ML20071D674
| ML20071D674 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 03/09/1983 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| JFN-013-83, JFN-13-83, MFN-048-83, MFN-48-83, NUDOCS 8303110294 | |
| Download: ML20071D674 (93) | |
Text
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GENER AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE.. SAN JOSE. CALIFORNIA 95125 fiFN 048-83 MC 682 (408) 925-5040 JNF 013-83 March 9, 1983 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Mr. D.G. Eisenhut, Director Division of Licensing Gen,tlemen:
SU8dECT: IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
DOCKET NO. STN 50-447 REVISED DRAFT RESPONSES AND MATERIALS UPDATE Attached please find revised final draft responses to selected questions of the Commission's August 25, 1982 and November 15, 1982 in formation requests and the Mechanical Engineering Branch draft SER meetings. Only modifications (new or revised) to the responses of the referenced letters are provided. Also attached is a draft of a GESSAR II update relative to the latest revisions to Regulatory Guides 1.31 and 1.44 and NUREG-0313.
Sincerely, Glenn G. Sherwood, Manager Nuclear Safety & Licensing Operation Attachments cc: F.J. Miraglia (w/o attachments)
D.C. Scaletti C.0. Thomas (w/o attachments)
L.S. Gifford (w/o attachments) 8303110294 830309 PDR ADOCK 05000447 !
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J LISTING OF ATTACHMENTS PROVIDED Attachment Number Subject 1 Draft Responses to Structural and Geotechnical Branch Questions 2 Draft Responses to Power Systems Branch Questions 3 Draft Responses to Meterology and Effluent Treatment Branch Questions 4 Draft Responses to Reactor Branch Questions 5 Draft Responses to Mechanical Engineering Branch Questions 6 Draft Update of GESSAR II Relative to R.G.'s 1.31 and 1.44 and NUREG-0313 l
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4 Attachment No. 1 Draft Responses to Structural and Geotechnical Branch Questions
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l 220.04 You state in Section 3.5.3.2 of your FSAR that you use an analysis (3.5.3) procedure stellar to that in Reference 5 (Williamson & Alvy) to deterutne an equivalent static load representing the tornado missile. ,
Describe the actual procedure by which tornado generated missiles are transformed into effective loads. Verify that your proposed design i procedure produces static loads comparable to those determined using the Williamson & Alvy formula.
220.05 ' Submit details of the methods and assiamptions which you use in the (3.5.3) evaluation of the overall response of concrete and steel barriers
. .- subjected to impactive and tapulsive loadtt. If you use the ductility ratio concept, indicate the ductility ratios you assume and verify that you meet the criteria delineated in Appendix A of Section 3.5.3 Revision 1, of the SRP.
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The structural response to this load is evaluated using equivalent '
I static forces obtained by the procedure in Reference 6 for rigid missiles, aind the procedure in Reference 7 for deformable missiles a tad ut,g w ftcAun o& sh u. s StJ4g cda do ut tual lo. k A M wJ1 lt .
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220.08 In Sectfon 3.7.1.3 of your FSAR you correctly quota our
\ .(3.7.1) ,
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position in Section C.3 of Regulatory Guide 1.61. However, it is e,
not clear whether you have complied with our position on this matter.
Accordingly, clearly state whether you comply with this portion of 1 Regulatory Guide 1.61. If so, indicate the mechanism used to assure this compliance. If not, justify your position.
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The damping factors indicated in Table 3.7-1 were used in the response analysis of various structures and systems and in preparation of floor response spectra used as forcing inputs for piping and equipment analysis or testing and presented in Section
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3.10. N These values are consistent with those given in NRC Regulatory Guide 1.61.
When developing seismic design data for the SSE, the higher damping values of Regulatory Guide 1.61 were not used. The SSE data was obtained by doubling the OBE values which were based on the lower damping values. In the design process, the stress levels have been assessed and found sufficiently high to justify the use of the damping factors in Table 3.7-1.
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Tab *.e 3.7-1 ?
CRITICAL DAMPING RATICS TOR DIFFEREhiT MATERIALS . .
h Percent Critical Damping _
p Item (OBlJefdition ,
5SE g tion Reinforced concrete structures 4.0 1 7.0 2.0 4 welded structural assemblies Equipment 2.0 3.0 '
Bolted or riveted stru:tural .,4.0 7.0 assemblies
)
vital piping systems i
- diameter greater than 12 in. 2.0 3.0
- diameter less than or equa'. 1.0 .
to 12 in. .
Reactor pressure vessel, 2.0 support skirt, shroud head, separator .
Guide tubes and CRD housings 1.0 -
2.
. Fuel 6.0 .0-Steel frame structures 2.0 ,
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220.10 In Section 3.7.2.1.5.1.1 of your FSAR. you state that a study has been (3.7.2) conducted which shows that the interaction between the steel containment
. vessel and the polar crane can be ignored and that the crane mass can be lumped into the containment model at that level. Provide this study.
220.10 The report on the study of polar crane interaction with the stee containment is W . frovtd.R.M k lwa.
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Ganorel Eloctric CRANE GIRDER-CONTAINMENT INTERACTION 'Q Scn Joco TVA STRIDE Novsmber 5, 1974 DYNAMIC INTERACTIOC DETWEEN CONTAINMENT AND POLAR BRIDGE CRANE GIRDER
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CONCEPT Dynamic interaction between any two structural systems depends on their relitive masses and stiffnesses.
The structural system in question, namely, the steel containment with the crane girder was divided into two systems. A main system consisting of the containment alone, and a sub-system consisting of the polar crane g/
and the crane bridge. The main system was idealized as a 3-mass system ,
with masses concentrated at the c.g. of the containment ellipsodial head, crane girder level and an intermediate level. The sub-system was idealized as a 2-mass system with masses concentrated at the center and at an extreme trolley position, the former representing the mass of the crane bridge and the later representing the trolley with L.L. To study dynamic interaction of the two systems in all possible modes of excitation, three different types of excitation were considered. They were vertical excitation, horizontal lateral excitation, and torsional excitation. For each of these excitations, the two systems were reduced to corresponding equivalent single d.o.f. systems by condensing out the non-juncture degrees of freedom.
These effective masses and stiffnesses yielded the frequencies for the main system and for the sub-system for each of the three modes of excitation. Using the existing literature and- the developed mass and frequency ratios, the percent error involved in decoupling the two systems and modifying the main system wj.th the mass of the sub-system lumped into it was studied.
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- , C F DRAUN O CO Page 3 M P Badani General Electric CRANE GIRDER-CONTAINMEt
- T INTERACTION November 5, TVA STRIDE 1974 San Jose f.- In conclusion, interaction between the steel containment and the crane can be ignored and the mass of the. crane etc can be lumped into the containment model at that level for all types of excitation.
OVALING MODES OF CRANE GIRDER Due to the non-axisymmetric point loads resulting from the polar bridge crane, the crane ring-girder and the steel containment shell can exhibit ovaling modes of vibration.
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The frequencies of these modes have been computed using standard formulac. The exact shape of a given ovaling mode of vibration ,
consists of a curve which is a sinusoid on the developed circumference of the ring. For these computations the ring-girder is assumed to act l as a structural composite with a tributary shell section. The results i are summarized below.
OVALIMG MODES MODE OF VIBRATION WITHOUT CRANE WI'PH CRAME
. RAD /SEC CPS RAD /SEC CPS n* = 2 25.4 4.04 18.8 3.00 n =3 71.9 11.44 53.3 8.48 n =4 138.0 21.96 102.0 16.23
- n = number of full sine waves along the circumference.
In conclusion, judging from,the high frequencies and nature of the ;
respective mode shapes, the ovaling modes have very little modal responses as well as very small participation factors and hence are
((j, not significant. In addition, the ovaling modes have been found to have hardly any coupling with the beam modes of vibration.
REFERENCES:
1 Pickel, T W, Jr, " Evaluation of Nuclear. System Requirements for Accommodating Seismic Effects", Nuclear Engineering and Design, Vol 20, 1972.
2 Den Hartog, J P, " Mechanical Vibrations", 4th edition, McGraw-Hill, 1956.
( 3 Bechtel Power Corporation, " Topical Report - Scismic Analyses of g
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(( , Structures and Equipment for Nuclear Power Plants", EC-TOP-4, Rev 2, June 1974.
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220.09 Our position regarding th'e soil-structure interaction is contained ;
(3.7.1) in Item II.4 of Section 3.7.2 of the SRP and states that in addition i to a finite element method of analysis, the elastic half-space [
method should also be used. Accordingly, provide in Section 3.7.1.4 t and Appendix 3A of your FSAR, your procedure and the results from an !
analysis using the elastic half-space approach, including a discussion '
on the effect of variations in soil properties. j
RESPONSE
The response to this question is being addressed in conjunction with question 220.44. An additional eight cases of SSI analyses are being performed using the elastic half-space method with R.G.I.60 motion applied at the foundation level to demonstrate that the current GESSAR 11 design envelopes and design parameters bound those produced by t.hese new analyses. The fundamental frequencies of structures, equipment and components will be limited above the low frequency range (64 Hz) within which the response spectra obtained from the elastic half-space method exceed the response spectra ebtained from the GESSAR 11 finite-element method. The results of these new analyses will be provided to the staf f in April 1983 to demonstrate that the GESSAR 11 finite element method analyses satisfies the intent of the staff's acceptance criteria regarding the input level of the motion.
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220.20 Provide the following information applicable to pool dynamic loads. --
. (3.8.2) their load combinations and the analysis of these loads: _ _ . . .
(3.8.3) 138A8.4) a. The procedures used to generate the in-structure response spectra - - -
_ . _ at critical locations such as the reactor pressure vessel supports. _ __
Discuss how the effects of soil-structure interactions are -
accounted for in this analysis.
_ _ _ _ _ - ._. .x _ _ _ _ _ . _ _ _ _ _ . .
. .. _ F a-s o 6 w s e a ._ _ ._ __ _ . _ . _ _ . _ _ . . . _ _ . _ _ _ . . . ._ ..... _ . _ ._
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. A finite element model h:2 h:: gused to represent the Reactor Building. Time history analyses h: : heen preformed for pool .. . . _
dynamic loads. Response acceleration time histories e aW - --
~ obtained at selected node points including the reactor
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pressure vessel supports. Soil elements are used in the math .
model to account for the soil structure interaction. -
_ . . . _ _ . _ _ _ _ _ _ _ . _ . ..b. The extent, if any, to which structures adjacent to the reactor -
building will experience the effects of these loads. ,_
. _ . _ _ _ _ 3 . . __.O_____.. _ .. . . _ . __._. . . _ . __.
3 The structure 3 adjacent to the Reactor Building are expected to _-
experience insignificant amount of impact from the pool dynamic --
loads. This is to be confirmed by the h~,
t _ _ . . _ . _ .
. _ _ - - _ , . _ _ . _ _ . - - . . _ _ . _ .._ . - - __. ._m._-_.. - . - - - _ - . - .-._.-.. -
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1
- c. Your procedures for combining static and alternating dynamic loads
) (Section 38A.8.4) do not agree with our positions on this matter. _ . _ _ _ . _ _ _ _ _ _ _
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(Refer to Sections 3.8.2 and 3.8.3 of the SRP.) Discuss the effect of this deviation. In addition indicate whether your method of -
analysis includes the effects of fluid-structure interaction in the manner specified in the last paragraph of Item II.3.a of Section 3.8.3 of the SRP; i.e., whether you comply with the Appendix to --- -
Section 3.8.1 of the SRP. (Refer to Question 220.25)
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In the pool dynamic load analysis, hydrodynamic mass has been I.T._1 used to cover the structure-fluid interaction. --G-ESS-PTL E ---
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Q,QO,QO(C) Rav. O 238 NUCLEAR ISIAND m
3.8.3.3.1.1 Loads (Continued) )
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{ Floor at El 11 ft., O in.
Floor at El (-) 5 ft.,
350 psf 250 psf 3 in.
Stairs and platforms ,
100 psf ,
Floor at RPV head laydown 1000 psf area Floor at equipment hatches , 1000 psf Floor in upper pool (in 1000 psf except for
{ , , ,
addition to water weight) 1500 psf in fuel storage area Pg ,
= normal operating or shutdown differential,' pressures between the inside and outside of tlie drywell "9 9
$ RV dischage,,
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= thermal eff ts uri no a operating conditions including ate or liner expansion, equipment and pipe
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reactions, and temperature graclients using the com-bination of internal and external temperatures which would produce the most critical transient or steady-state thermal gradient. The maximum normal opera-
' ting drywell temperature is 150 0F.
R = pipe reactions during normal operation or shutdown conditions based on the most critical transient or l
l steady state condition.
1
(; Py j = pressure difference between the drywell interior and exterior of the structure (i.e., containment) consid-ering both interior pressure changes due to heating or cooling and containment atmospheric pressure var-iations (0.8 psig) on positive exterior pressure.
Construction = loads applied to the structure from start to loads
{ completion of construction for construction load combinations (the definitions of D, L, and I T are applicable but the construction value is used).
I 3.8-58
6 neww, 220,2O (c) ut.ddax A A 238 NUCLEAR ISLAND R',v. 0 l
3.8.3.3.1.1 Loads (Continued)
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% drywell walls and slabs are calculated on the basis L-of the transient bulk gas and liquid temperatures on each side. (S g apter d or drywe temp ra-ture diagrams.) IncjucleS SRN ddSc/utrges.ThMrnM ,
Y 1 _
hCD.
= pipe reaction from thermal conditions generaim y R,
n the postulated pipe rupture, including Rg ,
L- . .
,Yr = reactions from high-energy pipe break. None are on the drywell as these pipes are not anchored.at th.e...
drywell.
Y = load on the structure due to jet impingement from a 3
ruptured high-energy pipe -
i Y, = the energy resulting from the impact of a ruptured high-energy pipe on the structure Ry = SRV discharge loads, vent clearing, chugging, pool
. stratification, pool swell loads, etc. (For details see Appendix 3B.)
H, *= water pressure resulting from internal containment flooding W = loads generated by the design wind 3.8.3.3.1.2 Shrinkage and Cyclic Effects The drywell wall is designed for all loads listed above, the atrains induceil by concrete shrinkage and for 500 cycles of temperature variation from 60 0 to 1500 F due to startup and shutdown during the 40-year life of the plant. Creep effects are addressed
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in Subsection 3.8.3.4.1.1.
3.8-60 f - - -
vuoonn 22 22A/007
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238 NUCLEAR ISLAND RI,v. 0 3.8.3.3.1.3 Load Combinations
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Structural concrete design load combinations are defined in this subsrection. Nonstructural concrete and pool liner design load combinations are the same'but a load factor of 1.0 is used through-out for .the applicable load cases and normal operating co,nditions including hot and cold shutdowns. .
3.8.3.3.1.3.1 Load Combinations for Sez'vice Load conditions
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The following load combinations are based on the working stress design method: .-
(1) test 1.0D + 1.0L + 1.0Pt + 1.0Tt I (2) construction 1.0D + 1.0L + 1.0T, + W; .
(3) normhl 1.0D + 1.0L + 1.0T + 1.0R + 1.0P ; and '
(4) severe environmental 1.0D + 1.0L + 1.0T g + 1.0F
+ 1.ORg + 1.0Py + R. y 3.8.3.3.1.3.2 Load Combinations for Factored Load Conditions (1) . Extreme environmental 1.0D + 1.0L + 1.0Tg + 1.0F egs
+ 1.0P y +R; g g #
(2) abnormal 1.0D + 1.0L + 1.5P, + 1.0T, + 1.0R, + .
y; (3) abnormal 1. 0D + 1. 0L + 1. 0P, + 1. 0T, + 1. 2 5R, + 1. OR ;
(4) abnormal / severe environmental 1.0D + 1.0L + 1.25P,
+ 1.OR, + 1.0 (Y r +Y
+ 1.0T, + 1.25 F,qg 3 + Y,)
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- d. Describe the analysis performed to determine the effects of negative ,
, pressures in the suppression pool on the containment and drywell lower liner plates, particularly when combined with the effects of high temperatures, seismic loads and cracking of the concrete.
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Finite element non-linear dynamic analysis (ANSYS) was made to determine the effects of negative pressures on the bottom - --. _
liner plate in the suppression pool". The effects were combined _ __ _
and with the effects of high temperatureg seismic loads, M ..
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220.25
- Provide in Section 3'.'8.2.4 of your FSAR, a discussion of the localized (3.8.2) deformations at penetrations in the steel containment vessel due to the internal pressure build-up resulting from postulated accidents. Discuss the effect of these internal pressure loads resulting from postulated accidents on the leak rates _at. the penetrations in the containment vessel. .
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RESPONSE
This response will be provided in June 1983 as part of the severe accident design review schedule.
220.31 Discuss, from a consideration of buckling, the effect of a postulated (3.8.2) pipe break in the annulus region between the shield building and the containment vessel. Indicate to what elevation this could flood the annulus, thereby causing an external hydrostatic pressure on the steel containment vessel.
Response
The GESSAR II design requirements for all of the piping in the annular region between the shield building and the containment vessel (the shield annulus) meet or exceed the requirements of BTP MEB 3-1 for fluid system piping in containment penetration areas for which breaks or cracks need not be postulated. The comparative requirements are:
- 1. For high-energy ASME Code Section III Class 1 piping,the GESSAR II requirements are the same as BTP MEB 3-1.
- 2. For high-energy ASME Code Section III Class 2 piping, GESSAR II design requirements are more conservative than BTP MEB 3-1 in that GESSAR II adds the additional require-ment that the piping run is straight.
- 3. For moderate-energy ASME Code Section III Class 2 piping, GESSAR II conservatively applies the same requirements as for ASME Code Section III Class 2 piping.
In addition to meeting these stringent requirements, guard pipes are provided for high-energy piping in the shield annulus. Thus, even in the unlikely event that one of these pipes did fail, its guard pipe would direct the fluid into the drywell and there would be no flooding in the shield annulus.
All of the piping runs in the shield annulus are straight except for one 2-inch moderate energy line. Although this line meets the "no crack" requirements of BTP MEB 3-1, applying the more conservative GESSAR II design requirements, it would be necessary to postulate a crack. However, a crack in this line would be accommodated by the shield annblus drain. Hence, the flooding elevation would be negligible and there would be no threat of containment buckling.
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1 220.43 (3.8.5) Your calculated factors of safety for seismic Category I structures
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against of your FSAR.sliding, overturning and floatation are given in F15are 3.8-75 We note that you state the factors of safety against 1.02 and 1.04, reactor, sliding for the the auxiliary and the control buildings are 1.01, respectively.
Inasmuch as the n values are below our minimum acceptance criteria of 1.1, we fina them unacceptable. Accordingly, revise your proposed design fr.d as:nonstrate with calculations, including all your assumotien, that you satisfy our acceptance criteria on this I
l matter. Coordinate to the Question your response to this question witt. your response 220.42.
(This question is similar to and replaces
- ' Question 241.11. Accordingly your response to Questien 241.11 should cross-reference your response,to this question.)
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See attached revised first paragraph of 3.8.5.4.1 and revised Figure 3.8-75. The design meets the requirements of SRP 3.8.5 Item II.5.
For the reactor building, uc havc additionally used-the passiv pressure resistance acting on that part of the foundation mat which is deepe t auxiliary and fu 1 building .
foundatio Mh1[1eman tg a ad a ep'h acd't e safety factor.ty 1.19 in th pseudo-static calculation.
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, i For the auxiliary and control buildings the minimum safety factor of 1.1 is obtained using time-history approaches. The ;
method is outlined in San Onofre 2 and 3, NRC question 131.20,l and Appendix 3.7C-G of the FSAR.
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GESSAB II 22A7007 230 NUCLEAR ISLAND Rev. 8 ---
3.8.5.3.2 Auxiliary, Fuel, Control, Radwaste, and Diesel Generator Buildings Foundations The foundation loads and load combinations for these structures are discussed in Subsection 3.8.4.3. !
3.8.5.4 Design and Analysis Procedures
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The soil and structural settlements have been calculated at various points of each building. These values are included in the piping design specification for each individual piping system. In each specification, there is a table listing load combination requirements and stress limits. The differential [*
settlement has been listed as a loading condition. The pip-ing systems are designed for *.his settlement. The acceptance criteria are specified in the form of stress limits. They are in accordance with the appropriate sections of the ASME Code which is identified in the piping specifications. _
3.8.5.4.1 Reactor Building Foundation The design of the Reactor Building foundation is concerned primarily with determining shear and moments in the reinforced concrete and determining the interaction of the substructure with the underlying foundation. For a reactor building foundation supported on soil or rock, the pertinent aspects in the design are to maintain the bearing pressures within allowable limits, particularly due to overturning forces, and to ensure that there is adequate frictional 3 .
I force to prevent sliding of the structure when subjected to lateral
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loads.
a (nau wm The design loads considered for an lysis of the base slab foundation are the worst resulting forces from superstructures due to static and dynamic load combinations and such loads directly applied on the base slab as dead, live, seismic, hydrostatic, internal pres-sure, and temperature loads. The post-LOCA flooding condition has .
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l l 3.8-136 l
l
GESSAR II 22A7007
~~ 238 NUCLEAR ISLAND Rev. O OVERTURNING SLIDING BUOYANCY REMARKS BUILDING S. F. ( I S.F.(2) S.F.(3I REACTOR BUILDING 10.O l. I 9 4.6 AUXILI ARY BUILDIN G
'8 I"IO (4) I* 8
+4 i.is 2.T BULDENG CONTROL SHALLOW
- 10. B I,1 O B'2 BUILDIN G FOUNDATION RADWASTE 6.7 1.74 1.5 BUILDING DIESEL SHALLOW GENERATOR 22.9 1.40 6.0 FOUNDATION BLDG DIV I SHALLOW GE E OR 10.3 1.30 6.6 FOUNDATION BLDG DIV E8]Il
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(I) OVERTURNING S. F. =
OVERTURNING KINETIC ENERGY DUE TO SSE ,
=
1 ORCE (2) SLIDING S. F.
BUILDING SEISMIC SHEAR FORCE (SSE)
=
(3) BUOYANCY S.E BUOYANCY FORCE
($) no f A ON } W$ LG ONNd
) gtmwM b*0 } bc-\ A*% M " d-
[ FACTORSOF SAFETY FOR OVERTURNING
( SLIDING, AND BUOYANCY )g Figure 3.8-75. Factors of Safety - Overturning, Sliding n'^-*4-~ AA, ho q %
3.8-276
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I 220.44 In Section 3A.S.2(1) of your FSAR, you indicate use of a deconvolution '
(3A.S.2) analysis (i.e., FLUSH) to determine the motion which would have to be -
(Fig.3A-18) developed in an underlying bedrock formation to produce the specified '
control motion at the finished grade in the free field. We consider :
this approach not sufficiently conservative and, therefore, unac.ceptable.
- Our position on this matter is that the control motion should be applied ;
at the foundation level in the free field when performing a deconvolution ;
analysis. Indicate whether your analysis will conform to our position on this matter. (Refer to Item II.4.iii of Section 3.7.2 of the SRP.)
i' In responding to this question, cross-reference to your response to Question,220.09.
}
RESPONSE
See response to question 220.09 l
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.. _ ._ . ._ . . - .- - - - -.~ . - . . . .
-- With respect to the dr 11 design, list those provisions of CC-3000,Section III, 2 code which are not applic b:;e and w !
- j the technical basis for the non-applicability. (Ref: B.3. 2.
i 1.(1) D khhcR.
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l Provisions of CC-3000 (ACI359, ASME Sec III Div 2) i
'Not applicable to Drywell Design (Ref -
3.8.3.2.1) a) METALLIC LINER, CC-3121 GENERAL This paragraph is not applicable because the drywell face shells are not the nominal leakage-prevention liners that are addressed by the code. The shell plates are intended to carry design loads and are designed to act like steel reinforcement.
, , b) FOUNDATION REQUIREMENTS, CC-3561 (a) GENERAL This is a standard design covering a range of postulated sites.
Therefore the design need not be based on information e
provided by laboratory or field tests of some specific
, subsurface strata. The Applicant will show that his
. _._ __ . - - ' ~
site falls within the postulated range of sites. -
c) CC-3720 LINER This paragraph does not apply because, as noted for CC-3121 above, the shell plates are meant
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to act as reinforcement. Thus, shell plate design is controlled by allowable stresses (CC-3400) and load factors (Table CC-3230-1) l e=imie .-e-e -- e - e.-- -.--e=w= =
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[ With respect to the analysis of the equipment hatch opening of the Shield Building investigate the effect of not considering the curvature on the adequacy of design. Also discuss how the effects of the three components of earthquake motion are accounted for in the hatch opening analysis and the basis there.
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6;Pinite element methodh, is used to evaluate forces and moments for design of equipment hatch opening.
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- - - - . - .. A portion of shield buildi . cylindr'ica'l kheil was afnalyzed usino a curved math model.hb (t.klo 0 ;,wot. h J.Prt,cB Ik
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- --Displacements and rotations from '
shield building model output for various loads and loading combinations have been imposed on boundary nodes hr N O .
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M O W Lb Lc 's These displacements and rotations provide continuityj with the remaining shield building.
l Three independent X, Y'and Z components of seismic ce were incorporated in the computer input, and. combined ing omputer analysis with other loads for desired load combinations.
Either vertical upward o Idownward seismic load was combined with other loads so that maximum design force N alculated.
-Horizontal seismic loads were also added.with the above loads lvh A b hb 2. Iluvo ho obe e ,
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, ..lphemethoduti.lizestheenergybalance
' technique which involves using the strain energy of the target at maximum structural response to balance the residual kinetic energy of the target resulting from missile impact. A plastic collison is considered, meaning the missile remains in contact with the target. In the case of automobile impact experimental
- data are available to enable definition of a force-time function.
L The equation of motion is then solved for the maximum displacement
- ~ --~ '
The Bechtel topical report, BC-TOP-9A, on design of structures for missile impact was extensively used in this work.
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It is concluded that the exterior building walls sustain the crash of automobile missile during a tornado. -
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3 Attachment No. 2 Draft Responses to Power Systems Branch Questions l l
i l i ' i l i . ! l , j - i l' i i ; I , . ( i l i i ! l 430.04 The undervoltage relaying described in Section 8.3.1.1.7 of your (8.3.1) FSAR, by itself, will not protect the Class 1E equipment against a
, degraded voltage condition. Branch Technical Position PSB-1 contained in Chapter 8 of the Standard Review Plan (SRP) requires that a second 6
level of undervoltage protection be provided to protect Class 1E equipment against degraded voltage conditions. Describe your compliance with this position for Class 1E, Divisions 1, 2 and 3. - Igc.,6 - se. -- ---- [. l l l The M' Gess A destgevM-T (s b ased .. = al .bte grtel [
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vettage c n k h'e n s e W. m.v.im um flu.c h .a h o.v .S -r s 7. a t n e. _ in%ches points (2j.cio Wein 4 ,q,Sig.6.a.2 #c9eo v s/4. sos.4 s y e [ }
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1 : l l
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'( canadrons Av oyrra6'o.a,surasinaea. qn. men 6
'aw.t Y. pre m ale .l O ApVea,4[ s The kv d 1
.eenhs ad 4ke allowan . M' ar e pk:mtUMI9UC E ELMS
~
p 4. se4,nal h>.! '* Ja(e %tecko.J m sors med , r, a ss.cc.bl 4< ,n dela3 devices powded ~ bj-. h Apucanh
~
%3 ra p B.2 ,
I
? guirM W="Zy ":%
~4 . IlP&S (cliv 3) b~s fouls re-7"ve mM sLulky *
)[R'*k %*S_ a,e g
.J,61>4-stw^
WA vol5 drop caled bo,j0, all ==%3 rel. fed busas deo n % 4k. goe voit (..! , predde p .had prob,a d =3u a, )
, c.J nmous op.r.+o.o (eu 1- enstgiged loads , Mon. a. %'e m13 yett ge o@
45w t s- 7 9srat ru. .L s. 4
= _ %: _
This n eo m-<,de.d 4eshng proc du.re will L. l acide.ased 4~ 4be lre +u l- procectuvs. q eF<.r bbnd
}
l l. I i
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l, l
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I i
' \ ,
l 1 i b ! l l l 1 - ' 430.11 Provide the following additional infonnation regarding the loading (8.3.1) of the HPCS diesel-generator: i . , . j ; i i l
- c. Indicate the sequence of events if the diesel-generato. is on test
(' in parallel with the offsite source and the offsite source is lost. --- Indicate whether the HPCS bus will require re-energization by local manual control in a manner similar to the Divisions 1 and 2 buses. _ __ r . , , ResoawJe' l t
, i .__. _ _ _
During periodic testing of the HPCS (Div. 3) diesel generator, the D/G is paralleled with the offsite power system. During such a test should a loss of offsite power occur, a LOOP signal would probably not be generated because the D/G would attempt _ to provide power to the offsite system through the closed offsite power feed breaker. In this case the offsite feed breaker will trip either by offsite feeder overcurrent - or by generator overcurrent with voltage restraint (following a time delay). This
$', will automatically separate the HPCS bus from the offsite system. The D/G will t
continue operating with the governor control changed from droop to isochronous mode and the voltage regulator changed to automatic mode. With these, actions complete the D/G will be ready to accept its required load for LOOP conditon. I f l I
-{ ;
l 1 s 8
-_m m mem op - e*
memem - wm - l e-m -e e - e- - e *
=N 6- wm e -M e.6 h,,ew e.e--gme+4
-he
1 00ESTION 430.12 , (8.3.1) 's
~
The separation you describe in Sections S.3.1.4.2.3.1 and 8.3.1.4.2.3.2' of your FSAR for, the scram solenoid circuits and the main steam line (MSL) isolation valve circuits must be justified by analysis, based on tests, to show that there is no detrimental effect on Class 1E circuits, trith which these circuits are run. Additionally, demonstrate that the function of the scram solenoid circuits and MSL isolation circuits will
. not be tapaired by this arrangement. Explain how isolation is maintained between the Class 1E power supply feeding the "A" solenoids and the non-Class 1E power e,4 ply feeding the "S" solenoids since these circuits are run in a common conduit. *
.- 6 '.'
Explain the use of the D1 through D4 inputs shown in Figure 8.3-2#.of your FSAR, coming via isolators into the load drivers of the "B"' scram solenoid circuits. Pe 3 90* Sd GESSAR II design for the RPS scram and MSIV solenoid circuits is same as one used for Clinton Power Station. The specific parameters, such as wire siz( , circuit protective devices, conduit grounding and resistivity for GESSAR II design will be che same as ones used in the Clinton analysis. Furthermore the GESSAR II deeign provides a redundant circuit protection for each scram and MSIV solenoid power circuit. The scraq solenoid and itSIV circuits qp,e run in separate conduits. Cables fron other circuits are not run4these conduits. Within the PGCC the conduit is flexible and since the circuits are nor-divisional the flexible conduit is routed within non-divisional PGCC ducts. There is no mixing of divisions. - Optical isolaters have been provided for electrical isolation within the panel between IE and non-1E interfaces of the logic circuit. The power supply feeding the "B" solenoids is of the sane type as the one feeding the "A" solenoids. Solenoid "A" is fed fron non-IF bus A via an inverter and an EPA. Power is maintained within IE paraneters and the equipment is used for the power supply systen is of high quality, + 1/2; voltage regulation. Solenoid "B" is fed fron non-1E bus "B" power supply similar to bus "A". It is acceptable to run "A" and "B" solenoid power circuits tonether since the isolation is provided in the logic cabinets. Separate men-44e64 non-1E power supplies are provided to enhance plant availability. Figure 8.3-?5 will be corrected on the basis of the above discussion. See attached narked copy. For nore discussion of the NSPS Power Distribution re'er to Question 430.la response.
=_ - -
o GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1 9
~
oavsecosi @oR@ oeveSeoN2 hom h DeV888ow3 hoRh DiVISKW4 h oRh E_3 sn= sons @@ e E@ s [E i I ( EC i) C EC i ) < Tel > < c )
< 4 h . - . . . . . - . . .
3mTair E-i m E 8 - 3mTaic E 2mTaie E Locic raiP
~
J Lonic - E tonic E - LociC E - i.coic E - (e)- - @ 4 ;;; -,- -i-- e = -- -
= = -@
ACTUATI C DEVICES C LD LD LD LD LD LD LD LD i }, .se$ 0 4 h as , m k k'
,, a n. . ssa ss +
1 s s h _ _ s s - ( [C i ) ( C ) ( [C ]) ( ie] )
+ - . _
+ - , -
+ -
s - A 29 3A B A B A B l l - I SCRAM SOL SCR AM SQL SCR AM SOL SCRAM SOL GnouP 2 GROUP 3 GROUP 4 GROUP 1 LD = LOAD DRIVER SEE FIGUME 7.21 FOR h
- OtST ANCE SEeAMATBoN eER IEEE 384 h = CoNDUtTf5 TEEL EseCLoEURE .
h
- S ARRIERt5AFETY CLASS STRUCTupif
' @ . NooNSPECivisD F AIL 4Att ComCutTS SEPAmATiou nsove=E MENT E
- ISolAloR Figure 8.3-25. RPS Separation Scheme
_ _ _ ..m. _._ . - - . i i Attachment No. 3 Draft Responses to Meterology and Effluent Treatment Branch Questions
.. . . . . ~ . -
4 h 460.12 . ggg,gy Provide information on source terms for the following items:
- a. Provide the appropriate data for the items listed in Chapter 4 of NUREG-0016. Revision 1 (January 1979). For those items for which information has already been provided elsewhere, cross-references to the applicable sections are acceptable.
_ SA-5 Q) oWS-4 . ._ . . Get s A.\ 9 M a A vvs m p p e = T13 0_ wt . . _ . t L M .3 _proolweh w 5 50 Cc p gg.r.A a.. __ .. __so CL/plqt _ _ _ _ .2' ...__ _ . . . __ . _ . . . _ . .
- q. _ r.A 1. s. _h. . .s_ p t y ,. _ __.
= 1 s . w o % . A ~ ,- ,
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= r. n n om . ,
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-_ !--_ .d __ N o O f 'I!..vw.A_._.[ h h eu w Cq d A.h.
. - -._._ .-- .. O g a.J._.._ ..s y % ..%._0.. Q.0.\7 . k. c J. .
d o N_ . p...ps pe.._ (l a I. g h .g._= Lo m t. n . ). ..... ._ . _ _____ _2-)_. S e .Gs Ja,_.'sI . . s._cd g as _.i 1, g _ o r_
'.. d 9 3. Cv'tp_ 1ew . d ..Of h .ct.L._ S .s' 9 A .._.
..__ _._. . ._. I _ . . fN 4 N w cA -_.. ....._.
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- w .. A 5 .v .. d o .O . ot. 4 h.P.
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. .. . . . . . S .1. X \ 0. I CL/,3e.c.._.
~
. . _1.E ak W8% v "hvb.\ N. _ Cod g z. C. '3 C t.h v.
s.) .H = s . of%2re.J
.c . . h w oE. chm az.4[.= _ra 3 *e ,wa ow eb- sr*F K,g, ad c.bo w tow 1 c.\ .y = 2 o 3 2. c c. /g V bsr = 9 3 cc / 3ve.
Alw ev3. p c o FPy_r o s\S: Mk.so , {i i s,)*c)
._...eAo s-n v .s w3 A . -J G 4-L.dvk a. e - .
e s s m ._:tr s J em.. u.s .
Cl i i ; i i .' i , i 4 l
- g,74 e. % h c o .t 2 a ( c M n M
- i. ,
i i cms (co i 1J) i I i i
, s t ,'
. . 6
.GESSARD desgn do<a w o' nek M Mi(Mr
< .wws+s.If sp W ,+L A n - - ~ ~
.A-- % b u b .+. .
, r..om a % J c.-
- 4. 4., c o . m e . 4 . ek. ,6 4 .ws l ..L2.vt , AL-
~- -
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. I. Regreested. IrifoNnaIioh (et cetsdentaii. e/pg4 wtem wo.n c. re rep >t.d. m re,r.ue to g .t,i,, i w
(
- 2.
- Average" da.ily 4.leur rates foi.
he~g A e.* nductivt ty ahd, cheMica.) wa,ste ' Vt t * >icA. is tespense +a questrin po.13z. nog. Af cAs specific 60a.dc. inp;ts cah iceu, at - vu r.ms t Ans ud 4.> ske.1 dostient. r/ie daily - - Steuas re,pented vertered the expected gent 1 (hg es,TS etWCrd.qd$ en A ke'lly bnfif. ~
- 3. PC A Se> hde' chs; SoIM NM ~
.,, g s.fo a 7,.. ne.tr a n4 r[c.tcre b p e led r c of tiggig uieste Ate keysyted ih YAtt.E 12 Z~O 4, pesaqn kare oN's for PAdtoASie y>ecessth
. , , gi m e,d is t-e pa rted ih tA8t.E 11.2.-2, .
]
. a,nd, ol.iseggse) in Seedles ll.2.2 4 '
S.' Est:N a.T e d P e.l e a s e t
- f I ~g u fel '
The envivenma nt
- t,o
' is e(ircuS5c8 --
1 . .. - in 6e c,tio n 11.2. 3 -
.. . . G.
Devw m evali)er cowebs 1 (rev sechw32.2.s.) o.4e. o d sea l s4ea.- flow - i :
~ . . ss ts;4oo Ib/br .
i' of NuitsG-ools, -
-lb.s . se alia$ s kmw, _m . cows sd e.v-ed _ela.= d .~11 e.,
hoIdup b e. .t s h.o \*-ss...l.hav, 3o asu d .
..-..w _ -
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\
1
-------w--.-e. - - - - - m- e +- -wi-w --w- --- ---m----+--------------w- w-
4 c o. t 3 6 033 cl
- 4. $1nce the ffitered detergent westes may be directly discharged i.its the circulating water discharge canal, state tas fraction or detergent wastes that you espect to ha discharged in a year to the cfreulating water discharge canal.
I fesfonse: % 0 /3cl as otsauwS*L Sk it.2.s., Mo # h e a />> *y%ML Ql" "" i M sp w %m.1--->J se 4 +umWnaw doe ~5 A.a A =e4'A re La ' ' ~ ' ~ ' ' '
-l4f m. r,-
m
.,,a --
> X ) ..=, v o 1,s i...;=
1. 6-tj..... ._'_. '
~ ~ ~ ~ ~ ~
r as/4.a p4r r M, wt - e -? ._...i _.
~ ~ - hy w M
- m 2-- J ao e % = ff & 't, e,4 i .
M*fush p>4 p 2 me44 rat.44 24 W
)
4 . l
/ 5 W ' =; '=h 4 & *f?'='s' . .. . ...
C' E && 14 Cnweaf. . . . . . . . .. { X4. _ e;'O 4 .% **> = _ : = : _ {Jb y - A l4,w/ !
. i
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+ .;.
N W.I\0h bb k lk. == e p S
GESSAR II 22A7007 ! 238 NUCLEAR ISLAND Rev. 4 ; 13.2.1.2 Design Criteria (Continued) specified in 10CFR50, Appendix I. Liquid discharge to the canal may be initiated only from one excess water storage tank and require passage through an administratively closed and locked valve as well as manual initiation. No single error or failure will result in discharge. The design will maintain occupational exposure as low as practica-ble in accordance with Nuclear Regulatory Commission Regulatory l Guide 8.8 while operating with the design basis fuel leakage. ] 1 The pressure retaining process equipment in the liquid radwaste system is designed with a rated pressure and temperature of 150 psig and 212*F. The evaporator and heating element, however are designed for 350*F. The collection and storage tanks are atmospheric. The mixed-bed demineralizers and evaporators are l pressure vessels. In response to the requirements of General Design Criterion 60, ' the liquid radwaste system provides one discharge line to the canal for the release of liquid waste with the flow rate of this effluent stream controlled by a flow control valve with feedback from a flow measuring system. Additionally, radiation monitoring equipment is placed on this line to assure that excess activity is not allowed to be discharge in the event of an operational variance. A high radiation signal from this monitor will close the discharge line. The single discharge line is sourced only by q the Detergent Drain Tank, a very low level radioactivity source, d or one of the two Excess Water Tanks which usually contain con-densate quality water. These two sources may not discharge simul-taneously because of interlocks on the two valves allowing access to the discharge line. The Excess Water Tank serves as a large (100,000-gal surge capacity) to allow temporary holdup, again pre- : cluding the need for discharge of any waste volume for which there ; is no immediate room available in operating tankage such as Conden-sate Storage. t I f 11.2-2 iussn.r/
INSERT ON PAGE 11.2-2 l By administrative control, the discharge from this single discharge line to the canal is limited to a maximum of 5 gpm. This discharge can be increased up to 50 gpm provided the Applicant can demonstrate the discharge will meet the requirements of 10CFR20.
.- _.,q ,. - -_ , __ -._ - - - - - - . . . - -,
440./3 W 460.13
. g. Stace your Pl! diagens for the waste sLbsysta .s are for a fusi use t radmaste systes, Indicate hether the a:pt;1ent that you have Its:ed l en page 11.2-30 of your FSA4 is for both units or ne:ner it is on a per unit basis.
i
'Respon se : ,
z.w4., =d i.m n g.. tea. .a yage, n . 3.-so is 4., n. single ad radwaste sqrtes. . _.
The eynjo, enf /irt tu erreortially the .
l sanse for i slayle e alna.1an>7 radmite t
33 .e se .. .e se emanne m}' age see seg ege .sm-- e- en- a- === -- -.m .--- -- - - = -
-; seet .-
ne no es e. ea ee ees -l---. - - . - - . - - - sees - seee .e - -
- - - - , .- - -=>
p lSO suo .
*D .
pY
- e, u o = = = By
., . l= = u . . .. . , .. j .
. . . . .. . . . . . - .... _s- . . . . . . . . - . . . , - .
---. -- - - ee s -.e - a/ ene se e - - ese .- are .se oss pen sese 3.s .-
. . . . . . . . . e _ _ _ _ _ .. _ _ _ _. ._, e,
.. ... _ .A _ _ .. _ _ ._ _ . .. .. .- .
- -- - - - ---- - t.e e - .- a0 -- - .see # set 8 et f.ee t 't .
--. 't -
- - - - - - - 4960 - a-=- De - - see 1 - e ste e tte . ste .e 99 80 m e- -a. - ---== -- 98 =- au-m- 80 =e st 0.s 88 88 88 et .4 .* e4 s.
J _ _ _ . . ,, , .. _
==m-.-. - -m-- - - - - m-. -- - ---. - - ==-- 88 09 - G te e' 4. . O. *--. i F et 05 as. 9 . t 9. i ===.
Ste ses See see ate 04 30 0 sto Ble Dec 9 See # # # 40 fe# Me see ese - - - 4
.--mm- es -s==.- en--s. -a= - et as es .e -m-. -- -
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y
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.4 e.
e - n. m.m.
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2 t.a s. eusq gg.nas p 3
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m.
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'l 0,0 00 G A L
. ..m.mul-
. ..... . .....4 Figure 11.2-lb. Liquid Waste Management
. i .
ip. . i . System Process Flow Diagram 11.2-30
- - = - _. . . . . _ _ . .
l 1 ft.as p aws.a. Yo 460.i 3 ~c I 1
*: i
( - I t
~~ ~~ 1.
Provfde tank. the concentrations of radionuclides in the excess water 7_ __ in curfes, for I-131 and the total curies,in the conce tank given in Table 12.2-13 of your FSAR. l.. .. . . . . . . - . . ..
. hts po n3e Tk
.M. bC oh c.awkmb ow& of v-E d\ Ghwc \ d.n.o
*x t.5LFJ u sbA e ac M p rg b U^b3 iS y av'ke-d Souvea.
I n .t h.sAker e
.: e + eo aw
. C.4 J % ch.s m+4.e. m.bu -L.
6e eM L w% ei ow nwA.. tvt
/h Ewb3g sk M.
{- . . .
"T
$p o pp c.. cor recA-tow 4 4- T- \ 3 ) e_.w %b4t ow
.sw .-4Lt c.4w %4 d dwa.344 b le % s kw
.W4 AO. Aod "Tdh 12.. '2-l 3 t
E
. em.um .e
.m .
w . m ee-em.
-.-w 6 6.
+ . . .
1 i . . l . - . . . ,
r : l . J. sxcess waTan vnnns A8/s* l f:ourec Volume = #00000 gal. I Totat curies = 0.56 3 D t Italoegens_ soluble Fission Insoluble Fission Products Products Activation Isotopo _Curles Isotg Products Curies Isotope _Curles Isotope Curles y HH-53 3.5R 09 GH-49 FIR-84 I.8E-05* f .5 E-02 FR-95 /.? R-O f SR-90 f .3 E-0 J NA-24 3.4 R-0 9 ' DR-85 O. 7.R-97 4 0E-06 I-111 SR-91 34 E-0 3 ND-95 P *2 S .fE-0 f 9.O N-02, SR-92 'd .# E-O f CR-51 1-132 2.oE-02, 2.f C-0 3 Ril-103 9.oE-O f E.68-03
- Y-90 / .3 E-0 3 MN-54 2.FE-0T
- t-133 V.fE-02, RU-106 ,.VE- O f" MN-56 Y-91M f.rE-0 3 i.48-03 1-134 7.oE-0 V MO-99 RN-103M j .o E-0 K CO-50 I-135 /.6E-0 2, 2.fE-0 Y RH-106 i .VE-0 f E .95-01
.TC-99M f.fE-O J CO-60 3.L 8-0 3 IA-140 J.6E-02. PR-59 TC-101 3.g E-0 6 CE-141 4 65-09 TOTAL l.7E-0) TE-129M / .fE-0 3 CE-143
/ . fE-O S* NI-65 6.08-/o TE-132 */ .fE-0 5" 2N-65 /.2E-03"
/. 9 E-0 2. 'CE-144 C5-134 f.fE-04f PR-142
/ .?E-0 f 2N*69M f.tE-06 CS-136 3 8E-0 Y /.VE-0V AG-110N .3.SE-01 ND-147 f.fE-OF W-187 C5-137 /.VE-07 ?.fE-O f CS-138 f.Z E-D V TOTAL DA-137M /.3 E-0 3 8.75-02. TOTAL .T .8E-02. *-
' DA-139 f 2 E-03 DA-140 l 3.f E-0 2 i
RA-141 /.fE-0 f ! BA-142 / .oE-0 6 I NP-239 2.tE-0/, i . ( TOTAL f .2. E-0 /. t
Table 12.2-13 (Coritinued) l CONCENTRATED WASTE TANK A700 Source Volume = 3000 gal. normal, 25,000 gals. full Total Curies = 29 Soluble Fission Insoluble Fission Activation _Italogens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies BR-83 2.5E-04 SR89 9.4E-02 ER-95 1.2E-03 NA-24 4.0E-04 BR-84 1.9E-06 SR-90 8.9E-03 ER-97 7.5E-06 P-32 4.4E-04 BR-85 4.7E-10 SR-91 4.8E-03 NB-95 3.7E-03 CR-51 1.4E-02 u I-131 2.6 1 SR-92 2.6E-04 RU-103 5.2E-04 MN-54 1.5E-03 $ l I-132 6.2E-0 Y-90 8.9E-03 RU-106 8.7E-05 MN-56 9.5E-05 g
- s. I-133 1.5 - 0 Y-91M 3.3E-03 RH-103M 5.2E-04 CO-58 1.7E-01 c
" 1.0E-0 MO-99 7.3E-02 Ril-106 8.7E-05 CO-60 2.0E-02 I-134 N I-135 7.3E-02 TC-99M 1.7E-03 LA-140 1.8E-01 FE-59 2.6E-03 h y
w TC-101 2.7E-07 CE-141 1.0E-03 NI-65 5.6E-07 WX
^
TOTAL 2.8 1 TE-129M 7.9E-03 CE-143 3.1E-05 EM-65 8.4E-05 we TE-132 5.9E-02 CE-144 1.2E-03 EN-69M 5.0E-06 "" CS-134 6.lE-03 PR-143 6.9E-04 AG-110M 2.3E-03 h i CS-136 1.9E-03 ND-147 2.2E-04 W-187 1.6E-03 o CS-137 9.4E-03 CS-138 3.9E-06 TOTAL 1.9E-01 TOTAL 2.lE-01 BA-137M 9.4E-03
- BA-139 5.4E-05 BA-140 1.6E-01 BA-141 6.8E-07 BA-142 1.3E-07 Q - TOTAL 1.0 - 0 c
?$
52: of
(N oTS : CAST PA GE owW cuna saro) N14 Provide additional faformation on the following itens applicable to
., (11.3) the gaseous weste management systems:
(- a. Since your systes description, tables and figures in Chapter 9 of your FSAR do not clearly indicate whether there are provisions for both IEPA and r.harcoal adsorbers for the reactor building pressure control mode and purge exhaust, provide the appropriate information ~~ relating to filter units for the reactor building.
. . . R=.spaw.5-t a .
c . JTw. se ssat z A aw .J. s n . twe.L A,. -N. '
..E lbr .u- A., ( w s P A -L i b % )
.. & ..e L ).i -L.pn mal.c% b r A enry. c u re .I :
. w u 6 c r. s'p = u . a A l
.u gr.vsAa.JL
.Lo n nu +a .nekAa- .m b M u Acm ehu uwd-a s e. . .
1
~
, .ss i h & h wwh m pv,m :- 3
..aw4%c n.=i ns
. % pr.y~I nk+ .g. .
ha.,s i- D ~
. wb hJ e -4.sTss.J
- t. -4. r u.w I j t% k twd Ah '_
.. N .*.. E h u w h b
-. C o A m tFkTovw G L1.WAy7WWA cad 54W ..
J -
-. . The Nuclear Island HVAC design no'l provide space, e.c. for installing -
exhaust
--' ' system. filter units, other than the primary contairunent purge and exhaust All ventilation exhaust have process radiation monitors in -
.._ the exhaust stream that will detect the release of radioactivity.
In event that high level of radioactivity is detected, the ventilation
~
exhaust will automatically be shut off and the Standby Gas Treatment ~~
... System v111[Iutomatica11y actuated to ventilate that area. _
. The above applies to the secondary containment buildings; that is. -
Shield Butiding Annulus. ECCS/RWCU Pump Rooms of the Auxiliary
--' Building and the Fuel Building and the primary containment. -
. - . = .
__t da._cd cLn:re> .N aA.hr _.l.sb.n e.l e.>s w.) 4.s ., _6%_N e,.e hee % ,%
238 NUCLEAR TE MND R3v. ( ,_ y w_ h + w & u o . @ _ 9.4.3.4 Inspection and Testing Requirements (Continued) ( l filters, fans and redundant components to assure system avail-ability. 'the tests include determination r f diffarential pressures and filter efficiencies, control setpoints and signals, alarm fune-tioning, modulation valve performance, airflow rates, damper func-tioning, airflow switch operation, isolation butterfly valve functioning and thermal performance of heaters and coolers. Test connections are provided for sampling and monitoring the above-
'noted categories of performance.
The balance of the system is proven operable by its use during operation. Standby equipment can be tested to ensure proper opera-tion on demand. Equipment layout provides easy access for inspec-tion and testing. 9.4.3.5 Instrumentation Application Instrumentation and controls for the Auxiliary Building pressure I control systems [ Figure 9.4-3 (K-163)] are designed for automdtic operation. The system fans are started from manual pushbutton stations in the main control room. Airflow failure, sensed by an airflow switch, actuates an alarm, which starts the standby fan and repositions the associated dampers. Geva Ma. Au xiIt av Bub ECCS A,ea Nase A4.l sb Exhaustairgiscontinuouslymontoredfo radioactivity. A high level of activity or an ECCS operating signal automatically starts ' the SGTS, stops the supply and exhaust f ans, closes their asso-ciated dampers, closes the air supply isolation valves and directs the exhaust air to the SGTS. The ECCS recirculating fan coil cooling units for RHR pump rooms A, B and C, RCIC, HPCS and LPCS pump rooms are interlocked to start ) when the pump they protect is started. Also, manual override from pushbutton stations in the main control room is provided. t 9.4-42
__ ~
~ ' ~ . - - .
. . " a ~ ~ ~, . ._ . ,
44o, m q I4 ,
] ,, 11.5.1.1.2 Systems Required for Plant Operation (Continued) l The radiation monitoring systems (RMS) provided to meet these objectives are (1) for gaseous effluent streams -
(a) plant vent discharge, l (b) offgas exhaust vent, l (c) radwaste building ventilation RMS, and 4 (d) turbine building ventilation RMS; (2) for liquid effluent streams - (a) radwaste effluent RMS and ( (b) service water effluent to cooling pond RMS: (3) for gaseous process streams - (a) offgas pretreatment RMS, (b) offgas post-treatment RMS, and (c) carbon bed vault RMS; and l (4) for liquid process streams - (a) RHR service water system RMS (loops A and B) and (b) closed cooling water RMS. s ( $ Ap 1 Caw WS f ChJ)bs 11.5-3 1
_- . . ~ . .
..n ,
238 NUCLEAR ISLAND Rav. l 11.5.2.1.4 Auxiliary Building Exhaust Radiation Monitoring (, This system monit the r diat o - Buildingventilatone.paus ka s e m evelehteriortothe SCC Area Ness ctg gesystemconsistso ukiLi(rg4e g , two , redundant, instrument subsystems,, channel A and channel B, which are physically and electrically independent of each other. Each ! channel consists of a local detector, a convt ter and a main control room radiation monitor. Power for chan.'.sl A is supplied from ! 120-vac RPS Bus E. Power for channel B is supplied from 120-vac
- RPS Bus F.
I Each radiation monitor provides two trip circuits: one for upscale (high) radiation or an inoperative circuit and one for downscale. The upscale / inoperative trip of channel A initiates opening of the exhaust to the SGTS valve, closing of the exhaust to the plant vent valve, closing of the ECCS corridor exhaust valve, and the closing of the RWCU corridor supply valve for Division 1. The same trip also initiates startup of the SGTS, Division 1. The trip of channel B monitor initiates the actuation of the corre-I sponding valves for Division 2 and startup cf the SGTS for - Division 2. High radiation and downscale control room annunciators are actu-ated by the signals from the monitors. Each control room radiation monitor visually displays the radiation level. ]N5tRT- s 7 ?FRoM 11.5.2.1.s M standby Gas Treatment Radiation Monitoring ~ System PA This system monitors the radiation level at the SGTS exhaust duct. The detectors are physically located downstream of the exhaust and heat removal fans and dampers on the exhaust ducts for Division 1 and Division 2. i 11.5-11 i
INSERT ON PAGE 11.5-11 The exhaust from the Auxiliary Building electrical areas, corridorsM g :ter t;;;d and Elevator Tower HVAC System (Figure 9.4-4a) is throegh two louvered roof vents and is not monitored. Only the steam tunnel has agtentgl fgrysegr,adigge releases requiring monitoring. Ksteam unneligMo ted rge regt of g gg , A uxiliary T !urt':: BuildinggsEam tunnel h${a the Seismic Interface Restraint Structure. Monitoring a gaseous releases from this area will be accomplished by Turbine Building vent monitoring. The Turbine Building vent monitoring is the responsibility of the Applicant. The control rod drive maintenance area source has been determined to be not significant. The remaining areas exhausted by this system contain no radioactive sources and are isolated from the potentially radioactively ; contaminated areas of the Auxiliary Building. l 1 I l l l k l l c
........u_._;.. . . . ..
3 Attachment No. 4 Draft Responses to Reactor Systems Branch Questions i l 1 l i I 1 l l
I 440.01 Indicate whether the design of your proposed 238 Nuclear Island conforms to the LRG-II positions. If there are any known exceptions at this time, so indicate.
Response
As described in Appendix IE (Sections 1E.1 through IE.13), the GESSAR II positions conform to all of the Reactor Systems Branch LRG-II positions with one minor exception; 5-RSB. The following will replace the GESSAR II Position on page IE.5-2:
" Leak detection capabilities are discussed in the response to NRC question 480.27. Each ECCS room is separate and water-tight. Any suppression pool water loss is therefore limited to the flooding of the largest volume room and redundant equipment in other rooms is protected from flooding. Any leakage from the first isolation valve will not result in a long term equilibrium suppression pool level below the NPSH requirements for the RHR system. The strainers on the intake of the suction line will remain submerged to a depth greater than 7 feet which has been determined to be sufficient for continued operation of the RHR pumps.*
440.20 We state in the SRP (e.g., in Section 15.1) that for anticipated
- (15.0) transients, the most limiting plant systems single failure shall be identified and assumed in the analysis. Accordingly, describe the worst single failure for each events analyzed in Chapter 15 of your FSAR. Provide analyses including these postulated failures for the five most limiting events identified in your FSAR.
Response
The five most limiting analyzed Chapter 15 transients are:
- 1. Loss of Feedwater Heater-Manual Flow Control (Subsection 15.1.1)
- 2. Feedwater Control Failure-Maximum Demand (Subsection 15.1.2)
- 3. Pressure Regulation Downscale Failure (Subsection 15.2.1)
- 4. Generator Load Rejection with Failure of Bypass (Subsection 15.2.2)
- 5. Turbine Trip with Failure of Bypass (Subsection 15.2.3)
In reviewing the expected sequence of events utilized in simulating the plant performance for each of these transients, it was determined that postulating a single active safety-related component failure does not alter the transients. For the feedwater control failure - maximum demand transient in which credit is taken for full turbine bypass capacity, a single active component failure would result in the loss of one of the turbine bypass paths. However, the consequence of loosing one bypass path is not expected to result in fuel failure.
4 Attachment No. 5 Draft Responses to Mechanical Engineering Branch Questions
l
)
I s QUESTION 48 Which operational transients will be used for preoperational testing of the non-NSSS piping systems? Which system will be monitored and what locations will be instrumented?
RESPONSE
The operational transients to be used for peroperational testing of the non-NSSS piping systems will be provided by the Applicant. The systems to be monitored and what locations to be instrumented are addressed in
- the revised subsection 3.9.2.1.2. (Attached.)
1 l
= e N CkA]-4. FoQwgh 7 __
- 3. 9.2. f.2 PREOPERATIONAL TESTING OF NON-NSSS PIPING s
3.9.2.1.2.1 PREOPERATIONAL VIBRATION TESTING This subsection defines the general requirements for vibration testing of
~
piping systems as specified in Regulatory Guide 1.68 "Preoperational and Ini.tf al Startup Test Programs for Water-Cooled Power Reactors".gpecific vibration testing requirements are defined in ANSI /ASME OM3-1982,
" Requirements for Preoperational and Initial Startup Vibration Testing of Nuclear Power Plant Piping Systems". An outline of that standard is given here. Preparation of detailed test specifications by the Applicant will require consulting the complete standard. Instrumentation locations will be We provided by the Applicant in accordance withgvibration monitoring group selection (Subsection 3.9.2.1.2.1).
Piping systems to be tested are classified into three vibration monitoring groups, according to required degree of test sophisications. Vibration Monitoring Group 1 (VMGI) requires precise- test instrumentation plus some degree of mathematical analysis, where as VMG3 ma'y requiie only visual observations by competent personnel. VMG2 permits simpler instrumentation than VMGl. such as hand-held on temporarily mounted displacement meters or accelerometers. 3.9.2.1.2.1.1 VIBRATION MONITORING GROUP SELECTION In selecting monitoring graups for various systems or piping configuration, a general rule is that the most rigorous testing (VMG1) should be applied to systems where vibration has the greatest safety implication. Af an , example, VMG1 is applied to all Safety Class 1 Piping. VMG1 may also be applicable to Safety Class 2 and 3, if vibration-producing elements (pumps, compressors, relief valves, etc) are present. Table 3.9-24 lists systems being preoperationally tested for vibration. The system classification for vibration teseing shall be made by applicant. Preliminary testing may require changes in group selection if unforeseen problems develop.
% A s sv s.5 c. d b de m w k e ,5 oE M Swb
- sy 34e2.s ( sc_ v M ) why k d J h% P .
s +. Pm d 6 s) q , M o d ta on D ese be s % cttis82.
.g ee . . . . ........4.. . = = . e. . .
e
3.9.2.1.2.1.2 ACEPTANCE CRITERIA -- VMG1 - For steady - state vibrations, the maximum calculated alternating stress intensity S should be limited to: alt For ASME Class 1 piping: M CK64 gS al+ 2 1 2 el Where C2 = Secondary stress index as defined in ASME Code. K2 = Local stress index as defined in ASME Code. M = Maximum zero-to-peak dynamic moment due to ' vibration only, or in combination with other loads as required by system design specification. F, = Factor of safety applicable to class 1 piping (=1.3). Where 4be. 4k user d.c w w s+c.4es am\ 34tc.\\3 ov- \.3 ,p.n *w u. A4 4b V G - 2. vw.eAk e as o.a s n 6%4h zoss enehm L sk \ad r Frekw o F t.s ,-+ ke 4 s or o f s o $ e.h > f= S y 3 w e.e.d w el be m edw do-cl t w -h e a t e Abe o F Soi+ . a: o S fov- modened s cove d h Fs3ure I-s.l o E S s.chsow IIE of 4he. A 3 M s co de . ed =. O.G 4ov mahcv sod covevdt) lo3 Psep % I - 9.'2. o9 se.ch ur J4L A ses to cA e . S* - s wel www o. hvwd (.Sn)-Evom I:s$u s I-3.I or I - 9 . 2. of Sa.ckten M of 4h A S M s a de. . 7= S e.ckson h e ch41 wJ ok b psf 2 __y ,-, ,.m . - _ - e " " ~ ~
. I \ i l
For ASME Class 2 aiid 3, and ANSI B31: a S gs zc K2 Fs 6 a S g Where 22 C K =21, i = Stress intersification factor, as defined in : Subsections
") .
NC and ND of the ASME Code or 831.
~
For transient vibrations, the and maximum alternating stress should be limited as follows. . For ASME Class 1 piping, S,= allowable alternating peak stress value from Figure I-9.1 or I-9.2 using N, w}ere Ny = (ERC) EVLC = Equivalent number of maximum antiSipated vibratory load cycles. U, = Unused usage factor = 1-U U
= Cumulative usage factor from ASME Class 1 analysis, which excluded vibratory code.
The maximum alternating stress intensity S alt shall be limited to 0.8 SEL for carbon steel, or 0.6 S for stainless steel. For ASME Class 2, 34 and EL 831 piping, the stresses shall be evaluated the same as for steady state vibration. A Merwak'h +k a f P wpyta+e A w S T co de sbH be us e d +o Ev a k a44. 4k 4. .s ta s s e. s -foe 4% ne d vi kv-dso n s a C 3.9.2.1.2.1.3 gEPTANCE CRITERIA -- VMG-2 For testing utilizing deflection measurements, acceptance is based on the follow'ing equation.
'S = d at g Where S,j = allowabYzerNo'plak de lection limit l
% A 5, =
Valu2 of Mcti n obtained
'all 'other symbols are the same ,as given for VMG1. For testing utilizing volocity measurement, acceptance is based on'the following equations.
3.64 x 10 3 V,j j, = C3C4 , 3g C3 ps ,
. C 2b ,
Where V,jj,, = Allowable velocity, in/second
=
C 3 Correction factor to compensate for the effect of concentrated
/
weights along the characterestic span (sec. Fig.10 of OM3-1982)
=
C 3 a correction factor accounting for pipe contents and insulation
= 1.0 + W p + W INS W W Where W= weight of the pipe per unit length (Ib/ft)
Wp= weight of the pipe contents per unit length (Ib/ft) WINS =, 'the weight of the insulation per unit length (Ib/ft)
=
1.0 for pipe without insulation and either empty or containing steam C4= correctior, factor for end conditions different from fixed ends and for ccnfigurations different from stra'ight spans
=
1.0 for a straight span fixed at both ends, but conservative for any practical end conditions for straight spans of. pipe
=
1.33 for cantilever and simply supported pipe span
=
0.74 for equal leg Z-bend
- =
0.83 for equal leg U-bend. Appendix D of OM-3 presents examples of correction factors C and C for 1 4 typical piping spans along with a combination of these factors to provide an initial screening method.
- t. '
3.9.2.1.2.1.4 AJEPTANCE CRITERIA -- VMG3 1 The acceptability of piping is deterrined by visual observation, or employing simple devices such as rules, optical wedge, or spring scale. If the level of vibration is too small to be perceived and t,he possibility of damage is judged to be minimal, the system is acceptable. The judgement as to acceptability can be made only by evaluation of all the following facts as i to their effects on piping stress. a) Vibration magnitude and location ~ e ,e e e e e- e .ee e e , e ee # +=
s a - I 8 b) Proximity to sensitive equipment c) Branch connection behavior d) Capability of nearby component supports Any unique operational characteristics of the systems shall be considered in the evaluations. - If unacceptable vibration levels are indicated by the method listed above, the system must be reclassified as either VMG2 or VMG1. ' 3.'9. 2.1. 2.1. 5 CORRECTIVE ACTION , Should the piping vibration exceed the accaptance criteria, correction action must be taken to nake the system acceptable. This action may consist of adding supports, reducing . forcing functions, determining and modifying resonant sections, or changing operating conditions. After corrective action is taken, additional testing shall be performed to determine if the vibrations have been sufficiently reduced to satisfy the acceptance criteria. 3.9.2.1.2.2 PREOPERATIONAL THERMAL EXPANSION AND DYNAMIC TESTING Preoperational thermal expansion and dynamic testing is provided in Subsection 14.2.12.1.75. l l e
' Table 3.9-24 PIPING SYSTEMS TO BE VIBRATION TESTED / INSPECTED (PREOPERATIONAL) i l
Reactor Feedwater System Reactor Water Cleanup System Standby Liquid Control System Residual Heat Removal System ' Rea'ctor Core Isolation Cooling System Reactor Recirculation System Controlled Drive Hydraulic Systems Low Pietsure Cece Spray System - High Pressure Core Spray System - Fuel Pool Cooling and Cleanup System Leak Detection System Li.ould, and Solid Radwaste Sygterns, . Neutrori Men'1toring' System , Offgas System Upper Pool Storage System -- N I Chilled Water System
~
Demineralized Water (and Condensate Distribution Essential Service Water System . Heated Water Distribution HPCS Service Water System Suppression Pool Make-up System Suppression Pool Cleanup Essential Bldg. Chilled Water RHR Service Water Systems Condensate System " Rea b- Pr,4a.cMow q.slem O e e o 4
. e=q 4 9 9 * .me .
\ -
Attachment No. 6 l t i Draft Update of GESSAR II Relative to R.G.'s 1.31 and 1.44 and NUREG-0313 l l l l
GESSAR II 22A7007
* .' 238 NUCLEAR ISLAND Rsv. 14 1.8.31 Regulatory Guide 1.dl, Revision 3, Dated April 1978 ,
Title:
Control of Ferrite Content in Stainless Steel Wald Metal This guide describes an acceptable method of implementi'ng the h rekuirementsofGeneralDesignCriterion1ofAppendixAto l 10CFR50'and Appendix B to 10CFR50 when fabricating and joining o austenitic stainless steel components and systems. Evaluation As discussed in Subsection 5.2.3, the GESSAR II design complies with this guide but, as discussed below, uses the alternate sition for coated electrodes as defined in method of weld pad de ASME Section III NB . GESSAR II implements the provisions of this regulatory guide. As a result of discussions with the NRC staff, GE conducted a program to demonstrate that the controls applied to filler metal
- provide adequate delta ferrite in production welds. This concept was accepted by the NRC and is stated in Revision 3, with corollary scope and application requirements.
For GESSAR II plants, all austenitic stainless steel weld filler metal for Class 1,, Class 2, and core support components will be supplied with ferrite compositions as defined by Regulatory Guide 1.31, Rev. 3. The weld filler metal additionally must comply with
- the chemical analysis requirements of Section III of the American
- Society of Mechanical Engineers Boiler and Pressure Vessel Code.
The weld pad deposition te-hnique for coated electrodes stated in ASME Section III, NB2400 is considered an acceptable alternate to the AWS AS.4 technique rccommended in Regulatory Guide 1.31, Rev. 3. I __ y ' NEWj l.8.31-1/1.8.31-2 l
)
~~
GESSAR II 22A7007
. 238 NUCLEAR ISLAND Rev. 14 1.8.44 Regulatory Guide 1.44. Revision ~0.' Dated May 1973
Title:
Control of the Use of Sensitized Stainless Steel This. guide describes acceptable methods of implementing the requirements of GDC 1 and 4 of Appendices A and B to 10CFR50, with regard to control of the application and processing of stainless steel to avoid severe sensitization that could lead to stress corrosion cracking. This guide applies to light-water-cooled reactors.
~ Evaluation The GESSAR II design complies with this regualtory guide and with the guidelines of NUREG 0313, revision 1 as well.
All applications of nuclear grade stainless steel are specified as either 304L or 316L (or LN) grade. See revised GESSAR subsections 5.2.3.4 and 4.5.2 for additional discussion. As stated, the General Electric Company is complying with the intent of R. G.1.44 by controlling the application and processing of stainless s. teel to avoid severe sensitization that could lead to stress corrosion cracking through the use of IGSCC resistant materials. In addition, stress rulo evaluation is being performed on GE scope of supply to predict other areas where IGSCC might be possible due to high stress. This effort will allow appropriate modifications to be made where appropriate. ! Induction Heating Stress Improvement (IHSI) treatment on stainless steel wold i joints to preclude intergranular stress corrosion cracking will also be considered and implemented by GE and plant owners when approved the NRC.
~,
wsw 1.8.44-1
L GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 4.5 REACTOR MATERIALS 4.5.1 Control Rod System Structural Materials
. 4.5.1.1 Material Specifications
- a. Material List N !
The following material listing applies to the control rod drive mechanism supplied for this application. The position indicator and minor nonstructural items are omitted. (1) Cylinder, Tube and Flange Assembly Flange , ASME SA182 Grade F304 . (200 Plugs ASME SA182 Grade F304 - < 000 Cylinder ASTM A269 Grade TP 304 - 4 2 DD Outer Tube- ASTM A269 Grade TP 304 < 0 00 Tube ASME SA351 Grade CF Spacer ASME SA351 Grade CF ; (2) Piston Tube Assembly Piston Tube ASME SA479 or SA249 Grade XM-19 Nose ASME SA479 Grade XM-19 Base ASME SA479 Grade XM-19 Ind. Tube ASME SA312 Type 316 .(
- Cap ASME SA182 Grade F316 . .g (3) Drive Line Assembly Coupling Spud Alloy X-750 Compression Cylinder ASME SA479 or SA249 Grade XM-19 Index Tube ASME SA479 or SA249 Grade XM-19 O '
Piston Head ARMCO 17-4 PH or its equivalent j
. l 4.5-1 j
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4.5.1.1 Mat 0rici Specificctiena (Continued) ic l bxceg}3 4 Piston Coupling . ASTM A312 Grade TP 304 or {eno SY.h ASTM A269 Grade TP 304 Magnet Housing ASTM A312 Grade TP,304 or (gu) di
,j ,
ASTM A269 Grade TP 304 or ASTM A312, A249, or A213 ,; 2 TP 316L . (4) Collet Assembly Collet Piston ASTM A269 TP 304 or
- ASTM A312 TP 304 -
Finger Alloy X-750 Retainer ASTM A269 TP 304 Guide Cap ASTM A269 TP 304 (5) Miscellaneous Parts 4 Stop Piston ARMCO 17-4 PH or its equivalent 0-Ring Spacer ASTM A240 Type 304 - Nut ASME SA479 Grade XM-19 Collet Spring Alloy X-750 Ring Flange ASME SA182 Grade F304 Buffer Shaft ARMCO 17-4 PH or its equivalent Buffer Piston ARMCO 17-4 PH or its equivalent Buffer Spring Alloy X-750 Nut (hex) Alloy X-750 The austenitic 300 series stainless steels listed under ASTM /ASME specification number are all in the annealed condition (with the exception of the outer tube in the cylinder, tube and flange assem-l bly), and their properties are readily available. The outer tube is approximately 1/8 hard, and has a tensile of 90,000/125,000 psi, yield of 50,000/85,000 psi and minimum elongation of 25%. 4.5-2
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GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. O 4.5.1.1 Material Specifications (Continued) The coupling spud, collet fingers, buffer spring, nut (hex), and collet spring are fabricated from Alloy X-750 in the annealed or equalized condition, and aged 20 hours at 1300*F to prod'uce a ten-sile of 165,000 psi minimum, yield of 105,000 psi minimum, and elongation of 20% minimum. The piston head, stop piston, buffer shaf t, and buffer piston are ARMCO 17-4 PH (or its equivalent) in condition H-1100 (aged 4 hours at 1100*F), with a tensile of 140,000 psi minimum, yield of 115,000 psi minimum, and elongation of 15% minimum. These are widely used materials, whose properties are well known. The parts are readily accessible for inspection and replaceable if necessary. All materials, except SA479 or SA249 Grade XM-19, have been suc-1 cessfully used for the past 10 to 15 years in similar drive mech-anisms. Extensive laboratory tests have demonstrated that ASME SA479 or SA249 Grade XM-19 are suitable materials and that they are resistant to stress corrosion in a BWR environment.
- b. Special Materials No cold-worked austenitic stainless steels with a yield strength greater than 90,000 psi are employed in the Control Rod Drive (CRD) system. ARMCO 17-4 PH (or its equivalent) (martensitic precipita-tion hardened stainless steel) is used for the piston head, stop l
piston, buffer shaft, and buffer piston. This material is aged to the H-1100 condition to produce resistance to stress corrosion cracking in the BWR environments. ARMCO 17-4 PH (or its equiva-lent) (H-1100) has been successfully used for the past 10 to 15 years in BWR drive mechanisms. l l J 4.5-3 .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 4.5.1.2 Austenitic Stainless Steel Components
)
- a. Processes, Inspections and Tests
~
Two,special processes are employed'which subject selected 300 Series stainless steel components to temperatures in the sensitization range: (1) The cylinder and spacer (cylinder, tube and flange assembly) and the retainer (collet assembly) are hard surfaced with Colmonoy 6 (or its equivalent) . (2) The collet piston and guide cap (collet assembly) are nitrided to provide a wear-resistant surface. Nitriding is accomplished using a proprietary process. Components are exposed to a temperature of about 1080*F for about 20 hours during the nitriding cycle. Colmonoy (or its equivalent) hard-surf aced components have per-formed successfully for the past 10 to 15 years in drive mechanisms. Nitrided components have been used in CRDs since 1967. It is nor-mal practice to remove some CRDs at each refueling outage. At this time, both the Colmonoy (or its equivalent) hard-surfaced parts and nitrided surfaces are accessible for visual examination. In addi-tion, dye penetrant examinations have been performed on nitrided surfaces of the longest service drives. This inspection program is adequate to detect any incipient defects before they could become i serious enough to cause operating problems. Regulatory Guide 1.44 Discussion of the degree of conformance to Regulatory Guide 1.44 is provided in Subsection 4.5.2.4. -
)<
4.5-4 l l 1
l GESSAR II 22A7007 l
, 238 NUCLEAR ISLAND Rev. 0 4.5.1.2 Austenitic Stainless Steel Components (Continued)
- b. Control of Delta Ferrite Content Discussion of this subject and the degree of conformance to Regula-tory Guide 1.31 is presented in Subsection 4.5.2.4.
4.5.1.3 Other Materials These are discussed in Subsection 4.5.1.1.b. 4.5.1.4 Cleaning and Cleanliness Control 4.5.1.4.1 Protection of Materials During Fabrication, Shipping and Storage All the CRD parts listed above (Subsection 4.5.1.1) are fabricated
- under a process specification which limits contaminants in cutting,
(/ grinding and tapping coolants and lubricants. It also restricts all other processing materials (marking inks, tape etc.) to those which are completely removable by the applied cleaning process. All contaminants are then required to be removed by the appropriate - cleaning process prior to any of the following: (1) Any processing which increases part temperature above 200*F. (2) Assembly which results in decrease of accessibility for cleaning. (3) Release of parts for shipment. The specification for packaging and shipping the Control Rod Drive provides the following: O 4.5-5
GESSAR II 22A7007 23b NUCLEAR ISLAND Rt.v. 0 4.5.1.4.1 Protection of Materials During Fabrication, Shipping and Storage (Continued)
?
The drive is rinsed in hot deionized water and dried in preparation for shipment. The ends of the drive are then covered with a vapor tight barrier with dessicant. Packaging is designed to protect the
; drive and prevent damage to the vapor barrier. Audits have indi-
'.. cated satisfactory protection.
Semiannual examination of the humidity indicators of ten percent of the units is required to verify that the units are dry and in satis-factory condition. This inspection shall be performed with a GE-Engineering designated representative present. Position indicator probes are not subject to this inspection. Site or warehouse storage specifications require inside heated
-storage comparable to level B of ANSI N45.2.2.
The degree of surface cleanliness obtained by these procedures meets the requirements of Regulatory Guide 1.37. Regulatory Guide 1.37 General Compliance or Alternate Approach Assessment: For Commit-ment and Revision Number, see Section 1.8. 4.5.2 Reactor Internal Materials 4.5.2.1 Material Specifica'tions Materials used for the Core Support Structure: Shroud Support - Nickel-Chrome-Iron-Alloy, ASME SB166 or SB168. 4.5-6
teidsAnsx Rav. 0 238 NUCLEAR ISLAND i 4.5.2.1 Material Specifications (Continued) i Shroud, core plate, and grid - ASME SA240, SA182, SA479, SA312, , SA249, or SA213 (all Type 304L) . ~
~
Peripheral fuel supports - ASTM A317 Grade TP-304, A479 Type-316L, ASME SA312 Grade Type-304L I Core plate and top guide studs and nuts, and core plate (all wedges - ASME SA479, SA193 Grade B8A', SA194 Grade 8A Type-304) Control rod drive housing - ASME SA312 TP-304, SA182 Type-304, and ASME SB167 Type Alloy 600. Control rod guide tube - ASME SA358 Grade 304, SA312 Grade TP-304; ASTM A358 Grade 304, A312 Grade TP-304, A351 Grade CF8, A249 TP-304. I Orificed fuel support - ASTM A249 TP-304, A240 TP-316L, A479 TP-316L. Materials Employed in Other Reactor Internal Structures. i (1) Shroud Head and Separators Assembly and Steam Dryer Assembly All materials are TP-304, 304L or 316L stainless steel. Plate, Sheet and Strip ASTM A240, TP-304, 304L or 316L Forgings ASTM A182 Grade F304 or 304L Bars ASTM A276 TP-304 or 316L
- Pipe ASTM A312 Grade TP-304 i f
1 4.5-7
GESSAR II yplyv(sygy
- 238 NUCLEAR ISLAND RLv. 0 4.5.2.1 Material Specifications (Continued \
)
Tube ASTM A269 Grade TP-304 s Castings ASTM A351 Grade CF8 ,- .' (2) Jet Pump Assemblies The components in the Jet Pump Assemblies are a Riser, Inlet Mixer, Diffuser, and Riser Brace. Materials used for these components are to the following specifications: Castings ASTM A351 Grade CF8 and ASTM SA351 Grade CF3 ASTM A276 TP-304, Bars ASTM A479 TP-316L ASTM A637 Grade 688 Bolts ASTM A193 Grade B8 or B8M and ASME SA479 TP-316L ASTM A240 TP-304, and Sheet and Plate ASME SA240 TP-304L, 316L Pipe ASTM A358 TP-304, 316L and ASME SA312 Grade TP-304, 316L Forged or Rolled Parts ASME SA182, Grade F304, F316L, ASTM B166, and ASTM A637 Grade 688. i 4.5-8 l
GESSAR II h 1 238 NUCLEAR ISLAND Rsv. 0 4.5.2.1 Material Specifications (Continued) ; i Materials in the Jet Pump Assemblies which are not austenitic j stainless steel are listed below: - The Inlet Mixer Adaptor casting, the wedge casting, a. bracket casting adjusting screw casting, and the Diffuser l collar casting are hard surfaced with Stellite 6 (or its f I equivalent) for slip fit joints.
- b. The Diffuser is a bimetallic component made by welding I
an austenitic stainless steel ring to a forged Alloy l l 600 ring, made to Specification ASTM B166.
- c. The Inlet-Mixer contains a pin, insert, and beam made of Alloy X-750 to Specification ASTM A637 Grade 688.
All core support structures are fabricated from ASME specified materials, and designed in accordance with requirements of ASME ' Code, Section III, Subsection NG. The other reactor internals are noncoded, and they are fabricated from ASTM or ASME specification materials. Material requirements in the ASTM specifications are identical to requirements in corresponding ASME material f specifications. I l 4.5.2.2 Controls on Welding {
/
Core support structures are fabricated in accordance with require- l ments of ASME Code Section III, Subsection NG. Other internals are Requirements of ASME re;;t required to meet ASME Code requirements. Section IX BPV Code, are followed in fabrication of core support structures. I J l 4.5-9
& U m w nr GESSAR II Ray. 0 238 NUCLEAR ISLAND Nondestructive Examination of Wrought Seamless }
4.5.2.3 Tubular Products a Wrought seamless tubular products for CRD housings, and peripheral fuel supports, were supplied in accordance with ASME Section:III, l ClassI CS, which requires examination of the tubular products by radiographic and/or ultrasonic methods according to para- I graph NG-2550. Wrought seamless tubular products for other internals were supplied in accordance with the applicable ASTM or ASME material specifica-tions. These specifications require a hydrostatic test on each ; length of tubing. 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel - j Regulatory Guide Conformance l l Regulatory Guide 1.31: Control of Stainless Steel Welding ;
) !
1 tcld-worked stainless steels are not used in the reactor internals except for vanes in the steam dryers. The delta ferrite content l for weld materials used in welding austenitic stainless steel assem-l blies is verified on undiluted weld deposits for each heat or lot l of filler metal and electrodes. The delta ferrite content is l defined for weld materials as 5.0 FN minimum and 8.0 FN average ; This ferrite content is considered adequate to (Ferrite Number) . in austenitic stainless prevent any micro-fissuring (Hot Cracking) steel welds. This procedure complies with the requirements of Regulatory Guide 1.31. l Regulatory Guide 1.44: Control of the Use of Sensitized Stainless l Steel Proper solution annealing of the 300 series austenitic stainless steel is verified by testing per ASTM-A262, " Recommended Practices for Detecting Susceptibility to Intergranular Attack in Stainless i 4.5-10
GESSAR II 22A7007
. ',, 238 NUCLEAR ISLAND Rnv. 0
'.' 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel - Regulatory Guide Conformance (Continued)
Steels." Welding of austenitic stainless steel parts is performed
~in accordance with Section IX (Welding and Brazing Qualification) and Section II Part C (Welding Rod Electrode and Filler Metals) of the ASME Boiler and Pressure Vessel Code. Welded austenitic stain-less steel assemblies require solution annealing to minimize the l po'ssibility of the sensitizing. However, welded assemblies are l i
dispensed from this requirement when there is documentation that welds are.not subject to significant sustained loads and assemblies have been free of service failure. Other reasons, in line with ! the regulatory guide, for dispensing with the solution annealing are that assemblies are exposed to reactor coolant during normal operation service which is below 200'F temperature or assemblies are of material of low carbon content (less than 0.025%) . These controls are employed in order to comply with the intent of the Regulatory Guide 1.44. Regu}atory Guide 1.37: Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants Exposure to contaminant is avoided by carefully controlling all cleaning and processing materials which contact stainless steel during manufacture and construction. Any inadvertent surface contamination is removed to avoid potential detrimental effects. Special care is exercised to insure removal of surface contami-nants prior to any heating operation. Water quality for rinsing,
- flushing, and testing is controlled and monitored.
The degree of cleanliness obtained by these procedures meets the requirements of Regulatory Guide 1.37. 4.5-11
GESSAR II 22A7007
. 238 NUCLEAR ISLAND RLv. 0 4.5.2.4 Fabrication and Processing of Austenitic Stainless ;
Steel - Regulatory Guide Conformance (Continued) Regulatory Guides 1.31, 1.44, and 1.37 , Geheral Compliance or Alternate Approach Assessment: For Commit-rent and Revision Number, see Section 1.8. 4.5.2.5 Other Materials Hardenable martensitic stainless steel and precipitation hardening stainless steels are not used in the reactor internals. Materials, other than Type-300 stainless steel, employed in reactor internals are: (1) SA479 Type XM-19 stainless steel; I (2) SB166, 167, and 168, Nickel-Chrome-Iron (Alloy 600); and (3) SA637 Grade 688 Alloy X-750. Alloy 600 tubing plate, and sheet are used in the annealed condi-tion. Bar may be in the annealed or cold-drawn condition. Alloy X-750 components are fabricated in the annealed or equalized condition and aged when required. Stellite 6 (or its equivalent) hard surfacing is applied to some austenitic stainless steel castings using the gas tungsten arc welding or plasma arc surfacing processes. All materials, except SA 479 Grade XM-19, have been successfully used for the past 10 to 15 years in BWR applications. Extensive laboratory tests have demonstrated that XM-19 is a suitable material and that it is resistant to stress corrosion in a BWR environment. 4.5-12
GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rsv. 0 4.5.3 Control Rod Drive Housing Supports All CRD housing support subassemblies are fabricated of ASTM-A-36 . structural steel, except for the following items:
Material Grid ASTM A441 Disc springs Schnorr Type BS-125-71-8 (or its equivalent) Hex bolts and nuts ASTM A307 6 in. x 1 in. x 3/8 in. ASTM A500 Grade B tubes For further CRD housing support information, see Subsection 4.6.1.2. i 4.5-13/4.5-14 l l l
(32SSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 5.2.3.3.4 Moisture Control for Low Bydrogen, Covered Arc 3 Welding Electrodes (Continued) Electrodes are distributed from sealed containers or ovens as required. At the and of each work shift, unused electrodes are returned to the storage ovens. Electrodes which are damaged, wet, or contaminated are discarded. If any electrodes are inadvertently left out of the ovens for more than one shift, they are discarded or reconditioned in accordance with manufacturer instructions. 5.2.3.4 Fabrication and Processing of Austenitic Stainless Steels 5.2.3.4.1 Avoidance of Stress / Corrosion Cracking 5.2.3.4.1.1 Avoidance of Significant Sensitization (- Regulatory Guide 1.44 addresses 10CFR50, Appendix A, GDCs 1 and 4, and Appendix B, requirements to control the application and proc-essing of stainless steel to avoid severe sensitization that could lead to stress / corrosion cracking. All austenitic stainless steel is purchased in the solution-heat-treated condition in accordance with applicable ASME and ASTM specifications. Cooling rates from solution heat treating temperatures are required to be rapid enough to prevent sensitization. Non-sensitization is verified using ASTM A262, Practice A methods. Material changes have been made to minimize the possibili,ty of w Fk intergranular stress / corrosion cracking (IGSCC) . All , wrought austenitic stainless steel in the reactor coolant pressure e ... ..e T m == boundary,'is low carbon ype 3 6,es= Stet with 0.02% maximum
- - L on LN ' l
/&b ull M Sust chW ta h s[e.v eur m en a=k*
5.2-39
Rsv. (6 238 NUCLEAR ISLAND 6- ._- 5.2.3.4.1.1 Avoidance of Significant Sensitization (Continued) a ,% y e.S * ,, carbon content g There is no piping which is sarvice sensitive re.ut s' ion ( , or nonconforming as defined in NUREG-0313 3
' "- d E _.'.t ' - :lds wie " ;== fin ==: m are -(maWLar.^
'_-t_' r"' __ 11 11 '- '
wge- - '""^ p h p 'l
-_C y ; 1;.L_1121. 3 r-g; .._ _ i. Lie... . - Men* *Wiraving u - 4%en.
I;/.k - sec " _~ yq - _ '_ r5- - r--
' -- t eI~ l ,~ .;;t _ a 4' ' ~"md r g"8 - -"* 7 --
ti==? t!. -_1- r i===e eI 1... wietf=44= h- T E = , ,.u. = -- -. . . ras win C . h ~/16 .uun; ,vi ana.. , - . ~ _._E j1- For machine,
..;'.;c . :/
2a-41ertrr'- crr: ri=. C cutomatic, and manual weldingg = , _ _ . ..._ _7.==rM'niit=ad
.- . i
_ . L. ;.m -_.,m.-_,--- 7_ t ~, :.. _ _ % i .__ knterpass temperature restricted to 350 F for all stainless cteel welds. High heat welding processes such as block welding and electroslag welding were not permitted. All weld filler metal and castings mee are required by specification to have a minimum of ; 5% ferrite. l s n ti ta 1 ss te 1 w s ea to ( W h e n o d- =- y/ ro gh n el o 1 t e a e r
' o ti n ea ea 'd.
c
, m te a w -
These controls were used to avoid severe sensitization and to f comply with Regulatory Guide 1.44, Control of the Use of Scnsitized ' Stainless Steel. For commitment and revision, see Section 1.8. i 5.2.3.4.1.2 Process Controls to Minimize Exposure to Contaminants Exposure to contaminants capable of causing stress /corrosiTon l cracking of austenitic stainless steel components was avoided by l I l { 5.2-40
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