ML19210A665
| ML19210A665 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/31/1978 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| GQL-0148, GQL-148, NUDOCS 7910300695 | |
| Download: ML19210A665 (5) | |
Text
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Metropolitan Edison Company 1/31/78 70:
Mr. R. W. Reid Rea ding, Pa.
J. G. Herbein 2/6/78
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ENCLOJJME OSSC.9tPTICM Consists of requested inf o. concerning the pot.ential for Lamellar tearing of steam ge.ierator & reactor coolant pump support i
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METROPOLIiAN EDISON COMPANY POST OFFICE BOX 542 RE ADING, PENNSYLVANI A 19603 TELEPHONE 215 - 929M601 January 31, 1973 GQL Olh8 7
Mr. R. W. Reid, Chief
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Operating Reactors Branch :To. h j
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\\D( "#HP;j'i7 U. S. Nuclear Regulatcry Ccesissica
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Washington, D. C. 20555
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Dear Sir:
.wh s 7 j h.
Three Mile Island :hclear Station Unit 1 (SII-1)
\\;h Operating License :To. DPR-50 Decket :To. 50-289 In response to your letter of September lb,19TT, and our letter of :Tovember 18,1977 (GQL 1529), we have attached the requested information concerning the pctential for Lamellar tearing of steam generatcr and resctor coolant pump support materials.
The TMI steam generators are suppcrted at the bottc= by the support skirt which is welded to the lower head. This skirt is attached to a base plate which is supported by the concrete foundation. The support near the top of the generators was not provided by B&W and is not being considered in this evaluation.
In evaluating the potential for Lamellar tearing, the skirt is not a primary concern as it is not subjected to major leading. This is because of the simplicity of its gec=etry and the minimi::ing of the large intersecting =er-bers. The weld areas of the skirt were preheated before welding; the finished assembly was post-heat treated and all welds were either magnetic pa_ticle and/or radiographically examined.
The TMI-1 reactor coolant pumps are completely supported by the reactor coclant piping which is supported from the reactor vessel and steam genera-tors. Supplementary supports for the piping or reactor coolant pumps are not used. The?;efore, no respense is required.
We have not submitted cur own evaluation of the fracture toughness of the steam generator and reactor coolant pump support materials per a telephone conversation of October 20, 1977 between R. Snaider of your effice and D. G.
Mitchell and E. G. Skuchas of Met-Ed.
M c - el y /y 1493 154 b
/ /
[//J.
Herbein 2.
Vice President-Generation U
Attachment JGE:DGM:cjg
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t to G2L 0148 At-t Quection 1:
Provide engineering drawings of the steam generator and reactor coolant pump supports sufficient to shcw the geccetry of all principal elements.
Provide a listing of =aterials of construction.
Fesponse:
Drasings enclosed:
a.
Number 13113CE - Assenbly and Detail of Support Skirt b.
Number 13110hE - List of Materfal/ Steam Generator c.
Number 131112E - Shell and Tubesheet Attachment Assembly Question 2:
Specify the detailed design loads used in the analysis and design of the suppc rts.
For each loading condition (normal, upset, emergency and faulted), provide the calculated maxi =um stress in each principal element of the support system and the correspcnding allowable stresses.
Re=psnse:
Normal & Upset Enerzency Faulted Lead (Ry Kip; Mr Kip in. ) 1801 ha,000 1911 96,000 20h6 306,000 Max. Stress (Kip /in2)
- Neg.
- Neg; *Neg.
h.3 20.h3 Allovable (Kip /in2) 5(3Sm) 8r 33.9
.5(Sy) er 18 5(1.257) or 21.6
.5(1.5Sy) or 27 5(1.8S7) or 32.h 5(1.357) or 32.h
- Indicates that the intersecting members are in a compressive state.
Question 3:
. Describe how all heavy section intersection s ther veldments were designed to minimize restraint and latellar tearing. Spec fy the actual section thick-nesses in the structure and provide details of 1*uical joint designs. State the maxi =am design stress used for the through-thickness direction of plates and elements of rolled shapes.
Response
There is no major leading on the steam generator skirt to show cause for concern about Lamellar tearing. See Item 1 for structure details. See No. 2 for design stress.
Question h:
Specify the minirum crerating temperature for the supports and describe the extent to which material temperatures have been measured and. at varicus points en the supports during the oteration of the plant.
1493 155
. Attachment to GQL Olhe
Response
No minimum operating temperature is specified for the steam generator supports. No temperatures have been measured on the supports of the steam generators during operatien of the plant.
Questien 5:
Specify all the materials used in the supports and the extent to which mill certificate data are available. Describe any supplemental require-ments such as melting practice, toughness tests and through-thickness tests specified.
Provide the results of all tests that may better define the properties of tne materials used.
Respon:c:
a.
The materials are listed in the List of Materials Drawing.
b.
Mark 96 has some mill certification information available which will be forwarded at a later date.
c.
This support material was ordered in accordance with the re-quirements en the ASME Code,Section II.
d.
No data from tcughness or through vall tests is available.
Question 6:
Describe the velding procedures and any special velding prccess requirements that were specified to minimize residual stress, veld and heat affected zone cracking and lamellar tearing of the base metal.
Respense:
a.
'4 eld Number Process Preheat (Min.)
61 Submerged Arc & Manual Metal Arc 2000F 6h Sdbnerged Arc 2000F 65 Flux Core 200 F 70 L 71 Flux Core 2000F b.
These velds all received a full Section III stress relief.
Questica 7:
Describe a:.1 inspections and non-destructive tests that were performed en the supports during their fabrication and installation, as well as any additional inspections that vere performed during the life of the facility.
Response
The Ncn-Destructive Examinations are shown en Drawing Nunter 131130E and described en Table I.
1493 156
/
TABLE I Recrance to flRC Question Ilo. 7 Ilon-Destructive Examinations For Welds In Support Skirt St.ructure N
ET)
Weld and Incation Drawing, View or Cection flon-Destructive Examination M
Tio. 61, Support skirt (96) to transition ring (95) 131112, Detail D Radiographic, Magnetic Particle 6
v fio. 6h, Cupport skirt (9 ) assembly 131130, Detail J Radiographic, Magnetic Particle 6
65, Cupport skirt (96) to base pinte (97) 131130, Detail M Magnetic Particle 4
70, Gusset pintes (98) to support skirt (96) 131130, Detail K Magnetic Particle No. 71, Gusset plates (98) to base plate (97) 131130, Section FF Magnetic Particle
/
l 9
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 asymetric 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 tenn 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 Ste 10,
., 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 1493 158
January 25, 1978 ENCLOSURE 1 BACKGROUND AND CURRENT STATUS OF THE NRC STAFF REVIEW OF ASYMETRIC LOCA LOADS FOR PWR FACILITIES On May 7,1975, the l'RC was informed by Virginia Electric & Power Company that an asymmetric 1(ading on t5e reactor vessel supports resulting from a postulated reactor coolant pipe ruptur - d a specific loMtion (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 an the vessel due to transient differential pressures 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 that postu-lated 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 vassel 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 ten-sion 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 decompr7ssion of the upper plenum.
1493 159
. 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 insidg the reactor cavity (i.e., at a nozzle), it has since become apparent that significant asymetric 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, tne 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 Combvtion 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 1975, the NRC fomed 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 asy/ metric 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 infomation needs. Several long-term experimental programs would be required to provide much of this infonnation. 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).
1493 160 d
/ A second - and equally important - reason for not accepting pr obability/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 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 duriag this interim period while additional evaluations can be perfonned 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 proba5ility of a reactor vessel itself has been estimated to lie between 10' 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 {or all pjpes sizes 6" and greater (with a lower and upper bound of 10- 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 lower bound is more appropriate since these pipes are more like vessels in size.
In addition, 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 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
, faciliti.es for operation can continue while this matter is being i Mesolveil. A -
1493 161 3
In support of the above, the staff has developed a short-term interim cri-terion to detennine 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 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 rangina 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 estimated to be 10'glds, the upper bound probability of pipe break wasThis value is thick-walled pipe w confirm rates of 10 gsh* whigh states that actual failure statistics tion by Dr. S. H. B to 10- per reactor-year in large pipes, with higher rates as the pipe size decreases. Considering these analyses, we conclude that our conservative estjmate on a pipe break in the primary coolant system is in the range of 10- 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 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 planf.s, 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 cavity.
It is expected that a similar meeting will be held in the near
- " Critical Factors in Blowdown Loads in the PWR Guillotine Nozzle Break (Volume 2 - the Asymmetric Load Problem)"mdated. lune 6,1977 1493 162
. 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 resulting from asymmetric LOCA loads. We believe that the results of this program which will ir '"de sensitivity studies, will provide significant confirmatory informa-
,n 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 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.
Ir. such cases, and only in such cases, we will be preoared to give further consideration to alternate approaches, such as probabi:ity/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.
1493 163 c.
8
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 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 g.
Reactor Internals h.
Biological Shield Wall and Neutron Shield Tank (where applicable) 1.
Steam Generator Compartmet, 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 s'ystems 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 hoi, and cold leg nozzle safe ends, pump discharge nozzle and crossover leg that re-sult in the most severe loading conditions for the above-mentioned 1493 164 systems.* Provide an assessment of the effects of asycretric pres-sure 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 as ess'~ tnt 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 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.
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 infor: nation is requested:
- B&W and CE plant licensees should also consider breaks in the hot leg at the steam generator inlet.
1493 165
3-1.
A complete description of the hydraulic code (s) used 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 simplifications and the numerical methot used to solve the final set of equations. Provide comparisons with experimental data, cover %g 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-d'mensional aspects of the reactor geometry.
If an approved vendor code is used to obtain the asynmetric 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.aay vary from plant to plant.
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) 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 codesdata input should be evaluated; e.g., loss coefficie'ntp imertPa! erms, vent areas, nodal volumes, and any t
'other inouh data where there may be variations from plant to plant or uncertainty for the given plant.
1493 166
' 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 sub-
'ompartments 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 max'..um structural loads, ano 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 in{erconnecting flow path areas. For each flow path, provide an L/A (ft- ) 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 vent loss coefficients and/or friction factors used to calculate flow between nodal volumes. When a loss coefficient con-sists of more than one component, identify each component, its value and the flow area at which the loss coefficient applies.
(5) Ucscribe 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 incluJe consideration of spatial pressure variation; e.g., pressure variation circumferentially, axially and radially within the compartment. The nadal model development studies should show that a spatially convergent differen-tial pressure distribution has been obtained for the selected evalua-tion model.
1493 167 o
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 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 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
& 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, 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 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 experimentel justification that vent areas will not be partially or complete 1/ plugged by displaced objects. Discuss how insulation for piping and components was considered in determining volumes and vent ar:as.
(7) Graphically show che 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 consorvatively 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 moments (e.g., Mx(t), M (t), Mz(t))x(t), f (t), f (t)) and transient the load forcing functions (e.g., f y
z as resolved about a specific, y
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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Metropolitan Edison Company cc:
G. F. Trowbridge, Esquire Shaw, Pi tt. man, Potts & Trowbridge 1800 fi Str s t, N.W.
Washington, D.C.
20036 GPU Service Corporation Richard U. Heward, Project flanager Mr. T. Gary Groughton, Safety and Licensing ilaneger 260 Cherry Hill Road Parsippany, l'ew Jersey 07054 Pennsylvania Elcatric Ccepany i;r. R. U. Ccnrad Vice President, Generation 1001 Broad Street Johnstc.en, Penasylvania 15907 iii ss !!c ry ','. Scuuard, Chai n..i Citizens for a Safe Environmnt P. O. Scx 40E Harrisburg, Pennsylvania 17103 Government Publications Section State Library of Pennsylvania Box 1601 (Education Building)
Harrisburg, Pennsylvania 7
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