class pt..e, r, * > j VI (Continund) 3
- < Mc % s-NW 9.
Electrical equipment and 3/3-~
.C N/A
-9/NA 6
- j'.7
, y ;.,{;%,f.,'/' f,y T k...
devices
..t' t,,;. ; -
..r
.:t,tts.sy };;.s.....- d
.. s
'.gie{,. J
' W 10.
Cable with safety function 2
C,X,A N/A B
I
-p
, as
,o,.
.. !'. ~, 8. '.
.,gg VII Neutron Monitoring System aM I
w M Electrical modules - IPRM -
2 A,D,R,X N/A B
I
' I ::,
NI P
y 8
and APRM
>; /
y 2.
Cable - IPRM and APRM 2
C,X,A N/A B
I,
$N 3.
Detector and tube assembly 2
D B
B I
VIII Remote Shutdown System This system is included under groups /MPL's, II/B21, XI/E12, XVI/E51, XXVI/G's'i, XXVII/H13 and XXXIV/P41.
IX Reactor Protection !Jistem 1.
Electrical modules 2
C,T N/A 3
'I 2.
Cable 2
A,C,T,X N/A B
I y
!W
. o O
b4 I
t.-
- 7
33 Tabh3.2-1
]
p EQUIPMENT CLASSIFICATION (Cont inued)
Quality Group Quality L ;ety Claset-Assurance Sei.2mic Principal Component" Classb Location fication Requirement Category Comments 4
XXV Fuel Pool Cooling and Cleanup System
,s 9
g ((.r)
..~.
1.
Vet.sels. - filter /demineralizers Other R
C B
14/is
.g 3
R C
B I
n' 2.
Vessels - oths.
3 r. ic. ~;,.*,
>.{ -
- ),.,,f/;,'
y 3
R C
B
'I MTr8,-
ed 3.
Ileat exchangers I., -
. :.;4.
i 3
R,C C
B 5/N/#,j,d:'--
g 4..
4.
Piping
- 4 ta 3
R C
B I
r 5.
Pumps and pump motors y
s 2
C.
B B
I d
'~
"e 6.
Valves and piping - containment 3
h 1
44 5 {3 isolation
-4/Other R,0,C C
ft:N 7*
5/N/A (r) 7.
Makeup system 3
A,R C
B I
8.
RilR connections - emergency cooling 3
R C
B I
9.
Electrical nodules and cables 4
l XXVI Suppression Pool Temperature Monitoring System 3
C,X N/A B
I 9
1.
Electrical modules with safety functions 3
C,X N/A B
I 2.
Cable with sarety function I
O O
O O
'O 34 Table 3.2.i EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality.
Safety Classi-Assurance Seismic Principal Component" Classb Mcation fication Requirement
- Category Consnents s
XXXI Standby Gas Treatment System i
l 1.
Filters 2
R N/A B
I 2.
Valves - ductwork 2
A,R,C N/A B
I 3.
Cable with safety function 2
R,A,C,X N/A B
I DJ co i
2 4.
Fans and motors 2
R N/A B
I g
l C
w O
XXXII NI Chilled Water Systems p<
1 w
- sf l
1.
Control Ruilding 3
D,C,A C
B I
s tn 2.
Electrical switch gear 3
A C
B I
O 3.
Other buildings Other A,R,W D
N/A N/A 4
l XXXIII HPCS Service Water System This system is included under groap/NPL XXXIV/P41.
i 1
1 XXXIV Essential Service Water System i
N E3 1.
Piping
, 2,3 0,A,C B/C B
I
(
94 2.
Pumps 3
P C
B I
4 3.
Pump motors 3
P N/A B
I i
i
O O
O o
Tablo 3.2 g EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Safet Classi d Assurance Seismicg Principal Component, Class Location fication Requirement, Category h r.ts-XXXVI (Continued) 3.
Condensate header - piping 2
A B
B I
. f,'.
, 3,, ' f,j,, '
and valves 4.
Piping and valves Other O
D N/A N/A
~
. f';'.:
u 5.
Other components Othec 0
D N/A N/A
%y) c.
i go 7
XXXVII Suppression Pool Nakeup System 2l r
1.
Valves 2
C B
B I
Q, w
d "o "
vi t H C
B B
I 2
2.
Piping s
0 l
XXXVIII Instrument and Service i
Air Systems f
1.
Vessels, accumulators, 3
A,T,C C
3 I
I supporting safety-related systems i
3 A,T,C C
B I
2., Piping in lines between acctuulators (item 1) and i
l safety-related systems b
McS l
3.
Pneumatic control equipment 1
A,T,C
-Jees&e&
B I
4*r-*)-
l g
4.
Piping and valves forming part 2
A.C B
D I
of containment boundary I
Table 3.2-1 EQUIPMENT CLASSIFICA N (Continu::d) h Quality Group Quality Safety Classi d Assurance Seismic g Principal Component, Classb Mcation fication Requirement, Category Comments XLIV Plant Electrical Systems 4
1 (Applicant to Supply)
XLV Auxil' ary AC Power System MbM (f) l 1.
All components with safety 2/3 A,C,X N/A B
I function Nh g
CO f
O td e ca Y
XLVI Diesel Generator Systems P,y WW j
1.
Day tanks 3
S C
B I
ys (A H 2.
Piping and valves - fuel oil 3
S,0 C
B I
h l
system i
i 3.
Piping and valves - diesel 3
S C
B I
]
service - water system j
4.
Pumps - fuel oil system 3
S C
B I
1 o j
5.
Pumps - diesel service - water 3
S C
B I,
i system i
'l 6.
Pump motors - fuel oil system 3
A,0 N/A B
I l
jl and diesel service-water system N
- o u m>
7.
Diesel ger. orators 3
S N/A B
I
- 4y o
l "4
i I
l t
___ t 1
I
.)
~
GESSAR II 22A7007
~
238 NUCLEAR ISLAND rov. O Table 3.2-1
% #p
~
EQUIPMENT CLASSIFICATION (Continued)
NOTES (Continued) e.
B = the equipment shall meet the quality assurance require-ments of 10CFR50 Appendix B in accordance with the quality assurance program described in Chapter 17.
O N/A = Quality Assurance Requirements not applicable tc this equipment.
f.
I = shall be constructed in accordance with the requirements of Seismic Category I structures and equipment as described in Section 3.7, Seismic Design.
N/A = The seismic requirements for the safe shutdown earth-quake (SSE)'are not applicable to the equipment.
2 i
Equipment that provides no safety function but which could damage Seismic Category I equipment if it failed is analytically checked and designed to confirm its integrity against colla'pse I
when subjected to seismic loading resulting from the SSE.
g id Au WsMHCkS U STM 1157 * '/ ' W /W*4* W '*L IE Y PI' 05"it WYNI" wmi 78a Wj ax wr:
1.
Lines one inch and smaller which are part of the reector M. E G coolant pressure boundary shall be Code Class 2 and Seismic
% 9-Category I.
2.
All instrument lines which are connected to the reactor 4
coolant pressure bocndary and are utilized to actuate and monitor safety systems shall be Safety Clast 2 from the outer isolation valve or the process shutoff *falve (root valve) to the sensing instrumentatior..
1!O i
3.2-44 tl
c.
. cg g -'
a-A' 2.t Questirn 5
- W M,.l.ST T e.;?c. !... f., ~ - o ;-
For the Reactor Water Cleanup System, justify the designation of S'afety Class 2 and Quality Group B for the containment piping penetrations and -
.m F-1.i ~
' ~)
" k; g.%,y.M':qSig%k %%p.;g ~. ' :
. 7. :;p73,.,,.-, c;,.,;,
z%
> -s valves.
.~
p,
..:.c
.-' *j
. =.d Ic -;;'C T^$h:.m...+2
, ~
~
- A
%C -..,'.:.:. 3;G_' ' ^.
w,."~' a.
,2
- ' q: : );c. p:-:.
-.-t m =-
~
. a. :
m.y.
ne.oon.e 5,j
.e 4
.....i..., _:...
w
,s g
' x ca u
4.-
~ L w a
o C'n '
12 -/
C p3 3zp27) ca.Af 49m/ cpw[
n dL k
~
~
x~
v.
O O
I
]
,--.-,v.,.--,-, - - -,,. -
e-- - - ' - --- --- - -=
w*--wwv------7==ov'----'v T
-v-v*--ww-vv=-v----*
w-w-rw-w-------me-e'====-
==--w=--^"*wT*-
w
O O
O O
o Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality
'" ' f Safety Classi d
^**"##"
'e Principal Component, Classb Location fication Requirement Category Comments i
XXIII (Continued) 4.
Piping and valves within outer-1 D,C A
B I
(g) most contaisunent isolation e
I valves on pump suction i
5.
Pump suction and discharge Other A
C N/A N/A (g) y co piping and valves from con-taisunent isolation valves back 2
to containment isolation CQ l
g HEB h tny valves a
l
'?
I k
W W!
U 6.
Containment piping part-zticr.;
7 C,A g
B I
(g)
'"'"""^ N AlsAA lu/k rg g,j,,g yy 2"'
7.
Filter /demineralizer and heat Other C
C N/A N/A (g) o exchanger piping and valves inside containmerit f
8.
Piping and valves returning 2
A B
B I
(g) from containment to RilR system 9.
Piping and valves from contain-Other A,T,W C
N/A N/A (g)
I ment to main condenser /radwaste 4
10.
Filter /demineralizer precoat Other C
D N/A N/A subsystem 1
11.
Nor. regenerative heat. exchanger Other C
D N/A N/A e4
)
chall and pipir.g carrying i
closed cooling water o
I i
Q'ESTION 6 J
O It is the staff's position that certain systems important not identified in Regulatory Guide 1.26 saould be classified Quality Group C, or its equivalent.
Among these systems are:
diesel fuel oil storage and transfer system,' diesel engine cooling water system, diesel engine lubrication rystem, diesel engine starting system, and diesel engine combustion air intake ar.d exhaust system.
Justify the absence of a quality group classification of portions of those systems listed below:
O Diesel Generator Cooling Wster System Diesel Generator Starting System i
1 Diesel Generator Lubrication System Diesel Generator Combustion Air Intake and Exhaust System
([)
RESPONSE
The fuel oil storage and transfer, cooling water, lubrication, starting air, combustion air intake and exhaust system are quality group C classification as shown in the attached marked-up Table 3.2-1 pages 3.2-38 and 3.2-39.
Compliance is by the Applicant.
The diesel generator including all integral components are per the engine marufacturers' and DEMA standards and therefore, quality group classification is not applicable.
Revised item 7, pump motors is not under the jurisdiction of the ASME
{)
Section III Code and quality group classification is not applicable.
O
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FSH:pab/J12117 12/11/82 ep
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te.. e hblo 3,2-1 O
O eaurensar etassir:oron (c::=tiauca) 7 Quality Group Quality Safety Classi-Assurance Eelsmic Principal Component" Classb Mcation fjration Requirement
- Category Consments XLIV Plant Electrical Systems (Applicant to Supply) 4 i
XLV Auxiliary AC Power System 1.
All components with safety 2/3 A,C,X N/A B
I u
function a
"4c:n XLVI piesel Generator pystems,
'Y% ~YN"'
[~
"N t=uel oil Sforog< ond +roasfee
- i 4
/
1.
Day mnka- *,$ teen 3
8.0 C
8 I
F3 (E)
(
i c oli vigt er
- >9 5 tom s
/
- 2. -D pi a -and >/alves-- fusl-oil-3 S/M C
B I
1 h'l (Z)
)
_tys u.c Lubrico t:an sphas
/
s
(.;)
\\
5 5.
Pipin9-and-valves - diesel-3 S
C B
I
/
servica -. water-systerc c o m bdi l a n d 'e 8'4 A
- h' O ^ J--
G.1,.
Pmaps -fuel-oil-syst.am-3 S.O C
B I
[j h
exhgw t-syst un 5_
Pumps-diosal-servioo--wates '
3 f
C a
5 Yb
-system
)
. y f..
Pump motors g fy g oil systemi 3
A,0 N/A B
I h
e.nl. dles e saevich-water system,ond.
i ful.,c. O.? 3*j hicav)
- )
- d M j
(
w 8 C.
Diesel generators 3
S it/A e>
I.
B-(r/ J
.I 4
44 6
q,.^~
M y.
.esN
- o
/,
e v
y 1
}
e l
mm
^h O
O b
A
,4 Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Safety classi-Assurance seismia Principal Ccyponent" Classb location" fication Requirement
- Category Comments d
XI.V tinuedy.
g-v s
q 8.
Electrical modules with safety 3
S N/A B
I functions d
/ (o 9.
Cable with s.aiety functions 3
- S,X N/A B
I
/
u
' 1n niusol fual-storaga tanke 3
O C-B I-W ceces'ars P'Any A*"' 4 Ed"d"$ 'O
- 3 n.
Sterting air tar.hu ald do g -
3 S'
C B
I (5) ci 3
w skraans piping and valves p
r
[
I i
4 D.
Startir.g air compressor and Other S
N/A N/A N/A
[o mctors y
vv A
fQ g
m 9
[
XLVII Combustible Gas Control System u
{
1.
Ilydrogen Recombiner System 2
C.
N/A B
I 2.
Containment /Drywell Purge 2
A,C,E B
B
.I System
\\
3.
Containment /Orywell Mixing 2
D,C B
B I
System 4.
ran 2
D
'B B
I 5.
ran motor 2
D B
B I
$4 6.
Electrical modales 2
D B
B I
o g
s
QUESTION 7 O
Information should be included on the piping and instrument diagrams to delineate in detail the system quality group classification boundcries and the boundary limits of those system portions designed seismic Category I.
RESPONSE
All pipelines on tne piping and instruments diagrams are identified with a system number and material class.
Included in the material class code is the quality group classification (see GEESAR Figure 1.7-4, NI Piping and Instrumentation Flow Diagram Symbols).
All quality groups A, B, and C are desi(ned seismic Category 1.
O
~
O O
FSH:pab/J12117 12/11/82
')
QUESTION,8, Are the reactor internals designed to NG?
RESPONSE
The reactor core support structures are designed to % as evidence in Table 3.2.2 and Section 3.9.5.3.f.
The further response of the request related to the reactor internt.ls is provided in revised section 3.9.5.3.5.
O rne. ove responses.eet ene require.ents or Sa, Section 3.
5.
O O
i O
i FSH:pab/J12117 '#
12/11/82 i
t
__--_.,______._.__..___.m_.__.___.
GESEAR II 22A7007-M-
238 NUCLEAR ISLAND R v. 0 3.9.5.3.5 Stress, Deformation', and Fatigue Limits for Engineered O
Safety Feature Reactor Internals (Except Core Support S tructure) (Continued)
Components ins:.de :he reactor pressurs vessel such as control rods which =ust scve during accident condition have been examined to determine if adequate clearances exis:'during emergency and faulted conditions.
No meenanical clearance problems have been identified. The forcing functions applicable to the rea= tor internals are discussed in Subsection 3.9.2.5.
- Is&T rh 3.9.5.3.6 Stress and Fatigue Li=its for Core Support Struc ures The ::::1:$4 1f f;;.;;:mclndcfhe j
& Co
-M -- Mr J - 4 " fer the core support s truc -
tures are in accordance vitt ASME Ccde Section III, Subsection NG.
and are su.ar.arized in Table 3.9-10 (6).
3.9.6 Inservice Testing of Pumps and Valves' t
Inservice testing of ASME Code Class 1, 2, and 3 pumps and valves
/
in accordance with Section XI of the ASME Boiler and Pressure vessel Code and applicable Addenda as required by 10CFR50, Section 55a(g), is readily feasible within the physical design.of the Nuclear Island.
Accessibility for this. inservice inspection has bee prov0: led by attentiveness of design personnel to ISI requirements (e.g., space is available for convenient placement of ISI equipment, insulatien is readily removable where required, and adequate lighting is providad).
Appropriate pressure taps are provided around pumps.
Position indicators are provided on valves.
i O
1 l
Details of the inservice testing program, including test schedules I
and frequencies, are the responsibility of the Applicant.
- Also, c
any applications for written relief from Section XI Addendum requirements, purruant to 10CFR50, Section 55a (g) (6) (1), shall be made by the Applicant.
<The design criter'i(loading conditions, and analyses that provide the
\\
basis for the design of reactor internals other than the re support 1 structures : L 'd meet the guidelines of NG-3000 and structed so as not to adversely affect the integrity of the core support structures fNG-ll22).
U
.. h 7. - H J.~-.
~
m Question 9 t
Q Why are the pipe restraints on the recirculator system safety Class 2't
~
Response
cl~:fiA f tL ;p ask:Js a s y k.4 p
0 b
au1 che. rL w p, + th < p d a predad i
+1 mis,J i ll 3 2 / c p -y 1 tr>
O'
' O O
lo
- 7 Table 3.2-1 EQUIPMENT CLASSIFICATION, (Continued)
Quality Group Quality Seismic Classi-Assurance f
Class Incation ficatioa Requirement
- Category Comments Safet Principal Component" II (Cont inued) other - isolation valves 1/2 C,D A/B B
I (a) 12.
- Valves, and within outermont isolation instrumeatation beyond 2/Other A
B/D B/N/A I/N/A (a)
[
valves 00 l
13.
Valves -
Z outermost isolation valves Cl 2
C N/A B
I g
t**
14.
Mechanical modules -
h!
instrumentation with sinfety Wi function H:
7 2
C N/A B
I (i)
H Electrical modules with safety 15.
z function O
2 C,D,A,X N/A B
I Cable with safety function 16.
III Reactor Recirculation Syt, tem 1
D A/B B
.I (g) 1.
Piping 1
D A
B I
Pipe surrension - recirculation 2.
MG8 line D
N/A
/
I
[
A//A A//jf
$f yh Pipe restraints - recirculation o M,y b
a 3.
line 1
D A
B I
O<
4.
Pumps O
O O
O O
QUESTION 10 1
O expi ia ta s is ic cat sary or tii. ru i rvic aui9 at? caatrai rad grapple?
l l
RESPONSE
1 The control rod grapple appears in XIX, In-Vez,sel Service Equipment.
This grapple and the general purpose grapple included in Item XVII are O
a=t s r tv =1 t d aad ar id atifi d with "ata r" aa ta att ca d mark up.
Also see ASB Question M10.31 on same topic for futher information i
(attached). #L_ d uf 4
g j, g y Al A W
do M& lm R, eh x
i-
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SY& & M,
& %d &
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FSH:pab/J12117 12/11/82 4
A o
40 o
es Table 3.2-1
~'
EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality
}
Safetg Classi d A"#'"**
8'i'"i Principal Component, e
f Class location fication Requirement Category Cosaments XVIII Reactor Vessel Servicing Equipment 1.
Eteamline plugs Other C
N/A N/A N/A f
2.
Dryer and separator sling and 2
-C B
B I
j head strongback M
W
(
m XIX In-Vessel Servicing Equipment a.
g CO P
1.
Control rod grapple
/
C,R 8
N/A N/A
'g HGf ng M
I f
-Q j a N (n N*
- o :d 3 XX Re[ur.11ng Equipment i
syg (A H I 1.
Re' fueling equipment platform 2
C N/A B
I hi j
assembly l
2.
Refueling bellows Other D
P/A N/A N/A j
3.
Fuel trenefer system 2/Other C,R B,D.N/A B
N/A (n) 4.
C,R B/D B
I (o) k 1
e----
d XXI Storage Equipment e
i 1
1.
Fuel storage racks 2
C,R N/A B
I
- u' lu N i
2.
Defective fuel storage 3
R N/A D
I 44 e>
container O
i O4 f
e I
l
R ID R W Christianccn General Electric GESSAR Project 6382-ROUND 1 QUESTION e
QUESTION /FESPONSE 410.31 (9.1.5)
QUESTION 410.31 For the fuel servicing equipment and cranes listed in Table 3.2-1 (Table 9.1-2) of your FSAR which are characterized as non-seismic Category 1, ver:.fy that they are designed not to be a missile sourr e Q as a result of a safe chutdown earthquake.
RESPONSE 410.31 s
The Fuel Prep Machine will be identified in Table 9.1-2 as Category 1 All other hoists, tools and equipment used for servicine shall either be removed during operation, moved to a location where it is not a potential hazsed to safety related equipment, or seismically restrained to 3 4 ok u n prevent 7: from becoming a missile. Sw
%%l. k.1 3 d
ra L 3.z-i
-~
13, C
9 e
6-9
j GESSAR 11 6miuus 238 NUCLEAR ISLAND RGv-4 Ot423 ru t s rvicias sauiv= ae Tho fuel servicing equipment described below has been designed in tecordance with the criteria listed in Table 9.1-2.
ITMS A/6T l.
!-H750 hs SEtSrvvc Gn 7d' gov 4 T Su c.h Pr5, Ho n sTS, WM R~O bin o-equv mer CJE.c <= cit settutud 6. Shat G 42.36 either be removaci during operation, moved to a location where it is not a potential hazard to safety related equipment, or <eismically restrained to
~
prevent it from becoming a missile.
i l
i O
)
O i
E Q 9.1-28a tr-
i O
~
O O
O O
Table 3.2-1
/0
[
EQUIPMENT CLASSIFICATION (Continued)
Quality l
Group Quality s"i""i Safetg classi d A"""#"" "o i
location fication pe<juireanent category Cous.or.t s j
Principal Component Class i :
l.
E XVI (Continue <3) i 4.
Pumps 2
A B
D I
5.
Pump motors 2
A' 8
D I
6.
Valves - outer isolation and 1,2 D
A/B B
I (9)
I within L
7.
Valves - return test line to Other 0,A D
N/A N/A (g)
):
condenaata storage beyond l'
i second isolation valve and vacuum punip dische.rga lina to j
containsent isolation valves
)
8.
Valves - other 2
C,A 3
D I
(g)
I 9.
Turbine 2
A N/A D
I imi 10.
Electrical modules with safety 2
A,y N/A B
I function 1
l 11.
Cable with safety function 2
D.A,X N/A b
I l
XVII Fuel Servicing Equipment 1.
Fuel preparations machirie Other C,R N/A B
~1 4D.M
]
2.
Gener'al purpose grapple Other C,R N/A B
N/l.
a A
4
i
\\
QUESTION 11 O
Explain the Quality Group Classification of the Hydrogen Recombiner System.
RESPONSE
The hydrogen recombiner system quality group classification is B as shown g
in the attached.*evised Table 3.2-1 in compliance with SRP 6.2.9 II(7).
l l
.)
l h*
s.
s O
s, t
O s
s
~
4 O
FSH:pab/J12117 12/11/82 7
_ - - -.. - - - _ _ -,, -.. -. ~ ~. - - - -, _ _. _. _.
t 4 11.
y.
Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued) i Quality Group Quality Safety Classi-Assurance seismic Principal Component" Classb Location fication Requirement
- Category Comments C
d XLVI (Continued) 8.
Electric 51 modules with safety 3
S N/A B
I functions l
9.
Cable with safety functions 3
S,X N/A B
I
^
w 10.
Diesel fuel storage tanks 3
0 C
B I
W
$o 11.
Starting air tanks and down-3 S
C B
I w
stream piping and valves og N
{
12.
Starting air compressor and Other S
N/A N/A N/A yw e
motors
.gg tn H h
XINII Combustible Gas Control System y
l
[B 1.
Ilydrogen Recombiner System 2
C B
I
$1/
2.
Containment /Drywell P!rge 2
A,C,E B
B I
System l
3.
Containment /Drywell Mixing 2
D,C B
B I
s System 4.
Fan 2
D B
B I
j w
l 5.
Fan motor 2
D B
B I
yy
-4 j
6 Electrical redules 2
D B
B I
o4 6
9 9
9 9
QUESTION 12 O
Justify the classification of the polars and cask cranes.
RESPONSE
The polar crane and the cask crane have been classified as Safety Class 3, Quality Assurance requirement B and seismically qualified to retain O
its functionai capantiity and maintain normai performance under conditions of 1/2 SSE (OBE) and to maintain structural integrity under SSE conditions.
Since their use is not necessary to accomplish any of the safety furstions which could occur during plant operation, no higher safety class is deemed necessary.
Safety Class 3 has been assigned to these cranes because of their location in the plant and safety systems they will lift, as well as possible failure causing collision with some safety related components.
Quality Assurance requirement B construction in accordance with the QA requirements of 10CFR50, Appendix B, is necessary because of the correlation of safety classes with the other design requirements as shown in GESSAR Table 3.2-3.
Note (X) on page 3.2-53 is being changed to read -
The cranes are designed to retain functional capability under conditions of 1/2 SSE and to retain structural integrity under i
conditions of SSE.
Table 3.2-1 page 3.2-40 and 3.2-41 will be reivsed to add Seismic Category I for these cranes.
Mark ups are included in the attachment.
O FSH:pab/J12117 12/11/82 27
Oh
-HIL Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued)
Quality Group Quality Assurance Sehmic safety Classi d e
g Principal Component Classb Incation fication Requirement Category Comments XINII (Continued) 7 Cables 2
D B
B I
8.
Containment /Drywell Atmosphere 2
D,C N/A B
I Monitoring System uw to XIN I T I Fire Protection System g
CO E
1.
Isolation valves and piping 2
C R
R I
(t) h within outmost isolation valves yy
.e
- o :o O
2.
Other piping and valves Other A,R,C,X,5 D
B I
(t)
HH tn H 3.
Pumps Other O
D B
I (t)
ZD 4.
Pump motors Other O
D B
I (t) i 5.
Electrical modules Other A,R,C,X,S D
B I
(t) 6.
CO a tuation modules 3
S N/A B
I (t) 2 7.
Cables Other 5,R,C,X,S D
N/A I
8.
Sprinkles Other A,R,C,X,S D
N/A N/A N
- o N XLIX Miscellaneous Components cJ lp HGB
.o l.
Containment polar crane 3
C N/A B
.J.
(x) p / 2.
O o
e e
S Table 3.2-1 EQUIPMENT CLASSIFICATION (C'ontinued)
Quality Group Quality Assurance Seismic Safet Classi d e
g i
Prinelyal Component Class location fication Requirement Category Comments XI.IX (Continued)
HGB 7
tt) 2.
Cask crane 3
R N/A B
L Emergency Lighting Other A,C,X N/A N/A N/A ww co LI Ileating, Ventilating, and
$o Air Co;?nitioning Systems l
O t9 b
Y (Applicant to Supply) e H
WW e
HH tn H LII ECCS Equipment Area Cooling hy 1.
Fans, flow meters, ducting 3
A N/A B
I valves, heat exchanges, chilling units with safety I
function 2.
Motors 3
A N/A B
/I l
3.
Instrumentation and controls 3
A N/A B
I
)
with safety function 4.
Electrical modules 3
A N/A B
I u
5.
Cable 3
A H/A B
I gy
.o b
G.
G 9
O O
GESSAR II 22A7007 238 NUCLEAR ISIAND Rav a O
Table 3.2-1
- S f1 O
EauIPr1ENT CI2SSIPICATIoN <<:eneinued)
NOTES (Continued) 2.
A certification shall be,obtained from the manufacturer of these valves and steam leads that the quality control program so defined has been accomplished.
' O w.
The condensate storage tank will be designed, fabricated, and tested to meet the intent of API Standard API 650.
In addi-tion, the specification for this tank will require: (1) 100% surface examination of the side wall to bot Mm joint.
and (2) 100% vo'lumetric examination of the side wall weld joints.
d M
The cranes are designed to hold up their 1 a &,s under condi-x.
tions of 1/2 SSE and to mel;.t:n.+u u s f. 0 :.7the tur <
9n6 1.7 t pus m uu-v m the T j7.
h unta. under mnditions ofv5SE.
s e
l O
O 3.2-53/3.2-54 h.
-,---w--.,,---
,-,,,.---,.c.--
,g-,
-,,c,
,,-,_.,.,.,------,n,---7,
,,y,,,vy,,
,7--,
i QUESTION 13
' O What is the natural frequency of the charcoal absorber tanks, including the support elements? How are they affected by suppression pool load:;?
RESPONSE
This tank is located in the turbine building and is not affected by new Q
loads.
(See co^mmT F
for further detail).
1 i
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O O
O FSH:pab/J12117 12/11/82 U
j P
QUESTION 14
)
What is TEMA.
i i
l
RESPONSE
TEMA stands for Tube Exchanger Manufacture Association.
The C next to TEMA stands for class C.
lO 4
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~
O 1
f O
FSH:pab/J12117 12/11/82 lp l
l QUESTION 14a O
Provide a list of all high energy lines.
RESPONSE
A list of all high energy lines are given in Table 3.6-6 :nd 3.6-7 for high energy piping inside containment and outside containment respectively.
O O
i O
O FSH:pab/J12117 12/11/82 y;s
i QUESTION 15 O
Section 3.6.2.1.4.2 The design stress and fatigue limits for Class 1 piping in the containment penetration areas are not in compliance with Standard Review Plan 3.6.2 and BTP MLB 3-1.
If the maximum, stress range of Equation (10) exceeds 2.4 Se, both Equation (12) and (13) must be Less than 2.4 Sm.
The cumulative usage factor must ha Less than 0.1 even for Equation (10) Less O
'"'" 2.4 S..
RESPONSE
The response to this request is provided with the revised Section 3.6.2.14 to comply with SRP 3.6.2 and BTP MEB 3-1.
' O O
O u
v--
--vv-wrw--wwww--- - - - -"-
-ws-w------v,-v--r--------w*M--ww
,-ee* - -ww*
ww-w--www wwwwwm1M==*ww eew?s eg wyow-ew-m--F---w
-~~ ' -. ~ -
~
___ T m
GESSAR II 22A7007 238 NUCLEAR ISIAND R;v. 0 M,-
{'T 3.6.2.1.4.2 Piping in Containment Penetration Areas (Continued)
(\\
7he, ctagulATNO.dAdHrfACTOR. ' C /l be l.W5 h 8. /
g53 m ur (b)
Tf ae cula ed m 4
tress ran of Equa-g7 han t,
n 10 ce s2 b
s
.USc.
ot gr ater/
en ac ulat.ve u ge factor "s les\\s' thal
'1. t.
' I (c)
If the calculated maximum stress range of Equa-2+
tion 10 exceeds -S,4Sm, then the stress ranges cal-culated by both Equation 12 and 13 of NB-3553 do
\\
not exceed 2.4Sm.
e t-
'- 1::_ M r?. ~rr (d)
The maximum stress as calculated by Equation 9 of s -3652 under the loadings resulting from a postu-lated piping failure beyond the required restraints does not exceed 2.25Sm.
Higher stresses between the isolation valves and restraints were permitted provided. a plastic hinge was not formed and oper-ability of the valves with such stresses was assured.
(2)
For ASME Code Section III Class 2 piping, the fol16 wing stress and fatigue limits are not exceeded.
(a)
The maximum stress ranges calculated by the sum of Equations 9 and 10 of NC-3652 for normal and upset 5
plant conditions (including an operating basis Q
earthquake) does not exceed 0.8 (1.2Sh + sa).
(b)
The maximum stress as calculated by Equation 9 of NC-3652 under the loadings resulting from a l
postulated piping failure beyond the required restraints does not exceed 1.8Sh.
Higher stresses
\\
between the isolation valves and restraints were g
permitted provided a plastic hinge was not formed and operability of the valves with such stresses 3.6-17 Q
,y
-y yv.r------.*-r-'-
... - a-s.
QUESTION 16 O
it is stated in the 854a that weided attachments to the pi,e are avoided except where detailed stress analysis or tests were perfomed to demonstrate compliance with the stress limits given.
Provide a list of all instances where this exception was taken and welded attachments were used and the re >ults of the analyses.
RESPONSE
O i
In the containment penetration area between isolation valves, only pipe head fittino or containment nozzle is (ttached to the pipe to form an Gs.4tuaesmoT anchoro Both types of attachments are required to be analyzed by detailed analysis method and documented in an ASME Code Stress Report.
h ws C /. l. 4. Z eWG W
c6.
as MJ u A. cAJ mf Le G JT v& x c6 "W-
- L
' 'l ' '
O ud mA s g et c esset
,mo w.s O
=
0 20
s.
GESSAR II 22a7007 238 NUCLEAR ISLAND R,v.
0
- 1b 3
Piping in Containment Penetration Areas (Cor.tinued)
O
.6.2.1.4.2 was assured.
When the piping beyond the isolation valve is constructed in accordance with ANSI B31.1, this exception may be applied provided the pipe is either of seamless'tonstruction with, full radi-ography of all circumferential welds or all longi-tudinal and circumferential welds are fully O
radiographed.
(3)
The piping runs are stre.ic;ht.
(4)
Weldedattachmentsfg.rpipesupportsorotherpurposes g
ise taa.
s gyx -as.o. avoided crc:pt whc;c detailed stress analyses or ygL tests were performed to demonstrate compliance with the stress limits given in items (1) and (2).
l (5)
The number of circumferential and longitudinal piping welds and branch connections are minimized.
Where guard pipes are used, the enclosed portions of iping are of seamless construction and have no circumferential welds unless specific provisions for access is made to permit 100% inservice volumetric examination of all welds.
(6)
The length of these portions of piping arc reduced to the minimum length practical.
(7)
The design of pipe anchors or restraints (t g., connec-tions to containment penetrations and pipe whip restraints) do not require welding directly to the outer surface of the piping (e.g., flued integrally-forged pipe fittings may be used) except where such welds are 100% volumetrically examinable in service and a O
9 3.6-18
^
J 1-MTES M
i i
'e
=
i
/4 INSPECTABILITY OF WELDED FLUED HEAD DE$1GN ON MAIU $ TEAM LINE CONTAINMENT PENETRATION O.
ISSUE:
The inspectability of welded flued head design on main steam line contaircent penetration should be demonstrated via the fo11'owing activities:
O 1.
Verify that the plant configuration allows adequate accessability to the penetration to perform necessary inspections.
2.
Determine if the penetration weld was ultrasonically examined during manufacturing.
If to, report on examination results.
3.
Deternine if additional details exist on the welded flued head inspectability demonstrati,ons perfonned at the Associated Pipe and Engineering facility in 1976 and 1977 and documented in General Electric Company Topical Report NED0-23652, " Analysis on General Electric Designed Welded Flued Head Fitting at Containment Penetration Assently and Provisions for Nondestructive Examination of Flued K:ad Fitting to Process Pipe Weld for BWR/6 Mark III -
~
~
218, 238, 251 Plants".
LRG-II RESPCriSE:
The inscectability of the main steam containment penetration for LRG-II plants with welded flued head design (i.e., Clinton and Perry) has been verified as follows:
)
1.
Main steam line containment penetrations are readily accessible to allow the performance of inspection activities. Clearance around the penetrations is sufficient to permit the use of inspection equipment. Insulation on the rain steam line was designed to be removable in the area of the welds.
l
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MJA:1m/13K-1 1/25/82 l{
i
^ ~ ' '
1 1-MTES (Pege 2)
,r i
2.
Main steam line flued head penetrations were ultrasonically and Q
radiographically examined as part of their manufacturing process.
These examinations revealed no indications requiring repair.
3.
In July 1976. Cast ral Electric Company conducted use feasibility study of pulse-echo altrasonic testing (UT) to assure ful? volum coverage of the flued head attachment weld. This TIT exanination technique was repeated as a demonstration for utility, O
architect-eng4nor and ac representat4ves dur4ng my im and May 1977 at the Associated Pipe and Engineering facility in Compton, California. The results of the demonstration are documented in the draft report NED0-23652. No additional documentation on the demothstrations perfenned is available.
The main steam kine penetration of River Bend Station are not of the welded fhed head design; they are forged.
O O
HJA:1m/13K-2 O
t/25/s2 1
20
-~'~:J L -
QUESTION 17 The maximum stress range as calculated by the sum of Equation (9) and (10) 30 auld consider sustained loads, occasional loads and thermal expansion.
Have occasional loads been included as per BTP MEB 3-17
RESPONSE
ffg Yes, the occasional loads such as thrust from relief valve and safety valve, flow transient, and earthquake have been incleded as requirM by NC-3652 ASME Section III.
O s
O O
FSH:pab/J12117 Aq 12/11/82
QUESTION 18 O
Whenever two or more intermediate break locations are not selected based upon the stress or usage factor limits, a total of two intermediate locations should be selected based upon highest st..esa.
?covide assurances that highest strus was the criteria used to select these intermediate locations.
"55" "S" O
For Class 1 piping pipe break analysis procedure regnires al) intermediate break locations to be postulated by highest stress method.
The reasonable basis referenced in Section 3.6.2.1.4.3 will be reivsed to highest stress basis.
Q l
O d
O FSH:pab/J12117 f./
12/11/82
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O l
i 3.6.2.1.4.3 For ASME Code,Section III Class 1 Piping
%)
Breaks are postulated to occur at the terminal ends
- of the piping run or branch run.
In addition, breaks are assumed to occur at any intermediate locacion between terminal ends where:
(1)
The maximum stress range between any two loads (includ-ing the zero load set) as calculated by Equation 10 of
()
NB-3653 for normal and upset plant conditions (including an operating basis earthquake) exceeds 2.4Sm but is not greater than 3.0Sm and the cumulative usage factor is greater than 0.1.
(2)
The calculated maximum stress range of Cquation 10 exceeds 3.0Sm and the stress ranges calculated by either Equation 12 or 13 of NB-3653 exceed 2.4Sm or the cumu-lative usage factor exceeds 0.1.
p)
A-In tha event that two or more intermediate locations cannot be determined by stress or usage factor limits, a total of two intermediate locations are identified on a reasonable basis ** for each piping run or branch run.
- Terminal ends are extremities of piping runs that connect to structures, components, or pipe anchors that are assumed in the piping stress analysis to act as rigid constraints to piping thermal expansion.
A branch connection to a main piping run is a terminal end for a branch run except when the branch and main run is modeled as a common piping system during the piping stress analysis.
4 gj, j
- Reasonable casis is ^== or _cth
^#
-"e
- ^ 11maing :-
b (EF fitting lccrtienc, and/cr p if (2) 'ighest ;tzces vu usag: factar inc'
- 4 o n s,-
Where more than two such intermediate locations are possible, using the application of the reasonable basis, those two loca-tions possessing the greatest damage potential is used.
A break at each end of a fitting may be classified as two discrete break locations where the stress analysis is sufficiently detailed to differentiate stress at each postulated break.
)
kk 3.6-20
QUESTION 19 f
The paragraph labeled (1) indicates that breaks are postulated at ends tt;at are lightly stressed.
Please modify this paragraph to reflect a more reasonable criteria.
RESPONSE
(1) Breaks are postulated to occur at terminal ends.
In addition, breaks are postulated at each location whe.
the maximum stress ranges calculated by the sum of Equaticos 9 and 10 of NC-3652 for normal and upset plant cer:ditions (incir41rg and operating basis earthquake) exceeds 0.8 (1.2bSHfSa). At least two intermediate break locations are postulated based on highest stress criteria with the following exractions. Where the maximum stress ranges of a straight run without fittings, welded attachment, or <alves does not exceed 0.8 (1.25ShtSa), at least one intermediate break location is postulated based on highest stress criteria.
O O
l FSH:pab/J12117 g_.,
12/11/82
~'
GE55AR II 22A7007 238 NUCLEAR ISLAND Rsv. 4 h !h 3.6.2.1.4.6 Piping in the Steam Tunnel and Seismic Interface O,
a tr i== (c==*i=u a)
The criteria applied to the main steam, feedwater and branch
\\
piping outbcard of the containment isolation valves ar_e as p
s h addi-dea A 5
follows:
6thMd -
l(
Breaks are postulated to occur at terminal end where the maximum stress ranges calculated by the zum of Equations k'
{
9 and 10 of NC-3652 for normal and upset plant conditions l (including an operating basis earthquake) does not expeed At least e+ne intermediate break loca-
~
g 0.8 (1.25h + Sa).
Iv Mf *M EMS C'Ib" M d
tion is postulated M fesf c2
+, 4 4 9p a p nup nc.9 E(2)
Breaks are postulated in high energy branch lines connec-1{
tad to the main steam and feedwater lines.
All branch h
lines are restrained to the extent necessary to protect 3
i g g
,g the containment isolation valves from any loadings, jet o
3 1,
j impingement, pipe whip and other forces that could
- 8 M impair valve structural integrity and operability.
The criteria used to evaluate the seismic interface restraint are as follows:
(1)
Mechanistic pipe breaks are considered everywhere along the main steam and feedwater lines between the outbosrd
(
containment isolation valve and the seismic interface restraint.
(2)
Mechanistic pipe breaks are postulated at locations in the main steam, feedwater and branch line piping where stress limits and usage factor limits are exceeded.
For such postulated mechanistic breaks, the associated dynamic effects are evaluated including jet impingement, s
(
pipe whip, flooding and pressurization.
O 3.6-21a 05 e
e
,y
2L'ES110N 20 No breaks are postulated at the RHR branch line connecting to the feedwater line since this branch line is included in the piping dynamic analysis for the feedwater line. What are the maximum stress ranges caiculated by the sum of Equations (9) and (10) of NC-3652 for normal and upet plant conditions and e at is the cumulative usage factor at the intersection?
RESPONSE
The paragraph labeled (4) of Soction 3.6.2.1.4.6 will be labeled as (3) and moved to the section under the criteria applied to the main steam, feedwater and branch piping outboard of the containment isolation valves.
It will also be revised to read as follows:
breaks for RHR branch line are postulated according to the requirements of paragraph (1)
Terminal end is net considered at the RHR branch line connecting to the feed water line bacause it is included in the piping dynamic analysis of the feed. water line.
O O
1 O
FSH:pab/J12117 Op 12/11/82 l
GESSAR II 22A7007 238 NUCLEAR ISLAND R0v. 4 (l6.2.1.4.6 Piping in t*ce Steam Tunnel and Seismic Intarface Restraint (Continued)
(3)
If application of steps (1) and (2) does not result in postulating a mechanistic pipe break, a.n intermediate (non-mechanistic) break is postulated.
The steam tunnel is. designed to withstand only the environmental effects M
of the intermediate (non-mechanistic) break (i.e.,
hw _.,...
,,=_,c..
r-s.- --- y~ ~ --- - z r-e 3y --
(
au.
2 d _ : - -
- s-s* - -
rs, 7 piping g
hy +
4a.1=meie,cipe whin ma=1y=i,
= " h e c
.e
-4j 2 )
N s d,,,, #.,.,_ -
(2)
The dynamic force of the jet discha e at the breth
~
location is based on the % ;ti:. cross-sectional flow area of the pipe and on a'c'alculated fluid pressure as modified by an analytically-or experimentally-determinad s
O
- '"*'"****2'P-'.-=~==-*==*-^-1"~'---
~
Wr--6 la-=*4aajiinerestrictions,flowlimiters,posi-tive W y-c ntrolled flow, and the absence of energy reservoirs are taken into account, as applicable, in the reduction of jet discharge.
//r-fr arm % s 4 46 A//4,= 4444 4-/73 (3)
A^ rise time not exceeding one millisecona is usea zor A the initial pulse,-' c-- le ;e; c-r' p y 3.dw _%
l
,,, a......:
e-
--e s_.
3----
7 -- ~~~ w p
p
'/
J q ; ; ',--' {I.= 1 thev v _ '
--. 4 =- -
- m i' A *-
OElowdown forcing functions are determined by either of the two following methods:
(1)
The predicted blowdown forces on pipes fed by a pressur,e vessel are described by transien*,and steady-state O
3.6-26 tb
--.---r-c%,..,, _ - -,,,
-m g._
.g
,y-my.-
QUESTION 22 On Page 3.6-27, Paragraph (b), clarify when reflections at bends and elbows will be assumed.
RESPONSE
No reflections at bends and elbows are assumed. Wave reflections will Q
occur at the break end, and the pressure vessel until a standby flow condition is established. Vessel and free-space conditions are used as secondary conditions.
The pipe is modeled as a straight, uniform piping fixed at one end and subjected to a time-dependent thrust force at the other end. The pipe-bending-moment-deflection (or rotation) relationsd8 used for these locatica is obtained from a static nonlinear cantilever -
beam analysis.
Further clarification is also provided in revised Section 3.6.2.2.1.
O
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FSH:pab/J12117 12/11/82 If
.-_.,.--..,_.,,._-,------,y,,
mw..y--e-
.-.w-,-
GEsfML II __
22A7007 238 NUCLEAR ISLAND Rey, 4 l
I dd
- y, s..i. Analytical Methods to Define Blowdown Forcing Functions (continued) j forcing functions.
Simply, they are as fo1 Tows:
i (a)
The transient forcing functions at points along the pipe result from the propagation of waves (wave I
thrust) along the pipe and frost the rei.ction force due to the momentum of the fluid leaving the end of O
the,i,e (h1.wdown thrust).
(b)
The waves cause various sections of the pipe to be i
loaded with time-dependent forces.
It is assumed that the pipe is one-dimensional in that there is no attenuation or reflection of the pressure waves at bends, elbows, and the like.
Fc11owing the rupture, a decompression wave is assumed to travel from the i
break at a speeG equal to the local speed of a,ound within the fluid.
Wave reflections will occur at l peg the break end. ^ ";-- i cf +!. @ gc =4 gg
&c pr rrur
- c = ci w * * ** 2 :L-_i; f1-:+ __ ::_:_-
he-eembMehed, vessel and free-space conditions are used as boundary conditions.
The blowdown thrust causes a reaction force perpendicular to the pipe break.
(c)
The initial blowdown force on the pipe is taken as the sum of the wave and blowdown thrusts and is equal to the vessel pressure (P,) times the break area (A).
After the initial decompression period (i.e., the time it takes for a wave to reach the first change in direction), the force is assumed to drop,off..to the value of the blowdown thrust (1,607(A).
s.
i
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. +. -
j,/
O' 3.6-27 s
h4
.. -.....=. - - -.
QUESTION 23 A thrust coefficient of 0.7 was used which is less than 1.26 as specified in Standard Review Plan 3.6.2.
Please provids justification for this lower value.
RESPONSE
Q The response to this request is provided in revised Section 3.6.2.2.1.
O f
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l 0 FSH:pab/J12117 12/11/82 fj
238 NUCLEAR ISLAND Rev. 4 N
Y G, 9 3
Analyticel Mcthods to Define Blowdown Forcing Functions O
.6.2.2.1 (Continusa)
. forcing functions.
Simply, they are as follows:
s (a)
The transient forcing functions at points along the pipe result from the propagation of waves (wave thrust) along the pipe and from the reaction force O
au
<o **
mo atum or ta riuta 1 vias th aa or the pipe (blowdown thrust).
(b)
The waves cause various sections of the pipe to be loaded with time-dependent forces.
It is assumed tiat.the pipe is one-dimensional in that there is no l
attenuation or reflection of the pressure waves at bends, elbows, and the like.
Following the rupture, a decompression wave is assumed to tiravel from the break at a speed equal to the local speed of sound O
HI-9 within the fluid.
Wave reflections will occur at 21 i by, y the break end, W ;- p a a pMtJ6 f yr the r,r u ce verselustil-e s.taady=fictw-conditia i= --*=Flich D vessel and free-space conditions are used as boundary conditions.
The blowdown thrust endses a reaction force perpendicular to the pipe break.
(c)
The initial blowdown fcree on the pipe is taken as s
O
- ==== or th
=va ^=a 51ovao-a thr==== =aa i
aqual to the vessel pressure (P ) times the break g
area (A).
After the initial decompression period (i.e., the time it takes for a wave to reach the first change in direction), the force is assumed to drop off to the value of tne blowdown thrust 4g (i.e., 0.1 P A). Mit $sf la aff* tied of *.7 2. <6k:sa./
n M;& t$.s j ;g:/,,, [,) La,,
,;,,,, /,
.a. T Ajh ths pp l<lra.
e:
, S, c,,m Jgg.j,,
cAj,Lt yygn,,rt,.n a.
,.,L/_,_y, x
( v Lks h J) < F A 7.afrauf chal.4, /Jo. wyu./ 1,
- 1 nadL. p/ ;r.
u & te. d4 7 3.6-27 i D r:.
s 4
i s
N x
QUESTION 24
's
~
O
^.
Page 3.6-36, paragraph (7).
Staadard Review Plan 3.6.2 does not allow the use of a jet expansica zone for saturated or subcooled water bloMown.
Justify this discrepancy as evident in Figure 3.6-4. '
RESPONSE
O TheresponsetothigreqhestisprovidedinrevisedSection3.6.2.3.1.
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4 s
O t
w O
l FSH:pab/J12117 ' '
12/11/82 47
,-,r
,-,,_-__,-,..-m
,-,,em
-,.--.----,_m m,m q.m, n--_ve e,,
m,,
e m,
,,-,,-mwm g-
_m,p
-GESSAR II 22A7007
?
238 NUCLEAR ISLAND rov. 4 O.2.3.1
,ee Impingemene Ana1yses and Effects on Safety-Re1ated Components (Continued)
~
s O
(6)
The jet impingement force is equal to thw steady-state value of the fluid blowdown force 3 calculated by the methods described in Subsection 3.6.2.2.1.
The distance of jet travel is divided into two or three O
(7) regions.
Region 1 (Figure 3.6-4) extends from the break to the asymptotic area.
Within this region the discharg-ing fluid flashes and undergoes expansion from the break area pressure to the atmospheric pressure.
In Region 2 the jet remains at a constant diameter.
For partial-separation circumferential breaks, the area increases as the jet expands; therefore, it is assumed that Region 3 never occurs.
In Region 3 (except partial-separation circumferential breaks), interaction with the surrounding Q
environment is assumed to start and the jet expands at a I
half angle of 10 *. ( A#d h4 s, 'Frgs %7 d.(-e a 4 / c ),
NE8 i
6 dez:1;pl..*.yl., analytical model for esti-(8)
"^^Ay h===
det area for.L< h) krate *-4 s'
44eem, saturated mating the asymptoti
^
surw'.r a c.4rta4r t
-/=*=g=d~n : nditi:n:.
For fluids are.
- water, 2-discharging from a break which are below the saturation temperature at the corresponding room pressure or have a pressure at the break area equal to the room pressure, the ie4&e& free expansion does not occur.
- n th;;c eases,.65e jet ::n he ess a d te w ond om. 10" u.
f
=v-e.
Octtad lia.., Fir eugle sterting :t 'le bra u urca,3.' 4, e oud W.
(9)
The distance downstream from the break where the asymp-totic area is reached (Region 1) has been found by Moody (for circumferential and longitudinal breaks) to be O
aggroximete1y eauet to five give diemeeers.
assumine a linear expansion from the break area.to the asymptotic g
3.6-36
=
QUESTION 25 O
Pages 3.6-47 and 48.
Provide assurances that in no instances does the value of strain exceed that associated with 50% of the ultimate uniform strain energy absorption capacity as determined by dynamic testing at loading rates within 50% of the specified design loading rate as required by Standard Review Plan 3.6.2.
RESPONSE
The response of this recuest is provided in the revised Section 3.6.2.3.3.
7kitissf.r4 wi// 4llhin <fx kr %fsi v% of ff.g., $l<ull>4 Cltsay4 joadief Pabr Mf its d A, [gy }, f,1, A.C O
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O FSH:pab/J12117 12/11/82
c_
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~~
22A7007 GESSAR II j
R;v. O h@,*'.
238 NUCLEAR ISLAND Loading Combinations and Design Criteria for Pipe Whip i
.3.3
,f, destraints (Continued) uniform strain for all materials which will i
be used for Type I' components.
D4f 4. ire {
=m
__ y I The bgg Dy amic material mechanical grope mata selseted exhi ensile impact
,q g properties w te -ar's not less than 70% of the
./ -
static percent elcnga
. or 90% of the stat' Iydeterminedminimbtotalenergy N
/
Ni tion.
/
Type II restraints (e.g., clevises, brackets, pins)
(2)
(a)
Materials Material selection conforms to:
)
ASIM Specifications including consideration 1.
for brittle fracture control, or ASME Code Section III, Subsection NF, B&PV.
2.
Code if applicable.
i (b)
Inspection Inspection conforms to:.-
ASME/ ASTM requirements or process qualification 1.
and finished part surface inspection per ASTM methods, or ASME Code Section III, Subsection NF, B&PV 2.
Code if applicable.
3.6-48
-ff
~~
_ +
-='e-T* * =-~----w'v-P-**w-*'r~"D-e*7w---~y
-C*'*h
-wvy g-r=9 er-y-me-+
y-..
s-mp y
--w-%gsww----,w-w-
,w
,-,-+ww------v--i-,-+-m,--,w-, - ---*T "N
--W't-~*
~~ ~
QUESTION 26 O
Provide details of the seismic guides referred to in paragraph (5) on page 3.6-52.
RESPONSE
Q p~ad a c.p=y,_ g,-f uta.,
e n.- a w
(-< w " 4 Ly tl.< asc c/- ti r H E B d-r<f t-S ER uh p
C e
O l
O
QUESTION 27 Standard Review Plan requires the review of sketches showing the loc.stions of the postulated pipe ruptures, including identification of longitudinal and circumferential breaks, structural barriers, if any, restraint locations, and the constrained directions in each restraint. Also to be reviewed are the data developed to select postulated break locations l
including, for each point, the calculated strecs intensity, the calculated
{Q cumulative age factor, and the calculated primary plus secondary stress range.
Prcride these sketches and data in the FSAR.
RESPONSE (27 %
4 As noted in the 3/32/81 GESSAR FDA submittal letter (Sherwood to Denton),
one of the NRC concerns regarding FDAs includes anti-trust considerations.
The issue is the potential problem as whether the level of detail required by the staff during the FDA review will dictate the use of particular equipment vendors in C0/0L applications that reference the FDA.
This is O
of primary edncern particuiariy in the " buy" area which is mostiy outside the NSSS. General Electric has elected to be responsive to an.ti-trust concerns by identifying specific quantities as " Applicant to Supply" for those quantities which are equipment vendor dependent.
Further, General Electric has also elected to have the hydrodynamic loads
-handled on a site specific basis.
Thus, even if there anti-trust was not a issue, many of the equipment and piping quantities would not be available generally and these quantities would have to be supplied by the applicant.
In summary, the sketches and data requested by Question 27.and the information requested by Question 42 cannot be supplied at this time, and must be supplied by the Applicant in his FSAR.
This will be included in Section 1.9 as an interface requirement.
O FSH:pab/J12117 i 12/11/82 (L
QUESTION 28 Breaks in non-nuclear high energy piping not seismically analyzed (nor j
qualified) should be postulated at those locations which produce the greatest effect' on an essential component or structure irrespective of j,
the fact that the high stress or fitting criteria might not require that the break be postulated.
Provide assurance that the above criteria has been eet.
O
RESPONSE
17/t-switn Gib+PH
.3. &. Z. /. 4 S diU 8E REVrs50 &$ & 2:
.3. 4. 2. l. 4. 5 A AL A Ms i A 31. I t*/PINCr Rwk 1 ARE roosrulA TEo s 'r 7H& h LlosJ /H G l ockfl0A/S
/hl E4cH pint Rud en Rahuck !%uN :
U) 47 Tett HIMA L E+/o s o F-77/6 fu& iF-LecArs:o_ AuTAcsdr THN prersevolt Q
l srRuerune rn C2) AT &ACM /d7&24 Est4TE pob2 7,177?MG, WELOEO A77AchfMEN7"i MC VALVE rp. THE PvP/A/Cr / $ NOT SE/3 4* cal' Y AAAl'r E & D.
$957UlArEO BMso cA rNE.. SANE ces7ER/A Sp5CoHEO /^' P4240*APH
- 3. 4. 2./.4 4.
CHw(1s THC-E-xn7/N G P4eAmehpH 3. 4 -2. /. 4.. d To A NEsj' O
,iann
- 3. s. z. i. r.. r.
ADO A ^few' 3.4 2. / 9. S PAR 4rreAP/-l R. C. 2. /. 4.. d, AmlicAelE ro 3.C.z./.4.3,.R&.2.l.4.4, Aac
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- 3. 4. z.. /. 4. S 1 O FSH:pab/J12117 12/11/82 8/
1.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 3.6.2.1.4.4 For ASME Code Section III Class 2 and 3 Piping
(-O Breaks are postulated to occur at terminal ends of the piping run or branch run.
In addition, breaks are assumed to occur at any intermediate location between terminal ends where the maximum stress ranges calculated by the sum of Equations 9 and 10 of NC/ND-3652 for normal and upset plant conditons (' including an operating basis earthquake) exceeds 0.8 (12Sh + SA).
O If two or more intermediate locations cannot be determined by stress limits, a total of two intermediate locations are identified en a reasonable basin for each piping run or branch run.
3.6.2.1.4.5 For ANSI B31.1 Piping u &l L 1 E- @
~
ar=4"=1-ende vs v.ne piping runt Breaxs ase y N1_ated to occur at s are assumed to occur at each }
'or branch run.
Tn lo,
pe fitting, weld attachmen, ani ra_1ve. [
M z) 3.6.2.1.4./ Piping in the Steam Tunnel and Seismic Interface Restraint As described in Subsection 3.8.4.1.2, a seismic interface restraint The is provided for both the feedwater and main stec.ulines.
restraint is located in the steam tunnel between the shutoff valve (in the main steam and feedwater lines) and the Auxiliary Building /
Turbine Building interface (Figure 1.2-9).
The seismic interface restraint serves as a guide for the main steam and feedwater lines and provides two-dimensional lateral restraint and moment-carrying I
The seismic interface restraint provides a seismic capability.
category transitional point between the Seismic Category I steam and feedwater piping in the Auxiliary Building and the portions of these pipes in the Turbine Building which are not designed to Seismic Category I (see Figure 3.2-1).
l J
O 3.6-21 gj
,y-_---_--.m_
+~
_%.-.,w,,9
'ra n nr @ Ta Jap 9 6 -21 n a.? # 2 s.
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/3rwk.c a <-c. f-1d4 h./ n.b t/g/4wi.j. hed.4, f :/-R-Y4h a r b r e +,<k / u 4 -
E*
- k t h + ;u/ o. h.7 i L
,-8
.:,1 /u LJ a>
& d ' A c 2.hf
-l--o k.a. l raf.a e frv.nC hru ebur2.
(2) kf-eaef fA.r /n.<alls's a-ff*-l:lll"},W4{../.aj affahrJ, ud vala -y ;4 - p y:hy a ut s's ;e h, icJ, a f.e-..t.
U)
- /,v S z is 4,; a /l
^ 8 < l774,o y np.,
n f= rh,ad.-k y
bruh3 fos Nlhanl $ & ch h S a hs x 0 Y.a W:
is Sf' c il v-f E Yapyafk d.d 2 -l k V, 2'S 2 / % 4 Sja,Jr., sin cins w :d H.pl Gym L.;u s krAfre
-b-->
l % - G Ajyl:c-!!r
)., ;. f,,..j. g _
O S D _- } - u u m J >
f. >. / v.9=D A
If a structure separates a high energy line from an essential' h separating structure
$ddbedesignedto component, withstand the consequences of the pipe break in the high-energy line which produces the greates,t effect at the st$teture irrespective of the f act that t$$"McNtaria L 3 c2,.d-5 might not require such a break location to be postulated.
O 4
1
g,,
m,un, 238 NUCLEAR ISLAND R;v. 0 3.6,1.3.2.2 Separation (Continued) requiremeyits. M mage was assumed to occur due to jet pg impingemeitt' Wi:he impingement force becomes neg-Sq
~
ligible beyond 30 feet.
M.
n.
e reces-o r A
-g e r e er a. m a n../
i u[ma x. new.,atingh". N M P'N bsacasic=QW Tv 4 sserte AG0irv tr
~
~..g,g ej i A w-tree'4 Pier. A
- 77tur so a vr Rana smis TNe nax,or.a nec,r uwcyx m e ac swo 4; time oe
>,ge,,,e i m r-
~
Essential systems, components, and equip.ient at a N #"
(3) distance. less than 30 feet from any high-energy piping were evaluated to see if camage could occur to more than one essential division, preventing safe shutdown of the plant.
If damage occurred to only one division of a
-r'edundant system,'the requirement fcr redundant separa-tion was met.
Other redundant divisicas are available for safe shutdown of the plant and no further evaluation was performed.
(4)
If damage could occur to more than one division of a
_/~
redundant essentiai system within 30 ft of any high
(
energy piping, other protection in the form of barriers, shields, or embedments was used.
These method of pro-tection are discusse,d in. Subsection 3.6.1.3.2.3.
Due to the complexities of several divisions being adjacent to high-energy lines in the drywell and steam tunnel, the require-mants for separation could not be' evaluated using these simplify-ing assumptions.
For these areas, specific break locations were datermined in accordance with Paragraph 3.6.2.1.4.3.
If spatial O==9eracioaresu1=e=eae=(41=ee=c=e=e/ ora==ase=e==eo9eveae demage) were not met based on the evaluation of specific breaks, barriers, enclosures, shields, or restraints pas necessary.
These mathods of protection are discussed in Subsections 3.6.1.3.2.3 and 3.6.1.3.2.4.
p.
'\\L) 3.6-9 y
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 lll3.6.2.3.2.2 Pipe Displacement Effects on Safety-Related Structures, Other Systems, and Components (Continued) pipe.
Otherwise, the impacted pipe is assumed to be ruptured.
'~
e4 (3)
If the whipping pipe impacts other compenents (valve N
actuators, cable trays, conduits, etc.), it is assumed that the impacted component is unavailable to mitigate h @)
the consequences of the pipe break event.
t>
blIS
'rNSi=R7@
3.6.2.3.3 Loading Combinations and Design Criteria for Pipe Whip 4I*9, Restraints
_g30 Pipe whip restraints, as differentiated from piping supports, are dcsigned to function and carry load for an extremely low-probability gross failure in a piping system carrying high-energy fluid.
The piping integrity.does not usually depend on the pipe whip
(__,1catraintsforanyloadingcombination.
When the piping integrity 10 lost because of a postulated break, the pipe whip restraint acts to limit the movement of the broken pipe to an acceptable distance.
Tha pipe whip restraints (i.e.,
those devices which serve only to -
control the movement of a ruptured pipe following gross failure) will be subjected to once-in-a-lifetime loading.
For the purpose of design, the pipe break is considered to be a faulted plant con-dition and the pipe, its restraints, and structure to which the rcotraint is attached are analyzed and designed accordingly.
thapipewhiprestraintsutilizeenergyabsorbingU-rodstoatten-uate the kinetic energy of a ruptured pipe.
A typical pipe whip rcotraint is shown in Figure 3.6-7.
The principle feature of these rcotraints is that they are installed with several inches of ennular clearance between them and the process pipe.
This allows for installation of normal piping insulation and unrestricted pipe Select critical locations inside primary con-()tharmalmovements.
ainment are also monitored during hot functional testing to pro-vide verification of adequate clearances prior to plant e seration.
sy 3.6-44
/
O M PR pLL.?e O
1Ncar@ v py 3.i-au t+> Danyc q ~,J,,o.J ut.y.y ?yn sg -
as YAlaked Slrachrzt, Ginf ban k s d syrb nr o Nor-tdu 9
& r y t-<,s o s a siad 2x 7 7-m b J
.l,
-e id-r ry nhg,y A.pl-n. -p; mskr e f-
-ecs u b / - p h m s p r.d i,y 7 y e L y or vs. s h e d r.
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4
QUESTION 29 Discuss how pipe whip and jet impingement effects were determined for those postulated breaks in high energy piping that are not restrained (unrestrained whipping).
RESPONSE
O For breaxs in unrestrained high energy piping, barriers will be designeo to withstand pipe whip and or jet impingement.
We will revise Section 3.6.2.3.2.2 and Sec. tion 3G.1 in Appendix 3G as shown attached.
O
~
O O
FSH:pab/J12117 (r 12/11/32
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 4 APPENDIX 3G l
PIPE FAILURE ANALYSES 3G.1 GENERAL
~
~
.x.
This appendix describer th.e specific ~high-energy. pipe-failure p%
cama,-f m pipe. mp or-1** u ewwer
- protection provided. to sat 2.sfy the' re'quirements of Subsec-
- y g
tion 3.6.1 It demonstrates that the essential systems, corspo-nents, and equipment are not adversely affected by pipe breaks to an extent that would impair the ability to shut down the plant or, mitigate the consequences of the postulated pipe failure.
r o.-
4:
The effects of spraying and flooding as a result of breaks or
~'
cracks are discussed in Subsection 3.4.l.1.
S. ^:.
For a detailed discussion of break and crack locations and types, break exclusion areas (no break zones), guard pipes, and whip restraints, refer to Subsection 3.6.2.
y
_#f' b
1 b
ac.1-inc.1-2 vy
.--,,.,,m
,-,.,,__.-_._,-_,._m,,
~
QUESTION 30 Provide assurance that the tip deflection of a restrained whipping pipe will not adversely affect nearby safety-related components from performing their safety-related function.
RESPONSE
O
==
YA.- Os 4T t b Yic h sf lMY bk S
-R Ys vis& wefroh
) d 2 -) M lv2k i Ys-Q, uft.4-Nbl-O O
O FSH:pab/J12117 12/11/82 28
QUESTION 31 0
Provide the loads, load combinations, and stress limits that were used in the design of pipe rupture restraints.
Include a discussion of the desing methods applicable to the auxiliary steel used to support the pipe rupture restraint.
Provide assurance that the pipe rupture restraint and supporting structure cannot fail during a seismic event.
RESDONSE The loads, load combinations and stress limits for the design of pipe rupture restraint are covered in paragraph 3.6.2.3.3.
The design method for structural steel used to support pips rupture restiaing is covered in paragraph 3.8.4.3.2.1 and 3.8.4.3.2.3.
The pipe rupture restraint and supporting structures are designed to withstand the combination of pipe rupture and seismic loads.
O O
O FSH:pab/J12117 12/11/82
>j
QUESTION 32
(
Provide the design criteria used for pipe rupture restraints that also i
support piping.
RESPONSE
No pipe rupture restraint is used to support pioing e2r:ept for SIRS which O
is a ioned to Atsc cad -
The governing design loads for the SIRS is pipe rupture loads, or pige 1
support loads are negligible.
i i
O l
i O
O FSH:pab/J12117 12/11/82 2l
.u.=
QUESTION 33 Provide the basis for assuring that the feedwater isolation check valves can perform its function following a postulated pipe break of the feedwater line outsida containment.
RESPONSE
Q The design criteria for the feedwater pipelines in the steam tunnel assurss that the structural integrity is maintained.
The maximum stress occurring between the two isolation check valves is limited to 2.25 Sm.
The restraint for a postulated pipe break outside containment is located so that a plastic hinge is not produced, assuring the desired action of the check valve can occur.
The design specificiation identifies these checks as isolation valves and requires the supplier to provide operability assurance to be verified by testing.
These valves are specified to be selected for non-slam characteristics. The leak integrity of the check valves sh.all be demonstrated by test on analysis under most sever operating O
transientconditions). The valve specification requires the feedwater check valves to be ANSI Class 900, tilting disc.
The tilting disc check valve is designed to close as quickly as possible to minimize the slamming that is caused when the high velocity reverse flow is allowed to build up before the completion of closing. The disc
'begins to close before the flow reverses and has completed the closing before the velocity has built up to a dangerous level.
O O
FSH:pab/J12117 12/11/82
)2
i QUESTION 34 i O 3.7.3.5.1 - What is the significance of this section?
1
RESPONSE
i The static analysis method to be used is provided in Subsection 3.7.3.8.1.5 as indicated on revised page 3.7-40.
The remainder of Subsection 3.7.3.5.1 is deleted.
Figure 3.7-31 is also deleted.
I 5
O O
O FSH:pab/J12117 12/11/82
>p
QUESTION 35 3.7.3.8.1.7 - Provide a list of all instances where the first footnote applies.
RESPONSE
Q GESSAR II presents criteria to be used, actual analysis is to be presented by applicant.
Therefore the list of piping subsystems for which the footnote applies is to be provided by applicant.
The footnote has been revised to reflect that it is to be criteria, rather than imply that it has already been used.
l O O
1 0
FSH:pab/J12117 12/11/82 80
^2'_,
~. _ _ _.
~
238 NUCLEAR ISLAND R;v. 0 l
3.7.3.8.1.7 Damping Ratio (Continued)
. y. -:
Damping Ratio Safe Shutdown Operating Basis Earthquake Component Earthquake
.(Percentage)
Large* diameter piping systems (pipe diameter greater than 12 in.)
,2 2**
Small diameter piping systems. (diameter equal to or less than 12 in.)
1 2
3.7.3.8.1.8 Effect of Differential Building Movements In most cases, piping subsystems are anchored and restrained to floors and walls of buildings that may have~ differential movements during a seismic event.
The movements may range frem insignifi-cant differential displacements between rigid walls of a ecm=cn building at icw elevations to relatively large displacements between separate buildings at a high seismicity site.
Differential enfpoint or restraint deflections cause forces and acments to be induced into the piping system.
The stress taus r.4 O
- If the piping subsystem consists of only one or two spans with p'3 u.se little structural damping,.va. lues for small diameter piping.aee.~
4y
- u. sed,
- Conservatively changed from 31 to 21, since 3% damping response 4
curves are not available.
g
.O 3.7-55 5)
QUESTION 36 O
The number of bolt up and unbolt events is listed in Table 3.9-1 as 40 each.
This is a reduction from about 120 listed in past FSARs by GE.
Explain the reason for this reduction in the number of cycles considered for these events.
RESPONSE
l O
"r-'
This quotation was withdr:wn and GE to clarify it with revised notes on
^
Table 3.9-1.
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O FSH:pab/J12117 12/11/82
>L
GESSAR II 238 NUCLEAR ISLAND
'22A7007 Rev. 0 O
eb rab1e 2.9-1 PLANT EVENTS A.
Plant Operating Events l
Operating No. of Condition 9 Events Y
t 10 l
1.
Bolt Up Normal 40 9%
2.
Design Hydrostatic Test Testing 40 a.
Leak Checks at 400 psig prior to power operation, 3 cycles /startup 0
3.
Startup (100 F/hr Heatup Rate)2 Normal 260 4.
Daily Reduction to 75% Powerl Normal 10,000 5.
Weekly Reduction 50% Powerl Normal 2,000 6.
Control Rod Pattern Changel Normal 400
. O 7.
u ss of Feedwater amaters Ugsee a
8.
Scram:
a.
Turbine Generator Trip, Feedwater Upset 220 On, Isolation Valves Stay Open, and Other Scrams b.
Loss of Feedwater Pumps, Isolation Upset 28 Valves Closed c.
Turbine Bypass, Single Ssfety or Upset 8
Relief Valve Blowdown O s.
aeauceton to os rower, noe seenaby, uPsee 252 0
Shutdown 100 F/hr Cooldown Rate)2 lei N 10.
Unboltl Normal 40 i +]L
(
11.-
Scram:
a.
Reactor Overpressure with Delayed Emergency 13 Scram, Feedwater Stays On, Isolation Valves Stay Open b.
Ar.tomatic Blowdcwn Emergency 13 3.9-129 gj
GESSAR II 22A7007 238 NUCIJ:AR ISLAND rov. O
([
Table 3.9-1 PLANT EVENTS (Continued) 1 NOTES:
i 1.
Applies to reactor pressure vessel only.
+
W.
2.
Bulk avwrage vessel coolant temperature change in any 1-hour period.
3.
The annual encounter probability'of a single event is <10-2 O
cor
= = rs==v
==,
==4 <to-4 cor u1e a a=-
4.
One CBE event includes 10 maximum lead or stress cycles.
i 5.
One stress or load reversal cycle of maximum amplituc*e.
6.
Applicable to main steam piping system only.
7.
The number of structural feedback vibratory load cycles on tlie reactor vessel and internal components is 13,200 cycles of varying amplitude during the 220 events of safety / relief valve actuation.
The main steam and recirculation piping l
system use 660 full range cycles and.880 half range cycles, which are comparable in effect to 13,200 cycles of varying magnitude.
The main steam piping system uses 5460 cycles to O
include additional effects of acoustic wave propagation in steam during the actuations.
8.
Table 3.9-2 shows the evaluation basis combinations of these dynamic loadings.
l 9.
The ASME Code Section II service limits of Levels A, S, C, D, or testing apply to these normal, upset, emergency, faulted, I
and testing operating conditions, respectively.
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=
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QUESTION 37 Verify whether the alternate shutdown cooling mode has been considered in the design of the SRV discharging piping.
Specifically, address the capability of the spring hangers to accommodate the additional weight of water.
RESPONSE
O The response to this request is provided in revised Section 5.2.2.4.1 (see attached page 5.2-19 for proper insert).
i i
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FSH:pab/J12117 12/11/82 ff-
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 0 5.2.2.4.1 Description (Continued)
(m L)
Category I.
SRV discharge line piping from the SRV to the suppression pool consists of two parts.
The first is attached at one end to the safety / relief valve and at its other end to a pipe anchor; and the main stgam piping, including this portion of the safety / relief valve discharge piping, is~ analyzed as a complete system.
The second part of the SRV discharge piping
(
extends from the anchor to the suppression pool.
Because of the upstream anchor on this part of the line, it is physically decoupled from the main steam header and is therefore analyzed as a separate piping system.
As a part of the preoperational and startup testing of the main steamlines, movement of the safety / relief valve discharge lines vill be monitored.
TACEkT The SRV discharge piping is designed to limit valve outlet pres-gjj)g sure to approximately 40 percent of maximum valve inlet pressure
- p y with the valve '
- ide open.
Water in the line more than a few feet above suppression pool water level would cause excessive pressure at the valve discharge when the valve is again opened.
For this reason, two vacuum relief valves are provided on each SRV discharge line to prevent drawing an excessive amount of water into the line as a result of steam condensation following termination of relief operation.
The safety / relief valves are located on the main steamline piping rather than on the reactor vessel top head, primarily to simplify the discharge piping to
(_)
the pool and to avoid the necessity of having to remove sections of this piping when the reactor head is removed for refueling.
l l
In addition, valves located on the steamlines are more accessible j
during a shutdown for valve maintenance.
The nuclear pressure-relief system automatically depressurizes the nuclear system sufficiently to permit the LPCI and LPCS 5.2-19 bl
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QUESTION 38 +38A O
i Standard Review Plan 3.9.1 requires thac for each transient loading condition or combination, an acceptable ASME Code service limit has been specified; i.e., Design, Lovel A, Level 8, Level C, or Level D.
Update Table 3.9-1 to show conformanc.e with these service limits.
RESPONSE
- O The response of this request is provided in the revised Table 3.9-1.
The response to GE's position on the 10 cycles / events (38a) used in OBE (item 15 of Table 3.9-1) is given in the attached reference document,S Q cc Tssua.:
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12/11/82 N
GESSAR II 238 KUCLEAR ISLAND 22A7007 Rev. O 2
Table 3.9 1
}
PLANT EVENTS i
AS m r ccot~
A. - Plant Operating Events sat.wic e t.. m r-
\\
Operating No. of Condition 9 ' Events
/
l 1.
Bolt Up
-K'-- ' t A 40 0
Td s t'ing-40 2.
Design Hydrostatic Test a.
Leak Checks at 400 psig prior to power operation, 3 cycles /startup 0
3.
Startup (100 F/hr Heatup Rateil Ncrmal-Ar 260 4.
Daily Reduction to 75% Poserl Nors:1-A-10,000 5.
Weekly Reduction 50% Powerl M0rl Ar 2,000
~
6.
Control Rod Pattern Changel McTmat A-400 7.
Loss of F'eedwater Heaters
- p;;t Il 80 V('T 8.
Scram:
a.
Turbine Generator Trip, Fetdwater Ueem i 75 220 On, Isolation Valves Stcy Open, and Other Scrams b.
Loss of Feedwater Pumps, Isolation
- p;ct(Il 28 Valves Closed 01 c.
Turbine Bypass, Single Safety or "pect 8
Relief Valve Blowdown
=
O' 9.
Reduction to 0% Power, Hot Standby, "p :t- (1 252 0
Shutdown 100 F/hr Cooldown Rate)2 10.
Unbolt'l
-Mer 21 At 40 11.
Scram:
'a.
Reactor Overpressure with Delayed Z.; r g e.n c y (_~
13 Scram; Feedwater Stays On, Isolation Valves Stay Open w
(~
(])
b.
Automatic Blowdown Emeese-"y_
13 3.9-129 jy
~
l GESSAR II 3
22A7007 238 NUCLEAR ISLAND R,v. 0 Table 3.9-1 N
N PLANT EVENTS (Continued)
A Sm tr c ooe-voc v w ee.t~
Operating No. of Condition 9 Events 3
C 12.
Improper Start of Cold Recirculation C.x ;;;.;i 13 Loop C
13.
Sudden Start of Pump in Cold Recircula-
-C w=ievy 13 tion Loop O.
C 14 Improper Pump Startup after Hot Scandby E.., mi.. :
13 with Reactor Drain Shut off B.
Dynamic Loading Events 8 operating Cycles /
Condition 9 Events
- 7 15.
Operating Basis Earthquake (OBE)4 rlp : t 10 at Event Sa Pressure Peak Cycles O
P 16.
Safe Shutdown Earthquake (CSE)5 at.
Pauli ;
13 Cycle Rated Power Operating Conditions R
17.
Turbine Stop Valve Closure (TSV) 6
- Q;e t.
660 During Event 8a Cycles' 3
19.
Safety Relief Valve Actuation (One, All --" sec 220 7
or Automatic Depressurs.zation System)
Events During Event 8a 19.
Pipe Rupture Accident D
Small Break Accident r;ul:.ed 13 O
t Intermediate Break Accident i;;100-2 13 D
Large Break Accident rcui::d.
13 G
3.9-130 0o
p__._.
GESSAR II 22A7007
~
238 NUCII.AR ISLAND R ;v. O Table 3.9-1 nij PLANT EVENTS (Continund) gT-(J T NOTES:
1.
Applies to reactor pressure vessel only.
2.
Bulk average vessel coolant temperature change in any 1-hour period.
3.
The annual encounter probabilit/ of a single event is <10-2
(^
for an amasgaae* event, and <10-4 for a Lialted event.
U VL/ C 1..syw]
4.
One CBE event includes 10 maximum load or stre s cycles.
5.
One stress or load reversal cycle of maximum a=plituda.
6.
Applicable to main steam piping system only.
7.
The number of structural feedback vibratory load cycles on the reactor vessel and internal components is 13,200 cycles of varying a=plitude during the 220 events of safety / relief valve actuation.
The main steam and recirculation piping system use 660 full range cycles and 880 half range cycles,
)O which are comparable in effect to 13,200 cycles of varying magnitude.
The main steam piping system uses 5460 cycles to include additional effects of acoustic wave propagation in steam during the actuations.
8.
Table 3.9-2 shows the evaluation basis combinations of these dynamic loadings.
~
/ 9.
The ASME Code Section II service limits of Levels A, B,C, D,
.f or testing apply to these normal, upset, emergency, faulted,j' i
and testing operating conditions, respectively.
' y O
O 3.9-131/3.9-132 4
980-)
A"
.. s Mems from:
I I#
PAUL C. YIN
\\
EXT. 51439 NUCLEAR ENERGY SUSINESS GROUP i
GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA
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3-NE8 08E STRESS CYCLES USED FOR NECHANICAL EQUIPMENT DESIGN Nes *>P'm O
155uf:
The RPV and internals fatigue evaluation is based on 10 peak Operating Sasis Earthquakes (08E) cycles. The Standard Review Plan (SAP) requires the evaluation to include contributions from 5 08E,'s with 10 cycles l
each. Justify this deviation from the SAP.
UiG II RESPONSE:
References:
1)
Letter, R. Bosnak (NRC) to R. Artigas (GE), subject,
" Number of 08E Fatigue Cycles in SWR M555 Design, dated December 2, 1981 2)
Letter, R. Artigas (GE) to R. Bosnak (NRC), same subject, dated December 3, 1981 The fatigue contribution of the 08E stress cycles, to the total fatigue usage, is small. Consideration of more than 10 peak accumulated cycles would not.significantly reduce the sargin. This is documented in O
Reference 2 in response to an NRC request in Reference 1.
In addition, GE has performed studies, as documented in Reference 2, that shcw a plant would expect to experience less than 10 equivalent
~
peak OBE cycles during its lifetime. Therefore, the GE fatigue evaluation of the RPV and internals with 10 peak 08E cycles is con-servative and appropriate.
9 0
NA:rf:csc/118A24 9)
f*.%
UNITED states
[*s g
NUCLEAR REGULATORY COMMISSION gg d 1
t WASMNG TON, O. C. 20905
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February 18, 1982
- EO FE82 3 g Mr. R. Artigas Manager BWR Projects Licensing a,qGA9 Nuclear Safety & Licensing Operation General Electric Company 175 Curtner Avenue San Jose, California 95125
! O
References:
- 1) Letter from R. Artigas to R.' Bosnak dated September 17, 198i.
- 2) Letter from R. Bosnak to R. Artigas dated December 2, 1981.
- 3) Letter from R. Artigas to R. Bosnak dated t
December 3, 1981.
Dear Mr. Artigas:
In Reference 1, General Electric documented its presentation given to the NRC at the GE Bethesda offices on September 15, 1981. The GE presentation provided the basis for using 10 peak OBE cycles for the seismic fatigue design of BWR NSSS components. In Reference 2, we responded to the GE presentation and September 17, 1981 letter and provided a justifiable framework to resolve the use of 10 OBE cycles for all BWR plants undergoing OL review. Subsequently, in Reference 3, GE provided 1) a response spectra comparison and 2) plant-O specific results which show the OBE fatigue contribution to the total cumulative usage factor. The results indicate that even for the " worst case" component, the fatigue contribution due to the OBE is almost negligible.
Results were not provided for BWR/4 plants (i.e. Limerick and Hope Creek) ~as indicated in your summary table. We will require that the plant-specific results for the BWR/4 plants undergoing OL review be provided to the staff when the fatigue calculations are completed.
For BWR/5 plants, GE provided fatigue results for the WMP-2 plant which is stated to have the highest usage factor amongst the BWR/5 plants.. Based on our review of the WNP-2 fatigue results and of the OBE as currently defined in the WNP-2 FSAR, we conclude that the 10 peak OBE cycles used.in the fatigue design of the WNP-2 plant NSSS components is acceptable.
. O ror BWR/s p w es, it is our understand 4n, that a generic BWa/s c - nent
[
fatigue design was perfomed using bounding values for the 10 peak OBE cycles.
These bounding values were based on a conservative OBE and typical plant-specific OBE values are much less than the OBE assumed in the bounding generic design.
In addition, we understand that for each BWR/6 plant, analyses of
~
the RPV components are performed using plant-specific data to confirm the adequacy of the generic design. Thus, based on the above understanding and on our review of the fatigue results for the generic BWR/6 design, we conclude
- O 4
SSc
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sa a l' that the use of 10 bounding design OBE peak cycles is acceptable for the BWR/6 generic fatigue design of NSSS components.
We will include our evaluation and findings, regarding the use of 10 peak OBE cycles for fatigue design, in the Safety Evaluation Report of each BWR plant undergoing OL review.
+
1 f
Robert J. Bosnak, Chief Mechanical Engineering Branch Division of Engineering Office of Nuclear Reactor Regulation O
O I
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4, QUESTION 39 NUREG-0300 requires that computer programs used in analyses of seismic Category I code and non-code items have the following information provided to demonstratt their applicability and validity:
a.
The author, source, dated version and facility Q
b.
A description and the extent and limitations of its application.
c.
Solutions to a series of test problems which shall be demonstrated to be substantially similar to solutions obtained from any one of sources 1 through 4, and source 5:
1.
liand calculations 2.
Analytical results published in the literature 3.
Acceptable experimental tests 4.
.By A MEB acceptable similar program O
s.
The ben =hmarx probiens prescribed in aenort NuaEG/Ca-1877
" Piping Benchmark Problems" Please demonstrate compliance with these requirements and provida sumary comparisons for the computer programs used in seismic Category I analyses.
RESPONSE
The re:ponse to this repest are provided in the revised Section 3.9.1.2.
O O
FSH:pab/J12117 y 12/11/82
_ ;_ _.= _. _-_._-..._..
3 GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 0
~ *M.
O 39 "zca^u:c^t svsrcxs ^xo coxeoucurs i
3.9.1 Special Topics for Mechanical Components 3.9.1.1 Design Transients 1
The plant events affecting the mechanical systems, components and equipment are summarized in Table 3.9-1 in twc groups:
(1) plant O
operating events during which thermal-hydraulic transients occur, and (2) dynamic loading events due to accidents and earthquakes during certain operating conditions.
The number of cycles for each event are listed.
The plant operating conditions are identi-fled as normal, upset, emergency, faulted, or testing as defined in S dsection 3.9.3.1.1.
Appropriate Service Levels (A, B,
C, D or testing) as defined in the ASME Code,Section III are designated for design limits.
The design and analysis of safety-related piping and equipment using specific applicable thermal-hydraulic transients which are derived from the system behavior during the events listed in Table 3.9-1 are documented in the design speci-i fication and/or stress report of the respective equipment.
Table 3.9-2 shows the loading combinations and acceptance criteria.
V 3.9.1.2 Computer Programs Used in Analyses The following sections discuss computer programs used in the analysis of all the major safety-related components.
Computer programs were not used in the analysis of all components, thus, not all components are listed (e.g., main steam isolation and safety / relief valves and recirculation gate valves).
The GE computer programs are maintained either by General Electric or by outside comp 1ter program developers.
In either case, the quality of the programs and the computed results are controlled.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O NkO f'
3.9.1.2 Computer Procrams Used in Analyses (Continued)
TA/L{.-4.Q D3{AfAd
' For each program, one or more engineers are assigned.
uties are to:
/
/
(1) keep a east of the capability, the software contents, j
and the tileory of the progranir' t
O (2) run test cases an ain't in the reliability of the program; and
/
s (3) advise use,rs on the proper usage of the program and N
the correct interpretation of com'puted results.
s f
All nectss ry modifications are coordinated and ve ied by the responsible engineers; thus, user confusion over the es is
~
avoided and the reliability of these programs is maintaine N O()
N i
~
I 3.9.1.2.1 Reactor Pressure Vessel and Internals The computer programs used in the preparation of the stress report for the pressure vessei (RPV) and internals stress report are identified.
3.9.1.2.1.1 Reactor Pressure Vessel Following are the computer programs used in the analysis to assure O
the seructuret end functionet inteerier of the aPv.
3.9.1.2.1,1.1 CB&I Program 7 GENOZZ The GENOZZ computer program is used to proportion barrel and double taper. type nozzles to comply with the specifications of ASME Code Section III and contract documents.
The program will either design such a configuration or analyze the configuration 3.9-2
??
. Computer Programs'Used in Analysos
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3.9.L.2 ANFkT A T% f. )' 9.- 2 )
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PIPING PROGRAMS su/ d4 per formicN,in.
The verification of the following programs'
~~
accordance with the requirements of 10CFR50, Appendix B,'M SA/ fuNcarhh-and methodology documented Tf the verification of pput,_.o.tttpytg oR. c.F BRA in General ElectricADesign Record Files.
TSFOR a.
PISY j.
PDA g'
- b.
LUGST k.
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BILDROL C F BRAUN & CO _
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PDA a.
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PD61 CB&I REACTOR VESSEL The verification of the following programs is assured by contractual requirements between GE and CB&I.
Per the requirements, the quality..
e assurance procedure of these proprietary programs used in the design of N-Stamped equipment is in full compliance with 10CFR50, Appendix B.
t.
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GENOZZ b.
NAPALM j.
EO962A r.
1666 k.
984 s.
1684 c.
1027/BIJLAARD 1.
GASP t.
E1702A d.
846 m.
DUNHAM'S u.
MESHPLOT e.
KALNINS n.
1335 v.
1028
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ASFAST o.
HAP w.
1038 g.
TEMAPR h.
PRINCESS p.
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PUMP AND MOTOR VENDOR PROGRAMS Applicant to provida.
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QUESTION 40 t 0 NUREG-0800 requires if experimental stress analysis methods are used in lieu of analytical methods, that the methods used meet the provisions of Appendix II of the ASME.
Provide assurance that this is indeed true.
RESPONSE
i O
The respaase to this reauest is provided ia revised Sectioa 3.9.1.3.
O O
O FSH:pab/J12117 12/11/82
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GESSAR II 22A7ao7 238 NUCLEAR ISLAND Rev. O Mt-) s u O!
3.9.1.2.6.3.3.1 Program Description (Continued) l
~
high energy piping.
PIPERUP is an adaptation of the finite element method to the specific requirements of pipe rupture analysis.
t i
Straight and curved beam (elbow) elements are used to mathematic-j ally represent the piping.
Axial and rotational springs are used i
to represent restraints.
The stiffness characteristics of piping Q
and restraints can reflect elastic / linear strain hardening material properties, and gaps between piping and restraints can be modeled.
i 3.9.1.2.6.3.3.2 Program Verification The PIPERUP progras has been developed and verifie'd by Nuclear Services Corporation.
General Electric's subcontractor is using i
Version 1.2 being run on the NOS 175 computer system.
3.9.1.3 Experim ntal Stress Anal.ysis O
The following subsections list those NSSS components for which gg experimental stress analysis is performed in conjunction with
$ 4:S A. c.r,/;,,a. uid +L p., c.9.uu,.4f,/~r/u 4 fcre S.fl.4, <rm.<,.f A y.,s;,.t of r L g g
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anslytical evaluation. %
- Ad 3.9.1.3.1 Experimental Stress Analysis of Piping Oceponents The following components have been tested to verify their design adequacy:
(1) piping seismic snubbers, and (2) pipe whip restraints.
Cascriptions of the snubber and whip restraint tests are contained in Subsection 2.9.3.4 and Section 3.6, respectively.
~
O 3.9-25
/62 a..--.-.. -. -.
,.- ~.r _.. - __ _ - -. _.. _ _ _ _
.. = - -
QUESTION 41 Please provide assurance that when Service Level D Limits are specified that the methods of anlysis used to calculate the stress and deformations conform to the methods outlined in Appendix F of the ASME Code.
QUESTION 41a)
S&M Q
NF supports and new load combination to include thermal and hpiping loads as primarily - etc.
(secac Awc *.t e ~r)
RESPONSE 41 It is required by piping analysis and support design specification that ASME Code requirements including Appendix F for faulted condition shall be met for Class 1 piping and support design.
Table 3.9-2(7) shows the requirements of Appendix F.
Further response to this request is provided in the reyised notes 9 on Table 3.9-1.
O
_ RESPONSE 41a
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GESSAR II 22A7007 238 NUCI. EAR ISLAND Rev. O h
Table 3.3-1 PLANT EVENTS (Continued)
U, g\\,
NOTES:
1.
Applies to reactor pressura vessel only.
2.
Bulk average vessel coolant temperature change in any 1-hour period.
p 3.
The annual encounter prebability of a single event is <10-2 for an e@y event, and <10-4 for a N eve.h.
v 4.
One CBE event includes 10 maximum lead or stress cycles.
5.
One stress or load reversal cycle of maximum amplitude.
6.
Applicable to main steam piping system only.
7.
The number of structural feedback vibratory load cycles on the reactor vessel and internal components is 13,200 cycles of varying amplitude during the 220 events of safety / relief -
valve actuation The main steam and recirculation piping system use 660 full range cycles and 880 half range cycles, which are comparable in effect to 13,230 cycles of varying magnit'ude.
The main steam piping system uses 5460 cycles to s
include additional effects of acoustic wave propagetion in steam during the actuations.
8.
Table 3.9-2 shows the evaluation basis combinations of these dynamic loadings.
9.
The AC^.
C2. 3.miga II creice 'elmits-of-Eeve4Me-%- C. - D, c" te-ting 2 % y-tc--;hu u uv&.uol, ess=L, emergency,--f2" I ted,
p-d *=-thg operating conditionsasspectir.41y,_
tlr;:
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Qt'ESTION 42 l
The uajority of the information in Tables 3.9-11 and 3.9-12 is listed as
" Applicant to supply". Most of this missing data appears to be items that should be included in the generic plant design FSAR such as GESSAR II.
Provide an explanation of why this information cannot be provided for the generic plant design.
- O neSPonSe Same response as given in Question #27.
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FSH:pab/J12117 12/11/82
/or 1
Question 43.(3.9.3)
Table 3.9-13 indicates compliance with Regulatory Guide 1.48 which has been superseded by NUREG-0800, Section 3.9.3.
Please update this table to show compliance with NUREG-0800.
R o s pon se T d le 3.9-is wiii bc.
d elele J. Ta 6,l c 3.9-21 0 us (\\
b.c u p cl a k cI da be c odoma r e e.
sw wth NvRsc _ os co Se cM on 3,S.3, j
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' QUESTION 44 O
verity that the ii.it, in Taoies 3.s-18, 3.s-1s, and 3.s-20 are in complisnce with NUREG-0800, Section 3.9.3.
Provide a list of all instances where the footnoted equations were used and provide the needed justification for their use.
RESPONSE
O.
These equations are left in the text uatil design is completed, as they ec c ei,5,.n aay i
are - m -r. used,S.t. hough rarely.
On design completion the non-applicable equations are deleted.
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l QUESTION 45.(3.9.3) l The functional capability for essential systems must be assured when they are subject tc loads in excess of those for which Service Level B limits are specified.
By essential systems are meant those ASME Class 1, 2 and 3 and any other piping systems which are necessary to shut down the plant following or to mitigate the consequences of an accident.
Please provide such criteria.
In particular, have the criteria in NEDO-21985 been met?
RESPONSE
The response to this request is provided in revised Section 3.9.3.1.1.6 and References in 3.9.8.
QUESTION 45a f
For safety related equipment, faulted conditions to use upsettallowableS O
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- 2. AU PIPtNC-OYSTEMS $ HALL $E ANALY3&O ~7b MGFT l t-V&l O Limi r.s udO6,9, PLANT R UI Te-o Coyotried Lo&os,
Tife fdicultNCr RAT /0NA-[M MR E TF/G 8,43/ $ t& iTUS To,CYId(2 THE.
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238 NUCLEAR ISLAND R;v. 0 3.9.3.1.1.4 Tautted Condition (Continued) dynamic : action associated with a safe shutdown ear hquake and I
hydrodynami: loads plus a loss of offsica power, or she safe shutd=wn earthquajce.
3.9.3.1.1.5 Correla-ton Of Plan: Conditions wiza Ivent Pr:cabill:7 The probabilit/ of an event oc=urring per :sact:
year associated i O wita eae 91 ant c=nditione fo11ews.
rhis ce==e1ati=n is used==
identify the appropriate plant condition for any hypothesized event or sequence cf events.
Event Encounter Probability Par Plant Condition Resetor Year
- =al 'plannedi Upse (mederate probabili 7) 1.0 P
'*l Imergency (low probabili y) 10-2 P
10-4 Taulted (ex::a=ely low pr:bability) 10-4 > P > 10-4 3.2.3.1.1.6 Safety Class Functional Criteria For any scrmal er upse design condition event, Safety 01 ass 1, [,
and 3 equip =ent shall be capable of acecmplishing its safety fun:-
-ions as required by the event and.shall incur no per=anent cha.ges
- hat could deterierata its ability to acecmplish its safety fun -
tions as required by any subsequent design cendition event.
1 M & M t'~
For any-emergency or faulted design condition event, Safety Class 1, 2, and 3 equipment shall be capable of ac:cmplishing ::s safety functions as required by the event but repairs could be required to ensure its ability te accomplish its safety ' unctions as_
required by any subsequen design condition event. } Esuf/4/ f rdca t 7
s sal,ik ar~ A, scar.taey -/w r).,f*denaa & faaboYeyyskj};p k fla. LV y af l
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3.9-69
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238 NUCLEAR ISLAND Ruv. 0 1
f a
Table 3.9-2 (1) 0:
'a^= co"='"^rro== ^== acc=>'^=c= cazrzar^ roa
- SAFETY-RELATED NSSS PIPING AND EQUIPMENT Operating Condition l
Categories Load Combination
- e$ignBasis Eval. Only N + SRV (all)
Upset Upset N + OBE Upset Upset N + SSE Faulted Faulted N+ (OBE + SRV (ALL))
E=argency
';pset N+ (SSE + hRV (ALL))
Faulted Faulte.1
)
Emergency Emerge:tcy N + (IBA + SRV (2))
Faulted Faulted N+ (SBA + SRV (ADS))
Emergency E=ergency N+ (SBA/IBA + SSE + SRV (ADS))
Faulted Faulted Faulted' Faulted
)
- N + (LOCAgy,7))
Faulted
- t: + (LOCA7 + SSE)
Faulted
- See Legend on the following pages for definition of terms and criteria for combining loads.
- From all initial conditions.
r
- From rated power initial conditions.
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t,r w F8IW J7Co evaiM3 Aft G" barticer n P
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-. m 238 NUCLEAR ISLAND Rav. O 3.9.8 References k
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(1)
Wilson, E.L.,
"A Digital Computer Program for the Finite Element Analysis of Solids with Non-Linear Material Properties," Aerojet General Corpetatien, Sacramento, California, Technical Memorandum No. 23, July 1965.
(2)
"PVRC Recommendations on Toughness Requirements for Ferritic Materials,* WRC Bulletin No. 175.
(3)
"BWR' Fuel Channel Mechanical Design and Deflection,"
NEDE-21354-P, September 1976.
O I
(4)
"BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA)
Loadings," NEDE-21175-P, November 1976.
(5)
NEDE-24057-P (Class III) and NEDE-24057 (Class I) Assessment of Reactor Internals.
Vibration in SWR /4 and BWR/5 Plants, I
Novems.er 1977.
Also NEDO-24057-P, Amendment 1, December 1979, and NEDE-2-P 24057 Amendment 2, June 1979.
(6)
" Design and Performance of G.E. SWR Jet Pumps," General Electric Company, Atomic Power Equipment Department, APED-5460, July 1968.
(7)
" Testing of I= proved Jet Pumps for the BWR/6 Nuclear Syste:,"
O.
General Electric Company, Atomic
- Power Equipment Department, NEDO-10602, June 1972.
(8)
" General Electric Company, Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K,"
Proprietary Document, NEDE-20566P, November 1975.
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=
@ESTION 46 0
What basis is used to determine allowable piping deflections during preoperational testing of piping? When a stress analysis based on a time history is performed, which code allowables were used to establish acceptability?
8
RESPONSE
O Question withdrawn.
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FSH:pab/J12117 12/11/82 fg
QUESTION 47 It is the staff's position that all essential safety-related instrumentation lines should be included in the vibration monitoring program during pre-operational or start-up testing. We require that either a visual or instrumented inspection (as appropriate) be conducted to identify any excessive vibration that will result in fatigue failure.
O Provide a iist of 11 s fety-rei ted so 11 bore pipins a4 iastrumentatioa lines that will be includad in the initial test vibration monitoring program.
The essential instrumentation lines to be inspected should include (but-are not limited to) the following:
a.
Reactor pressure vessel level indicator instrumentation lines (used for monitoring both steam and water levels.)
O b.
nain steam instrumentation iines for monitoring main steam fiow (used to actuate main steam isolation valves during high steam flow).
c.
Reactor core isolation cooling (RCIC) instrumentation lines on the RCIC steam line outside containment (used to monitor high steam flow and actuate isolation).
d.
Control rod drive lines inside containment (not normally pressurized by required for scram).
RESPONSE
We will revise Section 3.9.2.1.2.1.2 to include all safety related instrumentation lines that are to be included in the preoperational vibration monitoring program.
We will also include in the section the requirement of visual or instremented inspection.
O FSH:pab/J12117 12/11/82
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GESSAR II 22A7007 238 NUCI. EAR ISI.AND R v. 0 3.9.2.1.2.1.1 Program Description (Continued) j be approved to meat the requirements of Section 3.9.2 of Regulatory 4
med mea su vwmeat-Guide 1.70..
Allowable deflections should be developed after com-f g,
g plation of stress anal,ysis.
Piping exceeding Phase I acceptance j
limits will be treated as Phase II.
Phase II requires remedial action (add or relocate supports, etc)
/{Q or proceed with time-history analysi's.
Apply time-history analysis to determine whether additional corrections are required.
l 3.9.2.1.2.1.2 Measurements j
.i. d m. ~ M r.< / a o All safety-related piping aLI. De subjected to prelim 1"ary
~
vibration measurements.
These measurements shall be taken during pre-operational tests with machinery and fluid systems cperating M
under test conditions.
Any indication of persistent vibration Y k,7 lh shall be a11 owed by r9 corded measurements for subse ent analysis.
Q Visad. gJ s+rwnt-J.s.,r)i feh u.aeh s sh// Lt. <: eda $dh 1 i/ py kur.r.'va k w h Sh >t v.1.. / z. t. j 7. z
- Vtintla hht' u u mdf-
. o piping ttached to pumps, Special,attentio)n shall be gi en
- /,,,w.
compressors, and other rotating or reciprocating equipment.
Measurements shall be taken near isolaticn' valves, pressure con-i trol valves, and other locations where shock or high turbulence may be present.
Every masaurement record shall be accompanied by a sketch showing the location of the measurement point, plus a description of the system operating conditions at the time of measurement.
Measured data shall include actual deflections sad frequencies.
The time duration of measurement shall he sufficient O
to udicate waether the vu =auen a con e=uous or e=an u ent.
3.9.2.1.2.1.3 Corrective Action If the allowables are exceeded, two options are available, which-l' ever is deemed appropriate:
!4 i
(1) take reredial action (add cr relocate supports, etc)) or (2) perform time-history test of the piping syste=.
3.9-36
QUESTION 48 O
Which operational transients will be used for preoperational testing of the non-NSSS piping systems? Whir.h system will be monitored and what locations will be instrumented?
RESPONSE
i llO operaticisai treasieats are iisted in T ai 3.9-2c73 greoperationai test for cperitional transient conditions of the non-NSSS piping systems and measurement locations will be determined I
M.
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FSH:pab/J12117 fjf.
12/11/82
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QUESTION 49 s
NUREG-0800, Section 3.9.2 requires that'a' list be provided of selected q
locations in the piping systems at which visual' inspections and measure-ments (as needed) will be performed during the preoperational tests.
For each of these selected locations the appropriate criteria (such as permissible peak-to peak displacement) to be used show that the strest-i and fatigure limits are within the design levels should be provided.
O eiease suppiy this information.
RESPONSE
Acceptance Limits For steady-state vibration, the_ piping break stress due to vibration only (neglecting pressure) will not exceed 10,000 psi for level I criteria and 5,000 psi for Level 2 criteria.
These limits are below the piping material fatigue endurance limits as 6
defined in Design Fatigue Curves in Appendix I of ASME Code for 10 cycles.
For operating transient vibration,* the piping bending stress (zero to peak) due to operating transient only will not exceed 1.25, or pipe support loads are not expected to exceed the Service Level D ratings for Level I criteria. The 1.25, limit insures that the total primary stress including pressure and dead weight will not exceed 1.8S,, the new Code Service Level B limit.
Level 2 criteria are based on pipe stresses and supportloadsnottoexceeddesignbasispOdictions.
Design basis criteria rcquire that operating transients stresses and loads not to exceed any of the Service Level 8 limits including primary stress limits, fatigue analyses factor limits, and allowable loads on snubber and supports.
" Operating transient will never lead to a fatigue failure, as they are O
oniy exposed to 2 or 5 cycies oniy.
FSH:pab/J12117 12/11/82 67
w -.
n In the dynamic analysis, the location of highest peak stress is identified, U
and the modal stains and displacements as sensor locations are determined relative to the peak stress on normalized basis, such as highest peak sress in each mode is 20,000 psi.
This is the allowable stress range, twice the allowable amplitude.
The allowable level of vibration are obtained from the stress report, and therefore are not determined at this time.
O tocatioas or insaectioas and Oevices The main steam and recirculation piping are instrumented with transducers to measure temperature, thermal movement, and vibration deflections.
During preoperational vibration testings of recirculation piping, visual observation manual measurements by hand-held vibrograph are made to supplement the remote measurements.
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FSH:pab/J12117 12/11/82
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q QUESTION 50 Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and components, it is requested that maintenace records for snubbers be documented as follows:
Pre-Service Examination O
^ ar -= rvice x ia t.'oa shooid b mad aa 11 saubb r= iist d in T bies 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9.
This examination should be made after snubber installation but not more than six months prior to initial system pre-operational testing, and should as a minimum verify the following:
1.
There are no visible signs of damage or impaired operability as result of storage, handling, or installation.
2.-
The. snubber location, orientation, position setting, and configuration O
cattachments. extensions. etc.) are according to design drawings and specifications.
3.
Snubbers are not seized, frozen or jammed.
4.
Adequate swing clearance is provided to allow snubber movements.
5.
If applicable, fluid is to be recommended level and is not leaking from the snubber system.
6.
Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly.
O FSH:pab/J12117 12/11/82 4
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If the period between the initial pre-service examination and initial system pre-operational tests exceeds six months,due to unexpected situations, reexamination of items 1, 4, and 5 shall be perfonned.
Snubbers which are installed incorrectly or otharwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.
Pre-Operational Testing O
ourino pre-operat4on.i testino, snubber thermai movements for systems whose operating semperature exceeds 250*F should be verified as follows:
a.
During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement.
b.
For those systems which do not attain operating temperature, verify
~
via observation and/or calculation that the snubber will accommodate the projected thermal movement.
O c.
Verify the snubber swing clearance at specified heatup and cooldown intervals. Any discrepancies or inconsistences shall be evaluated for cause and corrected prior to proceeding to the next specified The above described operability program for snubbers should be included and documented in the pre-service inspection and pre-operational test programs.
O The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion.
This test program should be specified in Chapter 14 of the FSAR.
O FSH:pab/J12117 12/11/82
/g ;,.
O 50 Response The applicant will subnit the snubber pre-service examination and pre operational testing program. This progran will include maintence records for s)(nubbers as suggested above (January 15)
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QUESTION 51 Provide justification that the design c:' ancAcrs which separate seismically designed piping and non-seismic Categor} ! piping is acceptable.
RESPONSE
? L L d e s/
.TA S & T Y P/P /W& khhe>J t s, Cc lv??.TED 7"o TWE.Cqq-rY P /PtN<r-8 Y s+t ta7W4c5 0 4</cdo& tS d%:5A// c.A./1,y A&MY2.GD,
j SGI KMic. l 0401, (NbWCED G Y 807W 77lE-bfQ.7Y Lvb 17/G-NbA]-
CAFG 7y P rbiN C-Art & /Welvosa /M 77fF-OE;;,^) 0F-MC WK l
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O FSH:pab/J12117 12/11/82
/
QUESTION 52 O
Explain how in the design process the reinforcement thicknesses of branch connections are determined for both internal pressure and mechanical loads and incorporated into the fabricated piping.
Provide assurance that all branch connections that are decoupled from the main mn piping in the piping analytical model are designed and fabricated to the required reinforcement area.
O
RESPONSE
Branch fittings are specified in the piping material classes based on pressure / temperature requirements.
Additional reinforcement is added and shown on the procurement drawings if the stress analys.3 of the system arrangement requires it. The analytical model includes branch loads.
At the branch connection, a stress interisification factor is used to determine.the maximum stresses due to mecnanical loads.
If the maximum stress exceeds ASME Code limits a reinforcement such as weldolet, pad, etc will be added. When the branch line is decoupled from the run pipe, brancit connection must be evaluated considering all moments from three legs.
Reinforcement requirements are determined by the type of stres.s I
intensifications used to satisfy the code limits.
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l O FSH:pab/J12117 12/11/82 fg
f QUESTION 53(3.9.2)
Which plant is the prototype reactor for the GESSAR II plants?
RESPONSE
Per n is the prototype reactor for the GESSAR II.
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O FSH:pab/J12117 12/11/82
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QUESTION 54 Verify that the design and installation of the safety and pressure relief valves is accc-ding to ASME Code Section III, Subsection hB-3500 and Appendix 0.
RESPONSE
O 1.
swr Safety /Reifef vaives (Safety vaives with Aux. actuating devices and pilot operated valves) are designed and manufactured in accordance with ASME Code,Section III, Division 1 requirements.
Specific rules for pressure relieving devices are as specified in Article N8-7000; Code Case N-100 (spring-loaded pressure relief devices; and NB-3500 (pilot operated and power actuated pressure relief valves).
2.
The design of 3WR Safety / Relief Valves incorporates SRV opening nd pipe reaction louds considerations required by Appendix 0, and those identified under S W-Section N8-3658 for pressure and structural O-integrity, safety relief valve operability is demonstrated either by dynamic testing or analysis of similarly tested valves or a combination of both.
9 frov:dd A Further response,.= m ; the revision of test on Section 3.9.3.3.1 and 3.9.3.3.2 with wording similar to SRP of 3.9.3 4 44 a f<d., J.
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FSH:pab/J12117 l 12/11/82
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1.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 H. %3 3.9.3.3.1 Main Steam Safety / Relief Valves (Continued)
()
The method of analysis. applied to determine piping system response to relief valve operation is time-history integration.
The forces are applied at locations on the piping system where fluid flow changes direction thus causing momentary reactions.
The resulting loads on the SRV, the main steamline, and the discharge piping are combined with loads due to other effects as specified in Sub-
/~T section 3.9.3.1.
The Code stress limits corresponding to load V
combinations classification as normal, upset, emergency, and rub 7 faulted are applied to the steam and discharge pipe.
3.9.3.3.2 other Safety / Relief Valves Other Seismic Category I active SRV's are listed in Table 3.9-16.
Pressure relief valves are identified in the table by the valve type "RV INLET".
Vacuum relief valves are identified in the table by the valve type "VAC BREAKER".
()
~
The operability assurance program discussed in Subsection 3.9.3.2.5 applies to safety-relief valves.
The qualification of relief valves is.specifically outlined in Subsection 3.9.3.2.5.1.2.2 i
3.9.3.3.3 Nuclear Island Rupture Disks l
There are no rupture disk 3 in the Nuclear Island design which j
1 must function during and after an SSE including hydrodynamic loads.
()
i 3.9.3.4 NSSS Component Supports i
I 3.9.3.4.1 Piping j
l
}
Piping supports and their attachments are designed in accordance j
with Subsection NF of ASME Code Section III up to the interface
(}
of the building structure as defined in the project design speci-fications.
The building structure cooponent supports are designed 3.9-97 l
/-Cl i
-. ~. -. -
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7AfIEE
[4p.
] 7-77 v
Oh ht-s 6 O
BWR Safety / Relief Valves (Safety valves with Aux. actuating devices and pilot operated valves) are designed and manufactured in accordance Spect fic rules witt. ASE Code,Section III Division I requirements.
for pressure relieving devices are as specified in Article N8-7000; Code case N-100 (tpring-loaded pressure relief devices; and N8-3500 (pilot operated and power actuated pressure relief valves).
The design of BWR Safety / Relief Valves incorporates SRV opening and and pipe reaction louds considerations required by Appendix 0, those identified under Sub-Section 18-3658 for pressure and structural integrity, safety relief valve operability is demonstrated either by dynamic testing or analysis of similarly tested valves or a combination of bothn %jy a :,f Z y
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QUESTION 55.(3.9.3)
The last line on Page 3.9-91 makes reference to Tables 3.9.10(25) and 3.9-10(26). These references are in error.
Please provide the correct references.
RESPONSE
Q The correct references are 3.9-2(5) and 3.9-2(6) as shown in the revised Section 3.9.3.2.5.1.2.
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O FSH:pab/J12117 12/11/82 fpp
4JD CUCLEAR ISLAND g.sv. 0
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i 3.9.3.2.5.1 Procadures (Continusd)
(%
t C.
to accomplish its intended function are described in subsec-tion 3.9.3.2.5.1 3 3.9.3.2.5.1.1 Testa i
Prior to installation of the safety-related valves, the following tests are to be performed:
(1) s, hell hydrostatic test to ASME Code Section III requirements; (2) back seat and main seat leakage tests; (3) disc hydgostatic test; (4) functional tests to verify
'that the valve will open and close within the specified time limits when subject to the design differential pressure: and (5) operability qualification of valve actuators for the environ-mental conditions over ther installed life.
Environmental quali-fication procedures for operation follow those specified in Section 3.11.
The results of all required tests are properly docu-mented and included as a part of the operability acceptance documentation package.
3.9.3.2.5.1.2 Dynamic Load Qualification The functionality of a valve during and after a seismic plus hydrodynamic event may be demonstrated by an analysis or by a combination of analysis and test.
Valves shall be designed using either stress analyses'or the pressure trit.perature rating require-ments based upon design conditions.
An analysis of the extended structure shall be performed for static equivalent dynamic loads applied at the center of gravity of the extended structure.
See Cubsection 3.9.2.2 for further details.
The maximum stress limits allowed in these analyses have confirmed structural integrity and are the limits developed and accepted by the ASME for the particular ASME Class of valve analyzed.
Addi-tional details on. stress limits are listed in Tables 2.?-10(25)-
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i QUESTION 56 I
f Please provide your interpretation of jurisdictional boundaries as they pertain to NF supports.
Justify your position.
Provide tables detailing allowables for supports.
RESPONSE
O 1.
aurisdictieaai souadaries for Nr - Supports.
The jurisdictional boundaries for NF structural supports are defined in paragraph 3.0 of NF STRUCTURAL SUPPORTS specification 300-48.
The applicable paragraphs of NF-1131 and NF-1132 of the ASME Code are referenced, and typical examples of boundary delineation are presented. Additional discussion is found in specification 300-50, FIELD FABRICATION AND ERECTION OF SUPPORTS.
2.
Tables Detaining A11owables for Supports.
Allowable stresses of various materials used for support design are given in the Code, Subsection NF, Table NF-2121(a)-1, and Appendix I, Tables as referenced in NF-2121(a)-1. Additionally, a survey of the use of allowable stresses is given in pages 1 and 3 of the NF SUPPORT STUDY by Braun.
Manufactured components such as snubbers and U-bolts are load rated.
The supplier prepares Stress Report Certificates for these components as required by specification 400-20.
The allowable loads are listed in the supplier catalog.
Allowable loads of the standard support designs are given in the load tables of the standards.
They are based on the Code allowable stresses.
. ut 1;;d ce;binati;n; and stre;; limit; in th: CESSAR t;Me-O FSH:pab/J12117 12/11/82
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p.
)
i QUESTION 57 Explain how relief valve loads and piping reactions are calculated for those systems other than main steam safety / relief valves (SRV) discharge piping.
RESPONSE
O an in-nouse procedure based on Moody papers in use for caicuiating reifef valve loads.
Piping reactions are obtained froa flexibility analysis with input relief valve loads.
1)
F. J. Moody Flood Reaction and Impingement Loads -
2)
F. M. Moody Time-Dependent Pipe Forces Caused by Blowdown and Flood Stoppage.
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FSH:pab/J12117 12/11/82 fpj
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[ Questien 58.(3.9.3) l O g[
Please provide a statement es to the compliance with NUREG-0619. "BWR i
r <
- r ace <' aad caatrai aad or$v a tara '<a aazz w crac=4"s -
I
Response
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1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4
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3.9.3.1.1.7 Como.liance with Rec.ulatorv. Guide 1.48
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Regulatory Guide 1.48 delineates acceptable design limits and appropriate combinations of loadings associated with normal operation, postulated accidents, and specified seismic events for the design of the Seismic Category I fluid system components.
As shown in Table 3.9-13, the loading combinations and design
("]
limits for the GE-supplied NSSS equipment utilized in this facil-s-
ity are in complete compliance with or meet the intent of Regula-tory Guide 1.48 through the incorporation of the alternate approach.
3.9.3.1.2 Reactor Pressure Vessel Assembly The reac::: vessel assembly censists Of the reacter pressure ressel, vessel supper: skirt, and shroud supp:::.
The reac::: press 2re vessel, ressel sue.:.cr: skirt, arf shroud
' J sw support are constructed in accordance with the ASME Soller a.-d Pressur: Yessel Code Secti:n :::.
The shr:ud suppor: censists 2."
One shreud support pla e and the shroud support cylinder and its legs.
The reac ce pressure vessel assembly components are clas -
'saiied as an A3ME Safety Class 1.
C:mple:e stress reports on these : meonen:s have been crecared in accordance with A3ME C:de b ^5 Olso C*=j l-&
Yw& inglx
&dREC -OS/1 recuireme.
5.
T NrC:)h.
W The stress analysis is performed on the reac:ce pressure ressel, ressel supper: skirt, and shroud support for various plant oper-O ating conditions l
(including faulted condition:) by using the elastic metacds excep as noted in Subsection 3. 9.1. 4. 3.
Leading conditions, design stress limits, and methods of stress analysis for the : Ore support structures and other reactor internals are discussed in suosection 3.9.5.
o 1
3.9-70 1
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GESSAR II 22A7007 238 Nt' CLEAR ISLAND RLv. 0
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3.9.4.3 Design Loads, Stress Limits, and Allowable Deformations (Continued)
Table 3.9-11(1).
For the non-Code components, experimental test-ing was used to determine the CRD performance under all possible conditions as described in Subsection 3.9.4.4.
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3.9.4.4 CRD Performance Assurance Program The CRD test program consists of the following tests:
(1) development tests; (2) factory quality control tests:
(3)
Five-year maintenance life tests:
(4) 1.5X design life tests:
(5) operational tests:
(G) a :eptance tests; and (7) surveil'.r.n=e tests.
O All of the tests except (3) and (4) are discussed in Section 4.6.
A discussion Of tests (3) and (4) follows:
(2)
Five-Year !!aintenance Life Tests - Four control rod drives are normally picked a: randet fr:m the production s
ck each year and subjected to various tests under simulated reactor conditi:ns and 1/6 of the cycles specified in Subsection 3.9.1.1.
Upon completion of the test program, control rod drives must meet or surpass the minima.T specified perfermance requirements.
(4) 1.5X Design *ife Tests - When a significant design change is made to the components of the drive, the drive is subjected to a series of tests equivalent Oc 1.5 times the life test cycles specified in Subsection 3.9.1.1.
Two CRos have undergone such testing in 1976.
These 3.9-107
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GESSAR II 22A7007 238 NUCLEAR ISLAND R&v. 0
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3.9.8 References O
(1)
- Wilson, E.L.,
"A Digital Computer Program for the Finite Element Analysis of Solids with Non-Linear Material Properties," Aerojet General Corporation, Sacramento, California, Technical Memorandum No. 23, July 1965.
(2)
"PVRC Recommendations on Toughness Requirements for Ferritic Materials," WRC Bulletin No. 175.
(3)
"BWR Fuel Channel Mechanical Design and Deflection,"
NEDE-21354-P, September 1976.
(4)
"BWR/6 Fuel Assembly Evaluation ef Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident.(LOCA)
Loadings," NEDE-21175-P, November 1976.
(5)
NEDE-24057-P (Class III) and NEDE-24057 (Class I) Assessment of Reactor Internals.
Vibration in BWR/4 and BWR/5 Plants, November 1977.
Also NEDO-24057-P, Amendment 1, December 1978, and NEDE-2-P 24057 Amendment 2, June 1979.
(6)
" Design and Performance of G.E. BWR Jet Pumps," General Electric Company, Atomic Power Equipment Department, APED-5460, July 1968.
O (7)
" Testing of Improved Jet Pumps for the 9WR/6 Nuclear System,"
General Electric Company, Atomic Power Equipment Department, NEDO-10602, June 1972.
(8)
" Gene-al Electric Company, Analytical Model for Loss-of-Ccolant Analysis in Accordance with 10CFR50, Appendix K,"
Proprietary Document, NEDE-20566P, November 1975.
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QUESTION 59(3.9.3)
Using the guidance of NUREG-0609, provide the methodlogy used and the results of the annulus pressurization (AP) analysis (asymmetric LOCA loads) for the reactor system and affected components including the following:
1.
reactor pressure vessel and supports, O
2.
core supports and other reactor internals, 3.
control rod drives, 4.
ECCS piping attached to the reactor coolant system, 5.
primary coolant piping, and 6.
piping supports for affected piping The results of the above analysis should :;pecifically address the eff ects of the combined loadings due to annulus pressurization and an SSE.
RESPONSE
The results of the annulus pressurization (AP) analysis will be provided byv w, 1983.
The following is a brief description of the methodology.
h%jj 1.
Pressure-Time Histories The pressure time histories in the annulus region between the RPV and shield wall are generated from a feedwater line break and a recirculation line break. The RELAP code using nodalized mass and energy balance is used in this analysis.
2.
Concentrated Force-Time Histories O
FSH:pab/J12117 12/11/82 g
l The forcing function of jet impingement on the shield wall is O
o.tained fro. the break fiow transient cause by a feedwater iine break and a recirculation line break.
Forcing functions of jet reaction on RPV, jet impingement on kPV, :nd pipe whip restraint load on restraint anchors are obtained from the feedwater line break, the recirculation lir,e break, and main steam line break.
3.
Integrated Dynamic Analysis O
Beam and shell models are used to integrate pressure-time histories and concentrated force-time histories in determinirg the effects on i
the sheild wall pedestal, vessel support, core support and internals, and control rod drives. These dynamic analyses yield displacements, forces, stresses and moments.
9.
Attached Piping Analysis Acceleration time history from the integrated dynamic analysis is O
used to senerate respease spectra for the stress anaissis of the attached piping. This ar.alysis covers ECCS lines, primary coolant piping, and associated pi;;e supports.
5.
Load Combination for Vessel and Piping Asymmetric LOCA loads in combination with SSE by the SRSS methodology are treated as a faulted condition for evaluation against the ASME Code and functional capability requirements.
This is described in Table 3.9-2.
O FSH:pab/J12117 12/11/82
/4)
QUESTION 60 Describe the allowable stress limits used for bolts in equipment anchorage, component supports, and flanged connections.
4 1
RESPONSE
The allowable stress limits used for bolts are as follows.
O A.
Flange Connection 1.
N83230 and Table I.1.3 for ASME Class 1 Pip.'ng 2.
NC3658 and Table I-7.3 for ASME Class 2 & 3 Piping B.
Equipment Anchorage and Pipe Support 1.
NF3280 for NF Bolts O
~ AISC Stress Limits for AISC Bolts 2.
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FSH:pab/J12117 12/11/82 149
I O
o"'S " '2 Provide a discussion of the use of U-bolts for component supports.
l
RESPONSE
Due to the line contact between U-bolt and pipe they are not used to carry deadweight. Allowable loads are provided on the load capacity data sheets together with dimensions. The data sheets are published in the controlled issues of the supplier pipe support hardware catalog.
U-bolts are used to support both large and small piping.
Application details are shown in Braun's design standard and in specification 400-80.
Both vertical and lateral loads can be carried by U-bolts.
See the supplier catalog for allowables.
Normally they are used as guides.
If the lateral pipe movecent is negligible, they are only used as guides.
O U-bolts are designed for clearance all around the pipe, generally 16-inch around large bore and 1/32-inch around small bore.
Additional clearance is provided for high temperature and large pipe diameter designs.
The clearance is provided by installing a nut on both sides of the supporting member.
A double nut is provided on the far side to prevent loosening.
Installation and tolerances for the installed condition are given in specification 300-50.
U-bolts are.not preloaded, because they are not used for axial supports O
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practice is used.
O FSH:pab/J12117 12/11/82 jgf 1
QUESTION 62 Identify where in the plant high strength bolts have been > used.
RESPONSE
No high strength (1170 k::i ultimate) bolts are required.
The following are specified for bolting materials.
O SA193, GR B7 (125 ksi) Flanges SA193, GR 88
( 75 ksi) Flanges SA193, GR B16 (125 ksi) Flanges SA564, Tp 630 (140 ksi) Submerged Services SA325, GR1 (105 ksi) Supperts SA540, GR B21 CL 1 (165 ksi) Supports O
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FSH:pab/J12117 12/11/82
/44
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- Provide the inservice testing program for pumps and valves including any requests for relief from ASME Section XI requirements.
RESPONSE
Inservice testing program is by applicant.
Engineer provides a design O
s,ecific. tion c4oo-es) to insure testing progr.m con se m.iot.ined.its no relief required from ASME Section XI.
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FSH:pab/J12117 12/11/82
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1 __.-.
O o"'S " 5'-(...s.1.2.13 Describe those short-term and long-term actions being taken to preclude the occurrence of cracking in jet pump hold down beams as described in IE Bulletin 80-07.
RESPONSE
The response to this request is provided in revised Section 3.9.5.1.2.1.
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FSH:pab/J12117 12/11/82 jgj
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238 NUCLEAR ISI.AND Rev. 0 S
3.9.5.1.2 Reactor Internals 3.9.5.1.2.,1 Jet Pump Assentlies The jet pump assemblies are no:
- e supp=r: stru=tures but are discussed here to describe coolant flow paths in the vessel.
The jet pump assemblies are located in two semi-circular groups in the dcwn=omer annulus between the==re shroud and the reme.cr vessel wall.
The design and performance of the jet pump is
(])
covered in detail in References 6 and' 7.
Each stainless steel jet pump consists of driving nc:zles, suction inlet, throat er mixing secti n, and diffuser 'F yure 3.9-11).
The driving nozzle, suction inlet, and throm: are joined toge:her as a rs=cvable uni:
- nd the diffuser is permanen-ly ins alled.
High pressure water from the re:ir=ulatic pu=ps is supplied each pair of jet pu=ps through a riser pipe welded te the recirculation inle:
nczzle thermal sleeve.
A riser bra e ec= sis:s of =antilever bears we* ded := a riser pipe and to pads on the reactor vessel wall.
The noz:le entry section is connected to the riser by' s.
a metal-to-
~
metal, spheri:21-to-conical seal joint.
Firm contae is main-tained by a held-down cla=p.
The throat section is suppersed laterally by a bracket attached to the riser.
There is a slip-ft:
joint between the throat and diffuser.
The diffuser is a gradual conical section changing
= a straign: cylindri:al secti=n at the lower end.
G "T
.i 3.5.5.1.2.2 Steam Dryers O
The steam dryer assambly is not a core support stru::ure er safety class component.
It is discussed here to describe coolant flow paths in the vessel.
The steam dryers remove moisture frem the wet steam leaving the steam separators.
The ex:racted moisture flows down thm dryer vanes to the collecting troughs, then flows through tubes into the downcomer annulus.
A skir extends free.
the bottom of the dryer vane housin[to the steam separator stand-I pipe below the water level.
This skir forms a seal between the
~O 3.9-113 l_
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b INSER OF QUESTION 64
.J j-//J The preload on the beams will be reduced from 30 to 35 kips in accordanca with General Electric recommendation. The new heat treated beams will be needed. This increases the expected life of these beams without cracking to 19-40 years.
Inservice inspection of the jet pump holddown beam will be performed to detect cracking inspection frequencies will be based on a load plant experience and GE testing.
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FSH:pab/J12117 12/11/82
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aussTtoa as There are several safety systems connected to the reactor coolant pressure boundary that have design' pressure below the heated reactor coolant system (RCS) pressure. There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pretsure RCS and the low pressure systems.
The leak tight integrity of tnese valver must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressuae systans thus causing an inter-system LOCA.
Pressure isolation valves are required to be' category A or AC per IW-2000 and to meet the appropriate requirements of IW-3420 of Section XI of the ASME Code except as discussed below.
Q Limiting' Conditions for Operation (LCO) are required to be added to the tehenical specifications which will require correction action; i.e.,
shutdown or system isolation when the final approved leakage limits are not met.
Also sorveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical spec'ifications.
Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, after valve maintenance
- rior to return to service, and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed O
pa=ittaa #ai
> Justificatiaa is sivea-to t =tias iat rv 1 inauid
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average to be approximately one year.
Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, raintenance, etc.
The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and given an indication of valve FSH:pab/J12117 12/11/82 g
degradation ovar a finito period of time.
Significant increas.es over j
this limiting valve would be an indication of valve degradation from one O
test to another.
Leak rates higher than 1 GPM will be considered if the leak rate changes are below 1 GPM above the previous test rate or system design precludes measuring 1 GPM with sufficient' accuracy.
These items will be reviewed on a case by case basis.
The Class 1 to Class 2 boundary will be considered the isolation point which must be protected by redundant isolation valves.
In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves provide isolation, only two of the vsives need to be leak tested.
Provide a list of all pressure isolation valves included in your testing program along with four sets of Piping Instrument Diagrams which describe your reactor coolant system pressure isolation valves.
Also discuss in
' O det ii how your ieax testins prosram wiii conform to tne above staff position.
Response
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dB93AR II 238 NUCLEAR ISLAND 22A7007 R;v. 0 3.9.3.2.3.1.1 Hydrostatic Test
~
(
)
u All seismic-active pumps shall meet the hydroststic test regi -*-
ments of ASME Code Section III according to the class rating of the given pump.
The ASME Code existant at the time of bid award applies or the designated code of* record.
3.9.3.2.3.1.2 Leakage Test x_/
The fluid pressure boundary is examined for leaks at all joints, connections, and regions of high stress such as around openings or thickness transition sections while the pump is undergoing a hydrostatic test or during performance testing.
Leakage rates permitted in the design specification that are exceeded are eliminated and the component retested to establish an observed leakage rate.
The actual observed leakage rate, if less than permitted, is documented and made a part of the acceptable docu-mentation package for the ccraponent.
% pthw.r wus yr.nd tls-obtsif, of tL./~k kh Jju=yn M fl.n / rf r.ooers buhfios valns will da 2.,44 ^ pg^ w}'4) /'">+-
- /}9.3.2.3.1.3 3.
Performance Test Af & &
1% 6J-The vendor demonstrates that the pump is capable of meeting al1 hydraulic requirements while operating with flow at the total developed head, minimum and maximum head, NPSH, and otner param-eters as specified in the equipment specification.
Bearing temperature (except water cooled bearings) and vibration
(^)
levels are alao monitored during these operating tests.
Both are
\\/
shown to be below specified levels.
3.9.3.2.3.1.4 Dynamic Qualification The safety-related active pumps are analyzed for operability during an SSE and hydrodynamic loading event by assuring that the pump is not damaged during the seismic event and the pump con-fs tinues operating despite the SEE and hydrodynamic loads.
3.9-82
.I
]
QUESTION 66 Table 3.2-4 indicates that the inforination required in identifying pressure vessel components, piping, pumps and valves, and the corresponding component code, code edition, applicable addenda, and the component order dv e of each A24E Section III, Class 1 & 2 component within the reactor coolant pressure boundary is scheduled to be supplied by the individual applicants.
One of the advantages of the Nuclear Island concept would seem to be to minimize the amount of analysis and checking of ASME code lists to be done for each individual plant.
Various places in the SAR refer to 1974 and 1977 Code versions as used for design.
How will plans requiring later code versions treat the differences in code loading combinations and allowables? Will the ar,41yses have to be redone? If code allowables cannot be met without design modifications, how will these modifications be reconciled with the design presented in GESSAR II?
RESPONSE
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