Design of the Reactor Pressure Vessel Support System (Generic Letter 78-02)
GL 78-02
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555
January 25, 1978
All PWR Licensees (Except for Trojan, North Anna, Indian Point 3,
Beaver Valley and St. Lucie 1)
Gentlemen:
In October of 1975, the NRC staff notified each licensee of an operating PWR
facility of a potential safety problem concerning the design of the reactor
pressure vessel support system. Those letters requested each licensee to
review the design basis for the reactor vessel support system for each of its
PWR facilities to determine whether certain transient loads, which were
described in the enclosure to the letter, had been appropriately taken into
account in the design. Furthermore, these letters indicated that, on the
basis of the results of licensees' reviews, a reassessment of the reactor
vessel support design for each operating PWR facility may be required.
Licensee responses to that request indicated that these postulated asymmetric
loads have not been considered in the design basis for the reactor vessel
support system, reactor internals including the fuel, steam generator
supports, pump supports, emergency core cooling system (ECCS) lines, reactor
coolant system piping, or control rod drives.
Subsequently in June 1976, the NRC staff informed each PWR licensee that a
reassessment of the reactor vessel support system design for each of its
facilities was required. While the emphasis of these letters was primarily
focused on the need to reassess the vessel support design for transient
differential pressures in the annular region between the reactor vessel and
the cavity shield wall and across the core barrel, we indicated that our
generic review may extend to other areas in the nuclear steam supply system
(NSSS) and that further evaluation may be required.
For you information, Enclosure 1 is a summary of the background and current
status of our review efforts related to this generic concern.
7905150612
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All PWR Licensees - 2 - January 25, 1978
We have now determined that an assessment of the potential for damage to other
NSSS component supports (e.g., steam generator and pump supports), the fuel
assemblies, control rod drives, and ECCS piping attached to the reactor
coolant system due to loadings associated with postulated coolant system
piping breaks is required. Our request for additional information transmitted
to you in June 1976 has been revised both to clarify our original request and
to identify the extension of our concerns to other areas in the NSSS, as
identified above. A copy of this revised request for additional information
is provided as Enclosure 2.
The revised request for additional information identifies a requirement that
your assessment of potential damage to the reactor vessel and other NSSS
component supports, reactor vessel, fuel and internals, attached ECCS lines
and the control rod drives should include consideration of breaks both inside
and outside of the reactor pressure vessel cavity. This assessment should be
made for postulated breaks in the reactor coolant piping system, (secondary
systems are not to be included), including the following locations:
a) Reactor vessel hot and cold leg nozzle safe ends
b) Pump discharge nozzle
c) Crossover leg
d) Hot leg at the steam generator (B&W and CE plants only)
A number of licensees, have presented to the NRC staff alternate proposals,
other than to conduct a detailed analyses, to resolve this concern. Based
upon our review of these proposals, we have concluded that these alternative
proposals do not establish an acceptable basis for long term operation without
a detailed assessment of the risk resulting from these postulated transient
loading conditions. We have, however, concluded that the low probability for
occurrence of an event which could result in these loads establishes an
adequate basis to justify continued operation for a short term period.
The NRC staff will consider an analysis that is applicable to more than one
specific plant if it can be adequately demonstrated that such an analysis is
either representative or bounding for each plant concerned.
Additional guidance regarding loading combinations (safe shutdown earthquake
loads, loss of coolant accident loads), will be provided by about March 1,
1978, following the conclusion of staff investigations in this area.
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All PWR Licensees - 3 - January 25, 1978
Please respond within 30 days of receipt of this letter, indicating your
intent to proceed with an evaluation of the overall asymmetric loss of coolant
accident (LOCA) loads as described herein. In addition, please submit to us,
within 90 days, your detailed schedule for providing the required evaluation.
Your schedule should be consistent with our desire to resolve this problem
within two years and should clearly state your intent to demonstrate the
safety of long term continued operation.
We are transmitting information copies of this letter to the Westinghouse,
Combustion Engineering and Babcock & Wilcox Companies. If you have any
questions or want any clarification on this matter, please call your NRC
Project manager.
Copies of this letter are being sent to all addressees on the current service
lists for each docket.
Sincerely,
Victor Stello, Jr., Director
Division of Operating Reactors
Office of Nuclear Reactor Regulation
Enclosures:
1. Background and Current Status
2. Revised Request for Additional Information
cc w/enclosure:
See attached listing
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January 25, 1978
ENCLOSURE 1
BACKGROUND AND CURRENT STATUS OF THE NRC STAFF REVIEW
OF ASYMMETRIC LOCA LOADS FOR PWR FACILITIES
On May 7, 1975, the NRC was informed by Virginia Electric & Power Company that
an asymmetric loading on the reactor vessel supports resulting from a
postulated reactor coolant pipe rupture at a specific location (e.g., the
vessel nozzle) had not been considered by Westinghouse or Stone & Webster in
the original design of the reactor vessel support system for North Anna, Units
1 and 2. It had been identified that in the event of a postulated
instantaneous, double-ended offset LOCA at the vessel nozzle, asymmetric
loading could result from forces induced on the reactor internals by transient
differential pressure across the core barrel and by forces on the vessel due
to transient differential pressures in the reactor cavity. With the advent of
more sophisticated computer codes and the accompanying more detailed
analytical models, it became apparent that such differential pressures,
although of short duration, could place a significant load on the reactor
vessel supports and on other components, thereby possibly affecting their
integrity. Although this potential safety concern was first identified during
the review of the North Anna facilities, it has generic implications for all
PWRs.
Upon closer examination of this situation, it was determined that postulated
breaks in a reactor coolant pipe at vessel nozzles were not the only area of
concern but rather that other pipe breaks in the reactor coolant system could
cause internal and external transient loads to act upon the reactor vessel and
other components. For the postulated pipe break in the cold leg, asymmetric
pressure changes could take place in the annulus between the core barrel and
the vessel. Decompression could occur on the side of the vessel annulus
nearest the pipe break before the pressure on the opposite side of the vessel
changes. This momentary differential pressure across the core barrel could
induce lateral loads both on the core barrel and on the reactor vessel.
Vertical loads could also be applied to the core internals and to the vessel
due to the vertical flow resistance through the core and asymmetric axial
decompression of the vessel. Simultaneously, for vessel nozzle breaks, the
annulus between the reactor and biological shield wall could become
asymmetrically pressurized resulting in a differential pressure across the
vessel causing additional horizontal and vertical external loads on the
vessel. In addition, the vessel could be loaded by the effects of initial
tension release and blowdown thrust at the pipe break. These loads could
occur simultaneously. For a reactor vessel outlet break, the same type of
loadings could occur, but the internal loads would be predominantly vertical
due to more rapid decompression of the upper plenum.
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Although the NRC staff's original emphasis and concern were focused primarily
on the integrity of the reactor vessel support system with respect to
postulated breaks inside the reactor cavity (i.e., at a nozzle), it has since
become apparent that significant asymmetric forces can also be generated by
postulated pipe breaks outside the cavity and that the scope of the problem is
not limited to the vessel support system itself. For such outside-cavity
postulated breaks, the aforementioned concerns, such as the integrity of fuel
assemblies and other structures, need to be examined.
In June 1976, the NRC requested all operating PWR licensees to evaluate the
adequacy of the reactor system components and their supports at their
facilities with respect to these newly-identified loads. In response to our
request, most licensees with Westinghouse plants proposed an augmented
inservice inspection program (ISI) of the reactor vessel safe-end-to-end pipe
welds in lieu of providing an evaluation of postulated piping failures.
Licensees with Combustion Engineering plants submitted a probability study
(prepared by Science Applications, Inc) in support of their conclusion that a
break at a particular location (vessel nozzle) has such a low probability of
occurrence that no further analysis is necessary. A similar study has been
recently submitted by Science Applications, Inc. (SAI) for B&W plants.
When the Westinghouse and CE owners group reports were received in September
1976, the NRC formed a special review task group to evaluate these alternative
proposals. In addition, EG&G Idaho, Inc. was contracted to perform an
independent review of the SAI probability study submitted for the CE owners
group.
This review effort resulted in a substantial number of questions which
previously have been provided to representatives of each group. Based on the
nature of these questions and other factors to be discussed later in this
report, we cannot accept these reports in their present form as a resolution
for the asymmetric LOCA load generic issue. Based on our review, we have
concluded that a sufficient data base does not presently exist within the
nuclear industry to provide satisfactory answers to these information needs.
Several long-term experimental programs would be required to provide much of
this information. Although the probability study recently submitted by SAI
for certain B&W owners does respond to some of the informal questions raised
during our review of the SAI report prepared by CE plants, the more
fundamental questions remain. Therefore, this conclusion also applies to the
SAI topical report for B&W plants (SAI-050-77-PA).
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A second - and equally important - reason for not accepting probability/ISI
approaches as a solution at this point concerns our need and industry's need
to gain a better understanding of the problem. We consider it essential that
an understanding of the important breaks and associated consequences be known
before applying any remedy - be it pipe restraints, probability, ISI, or some
combination of these measures. Only in this way will we have a basis on which
to judge the importance of the remedy with respect to what it is designed to
prevent.
Although we have many questions on each of these topical reports, this does
not mean that we view the probabilistics/ISI approach as completely without
merit. In fact, the results of a probabilistic evaluation serves as the basis
for continued operation and licensing of nuclear plants during this interim
period while additional evaluations can be performed by vendors and licensees.
We believe that the justification for continued plant operation has as its
basic foundation the fact that the event in question, i.e., a hypothetical
double-ended instantaneous rupture of the main coolant pipe at a particular
location, has a very low probability of occurrence.
The disruptive failure probability of a reactor vessel itself has been
estimated to lie between 10(-6) and 10(-7) per reactor year - so low that it
is not considered as a design basis event. The rupture probability of pipes
is estimated to be higher. WASH-1400 used a median value of 10(-4) for LOCA
initiating ruptures per plant-year for all pipes sizes 6" and greater (with a
lower and upper bound of 10(-5) and 10(-3), respectively). We believe that
considering the large size of the pipes in question (up to 50" O.D. and 4-1/8"
thick), the lower bound is more appropriate since these pipes are more like
vessels in size. In addition, the quality control of this piping is the best
available and somewhat better than that of the piping used in the WASH-1400
study.
These factors, coupled with the facts that (1) the break of primary concern
must be very large, (2) it must occur at a specific location, (3) the break
must occur essentially instantaneously, and (4) these welds are currently
subject to inservice inspection by volumetric and surface techniques in
accordance with ASME Code Section XI, lead us to conclude that the probability
of a pipe break resulting in substantial transient loads on the vessel support
system or other structures is acceptably small such that continued reactor
operation and continued licensing of facilities for operation can continue
while this matter is being resolved.
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In support of the above, the staff has developed a short-term interim
criterion to determine if an acceptable level of safety exists for operating
PWRs under conditions of a postulated pipe break. This interim criterion is
based on a simplified probabilistic model that incorporates elastic fracture
mechanics techniques to estimate the probability of a pipe break. Critical
flaw size and subcritical flaw growth rates were determined assuming the
presence of a surface flaw located in a circumferential weld of a thick-walled
pipe. Determination of the critical flaw size was based on an estimated
fracture toughness value of KIC at a minimum temperature of 200 F and a
uniform tensile stress equal to the consideration of various operating
conditions producing elastically calculated stresses ranging in value from 1
to 3 times the material minimum yield strength.
Then, using the calculated critical flaw size, the subcritical growth rate,
and an estimated probability distribution of an undetected flaw in
thick-walled pipe welds, the upper bound probability of pipe break was
estimated to be 10(-4). This value is also supported by a recent publication
by Dr. S. H. Bush* which states that actual failure statistics confirm rates
of 10(-4) to 10(-6) per reactor-year in large pipes, with higher rates as the
pipe size decreases. Considering these analyses, we conclude that our
conservative estimate on a pipe break in the primary coolant system is in the
range of 10(-4) to 10(-6). This estimated pipe break probability is
considered acceptably low to justify short-term operation of nuclear power
plants.
In view of all previous discussions concerning this issue, the NRC staff has
concluded that an evaluation must be undertaken to assess the design adequacy
of the reactor vessel supports and other affected structures and systems to
withstand asymmetric LOCA loads, including an assessment of the effects of
asymmetric loads produced by various pipe breaks both inside and outside the
reactor cavity. On performing these evaluations the staff will permit the
grouping of plants, where adequate justification for such grouping exists, in
order to limit the number of plants to be analyzed. Alternatively, the staff
will permit the analyzing of a "prototypical" plant, which is sufficiently
representative of a group of plants, to provide the necessary information.
Both of these concepts have been discussed with the Westinghouse and CE Owners
Groups, and we believe that such approaches could save a significant amount of
time and effort in obtaining results on which to base any needed corrective
measures. The NRC staff is prepared to meet with PWR licensees to discuss
such approaches, and has already done so. For example, we met with the
Westinghouse owners group on October 19, 1977, for the purpose of discussing a
generic solution for breaks outside the reactor cavity. It is expected that a
similar meeting will be held in the near future to address breaks
______________________
*"Critical Factors in Blowdown Loads in the PWR Guillotine Nozzle Break
(Volume 2 - the Asymmetric Load Problem)" dated June 6, 1977
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located inside the cavity. This "phased" approach is acceptable to us,
provided that it sheds light on and serves to expedite consideration of the
more limiting inside-cavity breaks.
For your information, the NRC has a technical assistance contract with EG&G
Idaho, Inc., to independently model representative Westinghouse, B&W, and CE
plants for the purpose of assessing the loads on all major structures and
components resulting from asymmetric LOCA loads. We believe that the results
of this program which will include sensitivity studies, will provide
significant confirmatory information related to this generic safety concern.
Although, as stated earlier, we believe that continued operation and licensing
of facilities for the short-term is justified, we also believe that efforts to
resolve this issue should proceed without delay, with the objective of both
completing the necessary assessments and installing any necessary plant
modifications within two years. In making this statement, we wish to make it
clear that plant modifications, if indicated by licensee assessments, is the
preferred approach. At the same time, we recognize that there may be cases
wherein appropriate modifications may be judged to be unwarranted based on the
consideration of overall risk. In such cases, we will be prepared to give
further consideration to alternate approaches, such as probability/ISI. We
feel, however, that ISI techniques as they exist today could be considerably
improved, and, to the extent that such improvements could have a direct
bearing on this problem as well as an impact of nuclear safety in general, we
would welcome their development.
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January 25, 1978
ENCLOSURE 2
REVISED REQUEST FOR ADDITIONAL INFORMATION
Recent analyses have shown that certain reactor system components and their
supports may be subjected to previously underestimated asymmetric loads under
the conditions that result from the postulation of ruptures of the reactor
coolant piping at various locations. It is therefore necessary to reassess
the capability of these reactor system components to assure that the
calculated dynamic asymmetric loads resulting from these postulated pipe
ruptures will be within the bounds necessary to provide high assurance that
the reactor can be brought safely to a cold shutdown condition. For the
purpose of this request for additional information the reactor system
components that require reassessment shall include:
a. Reactor Pressure Vessel
b. Fuel Assemblies, Including Grid Structures
c. Control Rod Drives
d. ECCS Piping that is Attached to the Primary Coolant Piping
e. Primary Coolant Piping
f. Reactor Vessel, Steam Generator and Pump Supports
g. Reactor Internals
h. Biological Shield Wall and Neutron Shield Tank (where applicable)
i. Steam Generator Compartment Wall
The following information should be included in your reassessment of the
effects of postulated asymmetric LOCA loads on the above-mentioned reactor
system components and the reactor cavity structure.
1. Provide arrangement drawings of the reactor vessel, the steam generator
and pump support systems to show the geometry of all principal elements and
materials of construction.
2. If a plant-specific analysis will not be submitted for your plant, provide
supporting information to demonstrate that the generic plant analysis under
consideration adequately bounds the postulated accidents at your facility.
Include a comparison of the geometric, structural, mechanical and thermal
hydraulic similarities between your facility and the case analyzed. Discuss
the effects of any differences.
3. Consider postulated breaks at the reactor vessel hot and cold leg nozzle
safe ends, pump discharge nozzle and crossover leg that result in the most
severe loading conditions for the above-mentioned
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systems.* Provide an assessment of the effects of asymmetric pressure
differentials on these systems/components in combination with all external
loadings including asymmetric cavity pressurization for both the reactor
vessel and steam generator which might result from the required postulate.
This assessment should consider:
a. limited displacement break areas where applicable
b. consideration of fluid-structure interaction
c. use of actual time-dependent forcing function
d. reactor support stiffness
e. break opening times.
4. If the results of the assessment required by 3 above indicate loads
leading to inelastic action in these systems or displacement exceeding
previous design limits provide an evaluation of the following:
a. Inelastic behavior (including strain hardening) of the material used in
the system design and the effect on the load transmitted to the backup
structures to which these systems are attached.
5. For all analysis performed, include the method of analysis, the structural
and hydraulic computer codes employed, drawings of the models employed and
comparisons of the calculated to the allowable values.
6. Provide an estimate of the total amount of permanent deformation sustained
by the fuel spacer grids. Include a description of the impact testing that
was performed in support of your estimate. Address the effects of operating
temperatures, secondary impacts, and irradiated material properties (strength
and ductility) on the amount of predicted deformation. Demonstrate that the
fuel will remain coolable for all predicted geometries.
7. Demonstrate that active components will perform their safety function when
subjected to the postulated loads resulting from a pipe break in the reactor
coolant system.
8. Demonstrate functionability of any essential piping where service level B
limits are exceeded.
In order to review the methods employed to compute the asymmetrical pressure
differences across the core support barrel during subcooled portion of the
blowdown analysis, the following information is requested:
_____________________
*B&W and CE plant licensees should also consider breaks in the hot leg at the
steam generator inlet.
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1. A complete description of the hydraulic code(s) used including the
development of the equations being solved, the assumptions and simplification
used to solve the equations, the limitations resulting from these assumptions
and simplifications and the numerical methods used to solve the final set of
equations. Provide comparisons with experimental data, covering a wide range
of scales, to demonstrate the applicability of the code and of the modeling
procedures of the subcooled blowdown portion of the transient. In addition,
discuss application of the code to the multi-dimensional aspects of the
reactor geometry.
If an approved vendor code is used to obtain the asymmetric pressure
difference across the core support barrel, state the name and version of the
code used and the date of the NRC acceptance of the code.
2. If the assessment of the asymmetric pressure difference across the core
support barrel is made without the use of a hydraulic blowdown code, present
the methodology used to evaluate the asymmetric loads and provide
justification that this assessment provides a conservative estimate of the
effects of the postulated LOCA.
A compartment multi-node, space-time pressure response analysis is necessary
to determine the external forces and moments on components. Analyses should
be performed to determine the pressure transient resulting from postulated hot
leg and cold leg reactor coolant system pipe ruptures within the reactor
cavity and any pipe penetrations. If applicable, similar analyses should be
performed for steam generator compartments that may be subject to
pressurization where significant component support loads may result. This
information can be provided to encompass a group of similarly designed plants
(generic approach) or a purely plant specific (custom plant) evaluation can be
developed. In either case, the proposed method of evaluation and principal
assumptions to be used in the analysis should be provided for review in
advance of the final load assessment.
For generic evaluations, perform a survey of the plants to be included and
identify the principle parameters which may vary from plant to plant. For
instance, this should include blowdown rate and geometrical variations in
principle dimensions, volumes, vent areas, and vent locations. A typical or
lead plant should be selected to perform sensitivity and envelope
calculations. These analyses should include:
(1) nodal model development for the configuration representing the most
restrictive geometry; i.e., requiring the greatest nodalization;
(2) the most restrictive configuration regarding vent areas and obstructions
to flow should be analyzed; and,
(3) sensitivity to code data input should be evaluated; e.g., loss
coefficients, inertia terms, vent areas, nodal volumes, and any other input
data where there may be variations from plant to plant or uncertainty for the
given plant..
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These studies should be directed at evaluating the maximum lateral and
vertical force and moment time functions, recognizing that models may be
different for lateral as opposed to vertical load definitions.
The following is the type of information needed for both generic and custom
plant evaluations. Although this request was primarily developed for reactor
cavity analyses it may be applied to other component subcompartments by
general application.
(1) Provide and justify the pipe break type, area, and location for each
analysis. Specify whether the pipe break was postulated for the evaluation of
the compartment structural design, component supports design, or both.
(2) For each compartment, provide a table of blowdown mass flow rate and
energy release rate as a function of time for the break which results in the
maximum structural load, and for the break which was used for the component
supports evaluation.
(3) Provide a schematic drawing showing the compartment nodalization for the
determination of maximum structural loads, and for the component supports
evaluation. Provide sufficiently detailed plan and section drawings for
several views, including principal dimensions, showing the arrangement of the
compartment structure, major components, piping, and other major obstructions
and vent areas to permit verification of the subcompartment nodalization and
vent locations.
(4) Provide a tabulation of the nodal net-free volumes and interconnecting
flow path areas. For each flow path, provide an L/A (ft(-1)) ratio, where L
is the average distance the fluid flows in that flow path and A is the
effective cross sectional area. Provide and justify values of the vent loss
coefficients and/or friction factors used to calculate flow between nodal
volumes. When a loss coefficient consists of more than one component,
identify each component, its value and the flow area at which the loss
coefficient applies.
(5) Describe the nodalization sensitivity study performed to determine the
minimum number of volume nodes required to conservatively predict the maximum
pressure load acting on the compartment structure. The nodalization
sensitivity study should include consideration of spatial pressure variation;
e.g., pressure variation circumferentially, axially and radially within the
compartment. The nodal model development studies should show that a spatially
convergent differential pressure distribution has been obtained for the
selected evaluation model.
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Describe the justify the nodalization sensitivity study performed for the
major component supports evaluated, if different from the structural analysis
model, where transient forces and moments acting on the components are of
concern. Where component loads are of primary interest, show the effect of
noding variations on the transient forces and moments. Use this information
to justify the nodal model selected for use in the component supports
evaluation.
If the pressurization of subvolumes located in regions away from the break
location is of concern for plant safety, show that the selection of parameters
which affect the calculations have been conservatively evaluated. This is
particularly true for pressurization of the volume beneath the reactor vessel.
In this case, a model which predicts the highest pressurization below the
vessel should be selected for the evaluation.
NOTE: It has been our experience that for the reactor cavity, three regions
should be considered (i.e., nodalized) when developing a total model.
These are:
(1) the volume around or in the vicinity of the break location out to
a radius approximated by the adjacent nozzles, and including portions
of the penetration volume for some plants;
(2) the volume or region covering the upper reactor cavity, primarily
the RPV nozzles other than the break nozzle; and
(3) the region encompassing the lower reactor cavity and other
portions of the reactor cavity and other portions of the reactor cavity
not included in Items (1) and (2).
(6) Discuss the manner in which movable obstructions to vent flow (such as
insulation, ducting, plugs, and seals) were treated. Provide analytical and
experimental justification that vent areas will not be partially or completely
plugged by displaced objects. Discuss how insulation for piping and
components was considered in determining volumes and vent areas.
(7) Graphically show the pressure (psia) and differential pressure (psi)
response as functions of time for a representative number of nodes to indicate
the spatial pressure response. Discuss the basis for establishing the
differential pressure on structures and components.
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(8) For the compartment structural design pressure evaluation, provide the
peak calculated differential pressure and time of peak pressure for each node.
Discuss whether the design differential pressure is uniformly applied to the
compartment structure or whether it is spatially varied. If the design
differential pressure varies depending on the proximity of the pipe break
location, discuss how the vent areas and flow coefficients were determined to
assure that regions removed from the break location are conservatively
designed, particularly for the reactor cavity as discussed above.
(9) Provide the peak and transient loading on the major components used to
establish the adequacy of the support design. This should include the load
forcing functions (e.g., f(x)(t), f(y)(t), f(z)(t)) and transient moments
(e.g., M(x)(t), M(y)(t), M(z)(t)) as resolved about a specific, identified
coordinate system. The centerline of the break nozzle is recommended as the X
coordinate and the center line of the vessel as the Z axis. Provide the
projected area used to calculate these loads and identify the location of the
area projections on plan and section drawings in the selected coordinate
system. This information should be presented in such a manner that
confirmatory evaluations of the loads and moments can be made.
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