ML20083F276
| ML20083F276 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/30/1983 |
| From: | Ma W, Mcinerney J, Swamy S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19268E619 | List: |
| References | |
| MT-SME-3179, NUDOCS 8312300235 | |
| Download: ML20083F276 (58) | |
Text
- _ ___ _ _
i.
MT-StiE 3179 ENCLOSURE B TECHNICAL BASES FOR ELIMINATING LARGE PRIMARY LOOP PIPE RUPTURES AS THE STRUCTURAL DESIGN BASIS FOR CATAWBA UNITS 1 AND 2 Prepared by:
Winston Ma S..A. Swamy J. J. McInerney November, 1983 Approved:
b)
Approved:
J. 5. Chirigos, Manage *r
- E. R. Johnson, Manager Structural Materials Structural and Seismic
~ ' ~
Engineering Development Approved:
~ hf h W. S. Brown, Manager Mechanical Equipment and Systems Licensing
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A TABLE OF CONTENTS Section Title Page
1.0 INTRODUCTION
1 2.0
-0PERATION AND CHEMICAL STABILITY OF THE PRIMARY 4
COOLANT SYTEM
-3.0-PIPE GE0 METRY AND LOADING ~
6 4.0 FRACTURE MECHANICS EVALUATION 8
5.0 LEAK RATE PREDICTIONS 12 6.0 FATIGUE CRACK GROWTH ANALYSIS 13
7.0 CONCLUSION
S 15
8.0 REFERENCES
16
- APPENDIX A 18
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0455e:1/11883-iii
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1.0 INTRODUCTION
-1.1 Purpose _,
The current structural design basis for the _ reactor coolant system (RCS) primary' loop requires that pipe breaks be-postulated as defined in the approved Westi~nghouse Topical Report WCAP 8082, Reference 1.-
In addition, protective measures' for the dynamic effects associated with RCS primary loop pipe breaks have been incorporated in the Catawba plant design. -However,
- Westinghouse has demonstrated on a generic basis that RCS primary loop pipe breaks are highly unlikely and should not be included in the structural design basis of Westinghouse plants (see Reference 2). The purpose of this report is to demonstrate that the generic evaluations performed by Westinghouse are applicable to _the Catawba plant.
In order to demonstrate this applicability, Westinghouse has performed a comparison of the loads and geometry for the Catawba plant with envelope parameters used in the generic analyses (Section 3.0); fracture mechanics evaluation (Section 4.0); determination of leak rates from a-through-wall crack (Section 5.0), fatigue crack growth evaluation (Section6.0);andconclusions(Section7.0).
1.2 Scope
. This report applies' to the Catawba plant reactor coolant system primary loop
- piping.
It is-intended to demonstrate that specific parameters for the Catawba plant are enveloped by the generic analysis peformed by Westinghouse in_WCAP-9570 (Reference 3) and accepted.by the NRC-as noted in a letter from HaroldLDentondatedMay2,1983(Reference 4).
1.3 Objectives The conclusions of this report (Reference 3) support the elimination of RCS primary ' loop pipe breaks for the Catawba plant.
In order to validate this conclusion the following objectives must be achieved.
1 I
OdBt 1/11 AA'.1...
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- a.. Demonstrate that Catawba plant parameters are enveloped by generic Westinghouse studies.
b.. Demonstrate that margin exists between the critical crack size and a
-postulated crack which yields a detectable leak rate.
- c. : Demonstrate that there is sufficient. margin between the leakage through a postulated crack and the leak detection capability of the Catawba plant.
d.
Demonstrate that fatigue crack growth is negligible.
1.4-Backaround Information Westinghouse has performed considerable testing and analysis to demonstrate that RCS primary loop pipe breaks can be eliminated from the structural design basis of all Westinghouse plants. The concept of eliminating pipe breaks in the RCS primary loop was first presented to the NRC in 1978 in WCAP 9283 (Reference 5).
This Topical Report employed a deterministic fracture mechanics evaluation and 'a probabilistic analysis to support the elimination of RCS primary loop. pipe breaks.
~
This approach was then used aY,a means of addressing Generic Issue A-2. and Asymmetric LOCA Loads. Westinghouse performed additional testing and analysis to justify the elimination of'RCS primary loop pipe breaks. As a result of this effort, WCAP 9570 was su'bmitted to the NRC. The NRC evaluated the technical merits of_this. concept and prepared a draft SER in late 1981 endorsing this concept. xddi.tionally, both Harold Denton and the ACRS have endorsed the technical acceptability of the Westinghouse evaluations.
.Specifically, in' a May 2,1983' letter (Reference 4) Harold Denton states that
"... it is technically satisfied with Westinghouse Topical Report 9570 Rev.
2...." Additionally, the ACRS stated in a June 14, 1983 letter (Reference 6) that "... there is no known mechanism in PWR primary piping material for developing a large break without going through an extended period during which the crack would leak copiously."
0455e:1/11883 2
~
y The NRC funded research through Lawrence Livermore National Laboratory (LLNL) to address this same issue using a probabilistic approach. As part of the LLNL research effort, Westinghouse performed extensive evaluations of specific plant loads, material properties, transients, and system geometries to demonstrate that the analysis and testing previously performed by Westinghcuse and the research performed by LLNL applied to all Westinghouse plants including Catawba (Referen'ces 7 and 8). The results from the LLNL~ study were' released at'a March 28, 1,983 ACRS Subcommittee meeting. These studies which are applicable to all Westinghouse plants east of the Rocky Mountains, determined the mean probability of a direct LOCA (RCS primary loop pipe break) to be 10-10 per reactor year and the mean probability of en indirect LOCA to be 10-7 per reactor year. Thus, the results previously obtained by Westinghouse (Referencei5) were confirmed by an independent NRC research study.
The above studies estabJish the technical acceptability for eliminating pipe breaks from the Westinghouse RCS primary loop. The LLNL study has been shown applicable to Catawba plant by inclusion of plant specific data. This report will demonstrate the applicability of the Westinghouse generic evaluations to the Catawba plant.
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3 0455e:1/11883
2.0 OPERATION AND CHEMICAL STABILITY OF THE PRIMARY COOLANT SYSTEM
~ The Westinghouse reactor coolant system primary loop has an operating history (over 400 reactor years) which demonstrates its inherent stability characteristics. Additionally, there is no history of cracking in RCS primary
- loop piping.- In addition to the fracture' resistant materials used in the piping system, the chemistry of the reactor coolant is tightly controlled and variations in temperatures, pressure and flow during normal operating conditions'areLinsignificant.
As-stated above, the reactor coolant chemistry is maintained within very specific limits.
For-example, during normal operation oxygen in the coolant is limited to less than [.
]* This stringent oxygen limit is
+a,c, achieved by controlling charging flow chemistry and maintaining hydrogen in the reactor coolant at a concentration of [.
]+ The
+a,c, oxygen concentration in the reactor coolant is verified by routine sampling and chemical analysis. Halogen concentrations are also stringently controlled by maintaining concentrations of chlorides and fluorides at or below
.[-
-]*
This concentration is assured by controlling charging flow
+a,c, chemistry and specifying proper wetted surface materials. Halogen concentrations are also verified by routine chemical sampling and analysis.
-In order to ensure dynamic system. stability, reactor coolant parameters are
-stringently controlled. Temperature'during normal operation is maintained within [
']+ by control rod position. Pressure is controlled by
+a,c,(
~
pressurizer heaters and pressurizer spray, to a variation of less than
[L
.-]+ for steady state conditions..The flow characteristics of the
+a,c,e system remain constant during a fuel cycle because the only governing parameters, namely, ~ system resistance and the reactor coolant pump characterisites are controlled in the design process. Additionally,
Westinghouse has instrumented typical reactor coolant systems to verify the flow characteristics of the system.
0455e:1/11883 4
The.~rea.ctor coolant system, including piping and primary components, is designed for normal, upset, emergency and faulted condition transients. The design requirements are conservative relative to both the number of transients and their severity.
2
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r 0455e:1/11883 5
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~
3.0 PIPE GE0 METRY AND LOADING A segment of the primary coolant hot leg pipe is shown in Figure 1.
This segment is postulated to contain a circumferential through-wall flaw. The inside diameter and wall thickness of the pipe are 31.0 and 2.61 inches, respectively. The pipe is subjected to a normal operating pressure of
[
]+ psia.- The design calculations indicate that the junction 0-f the [
+a,(
]+ is most highly stressed.. At
+a,c this location the axial _ load, F, 'and the _ total bending moment M, are [1849]+
+a,(
kips (including the~ axial force due to pressure) and [-
~]+in-kips,
+a,c; respectively. Figure 2 identifies the loop weld locations. The material-properties and the loads at these locations resulting from Deadweight, Thermal Expansion and _ Safe Shutdown Earthquake are indicated in Table 1.
The method i
of obtaining these loads can be briefly summarized as follows:
The axial force F and' transverse bending moments, M and M, are chosen y
z for each static load (pressure, deadweight and thermal) based on elastic-static analyses for each of these load cases. These pipa load components are combined algebraically to define the equivalent pipe static l oads F,, Mys, and Mzs. Based on elastic SSE response spectra analyses, amplified pipe seismic loads,F, Myd'5d are btained. The maximum d
pipe loads are obtained by combining the static and dynamic load components as L
follows:
_F =
F
+
F s
d 2
M
+M M --[
y Z-where M
+
M My-ys yd M
+
M Mz" zs zd 4
0455e:1/11883 6
a The corresponding geometry and loads used in the reference report'(Reference
- 3) are as follows:
inside diameter and wall thickness are 29.0 and 2.5 inches; axial load and bending moment are [
]+inchkips.
+a,c The outer fiber stress for Catawba is [.
]+ ksi, while for the reference
.+a, c report it is [
.]+ksi.
This' demonstrates conservatism in the reference
+a,c report which makes _it more severe than the Catawba project.
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4.0 FRACTURE MECHANICS EVALUATION 4.1 Global Failure Mechanism Determination of the conditions which lead to failure in stainless steel must be done with plastic fracture methodology because of the large amount of deformation accompanying fracture. A conservative method for predicting the failure of ductile material is the [
]+.Thismethodologyhasbeenshowntobeapplicabletoductile
+a,c piping through a large number of experiments, and will be used here to predict the critical flaw size in the primary coolant piping. The failure criterion has been obtained by requiring [
]+ (Figure 3) when loads are applied. The detailed development is
+a,c provided in Appendix A, for through-wall circumferential flaw in a pipe with internal pressure, axial for_c.e.and' imposed bending moments. The [
.]+ for such a pipe is given by:
+a,c
+a,c,e 0455e:1/11883 8
~*
+a,c.e The analytical model described above accurately accounts for-the piping internal pressure as well as bposed axial force as they affect [
.]+
Good agreement was found between the analytical predictions and the
+a,c, experimentalresults[9].
4'. 2 Local Failure Mechanism The 1. cal mechanism of failure is primarily dominated by the crack tip behavior in terms of crack-tip blunting, initiation, extension and finally crack-instability. Depending on the material properties and geometry of the pipe, flaw size, shape and lo.a,d,ingr the local failure mechanisms may or may not govern the ultimate failure.
- The stability will be assumed if the crack does not initiate at all.
It has been accepted that the initiation toughness, measured in terms of J from a IN J-integral resistance curve is a material parameter defining the crack initiation.
If, for a given load, the calculated J-integral value is shown to be less than J f the material, then the crack will not initiate.
If the IN initiation criterion is not met, one can calculate the tearing modulus as defined by the following relation:
dJ E
T,pp = 3 3
- f 0455e:1/11883 9
where T,pp =' applied tearing modulus E = modulus of elasticity f=[
]*(flowstress) o
+a,c.e
~a = crack length
[o,o
- yield and ultimate strengt'1 of the material, respectively.]+
+a,c,e y u In summary, the local crack stability will be established by the two. step criteria:
J<J IN T,pp < T if J'> J mat 73 i
4.3 Results of Crack Stability Evaluation Figure 4 shows a plot of the [
]+ as a function of through-
+a,c, wall circumferential flaw length in the [ cross-over leg]+ of the main coolant
+a,c, piping. This[
]+ was calculated for Catawba data of a pressurized
+a,c, pipe at [
]+ with
+a,c,
]F foperties. The maximum applied bending moment
+a,c,.
ASMECodeminimum[
p of[-
.]+ in-kips can be plotted 'on this figure, and used to determine the
+a,c,,
critical flaw length, which 15 shown to be [
]+ inches. This is
+a,c, considerably larger than the [
]+ inch reference flow used in Reference 3.
+a, c, -
r
+a,c, 0455e:1/11883
.10 l
r g.
lt
- \\.
,g gt.
+a.c.e
.. [ Therefore, it can be concluded that a postulated [
]+ inch
+a,c.e through-wall flaw in the Catawba loop piping will remain stable 'from both a local and global stability standpoint.
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... a, t s
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-e
-t0455e:1/11883 11
~ -:;
o 5.0 LEAK RATE PREDICTIONS Leak rate calculations were performed in Reference 3 using an initial through-wallcrack[
]+.
The
+a,c, computedleakratewas[
']+ based on the normal operating pressure of
+a,c,
[
]+. psi.
[
+a,c,
]+
+a,c This computed leak rate [,
~]+ significantly exceeds the smallest
+a c.
detectable leak rate for the. plant.
The Catawba plant has a RCS pressure boundary leak detection system which is consistent with the requirements of. Regulatory Guide 1.45 and can detect leakage of 1 gpm in one hour. There is a factor of ['
'.3+ between the calculated leak rate and the~ Catawba plant
+a.c leak detection systems.
,,. v '-
0455e:1/11883 12
6.0 FATIGUE CRACK GROWTH ANALYSIS To determine the sensitivity of the primary coolant system to the presence of smalll cracks, a fatigue crack growth analysis was' carried out for the [
3+ region of a typical system. This region was selected
+a,c' because it is typically one of the highest stressed cross sections, and crack growth calculated here will be conservative for application to the entire primary coolant system.
A finite element stress analysis was carried out for the [
+a,c
]+ of a plant typical in. geometry and operational characteristics to any Westinghouse PWR System. [
]+ All normal, upset and test
+a,c r
conditions were considered, and circumferentially oriented surface flaws were j
postulated in the region, assuming the flaw was located in three different L
locations, as shown in Figure 6.
Specifically, these were:
- Cross Section A:
~
Cross Section B:
+a,c.e; Cross Section C:
Fatigue crack growth ' rate laws were used [-
]+ The law 'for stainless steel was
+a,c, derived from Reference 11,'with-a very conservative correction for R ratio, the ratio of minimum to maximum stress during a transient.
0455e:1/11883 13
_m_.____-___._-._______m_
h = (5.4 x 10-12) g 1nches/ cycl e 4.48 eff where K,ff - K,,x (14)M
" mini max
+a, The calculated fatigue crack ~$rowth for semi-elliptic surface flaws of circumferential orientation and various depths is summarized in Table 2, and shows that the crack growth is' very small, regardless [
+a,c
.]+
0455e:1/11883 j4.
7.0 CONCLUSION
S This report has estaolished the applicability of the generic Westinghouse evaluations which justify the.elihiination of RCS primary loop pipe breaks for the' Catawba plant as follows:
The loads, material properties, transients and geometry relative to a.
the Catawba RCS primary loop are enveloped by the parameters of WCAP
'9570.
b.
The critical crack length at the worst location in the RCS primary
-loop is [
]+ This is significantly greater than the
+a,
[.
]+ inches stable crack used as a basis for calculating leak rates
+a, in WCAP 9570.
c.
The leakage througn a [
]+ crack in the RCS primary loop is [
]+
- a, based on WCAP 9570. The Catawba plant has a RCS pressure boundary leak detection -system which is consistent with the requirements of Regulatory Guide 1.45 and can detect leakage of 1 gpm in one hour.
Thus, there is a factor.of.[
.]+ between the calculated leak rate and
+a, the Catawba plant 1sak detection' systems.
~
d.
Fatigue crack growth was determined for postulated flaws and was found to be extremely small over plant life and, therefore, is considered insignificant.
Based on the above, it is concluded that RCS primary loop pipe breaks should not be considered in the structural design basis of the Catawba plant.
l 0455e:1/11883 15
8.0 REFERENCES
1.
WCAP 8082 P-A, " Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop," Class 2, January 1975.
2.
Letter from ~ Westinghouse (E. P. Rahe) to NRC (R. H. Vollmer) dated May 11, 1983.
3.
WCAP 9570, Rev. 2, " Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack," Class 3, June 1981.
l 4.
Letter from NRC (H. R. Denton) to AIF (M. Edelman) dated May 2,1983.
l S.
WCAP 9283, "The Integrity of Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," Class 2, March, 1978.
6.
Letter from ACRS (J. J. Ray) to NRC (W. J. Dircks) dated June 14, 1983.
7.
Letter from Westinghouse (&.
P'. Rahe) to NRC (W. V. Johnston) dated April 25, 1983.
8.
Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated July 25, 1983.
9.
Kanninen, M. F., et al, " Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks" EPRI NP-192, September 1976.
- 10. Bush, A. J.,Stooffer, R. B. " Fracture Toughness of Cast 316SS Piping Material Heat No. 156576, at 600*F", f' R&D Memo No. 83-5P6EVMTL,'41, March 7,1983, Westinghouse Proprietary Class 2.
0455e:1/11883 16
y 11.: -Bamford,-W. -il., " Fatigue Crack Growth of. Stainless Steel Piping in a Pressurized Water Reactor. Environment" Trans. ASME Journal of Pressure-Vessel Technology Vol. 101, Feb. 1979.
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+a,c.e Y '~
0455e:1/11883 17
(
s APPENDIX A
+a,c 0455e:1/11883 18
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TABLE 2 FATIGUE CRACK GROWTH AT [
]+ (40 YEARS)
+a,c,e FINAL FLAW (IN)
INITIAL FLAW (IN)
[
]+ [
]+ [
]+
+a c,e i
0.292 0.31097 0.30107 0.30698 0.300 0.31949 0.30953 0.31625 0.375 0.39940 0.38948 0.40763 0.425 0.45271-0.4435 0.47421
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eference Fatigue Crack Growth Law for Inconel 600 ater Environr::ent at 500F.
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. Enclosure C CNS
~
.Duplicatio'nandphysicalseparationofcomponentstoprovideredundancyagainst other' hazards also protects against simultaneous failures due to local fires.
The Fire Protection System provides. fire detection eq'uipmer. for areas where potential for fire is. greatest or areas not. normally occupied by personnel.
Also, reliableisources of'either water, carbon dioxide or halon are provided to
. appropriate parts of;the station.
Reference:
Section 9.5.1 CRITERION 4 -LENVIRONMENTAL AND MISSILE DESIGN BASES Structures, systems and components important to safety shall be de-l signed toLaccommodats the effects of and to be compatible with the environmental conditions associated with normal operation,.mainte-nance -testing, and postulated accidents, including loss-of-coolant accidents.
These structures, systems and components shall be appro-priately protected against. dynamic effects, including the effects of
. missiles,. pipe whipping, and discharging fluids,.that may result
- from equipment failures and from events and conditions outside the nuclear power unit.
2 DISCUSSION:
I Structures,'sys'tems and components important to safety are designed to func-
. tion in.a manner which assures' public safety at all times. These structures,
'lsystemsandcomponentsarepostulatedfor;allworst-caseconditionsbyappro-priate missile barriers, pipe restraints, and station layout. The Reactor
- Building is capable of withstanding the effects of missiles originating outside
.the Containment such that no credible missile can result-in a loss-of-coolant accident.. The control room is designed to withstand such missiles as may be directed toward.it and still maintain the capability of cont' rolling the units.
i
-Class.1E electrica'l equipment is designed and' qualified to perform its. safety function (s) under.the harsh environmental conditions applicable to its location.
Emergency core cooling components are austenitic stainless steel or equivalent corrosion resistant material and hence are compatible with the containment atmosphereLover the full. range of exposure during the post-accident conditions.
Reference:
- Chapters 2.0,-3.0 and 6.0.
CRITERION 5 - SHARING OF STRUCTURES,~ SYSTEMS, AND COMPONENTS Structures, systems, and components important to safety shall not be shared.between nuclear power units unless it is shown that their ability to' perform their safety functions is nct significantly im-paired by.the sharing.
3.1-3-'
Rev. 9
's CNS Each rod' cluster control assembly is provided with a sensor to detect position-ting.at the bottom of its travel. :This condition is also alarmed in the Control Room.
Four ex-core,long ion chambers also detect asymmetrical flux distribu-tions indicative of rod misalignment.
Hovable in-core flux detectors and fixed in-core thermocouples are provided as operational aids'to the operator.
Chapter 7 contains further details on instrumentation and controls.
Information regarding the radiation monitoring
. system provided to measure environmental activity and alarm high levels is
. contained-in Chapter 11.
0verall reactivity control is' achieved by the combination of soluble boron and
. rod cluster control assemblies.
Long term regulation of core reactivity is Laccomplished by adjusting.the concentration of boric acid in the reactor coolant.
Short term reactivity control for power changes is accomplished by
.the Rod Control-System which automatically moves rod cluster control assem-blies. This system uses input signals including neutron flux, coolant tempera-ture, and turbine load..
Reference:
Chapters 7.0 and 11.0.
CRITERION 14 - REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary shall be designed, fabricated, erected,' and tested so as-to have an extremely low probability of abnormal. leakage, or rapidly propagating failure, and of gross rupture.
fDISCUSSION-The reactor coolant pressure boundary is designed to accommodate the system
. pressures and temperatures attained under all expected modes of plant opera-tion,: including all anticipated. transients, and:to maintain the stresses within applicable stress limits.
In addition to the loads imposed on the-
--piping under-operating conditions, consideration is also given to abnormal
-l loadings,such.as pipe rupture where. postulated and seismic loadings as dis-cussed inTSections 3.6 and 3.7.
The piping is protected from over pressure by means of pressure relieving devices as required by applicable codes.
Reacter coolant ~ pressure boundary materials selection and fabrication tech-niques assure a low probability of ' gross rupture or significant leakage.
The materials of construction of the reactor coolant pressure boundary are
~
.protectea by control of coolant chemistry from corrosion which might otherwise reduceLits structural integrity during its service lifetime.
The reactor coolant pressure boundary has provisions for inspections, testing and surveillance of. critical areas to assess the structural and leaktight integrity.
3.1-7 Rev. 9
CNS 3.6' PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING General Design Criterion 4 of Appendix A to 10CFR50 required that structures, systems, and components important to safety be protected from the dynamic effects of pipe failure.
This section describes the design bases and design measures to ensure that the containment vessel and all essential equipment in-side or outside the containment, including components of the reactor coolant pressure boundary, have been adequately protected against the effects of blow-down jet and reactive forces and pipe whip-resulting from postulated rupture of piping.
Criteria presented herein regarding break size, shape, orientation, and loca-tion are in accordance with the guidelines established by NRC Regulatory Guide 1.46, and include considerations which are further clarified in NRC Branch Technical Position MEB 3-1 and APCSB 3-1 where appropriate.
These criteria are intended to be conservative and allow a high margin of safety.
For those pipe failures where portions of these criteria lead to unacceptable consequences,
-l~ further analyses will be performed.
However, any alternative criteria will be adequately justified and fully documented.
.3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS INSIDE AND OUTSIDE CONTAINMENT
- 3. 6.1.1 Design Bases 3.6.1.1.1 Reactor Coolant System The Reactor Coolant System, as used in Section 3.6 of the Safety Analysis Report, is limited to the main coolant loop piping and all branch connection nozzles out to the first butt weld.
Dynamic effects are only considered for pipe breaks postulated at branch connections. The particular arrangement of the Reactor Coolant System, building structures, and mechanical restraints preclude the formation of l plastic hinges for breaks postulated to occur at the branch corrections.
Con-sequently, pipe whip and jet impingement effects of the postulated pipe break MEBl at these locations will not result in unacceptable consequences to essential Q102 components.
This restraint configuration, along with the particular arrangement of the Reactor Ccolant System and building structures, mitigates the effects of the jet from the given break such that no unacceptable consequences to essential components are experienced.
The application of criteria for protection against the effects of postulated breaks at the branch connections results in a system response which can be accommodated directly by the supporting structures of the reactor vessel, the steam generator, and the reactor coolant pumps. The design bases for postulated breaks in the Reactor Coolant System are discussed in Section 3.6.2.1.
3.6-1 Rev. 9
. =-
4 CNS U-
. Systems which do'not contain mechanical' pressurization. equipment are excluded from moderate-energy classification (e.g., systems without purps, pressurizing tanks, boilers, or.those which ' operate only from gravity flow or~ storage tank l
water head), however,. limited failures are assumed to occur for the purpose of considering theLeffects of flooding, spray, and wetting of equipment in the station analysis.
c The identification of piping failure locations will be performed in accordance with Section 3.6.2.
3.6.1.1.2.1 Interaction Criteria
- The following criteria define how interactions shall be_ evaluated..The safety evaluation of each interaction is described in Sections 3.6.1.3 and 3.6.-l.1.5.
a)
Environmental. Interaction An active component (electrical, mechanical, and instrumentation and control) _is assumed incapable of performing its function upon experiencing environmental. conditions exceeding any of its environmental ratings.
b)
' Jet Impingement ~ Interactions Active components (electrical, mechanical, and instrumentation and control) subjected to a jet are assumed failed unless the active component'is en-closed in a qualified enclosure, the component is known to be insensitive to such_an environment, or unless shown by analysis that the active function will not be impaired.
c)
Pipe Whip. Interaction
'A whipping pipe is not be considered to inflictfunacceptable damage to other pipes of equal or greater size'and wall thickness.
L L
A whipping pipe is only considered capable of developing through-wall leakage cracks in other pipes of equal or greater size with smaller wall thickness.
An active component (electrical. mechanical, and instrumentation and control) is assumed incapable of performing its active function following' impact by any whipping pipe unless an analysis or test is
-conducted to show otherwise.
i
-3.6.1.1.3 Protective Measures 3.6.1.1.'3.1 Reactor' Coolant System b
l 1The fluid discharged 'from postulated pipe' breaks at branch connections will
' produce' reaction and thrust forces in branch line piping.
The. effects of these i:
3.6-3 Rev. 9 L
2
CNS loadings are considered in assuring the continued integrity of the vital components and the engineered safety features.
To accomplish this in the design, a combination of component restraints, barriers, and layout are utilized to ensure that for a lors of coolant, or steam or feedwater line break, propagation of damage from the original event is' limited, and the components as needed, are protected and available.
l For piping connected to the Reactor Coolant System (six inch nominal or larger) and all connecting piping out to the LOCA boundary valve (Figure 3.6.2-1) is restrained to meet the following criteria:
a)
Propagation of the break to the unaffected loops is prevented to assure the delivery capacity of the accumulators and low head pumps, b)
Propagation of the break in the affected loop is permitted to occur but is limited by piping separation and restraints so as not to exceed 20 percent of the area of the line which initially failed.
This criterion is voluntarily applied so as not to substantially increase the severity of the loss of coolant.
(See also paragraph K.3 of Section 3.6.2.1.2).
c) ~ Where restraints on the linet, are necessary in order to prevent impact on and subsequent damage to the neighboring equipment or piping, restraint type and spacing is chosen such that a plastic hinge on the pipe at the two support points closest to the break is not formed.
Additional pipe restraint design criteria are discussed in Reference 1.
In addition to pipe restraints, barriers and layout are used to provide pro-tection from pipe whip, blowdown jet and reactive forces for postulated l
pipe breaks.
Some of the barriers utilized for protection against pipe whip are the follow-
-ing.
The polar crane wall serves as a barrier between the reactor coolant loops and the Containment liner.
In addition, the refueling cavity walls, various structural beams, the operating floor, and the crane wall enclose each reactor coolant loop in a separate compartment; thereby preventing an accident in any l
loop branch connection from affecting another loop or the Containment.
The portion of the main steam and feedwater lines within the Cor.tainment has been routed behind M rriers to separate these lines from reactor coolant piping.
The barriers described above are designed to withstand loadings resulting from jet and pipe whip impact forces.
Other'than Emergency _ Core Cooling System lines, all Engineered Safety Features are located outside the crane wall.
The Emergency Core Cooling System lines which penetrate the crane wall are routed around and outside the crane wall and then penetrate the crane wall in the vicinity of the loop to which they are attached.
3.6-4 Rev. 9
CNS Table 3.6.1-1 provides a listing of high-energy systems.
Moderate-energy systems are listed in Table 3.6.1-2.
Control room habitability is discussed in Section 3.6.1.1.3.4.
3.6.1.3
. Safety Evaluation Safety functions are identified for each initiating event by.the failure mode and effects analysis discussed in Section 3.6.2.1.2.
For each postulated failure, every credible unacceptable interaction shall be evaluated.
In establishing system requirements for eacia postulated break, it is assumed that a single active component failure occurs concurrently with the postulated rupture.
3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.2.1 Criteria Used To Define Break And Crack Location And Configuration
'3.6.2.1.1 Postulated Piping Break Location Criteria for the. Reactor Cool-ant System l The design' basis for postulated pipe breaks includes not only the break criteria, but also the criteria to protect other piping and vital systems from the effects of the postulated break.
A loss of reactor coolant accident is assumed to occur for a pipe break in piping down to the restraint of the second normally open automatic isolation valve (Cass II in Figure 3.6.2-1) on outgoing lines (*) and down to and in-
'cluding the second check valve (Case III in Figure 3.6.2-1) on incoming lines normally with flow.
A pipe break beyond the restraint or second check valve does not result in an uncontrolled loss of reactor coolant assuming either of the two check valves in the line close.
Both of the artomatic isolation valves are suitably protected and restrained as close to the valves as possible so that a pipe break beyond the restraint does not jeopardize the integrity and operability of the valves.
Periodic testing is performed of the capability of the valves to perform their intended function.
This criterion takes credit for only one of the two valves performing its intended function.
For normally closed isolation or incoming check valves (Cases I and IV in Figure 3.6.2-1), a loss of reactor coolant accident is assumed to occur for pipe breaks on the reactor side of the valve.
(*)lt is assumed that motion of the unsupported line containing the isolation valves could cause failure of the operators of both valves.
3.6-7 Rev. 9
~
,e' CNS 3$6.'2.1.1.1 Postulated Piping Break. Locations and. Orientations
Reference:
1 defines'the' original basis for postulating pipe breaks in the reactor coolant' system primary loop.
Reference 1.a provides the basis for eliminating from certain aspects of' design consideration previously postulated reactor coolant system pipe-breaks, with the exception of those breaks at branch. connections.
See Table'3.6.2-1 and Figure 3.6.2-2.
- 3. 6. 2.1.1.' 2 :
Postulated Piping Break Sizes For a'circumferential break, the-break area is the cross-sectional area'of the pipe at the, break location, unless pipe displacement is shown to be limited by 1
analysis, experiment or physical restraint.
-l. 3.i6;2.1.1.32 Line' Size Considerations for Postulated Piping Breaks
-Branch' lines connected to the Reactor Coolant System'are defined as "large" for the purpose of. this criteria 'as having an inside diameter. greater than 4 inches
'l up to the. largest connecting line. WLere postulated, pipe break of these lines
- results in a rapid blowdown of the Reactor Coolant System and protection 'is basically provided by the. accumulators and the low head safety injection pumps (residual heat removal pumps).
- 3.6.2.1.2
- General Design Criteria for_ Postulated Piping Breaks Other Than l
Reactor Coolant System a): Station design considers and accommodates'the effects of postulated pipe
. breaks with' respect to pipe whip, jet impingement and resulting reactive
-forces for piping'both inside and outside Containment.
The analytical methods-utilized to assure that concurrent single active component failure and pipe break effects do notl Jeopardize the safe shutdown of the reactor are outlined in Section 3.6.~2.3.
t
- b). ' Station generaliarrangement and ~ 1ayout ' design of high-energy systems Jutilize the possible combination of ph;ysical separation, pipe bends,.
pipe whip restraints and encased or. jacketed piping for the most practical 1 design of the station.'"These possible design combinations decrease postu-L lated piping break consequences to minimum and acceptable levels.
In all cases, the design is ~of a nature to mitigate the consequences of the break so that.the reactor can be shutdown safely and eventually maintained in a cold shutdown condition.
c)'.The environmental' effects of pressure, temperature and flooding are con--
trolled to acceptable levels utilizing restraints, level alarms and/or other warning devices, and vent openings.
i i
~
[
r 1.
3.6-9 Rev. 9
.. -. ~. - -
CNS l
-l 3.6.2.1.3 Failure Consequences Associated with Postulated Pipe Breaks I
~
The interactions that are evaluated to determine the failure consequences are
- dependent on the energy level of.the contained fluid.
They are as follows:
a)
High-Energy Piping 1)
Circumferential Breaks and Longitudinal Splits a)
Pipe Whip (displacement)
_ EB b). ~ Jet. Impingement
-M Q23-c)'
Compartment Pressurization
'd)
. Flooding e)
Environmental Effects (Temperature, humidity, water spray) 2)
Throu'ghwall leakage cracks a)
Environmental Effects (Temperature, Humidity) b)
Flooding Ib) Moderate-Energy Piping 1)
Through-wall leakage cracks
. a)
Flooding MEB-b)
Environmental Effects (Temperature, humidity, water spray)
Q23 c)
Water Spray For high energy
- piping there are certain exceptions as detailed in Reference la.for the reactor coolant loop.
3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Reactor Coolant System Dynamic Analysis' This section. summarizes the dynamic analysis-as it applies to the LOCA result-ing from.the postulated design basis pipe breaks at main reactor coolant branch line connections.Further discussion of the dynamic analysis methods used to verify the design adequacy of the reactor coolant-loop piping, equipment and supports is given in Reference 1 as it pertains to postulated breaks at-branch connections.
The.particular. arrange.nent of the Reactor Coolant System for the Catawba Nuclear
-Station'is~ accurately modeled by the standard layout used in Reference 1 and l the postulated branch connection break locations do not change from those
-presented in Reference 1.
1] In addition, an analysis is performed to demonstrate that at each postulated branch connection break location the motion of the pipe ends' is limited so as to preclude unac-ceptable damage due to=the effects of pipe whip or large motion of any major. components.
The. loads employed in the analysis are based on full pipe. area discharge except where limited by major structures.
3.6-18 Rev. 9 l
CNS The dynamic analysis of the Reactor Coolant System employs displacement method, lumped parameter, stiffness matrix formulation and assumes that all components behave in a linear elastic manner.
The analysis is performed on integrated analytical models including the steam lgeneratorandreactorcoolantpump,theassociatedsupportsandtheattached piping.
An elastic-dynamic three-dimensional model of the Reactor Coolant System is constructed.
The boundary of the analytical model is, in' general, the foundation concrete / support structure interface.
The anticipated defor-mation of the reinforced concrete foundation supports is considered where applicable to the Reactor Coolant System model.
The mathematical model is shown in Figure 3.6.2-4.
The steps in the analytical method are:
a)
The initial deflected position of the Reactor Coolant System model is defined by applying the general pressure analysis; lb)
Natural frequencies and normal modes of the broken branch connection are determined; c)
The initial deflection, natural frequencies, normal modes, and time-history forcing functions are used to determine the time-history dynamic deflection response of the lumped mass representation of the Reactor Coolant System; d)
The forces imposed upon the supports by the loop are obtained by multi-plying the support stiffness matrix and the time-history of displacement vector at the support point; and
_ l- )
The time-history dynamic deflections at mass points are treated as an e
imposed deflection condition on the ruptured loop branch connection, Reactor Coolant System model and internal forces,. deflections, and stresses at each end of the members of_the reactor coolant piping system j
are computed.
The results are'used to verify the adequacy of the restraints at the branch l
connections.
The general dynamic solution process is shown in Figure 3.6.2-5.
l In order to determine the thrust and reactive force loads to be applied to the Reactor Coolant System during the postulated LOCA, it is necessary to have a detailed description of the hydraulic transient.
Hydraulic forcing functions are calculated for the reactor coolant loops as a result of a postulated loss of coolant accident (LOCA) as a result of a postulated branch connection break.
These forces result from the transient flow and pressure histories in the l
The cal-culation is performed in two steps.
The first I
step is to calculate the transient pressure, mass flow rates, and other l
hydraulic properties as a function of time.
The second step uses the results obtained from the hydraulic analysis, along with input of areas and direction coordinates and is to calculate the time history of forces at appropriate l
locations in the reactor coolant loops.
i l
3.6-19 Rev. 9
c..
CNS REFERENCES FOR SECTION 3.6 1.
" Pipe Breaks for the Loca Analysis of the Westinghouse Primary Coolant Loop", WCAP-8082-P-A, January,1975 (Proprietary) and WCAP-8172-A (Non-Proprietary), " January, 1975.
1.a. Letter from H. B. Tucker (DPC) to H. R. Denton (NRC), dated December 20,
.1983, transmitting Westinghouse report justifying elimination of RCS primary loop pipe breaks fro certain design considerations.
2.
" Documentation of Selected Westinghouse Structrual Analysis Computer Codes", WCAP-8252; Revision.1, May, 1977.
3.
Bordelon, F.M., "A Comprehensive Space-Time Dependent Analysis of Loss of Coolant (SATAN IV Digital Code)", WCAP-7263, August, 1971 (Proprietary) and WCAP-7750, August, 1971 (Non-Proprietary).
4.
American Institute for Steel Construction, " Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings",
Februrary 12, 1969.
l' t
l'.
I L
l I.
3.6-29 Rev. 9 t
p.
.c.
,-9
-y,,
,- p
,y a
.t
'o y
f
- c Table 3.6.1-3'(Page 1).
-I jN-Comparison' of Duke Pipe Rupture' Criteria ^ And :
/:
i 4 "3
^
. NRC Requirements of Branch Technical Positions 2.
4' APCSB 3-1.(November 1975), MEB 3-L (November -1975), and NRC Regulatory Guide 1.46 (May 1973)
'W c/,j*
.4 1,
~
s p
- g.
b NRC Criteria' Ii Duke ' Cr.i teria' l.
-n
.? '
~
APCSB 3-1, Section B.2.c 9
SAR Section 3.6.2 Section 8.2.c. requires.that piping between containment Duke criteria is generall'y equivalent to NRC
~
(
isolation valves be'provided with pipe whip restraints criteria as clarified below:
l capable of.resi: ting bending and torsional moments pro-duced by a postulated failure either upstream or down-The containment structural integrity;is provided'
~
stream of the valves.
Also, the restraints should be for'all postulated pipe ruptures. :In addition, designed to withstand the_ loadings from postulated for any postulated rupture classified as a loss
- failures so that neither isolation. valve operability of coolant accident, the design.leaktightness of nor the leaktight integrity of the containment will the containment fission product barrier will be be impaired.
maintained.
I Terminal ends should be considered to originate at a Penetration design is discussed in SAR Section point adjacent to the required pipe whip restraints.
'3.6.2.4.
This section also discussed penetra-tion guard pipe design criteria.
q Terminal ends are defined as piping originating-at structure or component that"act as rigid con-straint to the piping thermal expansion.
APCSB 3-1, Section B.2.d SAR Section 6.6 (1) The protective measures,-structures, and guard Duke criteria is different'than the NRC criteria-pipes should not prevent the access required to due to the code' effective date as described below:
conduct inservice inspection examination.
ASME Class 2 piping welds will be inspected in (2) For portions.of piping between containment isola-accordance with requirements given in SAR Section tion valves,'the extent of inservice examinations 6.6.
completed during each inspection interval should provide 100 percent. volumetric examination of circumferential and longitudinal pipe welds.
- Pipe breaks in the RCS primarly loop are not postulated for consideration in certain~ aspects of' plant design, as defined in Reference la.
Rev. 9
Table 3.6.2-1 MEB Postulated Break Locations For The Main Q36 Coolant Loop Location of Postulated Rupture Tyge
- 1.
Reactor Vessel Outlet Nozzle Circumferential
- 2.
Reactor Vessel Inlet Nozzle Circumferential
- 3.
Steam Generator Inlet-Nozzle Circumferential
- 4.
Steam Generator Outlet Nozzle Circumferential
- S.
Reactor Coolant Pump Inlet Nozzle Circumferential
- 6.
Reactor Coolant Pump Outlet Nozzle Circumferential
- 7.
50 Elbow on the Intrados Longitudinal
- 8.
-Loop Closure Weld in Crossover Leg Circumferential
-9.
Residual Heat Removal (RHR) Line/ Primary Circumferential (Viewed Coolant Loop Connection from the RHR line)
- 10. Accumulator (ACC) Line/ Primary Coolant Circumferential (Viewed Loop Connection from ACC.line) 11.
Pressurizer Surge (PS) Line/ Primary Circumferential (Viewed Coolant Loop Connection from the PS line)
- Reference 1 defines the original basis for postulating pipe breaks in the reactor coolant system primary loop.
Reference la provides the basis for eliminating this previously postulated pipe break from certain aspects of design consideration.
Rev. 9 l
CNS 3.8.3.1.14 NSSS' Support Systems -
The support systems for the reactor vessel, steam generators, reactor coolant pumps, and main loop piping are completely described in Section 5.4.14.
.3.8.3.1.15-Accumulator Wing Walls The accumulator wing walls are two foot thick radial walls on either side of the accumulator tanks. 'They are doweled to the crane wall, accumulator
[
3.8-21a Rev. 9
CNS i
I are increased by_40 percent for design purposes.
These increased design i
. pressures are also listed in Table 3.8.3-2.
-In addition to designing the individual structural components for pressure, the overall interior structure is designed for the maximum uplift, horizontal shear, and overturning moment.
Each break location in the lower compartment has been evaluated to establish the maximum uplift, horizontal shear, and overturning i
moments on the interior structure.
Table 3.8.3-3 lists the maximum values of uplift, shear and overturning moment, the time at which they occur and the break identification for which they occur.
The loadings described above were utilized in the design of the interior struc-ture.
Subsequent to this design a revised postulated pipe break criteria was introduced in Section 3.6.
The differential pressures and load resultants presented in Table 3.8.3-2 and 3.8.3-3 respectively, are not applicable as listed but represent an upper bound for loadings resulting from a postulaed pipe break.
The final differential compartment differential pressures are in all cases less than those used for design.
Many of the postulated pipe break locations are provided with restraints to limit movement and consequential damage as a result of the pipe break.
The structure is therefore designed for the reactions including dynamic effects associated with the pipe restraints.
.The interior structure is also designed for the jet impingement forces created when a pipe ruptures near the structure.
The dynamic effect of the suddenly applied jet impingement force is also considered.
Internally generated missiles are discussed in Section 3.5.1.2.
The interior structure is designed to withstand the impact of such internal missiles and the dynamic effects associated with them.
3.8.3.3.4 Other Design Criteria The NSSS supports are designed for the load combinations and criteria set forth in Section 5.4.14.
The steel portion of the divider barrier between the upper and lower compartments ( consisting of the steam generator enclosures) are designed in accordance with Section III, Subsection NE, of the 1974 ASME Code including addenda through the Summer of 1976. A further discussion of the steam generator enclosures is included in Section 3.8.3.4.
3.8.3.4 Design and Analysis Procedures The elements of the interior structure are designed on an individual basis.
The interconnection between elements is included by considering relative stiff-nesses of connected elements to determine boundary conditons.
In some cases, portions of adjacent structural elements are modeled along with the particular element being designed to obtain the proper boundary interaction.
For other cases a most conservative approach of designing for both fixed and pinned boundary conditons is used.
A complete description of structural models follows.
3.8-24 Rev. 9 I
g3 3.8.3.4.1 Base Slab
.The base slab at elevation 552+0 is designed for bending forces and uplift forces created by attachments such as the cross-over leg restraints.
Downward forces
.are taken directly through bearing onto the foundation slab without imposing any bending or shear stresses on the base slab.
The anchorage of the larger com-ponents is achieved by means of continuous steel connections through the -liner plate into the foundation slab without creating' stresses in the base slab.
Hand calculations are used for design since the loads'are simple and the flat za slab can easily be represented as a wide beam.
Temperature and shrinkage steel is provided in the slab.in areas where there are no applied loads and resulting
. stresses.
3.8.3.4.2 Reactor Vessel Cavity Wall The reactor vessel cavity wall is represented as a space frame model for ana-lysis purposes.
The major loads include compartment pressure from postulated pipe breaks pressure, seismic forces and support loads from the reactor vessel and steam generator lower lateral supports.
Other smaller loads are included for-pipe supports and restraints.
3.8.3.4.3 Upper Reactor Cavity and Refueling Canal The' refueling canal floor and walls along with the upper reactor cavity walls
.are analyzed as a space finite element model.
The design loads include seismic,
] ' internal and external compartment pressures, and pipe support and restraint loads.
Reactions from adjacent structural elements are included for the operating floor-'and the CRDM' missile shields.
3.8.3.4.4 Crane Wall The crane wall is analyzed as a space frame model. The model includes additional members and elements to represent the walls and slab.that' connect to the crane wall.: Thus, the' proper stiffness and interconnection with other elements is included.
The applied loads include seismic forces, pressures from postulated pipe breaks, equipment loads, pipe support and restraint reactions, and reactions
.from adjacent structural' elements.
The. crane wall is. divided into two sections for analysis.
Both the upper and g
lower sections are modeled as space frames using STRUDL.
For more. details con-220.44 cerning governing loads and~1oad combinations, critical design forces and the design of reinforcing bars, refer to Table 3.8.3-4.
- 3.8.3.4.5-
. Steam Generator Compartments The removable steels shell portions of the steam generator enclosures are designed in;accordance with Section III, Subsection NE of the 1974 ASME Code including addenda through the Summer of 1976.
The steel dome is analyzed as a thin shell of revolution employing Kalnins' computer program for axisymmetric shells. The cylindrical steel shell portion of the enclosure is modeled as a plane frame for
^
a typical horizontal section of the shell. The concrete portions of the enclo-3.8-25 Rev. 9
^
p g
e
}(
CNS sure are modeled using' space frame members..The stiffness of the concrete walls is so much greater than the thin. steel shell that no interaction is considered.
- The concrete' displacements are included as boundary loads for the steel shell,
.and the steel shell reactions are included.as. loads cn the concrete model.
The loads on the.' steel shell are from internal pressure due to a main ~ steam l line rupture or other postulated pipe break and also seismic forces.
The-forces on the concrete portion include pressure due to main steam line rupture
-or-LOCA,' seismic', and pipe support and restraint loads.
- 3.8.3.4.6 Pressurizer Compartment The pressurizer compartment'is designed for internal pressure due to pipe
- rupture, pressurizer support reactions, seismic forces, and jet impingement forces ~ associated with postulated pipe ruptures.
The compartment is modeled
. using space frame members and elements. -The roof slab is included in the space frame model.to represent the proper stiffness.
An additional plate bending model with more detail is used, however, to design the roof slab.
~3.8.3.4.7 Operating Floor-
' The operating floor is modeled using plate bending and stretching elements.
Both in plane and out of plane forces are included..The in plane forces are due to'suppo; t reactions from the steam generator. upper lateral restraints.
.The major out of plane. forces include differential pressure from a postulated pipe break and jet impingement from the associated pipe rupture.
Other forces such as dead, live, seismic, and equipment and pipe support loads are also-
-included.
'Two: separate analyses'are performed using different element layouts and differ-ent computer programs. The analyses are conducted by two independent.and se-
- parate. groups (A and B on Figure 3.8.3-4) of the Structural Section of the Civil /
~ Environmental Division of the Design Engineering Department.
Each of the inde-pendent analyses are checked by. qualified engineers within the respective groups and cnc comparison of results is reviewed for agreement by the Group Supervisors of_ each group 'and the Principal Engineer of the Structural Section.
One model is runLusing the STRUDL computer program and the other is run using the ELAS program.'
For comparison purposes, the two models are loaded with a unit' pressure. The models are illustrated in Figures 3.8.3-5 and 3.8.3-6.
A comparison of the results is shown in Figures 3.8.3-7 through 3.8.3-8.
The close comparison between the programs _ assures the validity of the results.
3.8.3.4.8-Accumulator Floor
- TheLaccumulator floor at' elevation 565+3 is modeled as a plane grid.
Three 7 separate modelsL are'used for the various similar panels of the floor.
One model represents the portion of floor between wing walls enclosing the accumulators.
LA~second model represents.the portion of floor inside the fan compartments.
The third model represents'the portion'of floor within the instrumentation room.
3.8-26 Rev. 9
CNS w
Each model includes the openings in the floor and spring supports to represent
.the structural steel columns supporting the perimeter.
The design loads include l-pressuresfromapostulatedpipebreak,seismicforces,equipmentandpipesup-port and restraint loads, dead, and live loads. -
3.8.3.4.9 Ice Condenser Floor The ice condenser floor at elevation 593+8 1/2 is subjected to regularly spaced l
uniform support loads from the lower support structure within the ice region.
l Therefore, a representative segment of the floor is modeled using a space frame l model. The loads include pressure from a postulated pipe break, seismic forces, ice condenser lower support structure reactions, dead and live loads.
3.8.3.4.10 CRDM Missile Shield and Refueling Canal Gate The CRDM missile shield beams and refueling canal gate sections are both simply supported one way spans.
The analysis is therefore performed using hand calcu-lations.
Both are subjected to differential pressure due to a postulated pipe break and seismic forces.
In addition, the CRDM missile shield beams are designed for dead, live,'and internal missile loads.
The missile loads are described in Section 3.5.1.2.
3.8.3.4.11 NSSS Support Systems The design and analysis of the NSSS supports is fully described in Section 5.4.14.
l 3.8.3.4.12 Accumulator Wing Walls and Ice Condenser End Walls These walls are modeled using plate bending elements.
The major load is dif-
-l-ferential pressure from a postulated pipe break.
Also included are equipment and pipe support reactions as well as seismic loads.
l 3.8.3.4.13 Computer Programs for the Structural Analysis The following computer programs are employed in the analysis of Category I structures:
1.
For the stresses, stress resultants and displacements produced in a thin shell of revolution due to static and seismic loads: A computer program written by Professor A. Kalnins of Lehigh University, Bethlehem, Penn-sylvania.
Refer to Section 3.7.2 and Section 3.8.2.4 for description of
_ program.
2.
For the stresses, stress resultants and displacements of a shell of revolution due to the transient dynamic pressures associated with a loss-of-coolant accident: A computer program originally written at the Univer-sity of California, Berkeley.
Refer to Section 3.8.2.4 for description of the program.
3.
For seismic response of structures that can be idealized as multi-mass systems:
A computer program based on the theory presented in Section 3.7.2.1 and 3.7.2.6.
3.8-27 Rev. 9
The temperature of the auxiliary spray water is depending upon the performance of the Regenerative Heat Exchanger.
The most conservative case is when the letdown stream _is shut off and the charging fluid enters the pressurizer unheated. Therefore, for design purposes, the temperature of the spray water is assumed to be 70 F.
The spray flow rate is assumed to be 200 gpm.
It is furthermore assumed that the auxiliary spray will, if actuated, continue for five minutes until it is shut off.
The pressure decreases rapidly to the. low pressure reactor trip point.
At this pressure the pressurizer low pressure reactor trip is assumed to be actuated; this accentuates the pressure decrease until the pressure is finally limited to the hot leg saturation pressure.
At five minutes spray-is stopped and all the pressurizer heaters return the pressure to 2250 psia as shown on the' graph.
Again if the pressurizer heaters were not in operation the pressure would remain at the value reached in five minutes.
For design purposes it is assumed that no temperature changes in the Reactor Coolant System with the exception of the pressurizer occur as a result of initiation of auxiliary spray.
The total number of occurrences of this transient during the 40 year design life of the plant is specified'as 10.
-8.
Operating Basis Earthquake The mechanical stresses resulting from the operating basis earthquake (0BE) are considered on a component basis.
Fatigue analysis, where required by the codes, is performed by the supplier as part of the-stress analysis report.
The earthquake loads are a part of the mechanical loading conditions specified in the equipment specifications.
The origin of their determination is separate and distinct from those transients resulting from fluid pressure and temperature.
They are, however, considered in the design analysis.
Faulted Conditions The following primary system transients are. considered Faulted Conditions.
Each of the following accidents should be evaluated for one occurrence:
l 1.
Reactor Coolant Pipe Break (Loss of Coolant Accident)
~2.
Large Steam Line Break 3.
Reactor Coolant Pipe Break (Large Loss of Coolant Accident) l Following a postulated rupture of a reactor coolant pipe resulting in a large loss of coolant, the primary system pressure decreases causing the primary system temperature to decrease.
Because of the rapid blowdown of coolant from the system and the comparatively large heat capacity of the metal sections of the components, it is likely that the metal will still be at or near the operating temperature by the end of blowdown.
It is
-conservatively assumed that the SIS is actuated to introduce water at a minimum temperature of 32 F into the RCS.
The safety injection signal will also result in reactor and turbine trips.
3.9-7 Rev. 9
[
CNS 4.
STHRUST - hydraulic loads on loop components from bicwdown information.
5.
WECAN - finite element structural analysis.
6.
- DARI
- WOSTAS - dynamic transient r'sponse analysis of reactor vessel e
and internals.
7.
-SATAN IV - Space time dependent analysis of loss of coolant accident that treats all phases of blowdown loads.
3.9.1.3 Experimental Stress Analysis No experimental stress analysis methods have been used for the Catawba project.
3.9.1.4 Considerations for the Evaluation of the Faulted Condition This section describes the faulted condition load combinations and analysis methods for reactor coolant system piping, components, and supports.
As noted in Section 3.6, pipe breaks in the primary loop RCS piping jhave been eliminated from consideration in certain aspects of the-plant design, as defined in Reference 16.
However, reactor coolant system piping (including Class 1 branch lines), primary components, and their supports have been designed and analyzed for the faulted condition SRSS load combination of SSE and LOCA (postulated pipe break in main RCS piping).
This approach provides considerable margin in the plant design.
The following sections describe the faulted condition analyses including the analysis methods used for LOCA.
3.9.1.4.1 Loading Conditions The structural stress analyses performed on the reactor coolant system consider the loadings specified as shown in Table 3.9.1-2.
These loads result from thermal expansion, pressure, dead weight, Operating Basis Earthquake (0BE), Safe Shutdown Earthquake (SSE), design basis loss of coolant accident, and plant operational thermal and pressure transients.
3.9.1.4.2
' Analysis of the Reactor Coolant Loop The reactor coolant loop piping is evaluated in accordance with the criteria of ASME III, NB-3650 and Appendix F.
The loads included in the evaluation result from the SSE, deadweight, pressure, and LOCA loadings (loop hydraulic forces, asymmetric subcompartment pressurization forces, and reactor vessel motion).
The loads used in the analysis of the reactor coolant loop piping are described in detail.below.
Pressure Pressure loading is identified as either membrane design pressure or general operating pressure, depending upon its application.
The membrane design pressure is used in connection with the longitudinal pressure stress and minimum wall thickness calculations in accordance with the ASME Code.
d 3.9-11 Rev. 9 n.
'CNS
~
t The reactor internals structures have been conservatively designed to withstand the stress.and be within deflection limits originating from a LOCA
~
(full double-ended RCS primary loop pipe break) even though such pipe breaks
=
are no longer considered for dynamic effects, according to Reference 16.
3.9-35a Rev. 9 New Page 1
CNS 7.'
Repeat Step 1
~
The sequence is repeated, as for rod cluster control assembly withdrawal, up to 72 times per minute which gives an insertion rate of 45 inches per minute.
Holding and Tripping of th'e Control Rods During most of the plant operating time, the control rod drive mechanisms hold the rod cluster control assemblies withdrawn from the core in a static position.
In the holding mode, only one coil, the stationary gripper coil (A), is ener-gized.on each mechanism.
The drive rod assembly and attached rod cluster con-trol assemblies hang suspended from the three-latches.
-If' power to the stationary' gripper coil is cut off, the combined weight of the drive rod assembly and the rod cluster control assembly plus the stationary gripper return spring is sufficient to move latches out of the drive rod assem-bly groove.
The control rod falls by gravity into the core. The trip occurs as the magnetic field, holding the stationary gripper plunger half against the
-stationary gripper pole, collapses and the stationary gripper plunger half is forced down by the weight stationary gripper return spring and weight acting upon the latches.
After the rod cluster control assembly is released by the mechanism, it falls freely until the control rods enter the dashpot section of the thimble tubes in the-fuel assembly.
3.9.4.2-Applicable CRDS Design Specifications For those components.in the Control Rod Drive System comprising portions of the reactor coolant pressure boundary, conformance with the General Design Criteria and 10CFR50, Section 50.55a is discussed in Sections 3.1 and 5.2-conformance with Regulatory Guides pertaining in Section 4.5 and 5.2.3.
Design Bases Bases for temperature, stress on structural members, and material compatibility are imposed on the design of the reactivity control components.
Design Stresses The-Control Rod Drive System is designed to withstand stresses originating from various operating conditions as summarized in Table 3.9.1-1.
The CRDS has i,een conservatively designed to withstand the stresses originating from a LOCA (full' double-ended RCS primary loop pipe break) even though such pipe breaks are no longer considered for dynainic effects according to Reference 16.
A11cwable Stresses:
For r.ormal operating conditionsSection III of the ASME Boiler and Pressure Code is used. All pressure boundary components are ana-lyzed as Class I components.
. Dynamic Analysis:
The cyclic stresses due to dynamic loads and deflections are
. combined with the stresses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total stresses of the Control Rod Drive System.
3.9-56 Rev. 9
ps CNS 3.9.5.3 Design Loading Categories The. combination of design loadings fit into either the normal, upset, emergency or faulted conditions as defined in the ASME Code,Section III.
Loads and deflections imposed on components due to shock and vibration are deter-mined analytically and experimentally in both scaled models and operating reactors.
The cyclic stresses due to these dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces and ther-mal gradients for the determination of the total stresses of the internals.
The reactor internals are designed to withstand stresses originating from vari-ous operating conditions as summarized in Table 3.9.1-1.
- The scope of the stress analysis problem is very large requiring many different techniques and methods, both static and dynamic.
The analysis performed depends on the mode of operation under consideration.
Allowable Deflections For normal operating conditions, downward vertical deflection of the lower core support plate is negligible.
For the loss of coolant accident plus the safe shutdown earthquake condition, the deflection criteria of critical internal structures are limiting values given in Table 3.9.2-2.
The corresponding no loss of function limits are included in Table 3.9.2-2 for comparison purposes with the allowed criteria. The reactor internals
-structures'have been conservatively designed to withstand the stresses originating
'from a LOCA (full double-ended RCS primary loop pipe break) even though such pipe breaks are no longer considered for dynamic effects, according to Reference 16.
The criteria for the core drop accident is based upon analyses which have to deter-mine the total downward displacement of the internal structures following a hypo-thesized core drop resulting from loss of the normal core barrel supports.
The initial clearance between the secondary core support structures and the reactor vessel lower head in the hot condition is approximately one half inch.
An addi-tional displacement of-approximately 3/4 inch would occur due to strain of the energy absorbing devices of the secondary core support; thus the total drop dis-tance is about 1-1/4 inches which is insufficient to permit the trips of the rod cluster control assembly to come out of the guide thimble in the fuel assem-blies.
.Specifically, the secondary core support is a device which will never be used, except.during a hypothetical accident of the core support (core barrel, barrel flange, etc.).
There are 4 supports in each reactor.
This device limits the fall of the core and absorbs much of the energy of the fall which otherwise
-would be imparted to the vessel.
The energy of the fall is calculated assuming a complete and instantaneous failure of the primary core support and is absorbed during the plastic deformation of the controlled volume of stainless steel, load-ed in tension. The maximum deformation of this austenitic stainless piece is limited'to approximately 15 percent, after which a positive stop is provided to ensure support.
3.9-66 Rev. 9
~
f CNS REFERENCES FOR SECTION 3.9 (cont'd).
116. : Letter from H.B. Tucker (DPC) to H.R. Denton (NRC), dated December 20, 1983,
' transmitting Westinghouse report' justifying elimination of RCS primary loop
-breaks for certain design considerations.
3.9-71 Rev. 9 New Page
Table' 3.9.1-i.ip.wre 7) -
lle*. l_gn I raer. lent.+.. f or A*.Hf cale Cl.c.*. I l*lp,Ing (4)
CHFMICAL' RESIDUAL.
AND
. REACTOR.
-HEAT SAFETY:
VOLUME.
COOLANT-UPPER HEAD ?
(1)
. (2)
(3)
REMOVAL INJECTION' CONIROL' ' PRESSURIZER, PRESSURIZER PRESSURIZER
.. DRAIN' INJECTION DESIGN TRANSIENTS.
CONDITION OCCURRENCES SYSTEM SYSTEM SYSTEM SURGE LINE RELIEF SPRAY RTD BYPASS : ' LINES.
LINEST
' Loss of Load without Immediate Upset 80
- X X-X X-X' NOTES 4,' 5 X X
X.
X Turbine or Reactor Trip Loss of Flow in One Loop Upset
. 80 X-X.
X X-X-
X X
X Reactor Trip with Cooldown Upset
- 10 X
.X X
X X.
X X,
X-X" and Inadvertent SIS Actuation Inadvertent RCS Depressuri--
Upset 20 X
X X
X
.X X
X X
X zation inadvertent SI Accumulator Upset 4
X_
Blowdown during Plant Cooldown High Head Safety Injection Upset 22.
X.
Boron Injection Upset 48 X
Large Steam Break Faulted 1
X X
X X
X X
X X
X l High Head Safety Injection Pipe Rupture Faulted 1
X X
X X-X X
X X.
X Faulted 2
X Boron Injection Faulted 2
X
-~
Turbine Roll Test Test 10
'X-X X
X X
X-X X
X Hydrostatic Test' Test 5
X X
.X X
_X X
X.
X X
' Primary Side Leak Test Test 50 X
X X
X X
X X
X X
Inadvertent Auxiliary Spray Test 1
X X
NOTES:
1.
Pressurizer surge line is analyzed for 80 occurrences of transient C-7, the final cooldown spray.
2.
Pressurizer surge line is analyzed for 150,000 initial fluctuations and 3,000,000 random fluctuations.
3.
These transients are conditions which can cause the PORV's to open. Although a total of 320 such transients are shown, the PORV inlet ifnes are analyzed for 100 such occurrances.
4.
For analysis of the safety valves 40 occurrences were assumed.
5.
Number of occurrences is 20,000,000.
Rev. 9 O
~_
- 2.
. Analysis of Accident Loads
- ~As shown in Reference 7, grid crushing tests and seismic and LOCA evaluations show that the fue1~ assembly will maintain a' geometry that is capable of being cooled'under the worst-case accident Condition IV event.
The seismic and LOCA evaluations given in reference 7 (which encompass the Catawba plant) are conservative when compared to the Catawba plant's design bases relative to' the structural' integrity of the reactor coolant' system (RCS primary loop).
As discussed in Section 3.6, the elimination of consideration of the dyna-h mic effects of pipe breaks in the RCS primary loop has been fully justified.
A prototype' fuel-assembly.has been subjected to column loads in excess of those expected-in normal. service and faulted conditions (see Reference 7).
No interference with control rod insertion into thimble tubes will occur during a Safe Shutdown Earthquake (SSE).
Stresses.in che fuel assembly caused a tripping of the rod cluster control assembly have little' influence on fatigue because of the small number of events during the life;of an assembly.
Assembly components and prototype fuel. assemblies made from production parts have been subjected to structural
-. tests to verify that the design bases requirements are met (Reference 7).
3.
Loads Applied in Fuel Handling The fuel assembly design loads for shipping have been established at 6 g's.
Accelerometers are permanently placed into the shipping cask to monitor and detect fuel assembly accelerations that would exceed the criteria.
Past history.and experience has indicated that loads which exce'ed the allowable
- limits rarely occur.
Exceeding the limits' requires reinspection of the fuellassembly for damage.. Tests on various fuel assembly components such as'the grid assembly, sleeves, inserts.and structure joints have been per-formed to assure that the shipping design limits do not result in impair-ment..of fuel assembly function.
4.2.3.6 Reactivity Control Assembly and Burnable Poison Rods
' 1.
Internal Pressure and Cladding Stresses During Normal, Transient and Accident Conditions The designs of the burnable poison, source rods and 8 C absorber rods 4
.' provide a sufficient cold void volume to accommodate the internal pres-sure. increase during operation.
'For the burnable poison rod, the use of glass in tubular form provides a central void volume along the length of the rods.
For the source rods, and 6
the B C absorber rod, a void volume is provided in the cladding in order to 4
limit the. internal pressure increase until end of-life (see Figure 4.1.1-12).
The. stress analysis of.the-burnable poison and source roas assumes 100 per-cent gas: release to the rod void volume in addition to the initial pres-sure.within the rod.
For the B C control rod a 20% gas release is assumed.
4 4.2-25 Rev. 9
L t
(
CNS 5.4.14.2.2 Reactor Coolant Pump The reactor coolant pump support system consists of vertical steel columns and a lateral steel frame.
Figures 5.4.14-3 through 5.4.14-5 show outlines of the support system of the reactor coolant pump.
5.4.14.2.3 Pressurizer
- The' pressurizer support system consists of vertical steel hangers from the oper-ating-floor to the base of the pressurizer, a lateral frame at the base anchored to the crane wall and tied to the vertical hangers, and an upper lateral steel ring anchored to'the crane wall and pressurizer enclosure walls.
Figures 5.4.14-6 through 5.4.14-8 show outlines of the pressurizer support system.
5.4.14.2.4 Reactor Vessel The reactor vessel supports are individual water-cooled rectangular box struc-tures beneath the vessel nozzles and anchored to the nrimary shield wall.
Fig-ure 5.4.34-9 shows an outline of a typical reactor vessel support.
5.4.14.3 Fabricaticn The fabrication of all steel component supports is in accordance with Subsection NF of Section III of the 1974 or 1977 ASME Code, depending on the contract date for the particular support.
A code stamp is not required.
5.4.14.4 Materials The materials used for all steel supports are listed in Table 5.4.14-2.
For all materials except the reactor coolant pump bolts (See Figure 5.4.14-3),
the materials meet the requirements of Article NF-2000 of Section III sf the ASME Code.
The reactor coolant pump bolt material is a high strength steel (modified
'4340) not defined in Appendix I of Section III.
This material is required to
_ pass Charpy V-notch impact tests.
In addition, the material is not subjected to stress corrosion cracking by virtue of the fact that a corrosive environment is not present and the bolt has essentially no residual stresses and does not experience any significant sustained loads during normal service.
Concrete support structures are constructed in accordance with the ACI Code 318-71_using grade 60 reinforcing and 5000 psi concrete.
(
5.4-47 Rev. 9
o s-Figures 5.4.14-10 thru -15 Deleted by Revision 9 Revision 9
r c
,c:
n
, 6. 2 CONTAINMENT SYSTEMS'
~6.2.1-CONTAINMENT FUNCTIONAL DESIGN
'6.2.1.1
~ Containment Structure 6.2.1.1.1-Design Bases-
.l
.The containment. vessel steel shell is~ designed for dead' loads, construction loads,- design: basis accident loads, external pressure, seismic loads and pene-tration loads as described in Section.3.8.2.3.' The applicable loading combi-
-nations considered are 11sted in Table 3.8.2-1.
'The design basis accident internal pressure is 15 psig. The effects of pipe rupture in the primary coolant system up to and including a double-ended rup-
.ture of the. largest pipe as well as rupture of the main steam line are consid-ered in determining the peak accidcnt pressure.
The maximum design external pressure is 1.5 psig.
This is greater than the
- internal-vacuum created by an accidental trip of a portion of the Containment SprayfSystem during-normal operation. The Containment Pressure Control System is discussed in Section 7.6.
.The internal' structures of the containment vessel are also designed for sub-compartment-differential' accident pressures. The accident pressures considered are due to-the same postulated pipe ruptures as described above for the con-l 'tainment vessel or. as described in Section 3.6, as appl.icable.
A 40 percent margin is applied to these calculated differential pressures.
A tabulation of the calculated as well as the design pressures (including the 40 percent
. increase) is given in Table 3.8.3-2.
-The other simultaneous. loads in combination with the accident pressures and the applicable load factors are given in Table 3.8.1-2.
For a further descrip-tion of these loads see Section 3.8.3.7.
~
.The functional design of the Containment-is based upon the following accident input source' term assumptions and conditions.
-(1) The design basis blowdown energy of 324.2 x 108 Btu and mass of 498,200 lb put into the Containment.
l (2) The hot metal energy is considered.
(3)- A reactor core power of-3526 MWt (plus 2%) used for decay heat generation.
(4) The minimum Engineered Safety Feature performance (i.e., the single fail-ure criterion applied to each safety. system) comprised of the following:
a.
The ice condenser which condenses steam generated during a LOCA
. thereby limiting the pressure peak inside the Containment (see Section 6.7).
.b.
The Containment Isolation System which closes those fluid penetra-tions not serving accident' consequence limiting purposes (see Section 6.2.4).
6.2-1 1
~.
-o.
3 CNS Refer to Section 6.2.1.5 for an analysis of the minimum containment pressure transient used in the analysis of the emergency core cooling system.
ICSB Instrumentation provided to monitor and record the containment pressure during
-Q11~
the course of an accident within the containment is discussed in Chapter 7.
Ice ~ condenser instrumentation is discussed in Section 6.7.15.
6.2.1.2 Containment Subcompartments 6.2.1J2.1 Design Basis Consideration is given in the design of the Containment internal structures to localized pressure pulses that could occur following a loss-of-coolant accident.
If'a loss-of-coolant accident were to occur due to'a pipe rupture in these rela-tively small volumes, the pressure would build up at a rate faster than the overall Containment, thus imposing a differential pressure across the walls of the structures.
These subcompartments. include the steam generator enclosure, pressurizer enclo-sure, and the reactor cavity.
Each compartment is designed for the largest blow-down flow resulting from the severance of the largest connecting pipe within the enclosure or the blowdown flow into the enclosure from a break in an adjacent region.
The-extent to which pipe restraints are used to limit the break area of pipe ruptures is presented in Section 3.9.
.The preliminary calculated differential compartment pressures are increased by a minimum. of 40 percent for the design of interior structure walls, slabs, and component supports. The final calculated differential compartment pressures and component. support loads due to final calculated differential pressures are in all cases less than those used for design.
The_subcompartment pressurization following a loss-of-coolant accident, was con-
. sidered in the design of the ' interior structure.
Subsequent to this design a revised postulated pipe break criteria was introduced in Section 3.6.
The sub-
[
compartment!pressurizations resulting from loss-of-coolant accident is not
- applicable,' as described in this'section, but represent an upper bound for i
loadings-resulting from a postulated pipe break. 'The final calculated differ-l ential compartment pressures and component support loads due to final calculated differential ' pressures are in all cases less than those used for design.
The basic performance of the Ice Condenser Reactor Containment System has been demonstrated for a wide range of conditions by the Waltz Mill Ice Condenser Test Program.- These results have clearly shown the capability and reliability of the ice condenser concept to limit the Containment pressure rise subsequent to a hypothetical loss-of-coolant accident.
6.2-19 Rev. 9 u-
e
- 3
?-
-DuxE POWER GOMPANY P.O. Box 33189 CHARLOTTE. N.C. 28242
' HAL B. TUCKER "
m.rresorm U~.-
' December 20, 1983 iMr. Harold R.lDenton, Director i {l* Q ]
40ffice of Nuclear Reactor Regulation U.-S.; Nuclear Regulatory Commission Washington, D. C.-'20555 WITHHEt.D FROM
- Attention:
Ms.' E. G.' Adensam, Chief P.U.BLIC' DISCLOSURE
' Licensing Branch No. 4.
.Re: ' Catawba Nuclear Station Docket Nos. 50-413 and 50-414
~
References:
1)' Letter from W. H. Owen (Duke Power Company) to W. J. Dircks-(NRC), dated September 19, 1983
- 2) Letter from.H. R..Denton (NRC) to W. H. Owen (Duke Power Company), dated 0ctober 17, 1983 3)' L~etter from H. -B. Tucker (Duke Power Company)
.to H. R. Denton_(NRC), dated November 18, 1983
Dear Mr. Denton:
References'_l) and 3) infonned the NRC that Duke Power Company was evaluating
..the technical feasibility and potential benefits of eliminating postulated
- pipe breaksLin the Reactor. Coolant System (RCS) primary loop from the structural design basis of the Catawba Nuclear Station. As a result of.
efforts by Westinghouse, the NRC, and' Duke Power, we have concluded that it is. technically feasible to eliminate these postulated pipe breaks.
In addition, Westinghouse has assured Duke Power Company that the generic linformation previously submitted to the NRC to justify the elimination of LRCS ' primary. loop pipe breaks is applicable to the Catawba Nuclear Station.
As~a result of the above developments, and in-accordance with the statement in Reference-2) that' applications related to the leak-before-break pipe' failure concept will be permitted prior.to the NRC completing all of the changes in regulatory requirements, this letter is submitted. Duke Power hereby requests NRC approval for application of the " leak-before-break" concept to the Catawba Nuclear Station to eliminate postulated pipe breaks in the RCS primary loop from the plant structural design basis. A specific
. plant applicability ' report'is -included as Enclosure A to this letter. Because of the proprietary ndture of this report, Enclosure A has been provided only
- to the addressee and Mr. James P. O'.Reilly of the'NRC. A non-proprietary version of the-' specific-plant applicability report is included as Enclosure B
. and ~has-been provided to others on the attached distribution list.
As' Enclosure A contains information proprietary to Westinghouse Electric Corporation, it'is supported by the attached letter (Attachment 1) and
.: affidavit ~ signed' by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity A
a v
.,12 5 -
u....
j
- E' kMrsHaroldR.D:nton/Dirsctor
' ? December 20',1983 '
Page-2 the considerations -listed in paragraph '(b)(4) of 'Section 2.790 of the Commission's regulations. - Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public
- disclosure in accordance with 10 CFR'Section 2.790
- of the Commission's.
regulations. -Correspondence with respect to the proprietary aspects of.the Application for Withholding or the supporting Westinghouse' affidavit should reference, CAW-83-106, and should be addressed to R. A. Wiesemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation,
~
P. 0. Box 355, Pittsburgh, Pennsylvania 15230.
~ Implementation of the leak-before-break concept will have the following
- effects' on= the structural design for Catawba Nuclear Station:
- l);- Eliminate the need to postulate circumferential and longitudinal pipe breaks.in the RCS primary: loop (hot leg, cold leg, and cross-over leg _ piping).
2)
Eliminate the need for as' ociated pipe whip restraints in the RCS s
primary loop and eliminate the requirement to design for the structural effects associated with RCS primary loop pipe breaks (including jet impi_ngement.
4
- 3) Eliminate the ~need to consider dynamic effects-and loading conditions
~
associated.with previoGly ~ postulated primary loop. pipe breaks. These
~
' effects. include blowdown loads,' jet impingement loads, and reactor cavity.and subcompartment pressurization.
' Employment of.the-leak-before-break concept would not eliminate p_ipe breaks
=in the RCS primary loop 'as a design ~ basis for. the following:
1). Containment. design.
~
'2) Sizing.of: Emergency Core Cooling System 3)~ ~ Environmental qualification of equipment
- 4) Supports for heavy components-
~ The crac_k' sizes'and' result' ant flows from the leak-before-break analysis.will be used when reactor cavity land subcompartment pressurization data are revised.
.The impact =on-important design aspects of-implementing leak-before-break on
- Catawba-Nuclear Station has been evaluated by Duke' Power and is summarized. in-
- Attachment 2. 'A detailed. list of.affected pipe whip restraints is provided
-in Attachment 3.
Duke Power has Jalso evaluated the potential cost savings and operational benefits that result;from the elimination of postulated pipe
- breaks in the RCS. primary loop. - A summary of the potential benefits which can be' realized specifically.from the elimination of these pipe breaks for LCatawba Unit 2'is.provided in Attachment 4.
Note that these benefits total at:least $2:million'and involve an estimated 600 man-rem dose reduction over N
ithe. life of Unit 2.
Implementation of the ~1eak-before-break concept will therefore be cost-effective as well as technically justifiable while resulting in. improved overall plant safety.
I a
y
]
x w L EMr. Harold R. Dinton, Director.-
~
' December 20, 1983 Page 3
-Enclosure C consists of the revised Catawba FSAR pages associated with the.
~
A elimination of RCS : primary loop breaks, and -it will be included in Revision 9 fto;the FSAR. This current request is for' implementation on Unit 2 only;
. Duke Power willi submit additional.infonnation' prior to -implementation on--
Unit 1.
Construction: completion of the RCS primarylloop pipe whip restraints at Catawba' Nuclear Station Unit 2 is'.on hold pending an NRC ruling on' this
- proposal. LIn order to realize the maximum advantage from the elimination
- of RCS primary loop ruptures, we request a decision by February 15, 1984.
If I2canbe of further assistance,:or if a meeting with the Staff is deemed beneficial. for-'a final resolution of this matter, please contact me.
.Very truly yours,-
8.h.kM Hal~B. Tucker-if ROS/php.
' Attachment l
cc: { Mr. -James P. O'Reilly, Regional Administrator U.-S. Nuclear Regulatory Commission Region II-
.101 Marietta Street, NW,-Suite l2900 LAtlanta, Georgia 30303' D
NRC Resident Inspector
= Catawba -Nuclear Station
~
L Mr. Robert Guild,-Esq.
L' JAttorney-at-Law:
.P.'-0. Box 12097~
Charleston, South Carolina 29412 :
Palmetto Alliance if
<21351 Devine Street
' Columbia, South Carolina 29205 Mr. Jesse L. Riley
- Carolina Environmental ~ Study Group L854 Henley Place
^
p
' Charlotte,- North Carolina 28207 p
l^
(.:
D
ATTACHMENT 1 Westinghouse.
Water Reactor Nuclear Technology Division Electric Corporation Olvisions
,,,333 PmsDurgnPennsylvania15230 November 23, 1983 CAW-83-106
. Mr. Harold R. Denton, Director-Office of Nuclear Reactor Regulation U. S.. Nuclear Regulatory Commission Washington,-D. C.f 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
REFERENCE:
. Duke Power Company letter to. NRC dated November 1983 Decerter
Dear Mr. Denton:
.The proprietary material for which withholding is being requested in the
-reference-letter by Duke Power Company is further identified in an affidavit signed by -the owner of the proprietary information, Westinghouse Electric Corporation. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in para-graph (b)(4) of 10CFR Section 2.790 of the Comission's regulations.
The proprietary ma'terial for which withholding is being requested is of
~
the same technical type as that proprietary material previously submitted
.with application for withholding CAW-83-80.
Accordingly,thisletterautho'riYhstheutilizationofthe~ accompanying affidavit by Duke Power Company.
Correspondence with respect to "the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-83-106, and should _be addressed to the undersigned.
Very truly yours,
//M6 Robert A. Wiesemann, Manager
/bek Regulatory & Legislative Affairs cc:
E. C. Shomaker, Esq.
-Office of the Executive Legal Director, NRC W mpu&CE
e CAW-83-80 AFFI C ".". i COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared John D. McAdoo, who, being by me duly sworn according to'1aw, deposes and says that~he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact. set forth in this Affidavit are true and correct to the-best of his knowledge, information, and
. belief:
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D.McAdoo,Assiltr[ntManager Nuclear Safety Department Sworn to and subscribed before me this R M day of J d..d.4, 1983.
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- CAW-83-80 i
(1)
L am Assistant Manager, Nuclear Safety Department, in the Nuclear Techno-logy Division, of Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing or rule-making proceedings, and am authorized'to' apply for its withholding on behalf of the Westinghouse Water Reactor Divisions.
(2)
-I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of ttie Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.
(3)
I have personal knowledge of the criteria and procedures utilized by
-Westinghouse Nuclear Energy Systems in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4)
Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Coinnission's regulations, the foliosing is furnished for consideration by the Commission in determining whether the information sought to be with-
-held from public disclospre-should be withheld.
(i)
The information soughtr to be.vithheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
Westing-house has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hole certain types of information in confidence.
The application of that system and the substance of that system constitutes Westinghouse policy and~provides the rational basis required.
I
l 9
-3.
CAM-83-80 Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a) -The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advan-tage, e.g., by optimization or improved marketability.
(c)
Its use by a competitor would reduce his expenditure of resour-ces or improve his competitive position in the design, manufac-ture, shipment, installation, assurance of quality, or licensing a similar produc,t. -
~
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects.of past,'present, or future Westinghouse or customer funded development plans and programs of potential consnercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
)
. CAW-83-80 (g)
It is not the property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.
There are sound' policy reasons behind the Westinghouse system which include the following:
(a). The use of information by Westinghouse gives Westinghouse a competitive advantage over its competitors.
It is, therefore, withheld from disclosure to protect the Westinghouse competitive
-position.
(b)
It is information which is marketable in many ways.
The extent
-to which such information is available to comoetitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c) -Use by our competitor wod1d put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
.f.e -
'(d)
Each component of-proprietary information pertinent to a parti-cular competitive advantage is potentially as valuable as the I
total-competitive advantage.
If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle,.thereby depriving Westinghouse of a competi-l tive advantage.
L (e)
Unrestricted disclosure would jeopardize the position of promi-nence of Westinghouse in the world market, and thereby give a market advantage to the competition in those countries.
'.. a..,
' CAW'83-80 (f)
The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and main-taining a competitive advantage.
-(iii)
The iriformation is being transmitted to the Commission in confidence
_~j and, under the provisions of 10CFR Section 2.790, it is to be i
1 y
received in confidence by the Commission.
[
S
. iv)
"The information sought to be protected is not available in public
(
sources to the best of our knowledge and belief.
(v)
The proprietary information sought to be withheld in this submittal is that which is appropriately marked in " Technical Bases for Eliminating Large Primary Loop Pipe Ruptures as the Structural Design Bases for the; South Texas Project," dated September 1983, prepared by S. A. Swamy and J. J. McInerney.
The subject information could mnly be duplicated by competitors if they were to invest time and effort equivalent to that invested by Westinghouseprovided,theyhavetherequisitetalentandexperience.
Public disclosure of ttiis information is likely to cause substantial harm to the cgapetitive position of Westinghouse because it would simplify design and-evaluation tasks without requiring a comensurate investment of time and effort.
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Further the deponent sayeth not.
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t ATTACHMENT 2 m
. Impact of Elimination of Postulated Circumferential and Longitudinal Pipe Breaks t'
in the RCS Primary Loop s
STRUCIURES, SYSTEMS, COMP 0NENTS, PROGRAMS CONSIDERED FOR IMPACT IMPACT Primary Loop Pipe Whip Restraints Deleted from Design *
-Reastor Cavity / Primary Shield Wall /
Reduction in pressurization loading
.N Crane Wall / Operating Floor Steam Generator Su'-compartment No change o
RCS Component Supports / Heavy No change Component Supports Emergency Core Cooling Systems No change Containment Design No change RCS Pressure' Boundary Leakage No change
\\DetectionSystems
,h m.qs Environmental Qualification Program No change
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's
- Due to small hot gaps, the hot leg pipe whip restraints currently receive
-relatively small loadings from postulated main steam pipe breaks.
It has q-been shown that the Steam Generator calumn supports a.re adequate to support the additional load in the absence of the hot leg pipe whip restraints.
S-Also, an analysis is being performed to show that the reactor coolant loop loadings from the main steam pipe breaks will be acceptable without the M
.p hot leg pipe whip restraints.
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. ATTACHMENT 3
~
Postulated RCS Primary Loop Pipe Breaks and Associated Pipe Whip Restraints Per Unit Postulated Break Associated Whip Restraint Erection Status Locations Per Loop for Primary Loading Catawba Unit 2
.1.
Reactor vessel.
- 1.. Cold Leg Nozzle Break 1.
Structure installed inlet nozzle Restraint (wagon wheel) without shims 2.
Reactor. vessel.
2.
Hot Leg Nozzle Break 2.
Not installed
- outlet nozzle Restraint (wagonwheel)
- 3. ' Steam generator 3.
Hot leg pipe whip 3.
Structure installed inlet nozzle restraint without shims
- 4. '50' elbow in the 4.
Hot leg pipe whip 4.
Structure installed intrados(longitudinal restraint without shims slot) 5.
Steam' generator.
5.
Crossover leg pipe whip 5.
Structure installed outlet nozzle restraint (vertical run) with shims Crossover leg elbow Compression blocks restraints installed without shimming 6.
Reactor coolant' pump.
Crossover leg elbow 6.
Compression block 6.
inlet nozzle (pump restraini.s installed without suction) shims
- 7. ' Crossover leg closure 7
Crosscver leg elbow 7.
Compression blocks weld restraints installed without shimming
- 8 Reactor coolant pumu 8.
None outlet
ATTACHMENT 4 Estimated Cost Savings / Operational Benefits for Elimination of Primary Loop Pipe Breaks on Catawba Unit 2 Category Cost Savings (1983 rates)
Operational Benefit 1.
Elimination of RCS
$0.6M - Pipe whip restraint
-Substantial improvement pipe whip restraints installation cost
- in quality of ISI
$1.3H'- Occupational radiation
-Substantial improvement exposure over Unit 2 in personnel access life results in-dose reduction of 600 man-rem
- Simplifies plant design
-Improved access for by elimination of po-operation and maintenance tential interferences with piping, hangers,
-Reduced RCS heat loss to impulse tubing, etc.
containment at Whip restraint locations.
$0.1M - Eliminates additional hold points during
-Reduced risk of unanti-initial heatup for pated pipe restraint for verifying pipe-restraint thermal growth and seismic clearances movement.
-Improvement in overall plantsafety(NUREG/CR-2136) 2.
Simplification of
- Pressurization loadings
-Simplification of analyses analysis associated reduced on primary shield involving loadings due to with dynamic effects wall, crane wall, opera-future plant modifications, and loading conditions.
ting. floor, and subcom-partment analys s.
TOTAL
$2 Million 600 man-Rem cOf a total of 20 restraints, four have not been installed. Shimming work has not been performed on any of the restraints.
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