ML19220C835
| ML19220C835 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/18/1978 |
| From: | Treby S NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Johnson W, Rosenthal A, Sharfman J NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| Shared Package | |
| ML19220C836 | List: |
| References | |
| NUDOCS 7905150606 | |
| Download: ML19220C835 (2) | |
Text
Acril 18, 1978 Alan S. Rosenthal, Esq., Chairman Jerome E. Sharfman, Esq.
Atcaic Safety and Licensing Appeal Atomic Safety and Licensing Appeal Board Board U.S. Nuclear Regulatory Cocnission U.S. Nuclear Regulatory Cocnission Washington, D. C.
20555 Washington, D.C.
20L55 Dr. W. Reed Jchnson Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Corninfon Washington, D.C.
20555 In the Matter of Metropolitan Edison Company, et al.
(Three Mile Island Nuclear Station, Unit 2)
Docket No. 50-320 Gentlemen:
Enclosed for your infomation :re copies of the following documents regard-ing the referenced proceeding:
1) a letter, with enclosures, from Victor Stello, Jr.,
DISTRIBUTION:
Director, Division of Operating Reactors to all PNF.
H.McGurren Licensees (except for Trojan, North Anna, Indian G.{ess Point 3, Beaver Valley and St. Lucie 1) dated S.ireby January 25, 1978, requesting each licensee to pro-H. Silver vide specified information regarding the effects of Il6C-Phillips postulated asyncetric LOCA loads on certain reactor S.Varga system cocponents and their supports.
Il6-Phillips J.Norris - EP 2) a report entitled "An Evaluation of Environmental W.Regan - EP Data Relating to Selected Nuclear Power Plant Site.
H. Smith The Three Mile Island Nuclear Station Site" (ANL/
110-Phillips EIS-4). This report, which was prepared by Argonne H.jhapar National Laboratory under NRC contract, contains i. tngel,na rdt M.Grossman recoccendations for terrestrial monitoring which are different from the monitoring program adopted NRC Central File by the NRC Staff for TMI - Unit 2.
ELD F (2)
Chron.
3) a memorandum from Malcolm L. Ernst, Assistant Director for Environmental Tec.nnology, Division of Sita Safety and Environmental Analysis to Voss A.
Moore, Assistant Director for Environmental Projects, i17 013 omex w sunwa u c h Daft W NRC FORM 318 (9-76) NRCM 0240_
- u. s oovanwuany.nwyme ory,cas s ete -eas.ese 700515 0 (' Q' i f
h 2-Divisien of Site Safety and Environmental Analysis, dated February 13, 1978, cocmenting on item 2 identi-fied above.
The f taff does not consider matters addressed in the attached documents to be significant ner information related to any issues in controversy nor do they contain extraordinary new circumstances regarding matters not put into controversy in this proceeding.
Sincerely.
Stuart A. Treby Assistant Chief Hearing Counsel for NRC Staff
Enclosures:
As stated cc w/ enclosures:
Edward Luton, Esq.
Ms. Judith H. Johnsrud Mr. Gustave A. Linenberger Atomic Safety and Licensing George F. Trowbridge. Esq.
Board Panel Dr. Ernest 0. Salo Atomic Safety and Licensing Dr. Chauncey R. Xepford Appeal Panel Ms. Karin W. Carter Docketing and Service Section 117 014 N
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03L9ZL78 NRC FORM 318 (9 76) NIG 0240 W us s. sovsanunwt resemme orrics ien - ese.444
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UNITED STATES
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January 25, i978 All PWR Licensees (Except for Trojan, North Anna, Indian Point 3, Beaver Valley and St. Lucie 1)
Gent 1emen:
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. Further:nore, 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, reacter coolant system piping, or control rod drives.
Subsequently in June 1976, the NRC staff informed each PWR licensee l
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 i
across the core barrel, we indicated that our generic review may l
extend to other areas in the nuclear steam supply system (NSSS) and that further evaluation may be required.
For your infornation, Enclosure 1 is a summary of the background and current status of our review efforts related to this generic Concern.
117 015 7905150 6/2
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All PWR Licensees January 25, 1978 We have now detemined 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 infomation 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 revi ad request for additional information identi)ies a requirement that your assessment of potential damage to the reactor vescal 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 (BsW 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 tem 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 ajequately demonstrated that such an analysis is either representative or bounding for each plant concerned.
Additional guidance regarding loading combinaticns (safe shutdcwn earthquake loads, loss of coolant accident loads), will be provided by about March 1, j
1978, following the conclusion of staff investigations in this area.
I i
i 117 016 r
All PWR Licensees 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 Compcnies.
If you have any questions or want any clarification on this matter, please call your NRC Project Manager.
Captas of this letter are being sent to all addressees on the current service lists for each docket.
(Eje_'
Sincerely, Victor Ste lo,
., Of rec to r 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 117 017
January 25, 1978 ENCLOSURE 1 BACKGR0UND 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 Comoany that an asymmetric loading on the reactor vessel supports resulting from a postulated reactor cooldnt 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 inter-nals by transient differential pressure across the core barrel and by forces on the vessel due to transient differential pressure in the reactor cavi ty. 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 signi-ficant load on the reactor vessel supports and on other components, there-by 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 + hat postu-lated breaks in a reactor coolant pipe at vessel nozzles wert 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 reatcar vessel and other components. Fcr 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 cnanges. This i'omentary di f ferential 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 decenpression of the vessel. Simultaneously, for vessel nozzle breaks, the annulus between th.e 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 ten-sion release and blowdown thrust at the pipe break.
These loads could occur si:nul taneously. 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.
i17 018
_2 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 t? avaluate the adequacy of the reactor system components and their supports o 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 postulat~ed piping failures. Licensees with Combustion Engineering plants submitted a probability study (prepcred by Science Applications, Inc.) in succort of their conclusion that a break at a particular location (versel nozzle) has such a low probability of occurrence that no furtner 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 fern as a resolution for the asymmetric !CCA 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 infomation needs.
Several long-term experimental programs wcuid 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 tne SAI topical report for B&W plants (SAI-050-77-PA).
117 019
) A second - and equally important - reason for not accepting probability /ISI approaches as a solution at this point concerns cur.need and industry's need to gain a better understanding of the problem. We consider it essential that an understanding of the imoortant breaks and associated consecuences be known before applying any remedy - be it pipe restraints, probability, ISI, or some combination of these measures. Only in *. bis 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 tooical reports, this does not mean that we view the probabilistic/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 interia 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 probagility of a reactor vessel itself has been estimated to lie be' ween 10-and 10-7 per reactor year - so low that it is not considered us a design basis event.
The rupture probability of, pipes is estimated to be higher. WASH-1400 used a median value of 10-'
for LCCA initiating ruotures per plant-year for all ojoes sizes 6" and greater (with a lower and upper bound of 10-3 and 10
, respectively).
We believe that considering the large size of the pipes in question (up to 50" 0.D. and 4-1/8 thick), the Icwer bound is more appropriate since these pipes are more like vessels in size.
In audi uon, the quality con-trol 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 con-cern 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 'ith 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 operatinn and continued licensing of facilities for operation can continue while this matter is being resolved.
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4-In support of the above, the staff has developed a short-term interim cri-terion 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 frac-ture mechanics techniques to estimate the probability of a pipe break.
Critical flaw size and subcritical flaw growth rates were detemined assuming the presenct of a surface flaw located in a circumferential weld of a thick-walled pipe, Detemination of the critical flaw size was based on an estimated fracture toughness value of KIC at a ninimum temoerature of 200 F and a uniform tensile stress equal to the ccnsideration of various operating conditions producing elastically calculated stresses rangina in value fren 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 upoer bound orobability of pipe break was estimated to be 10'.
This value is also supported by a recent publica-confim rates of 10 ysh* which states that actual failure statistics tion by Dr. S. H. B to 10-per reactor-year in large pioes, with higher rates as the pipe size decreases.
Considerina these analyses, we conclude that our conservative estimate on a pipe break in the primary coolant system is in the range of 10" to 10.
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 adecuacy of the reactor vessel supports and other affected structures and sysu ms to'witnstand asymmetric LOCA loads, including an assessment of the ' effects of asymmetric loads produced by various oipe breaks both inside and outside the reactor cavity. On performing these evaluations the staff will permit the grouping of plants, where adequate justifica-tion 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 licen-sees 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 i
cavity.
It is expected that a similar meeting will be held in the near l
- " Critical Factors in Blowdown loads in the PWR Guillotine Nozzle i
Brea'K (Volume 2 - the Asymmetric Load Problem)"mdated. lone o,1977 i
117 021 t
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. future to address breaks 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, resul ting from asynmetric 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 state-ment, we wish to make it clear that plant modif': cations, 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 unwar* anted based on the consideration of overall risk.
In such cases, and only in such cases, we will be prepared to give further consideration to alternate approaches, such as probability /ISI. We feel, hcNever, that ISI techniques as they exist today could be considerably improved, and, to the extent that such aprovements could have a direct bearing on this problem as well as an
-Ict of nuclear safety in general, we would welcome their development.
l i
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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 asymetric loads under the conditions thai. esult 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 Founds necessary to provide high assurance that the reactor can be brought safely to a en'd shutdown condition.
For the purpose of this request for additioral infor-mation the reactor system components that require reassessment shall include:
a.
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 9
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 asymetric LOCA loads on the above-mentiened 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 gecmetry 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 nozzie safe ends, pump discharge nozzle and crossover leg that re-sult in the most severe loading cmditions for the above-mentioned
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systems.* Provide an assessment of the effects of asymetric pres-sure differentials on these systems /ccmoonents in combination with all external loadings including asy=etric 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 displacer.ent exceeding previcus design limits provide an evaluation of the folicwing:
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 struc-tural and hydraulic computer codes employed, drawings of the models employed and comparisons of the calculated to allowable stresses and strains or deflections with a basis for the allowable values.
C.
Provide an estimate of the total amount of permanent deformation sustainc i by the fuel spacer grids.
Include a description of the impact testing that was per~ormed in support of your estimate.
Address the effects of operating temperatures, secondary impacts, and irradiated inaterial properties (strength and ductility) en the amount of predicted defomation.
Demonstrate that the fuel will remain coolable for all predicted gecmetries.
7.
Demonstrate that active ccmconents will perform their safety function when subjected to the postulated loads resulting from a pipe break in the reactor coolant system.
8.
Demonstree functicnability of any essential piping where service level B limits are exceeded.
In order to review the methods empicyed to comcute the asymmetrical presnre differences across the core support barrel during subccoled portion of the blowdown analysis, the folicwing infomation is requested:
- B&W and CE plant licensees should also consider breaks in the hot leg at the steam generator inlet.
117 024
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. 1.
A complete description of the hydraulic code (s) Jsed including the development of the equations being solved, the assumptions and simplifications used to solve the equations, the limitations re-sulting from these assumptions and simolifications and the numerical methods used to solve the final set of equations. Provide comoarisons with experimental data, covering a wide range of scales, to cemonstrate 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 oressure 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 pres.;ure difference across the rnre supoort barrel is made without the use of a hydraulic blowdown code, present the methodolcgy used to evaluate the asym: etric lcads and provide justification that this assessment provides a conservative estimate of the effects of the pcstulated LOCA.
A compartment multi-node, space-time pressure response analysis is necessary to determine the external forces and moments on ccmoonents.
Analyses should be perfor ed to determine the pressure transient resulting from postulated hot leg and cold leg reactor coolant system pipe ructur ?s within the reactor cavity and any pipe penetrations.
If applicable, similar analyses should be perfcreed for steam generator coccartments that may be subject to pressurization where significant ccmoonent suoport loads may result. This information can be provided to encompass a group of similarly designed plants (generic approach) or a purely plant soecinc (custcm plant) evaluation can be developed.
In either case, the proposed method of evaluation and principal assumotions to be used ir the analysis should be provided for review in advaace 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 Diant.
For instance, this should include blowdown rate and geometrical varia-tions 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) nocal model development for the configuration representing the most restrictive geometry; i.e., requiring the greatest nodalization; (2) the must restrictive configuration regarding vent areas and obstructions to ficw should be analyzed; and, (3) sensitivity to cadr data inout shculd be evaluated; e.g., loss coefficients, inertia terms, vent areas, nodal volumes, and any other inout data ' here there may be variations from plant to plant or uncertainty for the given plant.
117 025
_4 These studies should be directed at evaluating the maximum lateral and vertical force and ecment time functions, recognizing that models may be different for lateral as opposed to vertical load d2finitions.
The following is the type of information needed for, both generic and custom plant evaluations. Althougn this request was primarily developeo for reactor cavity analyses it may be applied to other component sub-compartments by general application.
(1)
Proside 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, comocr.ent supports design, or both.
(2)
For each comoartment, provide a table of blowdcwn mass flow rate and energy release rate as a function of time for the break wnich results in the maximum structural load, and for the break which was used for the component -suoports evaluation.
(3) Provide a schematic drawing showing the compartment nodalization for the determination of maxi. cum structural loads, and for the component supports evaluation.
Provide sufficiently detailed plan and section drawings for several views, including princioal dimensions, showing the arrangement of the ccmpartment structure, major components, piping, and other major cbstructions and vent areas to permit verification of the subcompartment nodalization and vent locations.
(4) Provide a tabulation of the naal net-free volumes and injerconnecting flow path areas.
For each flow path, provide an L/A (f t- ) ratio, where L is the average distar.ce the fluid flows in that ficw rTth and A is the effective cross sectional area.
Provide and justify values of vent less coefficients and/or friction factors used to calculate flow between nodal volumes. When a losa coefficient con-sists of more than ona component, identify each comoonent, its value and the flow area at which the loss ccefficient acpiies.
(5) Describe the nodalization sensitivity study performed to detemine 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 circunferentially, axially and radially within the conoartment. The nodal meal development studies should shcw that a spatially convergent differen-tial pressure distribution has been obtained for the selected evalua-tion :rodel.
117 026
' Describe the justify the nodalization sensitivity study perfor"ed for the major component supports evaluated, if different from tne structural analysis codel, where transient forces and moments acting on the components are of concern. Where comoonent loads are of primary interest, show the effect of noding variations on the transient forces and mcments.
Use this information to justify the nodal model selected for use in the component supports evaluation.
If the pressurization of subvolumes located in regions way from the break location is of concern for plant safety, show that the selec-tion of parameters which affect the calculations have been conserva-tively evaluated. This is particularly true for pressurization of the volume beneath the reactor vessel.
In this case, a model which predicts the highest pressurization belcw 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 loca-tion 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, W imarily the RPV no::les other than the break no::le; and (3) the region encompassing the lower reactor cavity and other portions of the reactor. cavity not included in Items (1) and (2).
(6)
Discuss the manner in which movable costructions to vent ficw (such as insulation, ducting, plugs, and seals) were treated.
Provide analytical and experimental justification that vent areas will not be partially or completely plugced by displaced objects.
Discuss how insulation for piping and components was considered in determining volumes and vent areas.
(7) Grachically show the pressure (psia) and differential pressure (psi) response as functions of time for a representative number of nodes to indicate the spatial pressure resconse.
Discuss the basis for establishing the differential pressure on structures and components.
7 n')7 i
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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 ccrpartment 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 ficw coefficients were deternined to assure that regions removed ftcm the break locaticn are conservatively designed, particularly for the reactor cavity as discussed above.
(9) provide the peak and transient loading on the major ccmponents used to establish the adequacy of the support design. This should include the load forcing functions (e.g., f (t), f (t), f (t)) and transient x
y z
mcments (e.g., Mx(t), M (t), M:(t)) as resolved about a specific, y
identified coordinate systen. The centerline of the break nozzle j
is recomended as the X coordinate snd 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 ccordinate system.
This infor ation should be presented in such a manner that confirmatory evaluations of the loads and moment > can be made.
117 028