ML20234E570
| ML20234E570 | |
| Person / Time | |
|---|---|
| Issue date: | 10/26/1977 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML20234E460 | List:
|
| References | |
| FOIA-87-40 ACRS-GENERAL, NUDOCS 8707070613 | |
| Download: ML20234E570 (25) | |
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8 6" 0FFICIAL USE.0NLY DATE OF MEETING: 10/26/77
[ DATE ISSUED: 2/3/77 f -.
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M h, g h u M' g MINUTES OF THE SUBCOMMITTEE CONTAINS PROPRIETARY DATA ON FLUID / HYDRAULIC DYNAMIC EFFECTS MEETING RODEWAY INN PORTLAND, OREGON OCTOBER 26, 1977 The ACRS Subcommittee on Fluid / Hydraulic Dynamic Effects met on October 26, 1977 at the Rodeway Inn, 7101 NE 82nd Avenue, Portland, Oregon for the purpose of continuing its discussion on the effects of blowdown forces on reactor vessel supports. Notice of the meeting and its agenda was published in the Federal Register, Vol. 42, No.
196, p. 54894, on Tuesday, October 11,1977(Attachments). A copy of the Tentative Detailed Schedule is attached (Attachment B). A list of attendees is attached (Attachment C). Slides and handouts s
used at the meeting are attached as follows:
Attachment D: Babcock & Wilcox Presentation Attachment E: " Structural Analyses Reactor Pressure Vessel Supports", Combustion Engineering Attachment F: EG&G Presentation No requests for oral statements and no written statements were received from members of the public.
4 EXECUTIVE SESSION The meeting was called to order at 8:45. Present were:
Dr. M.S. Plesset, Subcommittee Chairman Dr. S.H. Btch, ACRS Member Mr. H. Etherington, ACRS Member Mr. J.C. Ebersole, ACRS Member Dr. Z. Zudans, ACRS Consultant Dr. I. Catton, ACRS Consultant Dr. S. Dong, ACRS Consultant Dr. L. Yao, ACRS Consultant Dr. Richard P. Savio, ACRS Staff, DFE Mr. R. Muller, ACRS Staff gg7o g a 87o62 PDR 1FFICIR HE MW .
THOMAS 87-40
4 0FFICIAL USE ONLY Dr. Bush inquired whether the utilities would be represented. He was advised that the B&W Owners' Group would be present. Several members indicated they had not received needed mail which had been sent over a week earlier. B&W had submitted 2 reports, BAW 10131, and BAW 10132.
MEETING WITH NRC STAFF, B&W, CE, AND CONSULTANTS The meeting was delayed to enable the hotel to produce an overhead projector and to activate the amplifying system (both of which had been prearranged). At 9:30 it was decided to proceed without an operating amplifying system. Dr. Plesset asked that speakers speak in a loud voice to pennit the reporter to hear the questions and presentations.
B&W PRESENTATION The first speaker was William R. Speight of B&W. He introduced Mr. H.J. Fortune, Dr. Felix Aguilar and Mr. L.M. Smalec who would discuss Structural Loading Analysis, the CRAFT 2 Model, and LOCA Loads Calculations using CRAFT-2, respectively. Mr. Speight called the Subconinittee's attention to the B&W Topical Reports listed on Slide 3(Attached).
- 1. Structural Loading Analysis f Using Slide 7, Jim Fortune, B&W, explained that he would focus first on the supports of the reactor vessel which were shown in green. He I showed the forces acting on the vessel during a LOCA with the help of Slide 8. In the event of cold leg break initially there would be a decompression wave starting at the break plane and moving back into the, reactor. It would tend to move around the internals and also to l propagate downward. Initially, there would be low pressure on the l break side of the internals and high pressure on the opposite side i
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4 0FFICIAL USE ONLY resulting in an asyrmetric loading on the interior of the vessel wall and on the shell of the internals. (Forces labelled F), and F2 on Slide 8). As the wave propagates there would be low pressure on the lower head of the vessel and high pressure at the upper head.
In the initial phases there would be an uplift force (F ) n the 3
reactor vessel. The other internal force on the core would be the dynamic response, or core bounce, as the core moves up and down. (F )
4 Mr. Fortune next described external forces due to the asymmetric sub-compartment pressure loadings 7(F ) caused by the initially high pressure in the external area near the break. This would propagate beneath the vessel and provide a vertical uplift force (F6 ). The thrust loading (F5 ) would be developed perpendicular to the pipe in the event a. split occurs. Any jet impingement loading on the side of the vessel resulting from a LOCA causing the pipe to drop or move up or laterally was described as F8. Mr. Fortune used Slide 9 (Attached) to describe the data flow in BAW-10127 which is currently under review by the NRC Staff. He explained two types of non-linearities - yielding of the materials and geometric nonlinearities, such as gaps. The pipe reaction hydraulic loads are a function of the break opening area and the break opening time. The structural and hydraulics people alternate in an iterative process to solve for the break opening area and break opening time.
Knowing these, the mass-energy release yields subcompartment pressure loadings. Mr. Fortune explained that the loading calculations are nonnally perfonned by the architect / engineer or Balance-Of-Plant designer.
In response to questions by Dr. Zudans, Mr. Fortune explained that changing flow is assumed, using upper bound deflections larger than would actually occur in practice.
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0FFICIAL USE ONLY Mr. Fortune indicated that there are 5 components primarily affected:
the reactor vessel and its supports, the internals, the core, the CRD mechanisms, and the service structure.
Next he showed a typical restraint (Slide 10).
He explained that a non-linear dynamic analysis leads to break opening areas. A sample is illustrated on Slide 11, which showed a reactor vessel outlet nozzle break. The ordinate is a percentage of full break area.
Mr. Fortune pointed out that vessel motion will not affect the motion of the pipe, but will affect the area that exists between the vessel and the pipe. The non-linear analysis is used to determine pipe motion.
The most conservative assumption is to assume no pressure drop. Then the pipe undergoes full system pressure times the area of the pipe, instantaneously, causing the pipe to impact the supports and restraints.
The reactor vessel moves on the order of 1/8" horizontally about 30 mills vertically, whereas the pipe moves on the order of an inch. ,
1 Dr. Bush observed that there were many different types of pipe restraints and the one selected by B&W may not be the worst case.
Mr. Fortune stated that for the hot leg case shown B&W had taken the worst case situation. He admitted that for the cold leg it was a bit more contract - unique, but he felt B&W had been successful to a degree in working with a/e's and suggesting adequate pipe support systems.
Mr. Ebersole inquired about a break on a 45 plane which would lose support for the vertical run of pipe causing an increased opening size.
Mr. Fortune stated that this would not provide a " worst case" because as you get further away from the vessel, as you would in this case, the internal pressures would be less.
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In response to Dr. Zudans question about hydraulic forces, Mr. Fortune explained that the hydraulic forcing function recJCes the loading with time, but as a conservatism it is assumed that it does not reduce with time.
Next with the help of Slide 13, Mr. Fortune discussed core bounce loads. He explained that having the break opening area,.one ts able to calculate the internal pressures. The tubes of the fuel assemblies are held together by friction, sliding friction between pins and grids, that is a function of the burnup rate. A fuel assembly is held down with a compression spring, such that under activation of a i.0CA, a l pressure differential would move downward, in the case of a cold leg' break, and the core would respond dynamically. It would impact on the upper grid. The fuel pins would move within the grillages which would help dissipate the energy. The core would then come back down and hit the internals. The loadings are then applied to the reactor vessel upper supports. It must be shown that all these loads are acceptable.
I In response to Mr. Ebersole's question, it was explained that in the loading calculations it was assumed the control rods never come in, despite the physical requirement that they have to get back in. Mr.
Fortune explained that the phenomenon discussed involved only the first 30 milliseconds.
l In response to Dr. Zudans question, Mr. Fortune explained that a large break opening was assumed, and as long as the calculated area was less than the original assumption, it was conservative. For some things, !
such as cavity pressure, the process was iterated to get the assumed break and the calculated break closer together. However, for internal pressure calculations, a full break was assumed and was not subsequently changed. This was done for scheduling reasons, and for future cost reductions the calculations may be iterated further.
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In response to Dr. Yao's question, Mr. Fortune explained that rotattonal ;
forces were very small and rotation was picked up as a function of the vertical model.
l Mr. Fortune stated that the flow is flat across the core permitting the use of an axisymmetric wave. Dr. Catton pointed out that if the break is en one side, the expansion wave would reach the bottom of the core barrel 4 on one side before the other. Mr. Fortune responded that the physical t features of the core symetrize the problem. The annulus is considered asynme tri c. The asymmetric effect is accounted for in on the motion of l
the internals. !
4 Mr. Fortune explained that there was a vertical model and a horizontal l model uncoupled from each other.
Dr. Bush pointed out that the discussion so far had concerned a cold leg b re ak . Results would be quite different for a hot leg break. Mr. Fortune concurred, adding that in the hot leg case flow would be out through the plenum. This would also be axisymetric as far as the core was concerned.
In response to Mr. Ebersole's question, Mr. Fortune stated that in the hot leg case the core uplift would be greater. The core springs, which are designed for normal operation, would bottom out in the hot leg case.
Mr. Fortune showed the isolated reactor vessel model used (Slide 14). l It is a beam model. All 205 fuel assemblies are modelled into one.
The hydraulic loadings developod are integrated over the interior wall and over the wall of the internals, and these are applied as concentrated loadings at the mass joints. The cavity pressures are similarly applied to the reactor vessel. 1 OFFICIAL USE ONLY
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0FFICIAL USE ONLY Using a plan model (Slide 15) Mr. Fortune explained how loading reactors F, thru F 7are applied, including F which 4
was developed from the bounce model. !
It was determined that the role of the steam generator in breaks in and around the reactor vessel was unimportant in tenns of dynamic response. j The shell was modelled as a finite element model, using an axisymmetric shell. Stiffness matrices were developed and these in turn were developed into beam equivalencies.
After inquiring into the degrees of freedom (at least 3 in each mode up to 6 at certain points such as the reactor vessel nozzle intersection),
Dr. Zudans said he felt it was a good model. Rotatory inertia was con-si de red. j Dr. Bush pointed out that architect / engineer influences the problem in such matters as the shield wall in the cavity which may have significant impact on the loads on the vessel supports, Mr. Fortune indicated that information is sent to the balance of plant designer, and taking ad-vantage of reduced break opening sizes, the a/e can calculate pressures on his concrete.
l l Mr. Fortune next covered the resulting Loading Specifications wtth Slide 16. Typical 205 Fuel Assembly Loadings for a cold leg break were
! illustrated with Slide 17.
A topical report has been filed on internals covering internals loading.
l A non-linear analysis was perfonned on the control rod drives, inputting l the time history motions developed both for LOCA and a seismic event.
l The drives are mathematically shaped to make sure none of the components have large plasticity which would impair insertion possibilities later on.
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0FFICIAL USE ONLY For fuel assemblies there is also a horizontal impact model. The ascertain that the internals probably do not have large structural defor-mations sufficient to prevent a coolable geometry from being maintained.
B&W has determined that for all cases there is an acceptable amount of de formation. This is documented in BAW-10133, for Mark C fuel. For Mark B fuel there is a non-proprietary version, BAW-10035.
Dr. Bush inquired about the loading on the large instrument package on the bottom of the reactor, expressing concern that a significant number of instrument lines were. sheared away there would be no knowledge of what is happening in the plant thereafter. Mr. Fortune indicated that pressure loading on that piping had not been provided. ?
Mr. Etherington pointed out that with the assumption of constant friction, one must use different equations for the positive and negative directions of movement which considerably complicates matters. Mr. Fortune concurred.
Mr. Fortune reported that B&W 10127 on pipe break criteria did not consider splits a credible break, only quillotine breaks are credible.
This is still under Staff review.
Following Mr. Fortune's presentation, Mr. Speight requested the. meeting be closed to discuss B&W proprietary data.
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i 0FFICIAL USE ONLY MEETING WITH B&W OWNERS GROUP (OPEN)
Dr. Plesset called on Mr. Louis Lanese of GPU Public Service Corporation, Chairman of the B&W Owners' Group Subcommittee on Reactor Vessel Supports.
The purpose of his organization was to determine if the required analysis was necessary in view of the low probability of the initiating event, and if it was, to participate in the development of the analytical tools and to derive whatever benefit possible by doing as much of the analysis as could be done on a generic basis.
The Group engaged Science Applications, Inc. to assess the probabilities of a disruptive break in the primary system in the reactor cavity. Dr.
David Harris of SAI wrote a report on this work and was available to discuss it. Based on this work the Group does not feel further analysis is warranted.
Mr. Lanese then reviewed data from the report.
In response to Dr. Zudans' question, Dr. Harris explained that the pro-bability of failure of the field welds is lower than of the shop welds because they are not as highly stressed.
i Mr. Lanese pointed out that for the SAI B&W study the probabilities of failure in the cavity were even less than the probabilities in the earlier SAI CE study because in the B&W system the cavity welds are very lowly stressed.
MEETING WITH NRC STAFF Dr. Plesset called on Mr. Vincent Noonan, NRC Staff, to lead off the Staff presentation. Mr. Noonan called on Mr. Shao who explained that this was one of about 30 Category 8 A generic subjects, which are on a high priority list.
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0FFICIAL USE ONLY Mr. Noonan discussed the operating plants. Meetings have been held with the W Owners Group and the CE Owners Group but.not yet with the B&W Owners Group. The Staff will require some sort of analysis from each. Break opening times in the range of 10-30 milliseconds have :
been discussed. Thennohydraulic structure interaction will be used if required and realistic damping values will be used. As-built material properties will be used. Strength properties of materials will be used on a dynamic vs. a static basis. The Staff has completed its analysis on Indian Point 3 and is 50 percent complete on AN0-1.
In response to Dr. Bush's connents, Mr. Shao, NRC Staff, explained ;
that some dynamic analyses had been done to show that under dynamic loading the factor of safety was greater than that under static load-ing based on the ASME Code.
Mr. Noonan stated that whereas earlier work was limited to the reactor supports, asynmetric loads were now being analyzed in other parts of the system. He cited fuel as an area of concern. He indicated the Staff would like to detennine that breaks inside the cavity are limiting breaks and that therefore breaks outside the containment would not need to be analyzed, j
In response to a question from Dr. Bush, Mr. Noonan indicated that the Staff would cover the possibility of loss of numerous instrument lines as a result of this type of accident, (which could exacerbate the accident), at a later date.
Mr. Etherington stated that he would be interested in the Staff's assessment of the probability aspects of such a break. Dr. Plesset l commented that a probabilistic approach must be used with caution.
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l Mr. Shao indicated that LOCA loads mostly greatly exceed the SSE loads. )
The Staff is examining the combination of the loads.
Dr. Bush cautioned that fatigue was but one fonn of pipe failure.
In response to a question from Dr. Zudans, Mr. Smalec reported that a ;
hot leg break displaced eight feet from the vessel nozzle reduced the loading components twenty to forty percent. Mr. Fortune pointed out that peak values of the various forces do not necessarily occur I I
simultaneously.
Mr. Fortune stated that if the break were outside the cavity the design would be acceptable. Dr. Plesset- conmented that it would be nice to see the calculations for the break outside the cavity. Mr. Fortune replied that they were available for the 145 and 205 plants but not for the 177 plants. -
In response to Dr. Bush's question, Mr. Fortune indicated that B&W has preferred to put the analysis in plant specific FSAR's rather than in a topical report. !
Dr. Catton observed that perhaps more time should be spent on the annulus forces than the forces on the internals. Mr. Noonan indicated that in the Indian Point 3 analysis the Staff would present data showing various loads and their time duration and the overall response of the system, then the differences and effects of e.g. the cavity load, and the thermal hydraulic load could be seen.
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0FFICIAL USE ONLY Next Ed Throm discussed the CRAFT applications. He indicated the Staff was reviewing BAW-10132. Questions were sent to BaW on Sept.15 and a meeting held with B&W on October 12. Some responses are expected in February and the remaining information by June 1,1978. The Staff is looking at the analytical development of the code, the application and systemic modelling (on four NRC standard sample problems), (for ex-perimental verification the Staff is accepting semi-scales), the pre-diction of results on the HDR German test facility, and the independent NRC analysis of the BSAR-205 plant at EG&G Idaho. The Staff expects an SER to be out about 2-3 months following receipt of the last infonna-tion in June.
Dr. Mattu informed the Subcommittee that the Staff had informed B&W of their concerns re BAW-10131, the reactor coolant system structural loading analysis, and was expecting clarification. In addition, the Staff is looking generically at new plants at the effects of assynnetric blowdown loads on the reactor vessel.
For the NRC Staff, EG&G is working on the mechanical response of the reactor coolant system to a pipe break for the Bellefonte plant. The information was received from TVA in September. An analysis of the CE furnished San Onofre 2 plant is complete for the LOCA transient loading.
Dr. Plesset indicated that time had been allotted to discuss thistafter the CE presentation.
Dr. Zudans pointed out one needs not only the forces but the point of application so that one can determine the overturning moment. Mr. Fortune indicated that the A/E would supply this information for the FSAR.
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l COMBUSTION ENGINEERING Next Gary King of CE indicated that CE had covered the thermohydraulic aspects of the problem at the May 25th Subcommittee meeting and would continue today the structural analysis of the reactor vessel supports.
He planned to cover generic methods and then the San Onofre 2&3 results.
Mr. Thomas Natan, CE, indicated that there had been no changes in structural methods since the May 25 meeting in Los Angeles. The support systems for all 3410 plants are not necessarily identical. Though all have the same roots, a plant specific analysis is planned for each plant. The CE presentation would discuss only the pipe rupture loads, and these would only be tied in with other (e.g. seismic) loads at the end of the presentation.
Bob Kassawara, CE, covered the generic part of the CE presentation with the aid of Slides CE-1 to CE-27 (Attached).
He pointed out on Slide CE-2, that the vessel, steam generators, and hot leg piping account for 90% of the total mass and stiffness of the system.
(Slide CE-3) The system allows unrestrained thermal expansion in a radial direction, and restrains motion in a horizontal and vertical direction f at the upper horizontal supports during an earthquake or a postulated pipe rupture.
Mr. Kassawara pointed out with Slide CE-9 that the model is 3-dimensional, though the internals are axisynrnetric. (Slide CE-ll) The model contains representations of the physical non-linearities, the gaps at the interfaces between the vessel and the internals.
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OFFICIAL USE ONLY In response to Mr. Etherington's question, Mr. Longo explained that core vessel support snubbers permit free radial expansion but snub the horizontal motion tangential to the radial.
Mr. Kassawara pointed out that the core support barrel is immersed in water, and any motion of the core support barrel relative to the vessel causes fluid to be accelerated around the annulus between the vessel and the core support barrel giving rise to annular hydrodynamic effects. This turns up in the analysis in the form of a non-diagonal mass matrix.
He stated that there were 67 mass degrees of freedom in the model and another 38 dynamic degrees of freedom.
In developing the forcing functions, the pipe breaks are postulated, then the pipe break forces are calculated, and the decompression forces inside the reactor vessel and the cavity pressurization forces outside the reactor vessel.
The most severe breaks for the reactor vessel and supports are the 100 square inch outlet nozzle guillotine, and the 350 square inch inlet nozzle guillotine. (Slide 16).
The stop on the steam generator sliding base is sized to take the pipe rupture load when the hot leg is no longer there. Opposite motion, if it occurs, is taken into account in the model, if there were significant motion of the reactor vessel.
Mr. Shao asked if CE had looked at a break near the generator. Mr.
Kassawara replied that they had and that a break at the vessel was more severe. Mr. Natan added that he could not state positively that the steam generator tubes had been analyzed in a similar way to the reactor internals.
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Mr. Kassawara described the sequence of events occurring on pipe rupture in the vessel cavity. The forces inside the vessel are calculated by l thermal hydraulic analysis. The variation with time are shown in Slide CE-22. The total force is shown in Slide CE-23. Mr. Etherington noted is it decaying to the right. i Slido CE-25 shows that the force on the reactor internals is mostly out of phase with the vertical force on the reactor vessel itself.
Slide CE-26 shows that the typical cavity pressure forces are not oscillatory as are the loads inside the vessel.
Dr. Zudans recommended that overturning monents be studied by CE also.
Dr. Catton asked how the forces in the annulus were calculated. Mr. Natan pointed out that this was plant-specific and he called on Bechtel to dis-cuss the San Onofre plant. Mr. Fried indicated that the COPDA Code was used. The method was presented in the San Onofre FSAR. Mr. Fried stated it involved quasi steady state methodology to calculate stagnation pressure.
-It allows choking in every node, and calculates internally whether or not a choke exists there. The COPDA Code starts with mass and enthalpy rates supplied to Bechtel by CE. Bechtel only looks at depressurization in the cavity and does not couple it with the primary coolant. The COPDA Code is a compressible fluid code.
Mr. Kassawara continued with a discussion of how the dynamic analysis is performed. (Slide CE-27) In response to questions from Committee members and consultants, he stated that the STRUDL, SH0CK, and CEDAGS codes had been published. The program handles all non-linearities present in the reactor coolant system and reactor internals system.
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OFFICIAL USE ONLY In response to Dr. Shao, Mr. Kassawara replied that the compressibility of the fluid and the volume changes in the annulus were not taken into account in. calculating pressures.
The STRUDL Code is not a non-linear code. It simply fonns the stiffness matrix which is used in the actual solution of the equations of motion. )
Thus the systems equation to be solved is generated and this is solved by the CE DAGS Code.
I Mr. Noonan asked about the crushability of the fuel grids. Mr. King in-dicated that the CE fuel analysts were not present. This will be addressed at a later date. Mr. Natan indicated that en the first plant CE has presented fuel analysis in the FSAR. It is under Staff review.
Mr. King added that CE had submitted a Topical report on fuel analysis methods over a year ago.
In response to Dr. Yao's question, Mr. Kassawara stated that the change in pressure of the fluid inside the reactor vessel due to motions of the structure is not included. Mr. Longo added that the pressures are cal-culates with a hydraulic code with rigid walls. These pressures are then applied to the structural code.
Mr. Longo explained that CE tried to distinguish between hydrodynamic mass and hydrodynamic coupling so that one could realize that more than the main diagonal terms were considered, that they actually had mass coupling. The procedure is in the FSAR.
Mr. King agreed to provide a copy of the paper on the CE DAGS Code.
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4 0FFICIAL USE ONLY Mr. Noonan stated small changes in frequencies resulted in many factors of 2 variation in load on fuel. He thought this should be addressed in a future meeting when hydrodynamic mass ef#ects are discussed. Dr. Bush added that this should be done with realistic rather than conservative loads.
SAN ON0FRE STRUCTURAL ANALYSIC 0F REACTOR VESSEL SUPPORTS Mr. Natan next reviewed the dynamic analysis of the reactor vessel supports. The two worst breaks were determined to be the hot leg terminal end of 100 in2 and the cold leg terminal end of 350 in.2 ,
The rupture is considered to be instantaneous over the entire circum-fe rence. The system is then allowed to move as it wants to move because of its mass, stiffness, supporting structure, etc., to determine what the pipe opening area will be. !
Slide CE-8 shows the modular arrangement, supplied by Bechtel and presented in the FSAR. At each of the nod'es there will be pressures as a function of time, and these are multiplied by area to obtain forcing functions.
In response to Dr. Zudans question, Mr. Natan indicated that moments, per se, were of no interest and were not'specifically calculated, although they could have been determined from the code.
Mr. Natan described 2 analyses. The first covered on idealized system where all the gaps were closed and there was a rigid connection between reactor vessel and core barrel. This would provide one limit-(In response to Dr. Catton's question, Mr. Fried of Bechtel indicated that the COPDA Code was covered in the San Onofre FSAR section 6.2 pages85-108. He reported that in general the results tend to yield higher numbers than the Staff's RELAP-4 Code. It is a finite difference, quasi steady state code. )
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0FFICIAL USE ONLY Mr. Fried indicated that only 13 of 17 nodes were shown on slide 28.
A homogenous frozen - flow equation was used in this case. The Code looks at the effective K. This is a 2 component, 2 phase system.
There are 2 separate flow regimes which converge. There are 2 flow equations depending on whether it is critical or subcritical. Mr. Fried concluded with a discussion of how tho equations for adjoining nodes are hooked together.
The result tends not to oscillate and it is therefore' called quasi steady state.
Mr. Fried indicated that analytical functions were available for superheat and, in response to Dr. Catton's question, explained that this was a homogeneous frozen flow equation which did not allow for slip flow, and that Moody says that for these kinds of equations the homogeneous frozen flow is a better type of model.
Mr. Shao observed that San Onofre was de:igned for 0.679 and wondered how the seismic load compared to the LOCA load.
Mr. Natan reviewed the results illustrated in Slides 30-39. Re Slica 39, Dr.
Plesset asked Mr. Natan to explain why one gets such large loads so late l
in the process (0.25 sec.). Mr. Natan explained it was due to the gap l between the reactor vessel and the internals. At the later time, the I internals bounce off the reactor vessel providing a sizable contribution to the support loading even though the net internal asynnetric pressure may be smaller at that time.
I Mr. Natan next discussed the results on Slide CE-40 and Slide CE-41.
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I Mr. Natan confirmed to Dr. Shao that even at 0.679 the seismic load is very small compared to the LOCA load, for this application.
The steam generator support by itself has the capability of absorbing more than 7 million pounds of load. In response to Dr. Zudan's question, Mr. Natan explained that the load is transferred from the reactor to a steam generator via an intact pipe. The vertical column is rectangular.
In tension the bolts govern. They remain totally elastic.
Mr. Natan explained that the bolts were pre-loaded because the uplift forces that are generated exceed the deadweight. No gap would open unless a higher load than the pre-load is applied, and in this applica-tion this is not the case.
Dr. Bush indicated ASME hadn't yet approved Appendix M, the dynamic load i approach. Static analysis is used here and it is more conservative.
NRC STAFF ANALYSIS OF SAN ON0FRE ,
Dr. Mattu indicated that Combustion Engineering had Cefined the forcing functions and the seismic loads as well as the LOCA forcing functions for the Staff analysis of San Onofre. NRC is developing its own loads and will redo the analysis with them.
Mr. B.F. Saffell, Jr. , EG&G, Idaho, Inc. , presented the study perfonned for the Staff. He used Slides EGG-2 and 3 to describe the system. He noted there was a gap of 35 mills hot between the support nozzle flange and horizontal support. (Slide 3). The analysis was a linear elastic analysis. No decision has yet been made as to whether the independently calculated NRC Staff load would be done non-linearly. The finite element model of one loop was shown in Slide EGG-4. The Preliminary Plant Model is shown in EGG-5 and EGG-6. Slide EGG-7 shows the PRV Internals Assembly, which is modelled in Slide EGG-8. EGG considered the core 0FFICIAL USE ONLY
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barrel to reactor vessel connection rigid whereas Combustion, in their '
analysis had assumed a flexible connection there. The EGG results are compared with Combustion's in Slide EGG-9. Dr. Zudans inquired about the factor 10 difference in calculated values for Ma. Mr. Saffell !
indicated that EGG had looked into the horiznntal force (H) differanca but not yet into Ma. With modification to similar models Combustion's value for H is now 3500 kips compared to 2900 for the Staff calculation. The l Staff analysis was linear and excluded the effect of the key. This, Mr. Saffell believes, accounts for the difference.
Slide EGG-10 indicates g that when the Staff includes the key, the gap between CE and Staff calculations is further reduced.
I Mr. Saffell reviewed the current status of the EGG work for the Staff with Slide EGG-ll. Future work includes evaluating further the differences noted. He explained wny he believes the linear analysis valid.
Mr. Shao asked Mr. Natan if the allowable stress P M n the last CE slide (CE-41) $ 50.8 Ksi is taken from ASME Code Section 3 Appendix F. Mr.
Natan replied that it is. Mr. Shao felt the allowable for components was much lower. Mr. Natan agreed to check into it.
There being no further questions, the Chairman adjourned the meeting at 6:00 p.m.
NOTE:
A complete transcript of this meeting is on file at the NRC Public Document Room at 1717 H Street, N.W. , Washington, D.C. or can be obtained from ACE Federal Reporters, Inc. , 415 Second Street, N.E.
Washington, D.C. (202)547-6222.
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M891 NOTKES May 31.1978, to permit the acquisillon WEDNESDAY, Octonta 20,1977 a request to do so prior to the meeting of ndditional emironmental information ,m ,,g g' identifying the topics and desired pre-en the cHects of ining this mode of cool- sentation time.so that appropriate ar- i p,;t It niso conforms the license with The Subcommittee, with any of its rangementa can be mnde. The Subcom.
carlier nettons taken by New !!ampshire consultants who may be present. si:1 mittee will receive ornl statements on and Vermont. rncet in Executive Session to explore topics relevant to als purview nt an ap.
The application for the nmendment their preliminary opinions rerarding piopriate time chosen by the Chairman, complies with the standards and require- matters which should be considered m (c) Further information regarding ments of the Atomic Energy Act of 1054. order to formulate a report and recom- topics to be discussed. whether the meet-e amended (the Act), and the Commis- mendations to the full Commtttee. ing has been enneelled or rescheduled, non's rules and regulations. The Com- ,y gm g g33 the Chairman's ruhng on requests for the ,
mwlon has made appropriate findings opportunity to present oral statements I ns required by the Act and the Commis. The Subcommittee will meet to hear *and t.he tiric allotted therefor can be non's rules and regulations in 10 CIR presentations by representatives of the obtained by a prepaid telephone call on Chapter I, which are set forth in the 11- NRC Staff, the Babcock and Wticox October 25. 1977 to the OfDee of the !
cense amendment. Prior public notice of Company, Combustion Engineering, and Executive Director of the Committee i tius amendment was not required since their consultants, and will hold discus- (telephone 202-634-1319, Attn: Dr. l the amendment does not involve a signi- sions with these groups pertinent to this Richard P. Savio) between 8:15 a.rn, and '
ficant hazards consideration. review. 5:00 p.m., EST.
The Commission has prepared an en- At the conclusion of this session, the (d) Questions may be asked only by vironmentalimpact appraisal for the re- Subcommittee may caucus to determine members of the Subcommittee, its con- i vised Technical Specifications and has whether the matters identified in the sultants, and the Staff.
concluded that an environmentalimpact initial session have been adequately cov- (e) The use of still, motion pleture, statement for this particular action is ered. and television cameras, the physical in-not warranted because th*re will be no Practical considerations may dictate stallation and presence of which will not significant environmentalimpact attrio- alterations in the above agenda or sched. interfere with the conduct of the meet-utable to the action other than that ule. The Chairman of the Subcommittee ing, will be permded both before and which has already been predicted and is empowered to conduct the meeting in after the meeting and during any recess.
described in the Commission's Final En. a manner that. In his judgment wilj The use of such equipment will be allowed vironmental Statement for the iacility. facilitate the orderly conduct of busi. while the meeting is in session at the dis- j For further detatls with respect to this ness, including provisions to carry over cretion of the Chairman to a degree that action see (1) the application dated Au- an uncompleted open session from or.e is not disruptive to the meeting. When gust 8,1977, (2) Amendment No. 38 to day to the next. use of such equipment is -permitted, ap-License No. DPR-28, and (3) the Com- The Advisory Committee on Reactor propriate measures will be taken to pro. l mission's Environmental Impact Ap* Safeguards is an independent group es. 'tect proprietary or privileged informa-praisal. All of these items are avadable tablished by Congress to review and re. tion which may be in documents, folders, for public inspection at the Commission's port on each application for a construe- etc. being used during the meeting. Re-Public Document Room.1717 H Street tion permit and ou each application for cordings will be permitted only during NW.. Washington, D.C. and at the an operating license for a reactor facu. those sessions of the meeting when a Brooks Memortal Library,224 Main St., ity and on certain other nuclear safety transcript is being kept.
Brattleboro, Vt. A copy of items (2) and matters. The Committee's reports become (f) A copy of the transcript of the (3) may be obtamed upon request ad* a part of the public record. Although portion (s) of the meeting where factual dressed to it e U.S. Nuclear Regulatory ACRS meetings are ordinarily open to information is presented and a copy of Commission, Washington, D.C. 20555. the public and provide for oral or wrn- the minutes of the meeting will be avail-Attention: Director, Division of Operat- ten statements to be considered as a part able for inspection on or af ter 1Novem-ing Reactors. of the Committee's information gather- ber 2,1977 and January 26,1978, respec-I Publje Document Dated at Bethesda, Maryland, this ing procedure concerning the health and safety of the public, they are not adju-
', "t 37[e 30th day of September 1977. D.C. 20555*
dicatory type hearings such as are con.
For the Nuclear Regulatory Commis- ducted by the Nuclear Regulatory Com. Copies may be obtained upon payment sion. mission's Atomic Safety & IJeensing of appropriate charges.
MoRToN B. FAIRmt, Board as part of the Commission's 11-Acffng Chic /, Operating Reac* Dated: October 4,1977*
censing process. ACRS meetings do not tors Branch No. 4, Dirision of .normally treat matters pertaining to Jouw C. Honr, Operating Reactors, environmental impacts outside the ra. Adefsory Committee -
[rn Doc.77-29600 Filed 101-77,8 35 am1 diological safety area. mnagement OAer.
With respect to public participation in f ra Doc.77-296H Filed 14-7-77;8:45 Am) the meeting, the following requirements
[7590-01] shan apply:
(R) Persons wishing to submit written [7590-01] !
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS. SUBCOM MITTEE ON statements regarding the agenda may dc (Docket No. 6c-341 A1 !
FLUID / HYDRAULIC DYNAMIC EFFECTS so by providing 15 copies to the Subcom-mittee at the beginning of the meeting. THE DET8OIT EDISON CO.* ET AL Notice of Meeting Comments should be limited to safety Receipt of Attomey General's Advice and In accordance with the purposes of related areas within the Committee's Time for Filing of Petitions to intervene !
Sections 29 and 182b. of the Atomic En. purview. on Antitrust Matters ergy Act (42 U.S.C. 2039, 2232b.), the Persons desiring to mail written com. The Commission has received, pursu- i ACRS Subcommittee on Fluid /Hydraulle ments may do so by sending a readily ant to section 105c of the Atomic Energy !
Dynamic Eff ects will hold an open meet- reproducible copy thereof in time fer Ac of 1954, as amended, the following I ing on October 26, 1977 at the Rodeway consideration at this meeting. Comments additional advice from the Attorney Gen. l Inn. 7101 NE. 82d Ave., Portland Oreg. postmarked no later than October 19' eral of the United States, dated Sep- I 97220. The purpose of this meeting is 1977, addressed to Dr. Richard P. Savio, tember 30,1971' to continue discussion on the effects of ACRS, NRC, Washington, D C. 20555, '
blowdown forces on reactor vessel sup- will normally be received in time to be pursua You have requested our further advice considered at this meeting. nt to section lose of the Atomic l porta. Energy Act of 1054, as amended, uth respect The acenda for subject meeting shall (b) Persons desiring to make an oral to the above-cited application. By letter to l be as follows: statement at the meeting shouhl make you dated August to,1971, we rendered ad- !
notaAt trotstER, VOL. 42, NO.196-TUf 50AY, OC70 sit 11,197y I