ML19352A716
| ML19352A716 | |
| Person / Time | |
|---|---|
| Site: | Hartsville  |
| Issue date: | 03/07/1980 |
| From: | Hanauer S NRC - TMI-2 UNRESOLVED SAFETY ISSUES TASK FORCE |
| To: | Plesset M Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML18025B195 | List:
|
| References | |
| FOIA-80-587, RTR-NUREG-0460, RTR-NUREG-460 NUDOCS 8104170477 | |
| Download: ML19352A716 (126) | |
Text
{{#Wiki_filter:-<e. a e pn ato ((},, s.,cf',g (o, UNITED STATES NUCLEAR REGULATORY COMMISSION g k(*. /s. [p/ E WASHINGTON, D. C. 20555 Sg, I MAR 7 C MEMORANDUM FOR: Dr. Plessett, Chaiman, ACRS FROM: S. H. Hanauer, Director Unresolved Safety Issues Program
SUBJECT:
DRAFT REPORT REGARDING ANTICIPATED TRANSIENTS WIDiOUT SCRAM I am delivering to you today twenty-one copies of a near final draft of NUREG-0460, Volume 4 " Anticipated Transients Without Scram for Light Water Reactors". This staff report will be issued in a few days at which time I will replace these copies with more elegant ones. I am giving you these drafts today, as we agreed, so that ACRS members and consultants can have plenty of time before the Subcommittee meeting scheduled for March 26. _. j 1 h S. H. Hanauer, Director Unresolved Safety Issues Program cc: H. Denton E. Case A. Thadani M. Singh 8104170%k
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o Mechanical Engineering Branch - ATWS Evaluation "Early Verification" i Information - Alternate 3 Plants BWR 4/5/6 We have reviewed the information provided by General Electric relative to mechanical component structural integrity and operability in topical reports NEDE-24222 Volume 1 dated May,1979 and NEDE-24222 Volume 2 dated December,1979. As discussed in these reports, the largest loads are imposed on Reactor Coolant System Components for the MSIV Closure Event. This event results in the highest Reactor Coolant System Pressures. Additionally, and not well described in these reports, some components in the reactor coolant pressure boundary and some components associated with safety systems that take suction or discharge from the suppression pool are also exposed to large vibratory loads resulting from the discharge of safety / relief valves to the pool. The high ATWS pressure loading occurs simultaneously with the safety / relief valve vibratory loads. Thus the i structural integrity and operability of these mechanical components must be evaluated for this combined loading environment. In Appendix A.3 of Volume 2 of the report G.E. has provided a fairly detailed tabulation of the maximum calculated reactor r aolant system pressures for many locations within the reactor coolant system for the " worst case" MSIV Closure Event. f l l
2-For G.E. supplied components at these various locations, a tabulation of ASME Code Level C Allowable pressures was provided. In general the Level C pressures were determined in a very conservative manner. The component design pressure was simply multiplied by the ratio of the applicable Level C stress criterion to the component design stress criterion,-(L:'!:' " 't). In some cases the " allowable pressures" were based on hydrostatic tests performed on the component in accordance with the requirements of the ASME Code or the similar requirements of some other standard as may be applicable for a specific component. Based on the tabulated results, it would appear that for most components the maximum calculated system pressure resulting from the MSIV Closure ATWS event is within no higher than, at the most, slightly above 10% over the component design pressure, considerably below the reported Level C allowable pressure. If the system static pressure resulting from the ATWS event were the only load to which the component is subjected, it could easily be concluded that the stresses in reactor coolant system componei.ir, supplied by G.E., were quite low under ATWS conditions, i.e. as noted above considerably below ASME Level C, in fact in most. cases close to Code Level B. However the reported results, both in Appendix A.3 and also to some extent, as qualitatively discussed in Sections 4 and 5 of Volume 2 of the report, are not adequate because they do not report or discuss component stress levels or operability capability for the combined loading case of the ATWS pressure load combined with'the SRV vibratory load. ,--m y --e
. In verbal discussions, G.E. has been attempting to provide justification that the existing " design basis" non-ATWS SRV loads, which inv 1 e lower system pressures and fewer SRV valve actuations and must be calculated using a "more conservative" methodology than is required for calculating ATWS SRV loads somehow envelope any SRV loads associated with ATWS. To date G.E. has not provided technical justification for this position. Although at least for the Reactor Coolant System Components 'abdressed in the Vol. 2 report there appears to bc some margin between the stress level resulting from the ATWS pressure alone and the Level C Code criteria. Until further information is provided by G.E. or Licensees and Applicants regarding the combined effects of ATWS pressure and SRV vibratory loads, we cannot complete an evaluation of BWR 4/5/6 component integrity and operability. Other than this basic problem with the G.E. Alternata 3 plant information, there are a few other areas where documentation has not been provided. Areas where additional information is required:
- 1) As noted above, no information has been provided regarding component structural integrity / operability for the combined loading case of the ATWS pressure plus associated SRV vibratory loads.
- 2) No information has been provided regarding structural integrity or operability of any B0P supplied components affected by ATWS loads.
For BOP supplied piping exposed to SRV vibratory loads, the evaluation
. of the effect of these loads on piping functional capability should be evaluated.
- 3) Although believed to be an oversight in the drafting of NEDE-24222, no specific confirmation of the Level C " allowable" pressure has been provided for any of the BWR 4/5/6 reactor vessels.
e o d i , _. _, _. _ - _,,, _.. - _. -, ~._.. - - ---, _ -,
Mechanical Engineering Branch Evaluation "Early Verification" Information - Alternate 3 Plants BWR/3 We have reviewed General Electric Topical Report NEDE-24223 foe information relative to component structural integrity and operability. Under the " worst case" ATWS event, the MSIV Closure, Reactor Coolant System Components are exposed to high system pressures. Additionally, Reactor Coolant System Components and some components associated with safety systems that take suction or discharge from the suppression are exposed to large vibratory loads resulting from the discharge of safety / relief valves to the pool. The only information proivded in NEDE-24223, is that Reactor Presscre Vessel Emergency (we presume Code Level C) limit is 1500 psi. No information has been provided regarding any other G.E. or BOP supplied components for either pressure capability alone or the effect or' the combined effects of the ATWS and the SRV vibratory loads. Either G.E. or each BWR/3 Licensee should be required to provi'de assurance of component structural integrity / operability for the combined loading case of the ATWS pressure in combination with the SRV vibratory loads. l
i ATWS ~ Mechanical Engineering Branch Evaluation (Preliminary) - B&W Alternate 3 RCPB Components - Structural Integrity / Operability We have reviewed the information relative to mechanical component structural integrity and operability supplied by B&W in the January,1980 draft report entitled " Analysis of B&W NSSS Response to ATWS Events". It is our under-standing that a final version of this report will be submitted to NRC the week of February 4, 1980. As discussed in the report the B&W Evaluation of Alternate 3 components, performed only for components within the B&W-scope of supply, fs not complete. The objective e Evaluation apparently is to demonstrate component integrity and operability for ATWS pressures of up to 4000 psi. From reviewing the referenced draft report, we conclude that there is a large amount of analysis still to be compl eted. By letter dated January 22, 1980 from J. Taylor to S. Hanauer, B&W has advised that a schedule for submittal of this on-going work is being developed. Although the Alternate 3 component evaluations are not complete, the results of " final" evaluations for a few components and the results of partial evaluations for some other components are discussed in the draft report. Our evaluation of the reported results is as follows.
. For major RCPB components such as reactor vessels, steam generators, pressurizei., and Reactor Coolant Primar Piping, B&W reports that they have utilized a simple but conservative calculational procedure to determine the minimum that the components could be exposed to without exceeding the ASME Code Level C Stress Criterion. The component design pressure, 2500 psi for all the components, is multiplied by the ratio of the Level C primary membrane stress limit to the Design primary membrane stress limit. For the materials of construction for B&W Alternate 3 RCPB components, the ratio of the Level C limit to the Design limit is 1.2 for austenitic stainless steel components and 1.5 for ferritic materials. The procedure thus yields minimum allowable Level C pressures of 3000 psi for the austenitic components and 3750 psi for the ferritic components. We are in agreement with B&W that the most conservative i.e. lowest Level C allowable pressures would be determined in this way. As B&W correctly poinM out, there can be other conservatisms that may have been included in the component's original design that could be determined by a more detailed review of the original design analysis or in some cases by performing a reanalyses of the component using techniques such as finite element analysis etc. that provides a more accurate under-standing of the component's true stress levels. By performing such additional in depth reviews and reanalyses, it is in many cases possible to show that a particular component can be exposed to much higher pressures without exceeding, the Level C stress limit, than would be indicated by using the simple stress l limit ratio method discussed above.
. In the draft report, B&W indicates that by more in depth review of existing design analyses it has been able to determine that a " majority" of components can be exposed to pressures of up to 4000 psi without exceeding the Level C limit. The report does not contain any detailed information about the reviews that were performed and how cor.clusions were reached from the existing analyses. For example it is not obvious that many of these components could be exposed to pressures as much as 1000 psi above the minimum level C value, calculated by the stress limit ratio method, and still in reality not be exposed to stresses above Level C. We assume that more detailed information about these "in depth" reviews aad the conclusions B&W has reached from them will be provided in a future submittal. Assuming that the B&W conclusions about component pressure capability is correct, the results would seem to indicate that many if not most B&W scope of supply i RCPB components are designed with margins of safety considerably higher than 1 required by the ASME Code. We believe that this factor would have to be specifically addressed in any final staff evaluation of the on-going B&W reviews. B&W further notes that for those portions of the major components, previously referred to, where review of existing design analyses did not indicate sufficient capability to withstand 4000 psi, without exceeding Code Level l C, additional analyses were being performed to attempt to verify such capability. ( l s, _
. Since B&W's discussion of " major components" included the reactor vessel and the pressurizer, but was not sufficiently detailed in nature to address specific areas of these components, there are two areas of concern regarding these vessels which we feel should be mentioned here and would have to specifically be discussed as a part of any final evaluation on B&W Alternate 3 plant capability. In 1977 B&W submitted component integrity information in an ATWS topical report, BAW 10099 Rev. 1. In that report B&W reported that analyses performed at that time to evaluate the pressure retention capability of B&W NSSS components indicated the pressurizer heater tubes would fail (pressure boundary collapse) at 3230 psi.. It was also reported that reactor coolant leakage through the reactor vessel closure gaskets would commence at 3250 psi for 177 FA plant reactor vessels and 3300 psi for 205 FA plant vessels. As was clearly shown during the recent, shutdown of the Three Mile Island 2 reactor, the operability of the pressurizer heaters may be needed to assure safe shutdown of the plant. Unless new information to the contrary is available, it would appear that the pressurizer heaters presently utilized in Alternate 3 plants cannot be expected to function after exposure to the ATWS pressure that is being calculated for these plants. Heater replacement with higher pressure capability heaters or lower system pressures based on this concern alone should be evaluated when determining the ATWS " fixes" for the B&W Alternate 3 plants. l
. As noted above, B&W had also reported reactor vessel closure leakage at pressures of 3250-3300 psi. We believe that at these pressures, leakage would literally be quite small 1.e. possibly drops per minute etc. and involve no gross loss of reactor coolant and complete lifting off of the closure head as has been reported by one other vendor at somewhat higher pressures. However, B&W has indicated as an objective to demonstrate component structural integrity at a pressure of 4000 psi. Certainly at 4000 psi, possible head lifting, deformations, gross closure leakage etc. become items of serious concern ar:d adequate assurance would have to be provided that any such phenomena would have no effect on safe shutdown. The draft report also provides some information on other B&W supplied Alternate 3 components other than those referred to thus far as " major components". Some very preliminary information is provided relative to analyses that are not complete for the reactor coolant pumps. A qualitative description is provided as to how each of the different pumps utilized in B&W Alternate 3 plants operates and generally at what physical location on the pump the highest stressed location has been determined to be. A table is provided that indicates how much of the pump ".acea", a term that is not defined, is exposed to stresses higher than Level C for both 3500 and 4000 psi. We have evaluated the brief description of the basic evaluation procedure that B&W reports is being utilized to determine pump acceptability. The only critical comment we have is that based on the information provided it is not clear that the high stressed areas and their associated deformations are being evaluated for their effect on pump operability 1.e. it may be necessary to run those pumps after the ATWS Pressure surge, to assure safe
. shutdown. From the report it appears, that other concerns of basic structural integrity, " leakage rates" (presumably seal integrity) are being evaluated. The draft report also includes some information apparently in " final" form with regard to RCPB valve pressure retention capability for those within the B&W scope of supply. It describes fairly extensive cataloguing effort whereby technical information for all of the valves supplied by B&W has been tabulated in a series of tahles included in the report. The information provided includes the manufacturer of each valve, the Cod; required hydrostatic test pressure for the valve, and a listing of the Ccde allowable stress values (design values) as given for each of the material types used in the valves, as specified in the 1965,1968,1971, and 1974 ASME Code editions for both the hydrostatic test temperature of 100 F and the ATWS temperature of 670 F. Apparently the Code editions used for this evaluation encompass design time period applicable to valves for Alternate 3 plants. With the exception of the safety and relief valves, the discussion of the data indicate that of all the B&W RCpB valves, the lowest Code hydrostatic test pressure used to qualify any one valve was 5400 psi which was applied at a temperature of 100 F. {
. 0 B&W has determined the ratio of the Code allowable stress value at 100 F to that at 670 F for all of the tabulated stress val es as noted above. Of all the ratios thus determined, the most conservative ratio 1.e. largest numerically, indicating greatest decrease in material strength from 100 F to 670 F,1.248 in this case, was used to conservatively calculate an " equivalent hydrostatic pressure" for the ATWS temperature of 670 F. The " equivalent" 670 F pressure thus calculated using the lowest hydrostatic pressure of 5400 and the above referenced material property reduction ratio 1.e. dividing 5400 by 1.248, yields a pressure slightly over 4300 psi. Although when using this pressure, there is no way to evaluate the exact state of stress within the valve at 4300 psi against a specific criterion such as the Level C stress limit, we are in agreement with B&W as to the basic structural integrity up to 4300 psi provided the valve is hydrotested in the same configuration that would be exposed to the ATWS pressure. It involves extrapolation of known valve-pressure containing capability data at 0 one temperature (100 F) to that at the ATWS temperature of 670 F. We believe this procedure can be used to provide adequate assurance at 670 F of the rather basic parameter, pressure retention capability without failure; that tne hydrotest already has demonstrated at a higher pressure at 100 F, as long as the Code recognized reduction in material strength with increasing temperature is adequately or conservatively taken into account. B&W has used the most conservative f.e. greatest reduction in strength with temperature l of any of the applicable materials, provided in any one of the editions of the Code applicable to the time period during which valves for Alternate 3 plants were designed. l
. We noted above one significant prerequisite for utilizing hydrotest data to provide assurance of pressure retention capability under ATWS conditions. Namely the configuration of the component as exposed to the hydrotest pressure must be the same configuration that would be exposed to the ATWS pressure. For example for most of the valves tabulated by B&W and for which B&W desires to use hydrotest data to demonstrate pressure retention capability, again excluding safety and relief valves, the standard industry practice would be to expose the valve body to the hydrotest pressure with the disc at least partially, if not all the way open. Such a test is valid for expolation to ATWS temperatures only for the valve body. Thus for valves that would be open during the ATWS pressure surge, extrapolation of such test data could be used to assure that the valve body would not be expected to pressure retention capability up to an ATWS pressure extrapolated per the procedure described above. Typically after the valve bodies have been tested as described in the preceding paragraph, the valves are tested with the disc in place at some lower pressure. In the draft report B&W, except for the safety and relief valves, does not l l discuss the position of the valve disc for the hydrotest. l l In the conclusion of the valve section of the draft report, they have stated that the hydrotest pressure extrapolation procedure provides assurance not only of valve body integrity under ATWS pressures but also of disc integrity. Conclusions about disc integrity are not acceptable in view of what has been stated about the standard industry procedures for testing such valves unless B&W can provide substantiation that these Alternate 3 plant valves were i
. tested to a sufficiently high pressure with the disc in the configuration it would be expected to be in if exposed to an ATWS pressure surge. We find the information provided relative to safety and relief valve pressure retention capability completely unacceptable. The draft report states the safety and relief valves of the type used in Alternate 3 plants have been exposed to very high hydrotest pressures, some as high as 9000 psi. B&W then wishes to apply a similar extropolation procedure to these valves to demonstrate basic pressure retention capability at ATWS pressures. However, for these valves B&W has attached to the draft report copies of portions of valve manufacturers hydrotest procedures which apparently were used in hydrotesting the Alternate 3 plant safety and relief valves. The procedures indicate that only portions of each of these valve types are exposed to such high pressures. The highest pressure used to test n assembled _ valve is 3750 which performed at 100 F cannot be extrapolated to an equivalent pressure l l at the ATWS temperature which would be high enough to provide assurance of valve structural integrity. Additionally, both safety and relief valves are exposed to large dynamic loads associated with the discharge of subcooled l liquid and two phase flow. For this reason extrapolation of hydrotest l l data even if it were available, for high pressures, would not provide assurance of safety and relief valve integrity. l A summary of specific areas where information has not been provided for l Alternate 3 plant components follows:
- 1) All B0P Supplied RCPB Components.
l
10 -
- 2) Additional on-going analyses for major components, reactor coolant pumps, end valve operability analyses.
- 3) Safety and Relief Valves and Associated Discharge Piping - Information in draft report regarding structural integrity of the valves is unacceptable.
No information was provided on safety and relief valve operability or structural integrity and functional capability of discharge piping. All of this information may be dependent on results of EPRI valve test ~ program in response to NUREG-0578. 4) Instrumentation - No information provided on pressure capability of primary system instrumentation.
- 5) Pressurizer heater operability and reactor vessel closure leakage must be specifically addressed as stated in the text above.
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Mechanical Engineering Branch Evaluation "Early Verification" Information - Alternate 3 Plants Combustion Engineering We have reviewed Combustion Engineering Topical Report CENPD-263-P for information relative to mechanical component structural integrity and operability. The report indicates that for the " worst case" ATWS transient i.e. Loss of Feedwater resulting in highest reactor coolant system pressure, the ATWS pressure, depending on the plant type, ranges from 3800-4300 psi. The report describes the results of analyses that C.E. has performed for some Alternate 3 plant reactor coolant system components within the C.E. scope of supply. C.E. has provided a general description of the numbers and types of components which were evaluated for the ATWS pressure loading environment. It appears from the description that some types of components were evaluated for each of a selected list of Alternate 3 plants and others chosen on the basis of being representative of one of three classes of Alternate 3 plants, i.e. 2560, 3410, or 3800 MWT. We find that it is not clear from the descriptions provided in the report exqctly which components have been evaluated and which have not. Additionally, C.E. has described a finite element analysis performed to determine the response of the reactor vessel closure to the ATWS pressure and has taken credit for leakage through the calculated flange opening for determining their peak ATWS pressures. For the components that have been evaluated, the report provides a brief general description of each of several methods used to evaluate component
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u 2-stress and in a few cases deformation level. The report states that methods used included scaling of originally performed design stress analysis results and elastic or inelastic finite element analysis for some components. With a few exceptions, it is not clear from the report exactly which method of analysis was used to determine the stress or deformation level for a carticular component. The report notes that of the components evaluated, t.even were found to exceed ASME Code level C by varying amounts, and thus have at least poe; ions of the component pressure boundary exposed to stresses above the material yield strength. For a couple of these high stressed components i.e. reactor coolant pumps and pressurizer heater elements, component operability may be required for safe shutdown of the plant. With the exception of a 12" surge line piping elbow there is little or no description of how these components were analyzed, whether deformations were evaluated for effect on operability, etc. At 4300 psi some of these components would probably experience a significant amount of permanent deformation, the effects of which should be evaluated using inelastic analysis methods of analysis. The inelastic analysis method recognizes the nonlinearities in the relation-ship of stress to strain in the material and computes the component structural behavior under the known loading environment considering the strain I i hardening characteristics of the actual component material (s), its permanent l deformations, and stress redistribution occurring in the component. l l l l l l
. C.E. has stated that they perfomed finite element inelastic analyses for some components, which are exposed to stresses above the material yield strength and also for the analyses performed to evaluate the amount of reactor vessel flange o-ring leakage. Our specific comments regarding the vessel flange analyses are provided separately below. However, regarding finite element inelastic analyses in general, we have the following remarks. The accuracy of the results are quite dependent on the characteristics of the model that is utilized for the analysis and a correct utilization of what is usually a very complex computer code which has the capability to take into account the changing material properties referred to above. This type of analysis and the computer codes used to perform finite element analyses using the method' have not been in widespread use for very many years, i.e. the technology is relatively new. Because of the many complexities involved in performing these analyses and the fact that because of its high cost it is only used in high stress / strain critical applications where it is less expensive to perform a high cost analysis to demonstrate adequacy than to replace equipment. We require that such analyses be reviewed in a fair amount of depth. For some applications, it is also preferable to have an independent analysis or a test performed where possible to confirm the result. It is to this and, namely to enable us to perform a detailed review, that the February 15, 1979 questions contained requests for specific information to be supplied when finite element inelastic analysis was used. The report is not responsive to the February 15 questions in that it does not contain sufficient detailed information to enable us to evaluate the adequacy of the analyses that were performed. A few components were specifically addressed in the report and.for these we have the following comments. The report states that a finite element analysis was performed for one retivel6 inch shutdown cooling isolation valve. It is stated that the analysis indicated that no plastiscity occurred in the valve body and no adverse effect on operability would be expected. The only " design" information provided about the valve analyzed, other than its nominal size and type, i.e.16 inch gate, is that the valve body is thicker than the pipe to which it is attached. The report then concludes that similar results, i.e. no adverse effect on operability, would be expected to occur for other active valves of all sizes and types provided that the valve body is thicker than the pipe to which it is attached. In our view, C.E. has provided essentially no information about the details of the anlaysis performed for the 16 inch valve. Also the extrapolation of results to active valves of other sizes and types solely on the basis
of valve body thickness, we do not believe can be supported technically, even if C.E. were to provide more detailed information on the one valve analysis. The information requested in the February 15, 1979 letter to the vendors for each type and size of active valve, possibly supplemented with test data, if available, is needed to verify operability In summary, C.E. has apparently performed evaluations for some reactor coolant system components within the C.E. scope of supply. However, as a result of the extremely brief summary format utilized in the report for 5 describing the analyses and the "results" there is not sufficient information available for us to meaningfully evaluate what has been done. Specific additional documentation requirements are listed below. A few comments should be made regarding the description in the report of the analyses, that were performed to determine the response of the reactor vessel closure to the ATWS pressure surge so that credit could be taken for the pressure mitigation effect of fluid leakage past the vessel gaskets. In addition to reviewing the information in the report we have also had verbal discussions with C.E. to obtain further clarification as to what was done. Parts of the analysis is considered proprietary by C.E. i Consequently, only a few concerns are noted here:
- 1) Finite element inelastic analysis methods, were used. Comments as
( made above for information requirements when this method of analysis is used are applicable. l l t .., -.. ~., -, - .a 2) It is our understanding that two " representative" vessel closures were analyzed to " represent" at least three different vessel designs. Insufficient information is available to evaluate the validity of this approach. 3) It is not clear to us that accepted tolerances on the amount of preload applied to the reactor vessel bolts has been conservatively taken into account. This is a significant parameter which can greatly affect the system pressure at which closure leakage would begin, and of course could thus greatly affect the final maximum system ATWS pressure. 4) It does not appear that the analysis has taken into account possible deformations and movements of the closure head dome itself. The control rod drive penetration housings are installed into the vessel head dom'e with relatively small partial penetration welds which by ASME Code rules are not to be exposed to any bending moment type loading. Assurance must be provided that at these high stress levels, closure head movement or deformation will not impose a severe enough moment loading on one or more of the control rod drive mechanism penetrations such that combined with the high vessel pressure housing to dome weld failure would result.
- 5) There is also a concern raised by an NRC consultant that with head lifting and leakage past both closure gaskets, that the head may " cycle" i.e.,
continue to alternately lift and reseat somewhat analagous to " chattering" of a safety valve. Assurance would have to be provided that this
D . would not occur. It would result in large dynamic loads being imposed on the cic5ure bolts and also fluid leakage would be much lower than assumed resulting in higher than calculated maximum system pressures. A sumnary listing of areas where adequate information has not been provided follows:
- 1) No information has been provided for B0P supplied components. Additionally it is not clear which C.E. supplied components have been evaluated versus those that have not.
2) It is not clear in all cases exactly where finite element elastic or inelastic analysis has been applied. The information requested in the February 15, 1979 letter to vendors is considered the minimum necessary to perform a review. For finite element inelastic analysis, information needs were further discussed above.
- 3) Vessel Closure Analyses - This analysis is the foundation on which the entire Alternate 3 evaluation for ATWS is based. As such, it must be thoroughly understood and demonstrated to have been performed in a conservative, not a realistic manner.
It is not completely clear, at this writing, exactly how much additional documentation or analysis would be required for us to find the concept acceptable from a regulatory perspective. As a minimum, the concerns noted in the text above must be satisfied, including the documentation mentioned for finite element analyses in the February 15, 1979 letter. It is also fairly obvious l. that for such a critical application an independent confirmatory analysis
. would be invaluable input to reaching a final decision.
- 4) Structural Integrity & Operability of Active Valves - Technical justification must be provided for each size and design, not arbitrary extrapolations from a single analysis.
- 5) Letdown HX Piping - The report indicates that this piping may fail.
Documentation indicating that the dynamic effects of this failure have been evaluated and shown to have no effect on safe shutdown has not been provided.
- 6) Pressurizer Safety Valves, Relief Valves (PORV's), and associated Discharge Piping - The information provided on safety valve structural integrity is lacking in detail as per comments made above for components where finite element analysis is used.
Information on Safety Valve operability is qualitative and probably somewhat speculative. Assurance of operability will probably have to wait for results of EPRI Safety and Relief Valve Test Program to be completed by July,1981 in response to Short Term Lessons Learned per NUREG-0578, C.E. has not provided l any information on structural integrity or operability of PORV's. I Additionally, no information has been provided on the structural integrity and functional capability of safety and relief valve discharge piping. It is expected that assurance of PORV operability and discharge piping integrity and functional capability will also be dependent on results of/!EPRI test program. ~_
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- 7) Non-Active Valves - The brief evaluation description provided appears to indicate that for these valves on extrapolation of capability was also made based on the one 16" active valve that was analyzed.
Sur.h approach may be technically feasible where operability is net a concern. However, the information in the report is too brief to provide the required technical justification for the validity of the extrapolation. 8) Instrumentation - Information presented is fairly detailed and indicates many " typical" instruments probably would need upgrading for Alternate 3 plants. It would appear that this equipment capability ultimately would have to be addressed on a plant specific basis by each Applicant or Licensee. t I
s ATWS Mechanical Engineering Branch Evaluation - W - Alt. 3 RCPB Components - Structural Integrity / Operability We have reviewed the information supplied by )[ in both the June,1979 and December, 1979 submittals. In these submittals h[ has reported that Reactor Coolant Pressure Boundary Components in the }[ scope of supply can be exposed to an ATWS pressure of 3200 psi without exceeding the ASME Code Level C Service Limit. In general based upon the information provided, we are in agreement with the )[ conclusions. For a few }[ supplied components, however, as specifically noted below, we have concluded that more documentation is required so that we can evaluate the conservatism of the analyses performed by h[ to arrive at the reported 3200 psi pressure. For some major pressure vessel components i.e. reactor vessels, steam generators, and pressurizers, }[ has stated that it has reviewed all of the designs used in Alternate 3 plants and has performed analyses to determine the pressure at which the Level C stress criteria are reached. 1 In the December,1979 submittal }[ has stated that it is reporting from its review and analysis, the allowable pressure for the " limiting" highest stress areas of these components, those with lowest allowable pressure at Code Level C Service Limit. As a conservatism, h[ has performed the evaluation for these vessels for an assumed component average temperature
. of 700 F. This is approximately 100 F higher than the maximum average vessel temperature that would occur at the time when the system pressure is at a maximum for the highest pressure ATWS event. For materials such as those used in a h[ Alternate 3 plant reactor coolant system component, the Code allowable stress limits decrease with increasing temperature. Determining a maximum allowable. pressure assuming a conservatively high material temperature provides additional assurance that the pressures reported by h[ are conservative. It is possible to calculate Level C allowable pressures using the following simple, but conservative, method. The ratio of the Level C limit to the design limi t f.c. t'e Ccde Leve' ^ ' 4 4 +, can be determined. This ratio, always larger than one, when multiplied by the component design pressure provides a conservative value for the Level C allowable pressure. Applying this conservative method and utilizing the information supplied by W'with regard to material types, highest stressed areas etc. for the " limiting" vessels we have, with the exception of the 4 loop reactor vessel closure bolts calculated that these vessels can be exposed to a minimum ATWS pressure of 3000 psi without exceeding the Level C Service limit. The analytical results reported by h[ indicate that the most " limiting" component for Level C allowable pressure is one of the reactor vessel designs used in h[ 4 loop plants i.e. nozzle safe ends and closure bol ts. Taking into account the fact that the h[ analyses were performed assuming very conservative material properties i.e. the vessel material 0 assumed to be in equilibrium at 700 F, and the fact that we do not have access to the specific dimensions for each of the h[ Alt. 3 plant reactor
. vessel safe ends and closure bolt configurations, in order to obtain additional assurance of the reported allowable Level C pressures for the reactor vessel safe ends and closure bolts, we request that each Applicant or Licensee confim the applicability of the Level C allcwable pressure as reported by E for his particular plant and the specific basis for that confir=ation i.e. safe end wall thickness etc. Based upon the evaluation made by 1 and our understanding of the nomal industry design and :.anufacturing practice for RCFB cceponents, we believe that no Alternate 3 E plant Licensee or Applicant will have any difficulty providing confir ation for the 3200 psi allowable pressure for reactor vessels, steam generators, and pressurizers. In the Dececber,1979 submittal, E has reported an allcnable Level C pressure of 3474 psi for Control Rod Drive Mechanis~s on the basis of the results of an " elastic stress analysis." As for the vessel nozzle safe ends, Licensees and Applicants should confim this and provide sufficient details of the analysis to specifically clarify the basis for the reported pressure. With regard to Reactor Coolant System Piping in the E Scope of Supply for Alternate 3 plants, E has reported in both the June and Dececber,1979 submittals that the ASME Code Level C Service Li=it allcwable pressure is in excess of 3700 psi. Based upon our understanding of the caterials used for E supplied RCFB pioing, we are in agreement that the Level C acceptable pressure for E supplied RCFB pioing is considerably above the previously renticned allowable pressure of 3200 psi re:orted for scr e other RCPS c:r ponents. E has reported a Level C allowable pressure sccewnat higner than 3200 psi for this ciping
. based upon the ASME Level C criterion which permits such piping to reach pressures of 1.5 times the design pressure under certain loading conditions. While we have not specifically attempted to confirm the allowable piping pressure reported by W i.e. above 3700 psi, we are in agreement that somewhere above 3200 psi 'n acceptable without exceeding Level C for the loading condition resulting from at ATWS event. With regard to RCPB valves within the W scope of supply for Alternate 3 plants, W has reported in the December,1979 submittal that all ASME Section III valves used in such plants have been hydrostatically tested at 100 F, as required by the Code, at 5625 psig. Taking into account the fact that the valves are subjected to temperatures of around 550 F during an ATWS event and recognizing the change in Code allowable material properties from 100 F to 550 F, W has determined an equivalent allowable pressure for the valves of 4725 psi, considerably in excess of 3200 psi. Based upon the fact that test results are usually preferable to analytical ~ results, when available, for component qualification and that the change in Code allowable material properties is based on considerable testing, we have concluded that the structural capability of W supplied RCPB valves has been adequately verified for at least the 3200 psi generically discussed in the June and December,1979 submittals for the Alternate 3 W NSSS components. With regard to the W supplied Reactor Coolant Pumps utilized in W Alternate 3 Plants, W has reported in the December, 1979 submittal that based upon a finite element analysis which has been perfomed, the R.C. pump used for
. Alternate 3 plants can withstand a pressure of 3231 psig from an ATWS event without exceeding.the Level C Service Limit. Based on verbal discussions that we have had with W regarding the type of analysis perfomed etc., we have concluded that the " allowable" pressure reported by }[ is acceptable. However, as noted below confirmatory documentation is required to confim these verbal discussions of analyses which have been perfomed. The structural integrity of the pressurizer safety valves and power operated relief valves have been confimed, at least for the static maximum pressure conditions associated with an ATWS event, in much the same way as described above for other W supplied valves. The structural integrity and operability of these valves under the dynamic fluid relieving conditions associated with an ATWS event is further addressed below. The operability and functional capability of a few RCPB components is essen-tial to obtaining ultimate safe shutdown after the ATWS event. In the December,1979 submittal, W did address the operability of W supplied ASME Section III isolation valves which must be operable so that such systems as RHR, HPCI etc. can function after the exposure to the peak ATWS system pressure when RCP pressures have returned to system design pressure or lower. For such valves W has stated that qualification tests are perfomed to pressurize the valves with disk in place at over 3700 psi at ambient (100 F) l temperature. Taking into account the change in valve material properties l l with temperature, this is equivalent to hydrostatic testing the valve at about
. 3150 psi at 550 F, the temperature at which the valve would be exposed to the maximum ATNS pressure. h[ has stated that the disks have routinely been exercised after the hydrostatic testing. We are in agreement and find acceptable the reported testing as justification of operability for such valves up to pressures of 3151 psi at 550 F. For some ATWS events, it may be necessary to be able to run the reactor coolant pumps to assure safe shutdown. In the December,1979 submittal, h[ has described the results of finite element analyses etc. used to check pump clearances etc. to evaluate the effect of ti.e ATWS pressure and any resulting deformations on pump operability. }[ has concluded that after exposure to a pressure of over 3200 psi, sufficient clearances etc. are maintained so that operability would not be impaired. As reported below, we require additional information on the details of this pump analysis to be able to evaluate the h[ findings. It may be necessary after an ATWS event to utilize the pressurizer heaters to assist in safe shutdown of the plant. In the December,1979 submittal h[ has specifically stated that the heater tubing integrity and the integrity of the pressurizer bottom head to heater welds have been evaluated and have been found to meet the Code Level C Service Limit. In the submittal, }[ has not specifically addressed the operability of the heaters. Based on. independent discussions that we have had with suppliers of these heaters, we believe that they probably would be functional at some pressure above 3200 psi. However, we require that Applicants and Licensees provide confirmation of thic.
. Based upon our review of the June and December,1979 h[ submittals intended to address structural integrity and/or operability of Alternate 3 plant Reactor Coolant System Components, we have reached the above conclusions, qualified in some cases with expressed need for confirmatory documentation. Additionally, there are some components for which considerably more documentation is required to verify component structural and/or functional adequacy. Listed below are those areas where additional' documentation is needed to confirm conclusions reported in the June or December,1979 submittals whether or not previously noted above.
- 1) Each Applicant or Licensee shall confirm the applicability of the h[ reported Level C allowable 3200 psi pressure for his plant.
The basis for the allowable reactor vessel safe end and closure bolt allowable pressures shall be specifically described in detail.
- 2) Each Applicant, Licensee or h[ shall provide detailed description of the Finite element analysis performed to determine Reactor Coolant pump structural integrity and operability. Additionally, a more detailed description of the elastic analysis performed for determining the Level C allowable pressure for the Control Rod Drive Mechanisms should be provided.
(Ref. the Feb. 15, 1979 questions).
- 3) The information presented regarding the operability of pressurizer safety valves and power operated relief valves and the structural and functional integrity of safety and relief valve discharge piping is inadequate. Each Applicant and Licensee shall establish the
9 operabil ty of these valves for ATWS conditions and similarly the integrity and functional capability of the associated discharge piping. It is expected that these areas cannot be adequately addressed until sometime after the completion of the EPRI Safety and Relief Valve and Piping Test Program about to'be undertaken in response to the Short Term Lesson Learned Recommendations contained in NUREG-0578. 4) Instrumentation - Structural Integrity of Primary Reactor Coolant Pressure Boundary Instrumentation in the h[ scope of supply necessary for Safe Shutdown has only been addressed in a cursory manner and functional capability of such instrumentation has not been addressed at all. Much more detailed information is required to complete our evaluation.
- 5) B0P Supplied Equipment - Neither the June nor December,1979 submittals address structural integrity and/or operability of any equipment outside the h[ scope of supply.
- 6) The December,1979 submittal addresses the structural integrity and operability aspects of h[ Supplied Reactor Coolant Pressure l
Boundary valves that were manufactured to ASME Section III require-ments. There are undoubtedly some Alternate 3 plants which were constructed prior to inclusion of valve design requirements in ASME Section III. Each Licensee or Applicant shall either confirm for its plant (s) the applicability of the Section III valve informa-
i !- i tion as described by W in that submittal or shall provide assurance of structural integrity and operability for non-ASME Section III W supplied valves.
- 7) Each Licensee and Applicant shall provide confirmation of pressurizer heater operability after exposure to a minimum ATWS pressure of 3200 psi.
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[7 Mechanical Engineering Branch ,ii .,c- ~ / Evaluation "Early Verification" Information - Alternate 3 Plants m a /, Combustior. Engineering f'- a. o We have reviewed Combustion Engineering Topical Report CENPD-263-P for information relative to mechanical component structural integrity and operability. The report indicates that for the " worst case" ATWS transic; c i.e. Loss of Feedwater resulting in highest reactor cooler.t system pressure, the ATdS pressure, depending on the plant type, ranges from 3800-4300 psi. The recart describes the results of analyses that C.E. has perfomed #cr some Alternate 3 plant reactor coolant system components within the C.E. scope of supply. C.E. has provided a general descripion of the numbers and types of components which were evaluated for the ATWS pressure loading environment. It appears from the description that some types of components were evaluated for each of a selected list of Alternate 3 plants and others chosen on the basis of being representative of one o' three classes of Alterrate 3 plants, i.e. 2560, 3410, or 3800 MWT. We find that it is not clear from the descriptions provided in the report exactly which components have beer evaluated ar.:: which nase not. Additionally, C.E. has described a finite element analysis performed to determine the response of the reactor vessel closure to the ATWS pressure and has taken credit for leat. age through the calculated flange o;.er.ing for deterai. ir.; tief e peak ATdS pressures. For the components that have been evaluated, the recort orovides a brief general description of each of several methods used to evaluate component
. stress and in a few cases deformation level. The report states that methods used included scaling of originally performed design stress analysis results and elastic or inelastic finite element analysis for some components. With a few exceptions, it is not clear from the report exactly which method of analysis was used to determine the stress or deformation level for a particular component. The report nctes that cf the compor.ents evaluated, seven were found to exceed ASME Code level C by varying arctnfs. ana thas have at least portions of the component pressure boundary exposed to stresses above the material yield strength. For a couple of these high stressed components i.e. reactor coolant pumps and pressurizer heater elements, component operability r;av be required for safe shutdown o' the plant. With the exception of a 12" surge line piping elbow there is little or no description of how these components were analyzed, whether deformations w ere evaluated for effect on operability, etc. At 4300 psi some of these ccr;oonents would probably exper:erce a sign?f t: ant amount of permanent deformation, the effects of which should be evaluated using inelastic analysis methods of analysis. l The inelastic analysis method recognizes the nonlinearities in the relation-i ship c' stress to st ain in the material and computes the component structural behavior under the known loading envir nment considerine the strain l i l hardening characteristics of tne actual coop:ntnt materialit }, its cercar.ent f defornations, and stress redistribution occurring i:1 the om7enent. l l l
. C.E. has stated that they performed finite element inelastic analyses for some components, which are exposed to stresses above the material yield strength and also for the analyses performed to evaluate the amount of reactor vessel flange o-ring leakage. Our specific comments regarding the vessel fiange analyses are provided separately below. However, regarding finite element inelastic analyses in general, we have the following remarks. The accuracy of the results are quite dependent on the characteristics of the model that is utilized for the analysis and a correct utilization of what is usually a very complex computer code which has the capability to take into account the changing material properties referred to above. This type of analysis and the computer codes used to perform finite element analyses using the method have not been in widespread use for very many years, i.e. the technology is relatively new. Because of the many complexities involved in performing tScse analyses and the fact that because of its high cost it is only used ir high' stress / strain critical applications where it is less expensive to perform a high cost analysis to demonstrate adequacy than to replace equipment. We require l that such analyses be reviewed in a fair amount of depth. For some l applications, it is also preferable to have an independent ar.alysis or a test pe-formed where possible to confirm the result.
. It is to this ant', namely to enable us to perform a detailed review, that the February 15, 1979 questions contained requests for specific information to be supplied when finite element inelastic analysis was used. The report is not responsive to the February 15 questions in that it does not contain sufficient detailed information to enable us to evaluate the adequacy cf + naly ses that were performed. A few components were specifically addressed in the report ard for these .. t 52.ve the follo tir g comertt. The report states that a finite element analysis was performed for one active 16 inch sht.tdcwn coalir.g isolation valve. It is statsd thtt the ant'y:is indicated that no plastiscity occurred in the valve body and no adverse effect on operability would be expected. The only " design" information provided about the valve analyzed, other than its nominal size and type, i.e.16 inch gate, is that the valve body is thicker than the pipe to which it i:. attached. -'le r port then concludes that similar rest.lt:, i.e. no ad erse effect on operability, would be expected to occur for other active valves of :1 sizes and t pes provided th.t t7e valve body 's thickcr tna'i the pipe to which it is attached. In our siew, C.E. hrs provided essentially na information about tne da vis .s e. Alsc the extr y '3Eic9 c ' ;! e an W s,'eri: r. c fo.- tne 16 L '1 of results to active valves of other sizes and types solely on the baris i
. of valve body thickness, we do not believe can be supported technically, even if C.E. were to provide more detailed information on the one valve analysis. The information requested in the February 15, 1979 letter to the vendors for each type and size of active valve, possibly supplemented with test data, if available, is needed to verify operability. In summary, C.E. has apparently performed evaluations for come reacter coolant system components within the C.E. scope of suppl H34 eve, as a result of the ertremely brief summary format utilized in the report for descricing the analyses and the "results" there is not sufficient informatior available for us to meaningfully evaluate what has been done. Specific additional dccurentation recuirerents are listed below. A few comments should be made regarding the description in the report of the analyses, that were performed to determine tne response of the reactor vessel closure to the ATWS pressure surge so that credit could be taken for t's pressure mitigation effect of fluid leatage pest tha ves el gasre_tr In addition to reviewing the information in the report we have also hac verbal discussions with C.E. t: ottain further clarificrtion as te what was done. Parts of the analysis is considerec proprietary by C.E. Co,se:uently, or.ly a 'ex concerns are noted here: 1, fr':e eiers't i.lastic analysis methofs, ware uter. Cora_.ts as made above for information recuirements when this retrod ci ar.alysis is used are applicable.
. 2) It is our understanding that two " representative" vessel closures were analyzed to " represent" at least three different vessel designs. Insufficient information is available to evaluate the validity of this approach.
- 3) It is not clear to us that accepted tolerances on the amount of preload applied to the reactor vessel bolts has teen conservat vely t i.en ir.to i
account. This is a significant parameter which can creatly affect the system pressure at which closure leakege would begin, ard of course could thus greatly affect the f*:nal maximum system ATWS pressure.
- 4) It does not appear that the analysis has taken into account possible diforJ,ations and covements of the closure head dome it. elf. The control rod drive penetration housings are installed into the vessel head dome witn relatively small partial penetration welds which by ASF.E Code rules are not to be exposed to any bending moment type loading. Assurance nust ce pravidad that at these hi.-5 stress leve'3, c1:t;~- baad e7ie art or deformation will not impose a severe enough moment loading on one or more cf the control od dri' n nec5srism 7tnetratf-- sect f rit
- ombined with the high vessel pressure housing to dome weld failure would result.
5) ' ere is a'so a ccacern r-ised by :r. NRC cc7n ite -t wite hced i 1 fting anc ieakage past botn closure gaskets, ;nat tra need may ":ycie" 1.e., continue to alternately lift and reseat somewhat analagous to " chattering" af a caf: y n?/e. 's2.:rar.ce u. 1 ' a.- tc te J. c < i e. d at tr is d
. would not occur. It would result in large dynamic loads being imposed on the closure bolts and also fluid leakage would be.much lower than assumed resulting in higher than calculated maximum system pressures. A summary listing of areas where adequate information has not been provided follows:
- 1) "' informatio has been provided for B0P supplied componerts. Additionally it is not clear which C.E. supplied components have been evaluated versus those that ha,e not.
2) It is r.ot clea-i;. ali cases exactly where fir.ite elemer.t elait,c or inelastic analysis has been applied. The information requested in the February 15, 1979 letter to vendors is considered the minimum necessary to perform a review. For finite element inelastic analysis, information needs were further discussed above.
- 3) Vessel Closure Analyses - This analysis is the foundation on which the entire Altcrrate 3 evaluation for ATWS is based. As such, it must be thoroughly understood and demonstrated to have been performed in a corse vative, not a rea'istic maar.er.
It is not completelf clear, at this writing, exactly how much additional documentation or analysis wculc be reagired for us to fir.d the ccncept acceptable frcm a ; agulatory ptrspectise. Ao n'n: rum, the coace-r.: noted in the ter: crove nust be satisfied, including the documentation mentioned for finite eiererit ?,31yses in toe February 15, 1979 letter. It is also fairly obizious . hat or sic) a critical applica tion en iridepardean cor.t i: matory analys.i r r
. would be invaluable input to reaching a final decision.
- 4) Structural Integrity & Operability of Active Valves - Technical justification must be provided for each size and design, not arbitrary extrapolations from a single analysis.
- 5) Letdown HX Piping - The report indicates that this piping may fail.
Documentation indicating that the dynamic effects of th:s failure have been evaluated and shown to have no effect on safe shutdown has not been provided.
- 6) Pressurizer Safety Valves, Relief Valves (PORV's), and associated Discharge Pipinc - The information provided on safety v11ve structural integrity is lacking in detail as per comments made above for components where finite element analysis is used.
Information on Safety Valve operability is qualitative and probably somewhat speculative. Assurance of operability will probably have to wait for results of EPRI Safety and Relief Valve Test Program to be completed by July,1981 in response to Short Term Letsons Learned per NUREG-0578 C.E. has not provided any information on structural integrity or operability of PORV's. Additionally, no information has been provided on the structural integrity and functional capability of safety and relief valve discharge piping. It is expected that assurance of PORV operability and discharge piping integrity and functional capability will 'also be dependent on results of. EPP.I test progran. l l l
. 7) Non-Active Valves - The brief evaluation description provided appears to indicete that for these valves on extrapolation of capability was also made based on the one 16" active valve that was analyzed. Such approach may be technically feasible where orerability is nct a concern. However, the information in the report is too brief to provide the required technical justification for the validity of the extrapolation. c) .ist umentatic" - Information presented :s fairly detai!ec and indicates many " typical" instruments probably would need upgrading for Alternate 3 pl a nt s. It vould 4ppear that this equipment capatility ultimatelj would hava t-5r -Md essed cn a piar.' specific be: 4s by each topiitant or Licensee. l .,. _ _ _. -,,...,,. _,. ~ 4 r
W 4 ) J. L ), .. l 7 4 7 r-N/ f l.!? / .) Mechanical Engineering Branch - ATWS Evaluation "Early Verification" Information - Alternate 3 Plants BWR 4/5/6 'We have reviewed the information provided by Ge'neral Electric relative to mechanical component structural integrity and operability in topical reports NEDE-24222 Volume 1 dated May,1979 and NEDE-24222 Volume 2 dated December, 1979. As discussed in these reports, the kargest loads are imposed on Reactor Coolant System Components for the MSIV Closure Event. This event results in the highest Reactor Coolant System Pressures. Additionally, and not well described in these reports, some components in the reactor coolant pressure boundary and some components associated with safety systems that take suction or discharge from the suppression pool are also exposed to large vibratory loads resulting from the discharge of safety / relief valves to the pool. The high ATWS pressure loading occurs simultaneously with the safety /re11af valve vibratory loads. Thus the structural integrity and operability of these mechanical components must be evaluated for this combined loading environment. In Appendix A.3 of Volume 2 of the report G.E. has provided a fairly detailed tabulation of the maximum calculated reactor coolant system pressures for many locations within the reactor coolant system for the " worst case" MSIV Closure Event.
. For G.E. supplied components at these various locations, a tabulation of ASME Code Level C Allowable pressures was provided. In general the Level C pressures were determined in a very conservative manner. The component design pressure was simply multiplied by the ratio of the applicable Level C stress criterion to the component design stre'ss criterion 'L.;l ' '+). 5g ...___________a e, g In some cases the " allowable pressures" were based on hydrostatic tests performed on the component in accordance with the requirements of the ASME Code or the similar requirements of some other standard as may be applicable for a specific component. Based on the tabulated results, it would appear that for most components the maximum calculated system pressure resulting from the MSIV Closure ATWS event is within no higher than, at the most, slightly above 10% over the component dcsign pressure, considerably below the reported Level C allowable pressure. If the system. static pressure resulting from the ATWS event were the only load to which the component is subjected, it could easily be concluded that the stresses in reactor coolant system components, supplied by G.E., were quite low under ATWS conditions, i.e. as noted above considerably below ASME Level C, in fact in most cases close to Code Level B. However the reported results, both in Appendix A.3 and also to some extent, as qualitatively discussed in Sections 4 and 5 of Volume 2 of the report, are not adequate because they do not report or discuss component stress j levels or operability capability for the combined loading case of the ATWS pressure load combined with the SRV vibratory load. l t
.. In verbal discussions, G.E. has been attempting to provide justification that the existing " design basis" non-ATWS SRV loads, which inv 1 e lower system pressures and fewer SRV valve actuations and rnust be caNulated using a "more conservative" methodology than is required for calculating lh p;* 5 ATWS SRV loads somehow envelope any SRV loads associated with ATWS. f i/ To date G.E. has not provided technical justification for this position, kh,l i g$ rY 0 $q.V Although at least for the Reactor Coolant System Components addressed in the gs o dp* cv Vol. 2 report there appears to be some margin between the stress level y resulting from the ATWS pressure alone and the Level C Code criteria. Until 4 further information is provided by G.E. or Licensees and Applicants regarding the combined effects of ATWS pressure and SRV vibratory loads, we cannot complete an evaluation of BWR 4/5/6 component integrity and operability. Other than this basic problem with the G.E. Alternate 3 plant information, there are a few other areas where documentation has not been provided. Areas where additional information is required:
- 1) As noted above, no information has been provided regarding component structural integrity / operability for the combined loading case of the ATWS pressure plus associated SRV vibratory loads.
2) No information has been provided regarding structural integrity or operability of any BOP supplied components affected by ATWS loads. For BOP supplied piping exposed to SRV vibratory loads, the evaluation
4 of the effect of these loads on piping functional capability should be evaluated.
- 3) Although believed to be an oversight in the drafting of NEDE-24222 no specific confirrnation of the Level C " allowable" pressure has been I
provided for any of the BWR 4/5/6 reactor vessels. i 1 l n l ..----.n, .--n,.,--,,n-- e.-,--. ,,-n.--.r-,.-
w - a. Mechanical Enaineering Branch Evaluation "Early Verification" Information - Alternate 3 Plants BWR/3 ^ We have reviewed General Electric Topical Report NEDE-24223 for information relative to component structural integrity and operability. Under the " worst case" ATWS event, the MSIV Closure, Reactor Coolant System Components are exposed to high' system pressures. Additionally, Reactor Coolant System Components and some components associated with safety systems that take suction or discharge from the suppression are exposed to large vibratory loads resulting from the discharge of safety / relief valves to the pool. The only information proivded in NEDE-24223, is that Reacter Pressure Vessel Emergency (we presume Code Level C) limit is 1500 psi. No information has been provided regarding any other G.E. or B0P supplied components for either pressure capability alone or the effect of the combined effects of the ATWS and the SRV vibratory loads. Either G.E. or each BWR/3 Licensee should be required to provide assurance of component structural integrity / operability for the combined loading case of the AfWS pi assure in conbination with the I SRV vibratory loads, i l c--
2-of 700 F. This is approximately 100 F higher than the =aximum average vessel temperature that would occur at the time when the systes pressure is at a maximum for the highest pressure ATd5 event. For raterials such as those used in a E Alternate 3 plant reactor coolant system component, the Code allowable stress ifmits decrease with increasing temperature. Deter =ining a maximm allowable pressure assuming a conservatively high nsterial temperature provides additional assurance that the pressures reported by y are conservative. It is possible to calculate Level C allowable pressures using the follcwing si=ple, but conservative, method. The ratio of the Level C Ifmit to the 4 9 5- ' Resign limit i.e.
- M N 1 w '. * r't, can be deter =ined. This ratio, always larger than one, when multiplied by the component design pressure provides a conservative value for the Level C allowable pressure.
Applying this conservative method and utilizing the infor=ation supplied by E with regard to material types, highest stressed areas etc. for the " limiting" vessels we have, with the exception of the 4 loop reactor vessel closure bolts calculated that these vessels can be exposed to a mini =um ATd5 pressure of 3000 psi without exceeding tre Levei C 5ervice limit. The analytical results reported by E indicate that the most " limiting" component for Level C allowable pressure is one of the reactor vessel designs used in E 4 loop plants i.e'. nozzle safe ends and closure belts. Taking into accour.t the fact tnat the E aulyses were :erforned assuming very conservative material properties i.e. the vessel caterial 0 assuned to be in equilibrium et 700 F, and the fact that we do not have access to the specific dimensions for each of the E Alt. 3 plant reactor t
. vessel safe ends and closure bolt cor. figurations, in order to obtain additional assurance of the reported allowable Level C pressures for the reactor vessel safe ends and closure bolts, we request that each Applicant or Licensee confirm the applicability of the Level C allowable pressure as reportad by E-for his particular plant and the specific basis for that confimation i.e. safe end wall thickness etc. Based upon the evaluation made by E and our understanding of the nomal industry desigr. and r.anufacturing practice for RCPB components, we believe that no Alternate 3 y plant Licensee or Applicant will have any diffi:ulty providing confimation for the 3200 psi allowable pressure for reactor vessels, steam generators, and pressurizers. In the December,1979 submittal, E has reported an allowable Level C pressure of 3474 psi for Control Rod Drive Mechanisms on the basis of the results of an " elastic stress analysis." As for the vessel nozzle safe ends, Licensees and Applicants should confim this and. provide sufficient details of the aralysis to specifically clarify the basis for the reported pressure. With regard to Reactor Coolant System Piping in the E Scope of Supply for Alternate 3 plants, y has reported in both the June and December,1979 submittals that the ASME Code Level C Service Limit allowable pressure is 1 in excess of 3700 psi. l l Based upon our understanding of the materials used for y supplied RCPB piping, we are in agreement that the Level C acceptable pressure for E supplied RCPB piping is considerably above the previously mentioned allowable pressure of 3200 psi reported for some other RCPB components. E has reported I a Level C allowable pressure somewhat higher than 3200 psi for this piping
4-based upon the ASME Level C criterion which permits such piping to reach pressures of 1.5 times the design pressure under certain loading conditions. While we have not specifically attempted to confim the allowable piping pressure reported by W i.e. above 3700 psi, we are in agreement that somewhere above 3200 psi is acceptable without exceeding Level C for the loading condition resulting fmm at ATWS event. With regard to RCPB valves within the W scope of supply for Alternate 3 plants, W has reported in the December,1979 submittal that all ASME Section III valves used in such plants have been hydrostatically tested at 100 F, as required by the Code, at 5625 psig. Taking into account the fact that the valves are subjected to temperatures of around 550 F during an ATWS event and recognizing the change in Code allowable material U properties from 100 F to 550 F. W has detemined an equivalent allowable pressure for the valves of 4725 psi, considerably in excess of 3200 psi. Based upon the fact that test results are usually preferable to analytical results, when available, for component qualification and that the change in Code allowable material properties is based on considerable testing, we ha/e concluded that the structural capability of W sup; lied RCPB valves has been adequately verified for at least the 3200 psi generically discussed in the June and December,1979 submittals for the Alternate 3 W NSSS components. With regard to the W supplied Reactor Coolant Pumps utilized in W Alternate 3 Plants W has reported in the December, 1979 submittal that based upon a finite element analysis which has been perfomed, the R.C. pump used for m ~
. Alternate 3 plants can withstand a pressure of 3231 psig from an ATWS event without exceeding the Level C Service Limit. Based on verbal disccssions that we have had with )[ regarding the type of analysis performed etc., we have concluded that the " allowable" pressure reported by )[ is acceptable. However, as noted below confirmatory documentation is required to confirm these verbal discussions of analyses which have been performed. The structural integrity of the pressurizer safety valves and power operated relief valves have been confirmed, at least for the static maximum pressure conditions associated with an ATWS event, in much the same way as described above for other )[ supplied valves. The structural integrity and operability of these valves under the dynamic fluid relieving conditions associated with an ATWS event is further addressed below. The operability and functional capability of a few RCPB components is essen-tial to obtaining ultimate safe shutdown after the ATWS event. In the December,1979 submittal, )[ did address the operability of )[ supplied ASME Section III isolation valves which must be operable so that such systems as RHR, HPCI etc. can function after the exposure to the peak ATWS system pressure when RCP pressures have returned to system design pressure or lower. i For such valves }[ has stated that qualification tests are performed to 0 pressurize the valves with disk in place at over 3700 psi at ambient (100 F) l temperature. Taking into account the change in valve material properties with temperature, this is equivalent to hydrostatic testing the valve at about )_ r l
. 3150 psi at 550 F, the te=perature at which the valve would be exposed to the maximum ATd5 pressure. E has stated that the disks have routinely been exercised after the hydrostatic testing. We are in agreement and find acceptable the reported testing as justif1 cation of operability for such valves up to pressures of 3151 psi at 550 F. For some ATd5 events, it may be necessary to be able to run the reactor coolant pumps to assure safe shutdown. In the December,1979 submittal, y has described the results of finite element analyses etc. used to check pump clearances etc. to evaluate the effect of the ATdS pressure and any resulting deformations on pump operability. E has concluded that after exposure to a pressure of over 3200 psi, sufficient clearances etc. are maintained so that operability would not be tapaired. As reported below, we require additional information on the details of this pump analysis to be able to evaluate the y findings. It cay be necessary after an ATd5 event to utilize the pressurizer heaters to assist in safe shutdown of the plant. In the December,1979 submittal y has specifically stated that the heater tubing integrity and the integrity of the pressurizer bottom head to heater welds have been evaluated and have been found to meet the Code Level C Service Limit. In the submittal, y has not specifically addressed the operability of the heaters. Based on. independent discussions that we have had with suppliers of these heaters, we believe that they probably would be functional at some pressure above 3200 psi. However, we require that Applicants and Licensees provide confirmation of this.
. Based upon our review of the June and December,1979 )[ submittals intended to address structural integrity and/or operability of Alternate 3 plant Reactor Coolant System Components, we have reached the above conclusions, qualified in s'ome cases with expressed need for confirmatory documentation. Additionally, there are some components for which considerably more documentation is required to verify component " uctural End/or functional adequacy. Listed below are those areas where additional documentation is needed to confirm conclusions reported in the June or December,1979 submittals whether or not previously noted above.
- 1) Each Applicant or Licensee shall confirm the applicability of the )[ reported Level C allowable 3200 psi pressure for his plant.
The basis for the allowable reactor vessel safe end and closure bolt allowable pressures shall be specifically described in detail.
- 2) Each Applicant, Licensee or )[ shall provide detailed description of the Finite element analysis performed to determine Reactor Coolant pump structural integrity and operability. Additionally, a more detailed description of the elastic analysis performed for deternining the Level C allowable pressure for the Control Rod Drive Mechanisms should be provided.
(Ref. the Feb. 15, 1979 questions).
- 3) The information presented regarding the operability of pressurizer safety,alves and power operated relief valves and the structural and functional integrity of safety and relief valve discharge piping is inadequate. Each Applicant and Licensee shall establish the
f 8-operability to these valves for ATWS conditions and.:imilarly the integ-ity and functional capability of the associated discharge piping. It is expected that these areas cannot be adequately addressed until sometime after the completion of the EPRI Safety and Relief Valve and Piping Test Program about to be undertaken in response to the Short Term Lesson Learned Recommendations contained in NUREG-0578. s.. 4) Instrumentation - Structural Integrity of Primary Reactor Coolant Pressure Boundary Instrumentation in the E scope of supply necessary for Safe Shutdown has only been addressed in a cursory manner and functional capability of such instrumentation has not been addressed at all. Much more detailed information is required to complete our evaluation.
- 5) BOP Supplied Equipment - Neither the June nor December,1979 submittals address structural integrity and/or operability of any equipment outside the E scope of supply.
- 6) The December,1979 submittal addresses the structural integrity and operability aspects of W Supplied Reactor Coolant Pressure Boundary valves that were manufactured to ASME Section III require-ment-There are undoubtedly some Alternate 3 plants which were constructed prior to inclusion of vahe design requirements in ASME Section III. Each Licensee or Applicant shall either confirm for its plant (s) the applicability of the Section III valve informa-
9-tion as described by 1 in that submittal or shall provide assurance of structural integrity and operability for non-ASME Section III y supplied valves.
- 7) Each Licensee and Applicant shall provide confirmation of pressurizer heater operability after exposure to a minimum ATWS pressure of 3200 psi.
4 t w.,---,-. ,..-r ---y---r- .,-.~.--..y.+.*_.- ,,,,-wm-,,----,m,-- .,r c
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l. . ). l. '.l g f fl ,._~. /. f c -H e r
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.gl cf t 7. f, y.; ! Jc. s *l J - y a -<- ..v, -r., g-Mechanical Engineering Branch -: ' r L. Evaluation - W - Alt. 3 RCPB .1 ,.., i Components - Structural Integrity / Operability We have reviewed the information supplied by y in both the June,1979 and / ~f December,1979 submittals. In these submittals y has reported that Reactor Coolant Pressure Boundary Components in the y scope of supply can be exposed to an AWS pressure of 3200 psi without exceeding the ASME Code Level C Service Limit. In general based upon the information provided, we are in agreement with the 1 conclusions. For a few g supplied components, however, as specifically noted below, we have cor;cluded that more documentation is required so that we can evaluate the conservatism of the analyses performed by E to arrive at the reported 3200 psi pressure. For some major pressure vessel components i.e. reactor vessels, steam generators, and pressurizers, y has stated that it has reviewed all of f the designs used in Alternate 3 plants and has performed analyses to I determine the pressure at which the Level C stress criteria are reached. In the December,1979 submittal y has stated that it is reporting from its review and analysis, the allowable pressure for the " limiting" highest stress areas of these components, those with lowest allowable pressure at Code Level C Service Limit. As a conservatism, y has performed the i evaluation for these vessels for an assumed component average temperature
9 477. July 5,1980 to All BWR Operating Station Superintendents from F. P. Felini
SUBJECT:
IE BULLETIN 80-17 -- FAILURE OF CONTROL RODS TO INSERT DURING A SCRAM AT A BWR (8 pages) 478. July 3,1980 Memorandum for Paul Check from James W. Pittman
Subject:
CONCLUSIONS ABOUT THE BROWNS FERRY INCIDENT OF 6/28/80 (4 pages) 479. Brown's Ferry Partial Scram - RSB Comments (4 pages) 480. Handwritten Notes, Important Characteristics of BFNP Event Consideration (7 pages) 481. MSIV Transient with Partial Scram and Recirculation Pump Trip (RPT) (2 pages) 482. October 10, 1979 Daily Report-RIII (1 page) 483. December 1,1980 Memorandum for G. C. Lainas, Thomas Novak and Robert L Tedesco from Paul S. Check
Subject:
BWR SCRAM DISCHARGE SYSTEM SAFETY EVALUATION, attaching Generic Safety Evaluation Report, BWR Scram Discharge System (214 pages) 484 e em r 80 Memorandum for William Dircks from John Ahearne
Subject:
ATWS L_
25 485. December 10, '980 Memorandum for R. F. Fraley from Gary G. Zech
Subject:
NRR RESPONSE TO ACRS RECOMMENDATIONS FOLLOWING THE 245TH ACRS MEETING, attaching October 22, 1980 Memorandum for Raymond F. Fraley from Paul S. Check
Subject:
RESPONSE TO ACRS QUESTIONS ON SCRAM SYSTEM MALFUNCTIONS, August 29, 1980 letter to Paul S. Check from R. H. Buchholz
Subject:
PARTIAL SCRAM
O Y /o L 5 m/ 7 l 1 l
/1,.....k UNITED STATES I 'er g NUCLEAR REGULATORY COMMISSION ji -l WASMtNGTON. D. C. 20555 %.v / ..a December 1, 1980 MEMORANDUM FOR: Gus C. Lainas, Assistant Director for Safety Assessment, 00L Thomas H. Novak, Assistant Director for Operating Rectors, DOL Robert L. Tedesco, Assistant Director for Licensing, DOL FROM: Paul S. Check Assistant Director for Plant Systems DSI
SUBJECT:
SkR SCRAM DISCHARGE SYSfEM SAFETY EVALUATION The enclosed report suarnarizes the results of our review and evaluation of the BkR scram discharge system. The report deals with Browns Ferry 3 i l partial scram event of June 28, 1980; subsequent investigations, tests, and analyses involving a number of operating BWRs; and failures of the scram level instruments at Brunswick and Hatch plants. Our review has considered licensee response to IE Bulletin requirements and the SWR l Owners' Group proposed criteria for scram discharge volume (SDV) designs. The report specifies acceptable bases for continued BWR plant operation and provides design criteria for the SDV system. The findings of this l report should be transmitted to BWR licensees and applicants for imlemen-tation. Also included as an appendix to the SER are plant-specific evalu-ations for each operating BWR. These evaluations provide the basis for continued operation while needed permanent modifications are being designed and i@lemented which will probably take about two years. Our review has identified two additional requirements beyond those proposed by the Owners Group. The first addresses the potential for fast fill of the SDV on decaying air system pressure. An automatic air header dum will i be required to initiate control rod insertion on low pressure in the con-l trol air header. This should prevent loss of scram function daring certain I low probability loss of air pressure events. This requirement is appli-caole to all licensees identified in Table 1 of the SER. It should De installed within about two months. Meanwhile the operator action to scram on low air pressu e alarm, backed up by the same action on rod drift alarms l and other indications, is acceptable. The second added requirement deals with the SDV level instrumentation and addresses potential corron-cause f ailures. It is described in Section 4.2.2.3 of the SER. This requirement together with the criteria for scram system design provide an acceptable basis for scram discharge system design, (new plants) ano design modifications (operating plants). 1 e\\hx e\\ q by 'l' , ~..
a . December 1, 1980 We have suggested acceptable ways of conplying with the requirerents and criteria. If a licensee or applicant chooses to employ these approved means, no fur +.her review by NRR is needed. Under separate cover, I will shortly transmit this SER to Dave Waters, Chairman of the BWR Owners Group. b 4 Paul S. Check Assistant Director for Plant Systems Division of Systens Integration
Enclosure:
As stated cc: D. Ross ACRS (3) D. Eisenhut A. Bates E. Jordan D. Zukor S. Hanauer F. Schroeder T. Ippolito R. Reid D. Crutchfield G. Lainas F. Coffman S. Rubin G. Lanik J. Huang D. Thatcher J. Wilson R. Youngblood l W. Mills J. Hannon T. Speis T. Loomis M. Mendenca V. Panciera M. Goocman i R. Satterfield A. Thomas A. Thadani W. Miners C. Graves J. Stolz N. Wagner
Contact:
V. Panciera x28164 L
GENERIC SAFETY EVALUATION REPORT BWR SCRM1 DISCHARGE SYSTEM CEC EER 1, 1980 $l b\\/ \\i NO 0 l
[p eergjo, UNITED STATES Action: NRR/SD .y {,g NUCLEAR REGULATORY COMMISSION Cys: Dircks Cornell s* l wAsHmoTom, o. c. aosas Rehm t
- +, '
/e December 10, 1980 Shapar Minogue CHAIRMAN ggpng,.g /AThadini,NRR WMinners,NRR El-Zeftawy, 50 MEMORANDUM FOR: William Dircks, Executive Director for Operations FROM: John Ahearne
SUBJECT:
ATWS RULE In order to address some issues I believe are necessary, please be prepared to discuss the attached alternative form of the ATWS rule at the December 18 meeting. It includes the following positions: 1. Alternate 2A For old plants under SEP - there should not be much argument on this. 2. Alternate 2A and Mitigation Systems Circuitry For Westinghouse operating plants and new plants. 3. Alternate 2B or plant specific analysis to demonstrate acceptable consequences for B&W and CE PWR's. 4. Alternate 2B For all BWR's. 5. In acceptance criteria, limiting to ASME Level C stress limits is probably too tight. I would suggest Level D plus strain limit on components (pumps and valves), a demonstration of their operability. In addition, please address hew special consideration could be given for plants near population centers.
Attachment:
1 As stated cc: Comissioner Gilinsky Commissioner Hendrie Comissioner Bradford SECY l OPE NRR SD ) R::c'd O'T. C - bS cau.. EC....en Mj'" T Tims...d. ]. s. ...,x.x .u... CY 9
g , /) + e, J / fiEMORANDUM FOR: R. F. Fraley, Executive Director Advisory Committee on Reactor Safeguards FR0": Gary G. Zech, Chief Technical Support Branch Planning f. Program Analysis Staff, NRR SUSJECT: NRR RESPONSE TO ACRS RECOMMENDATIONS FOLLOWING THE 245TH ACRS MEETING ~ In response to your memorandum of November 24, 1980 which sumarized Actions, Agreements, Assignments and Requests identified during the 245th ACRS meeting, the following is provided on those items assigned to NRR. 1 and 2 By memo dated October 22,1930 (copy attached), the staff provided the additional infor; nation requested by Dr. Okrent regarding (1) ATWS and BWR Scram System, and (2) GE's position regarding the use of the Automatic Depressurization System during an ATWS event. (60-09-03-001) and (80-09-04-Q01). .c Gary 3. Zech, Chief Technical Support Branch Planning & Program Analysis Staff, NRR i
Attachment:
As stated Dis tributicn_ N Central Files SS N TSB R/F GZech i ey HThompson {gyggp SCavanaugh ED0-9892 T [] ~ Q [d a 'f', on GErtter (ED0 9892 g{ PPAS I 7 Eisenhut "jjl h Grimes Hanauer U, Q/' ))77 NRR/TS / -NRR/fPAS/D ,,,,c, g), suaNavc k..GZech,;pa HL - on oarc),.12/.N./80 12 y /80 N^:c Fo AM 318 (9-76, NaCV C240 D U.S. Gov t p Nr.*ENT P AINTING OF FACE: 1979-239 369
- ...=.,,,
c. UNITED STATES NUCLEAR REGULATORY COMMISSION i, - E nasmNoTos.o c rosss 's.,,*' / OCT 2 21980 MEMORAZLH FOR: Raymond F. Fraley, Executive Director Advisory Coccittee on Reactor Safeguards FROM: Paul S. Check, Assistant Director for Plant Systems Division of Systems Integration
SUBJECT:
RESPONSE TO ACRS QUESTIONS ON SCRAM SYSTEM MALFUNCTIONS I am hereby transmitting GE's answers to ouestions on scram system failures. This fulfills a promise we made to the Concittee on September 6 during a discussion of the Browns Ferry event. i J Paul S. Check, Assistant Director for Plant Systems Division of Systems Integration
Enclosures:
As Stated cc: D. F. Ross, Jr. E. Jordan T. McCreless l T. Speis 'G. Fazetis V. Panciera A. Bates C. Graves B. Sheron W. Mills
Contact:
Farv Mendonca, NRR 49-27462 p ,., / ~ (a L, w\\ 2 0 l (j
t 0 ENCLOSURE 1 [ M i f3 L3 C L E A ll P O W t. H C Y O T I'. M O D I V i s t u t4 GLil8h% f.Lcf.TP.1C COMPAN*/,17s C1F.lfRR AVA. SAN JO3tl, CAUFOUNIA Sil:M. ~ MC 682, (408) 92b-3G22 t.ugust 29,19so U.S. Nuclear Regulatory Cornission Office of Huclear fleettor Rcgulation Vashington, DC 2055b Attention: Paul S. Check Gentlemen:
SUBJECT:
PARIIA1. SCRol A55E55HEtiT GENfRIC EVAt0ATIO!! ~ !!cference: Letter from Check, P. S., to G. G. Sherwood, dated 8/1/80 This letter provides the information on partial scracs requested by the NRC in the reference letter. General Electric believes se.veral of the NRC requested assur.ptlons used in this essessment are overly conservative and not appropriate extensions of the' observed partial scram. For exampic, the response of the operator at Ero,as Terry Unit 3 (Br3) on June 20, 1930, dec.onstrates the ability to take nu=erous actions in a sietter of a few minutes if a partial scram occurs. Also, extendiu0 the analysis to an isolation frorr. full power and in cumbinatiori with a postulated failure of one half nf the control reds is unrealistically conservative. Our analyses, which are based un the BW2/4 ev.11uated in the ATWS report (NEOF.-24222), indicate that for the most s4 vere fu11 power transient plus a DF3. type partial scram the suppre'sien pool temperature will not s exceed the NRC er,tablish suppression pool temperature limit even lf a sinolt stan@y liquid control system purrp is not started until thirty minutes into the event. The NRC also requcste.d that General Electric perform analyses of a postulate.d worst-case partial scram in which all of the rods in one-half of the core do not insert. For the ilRC requested ennditions, nur preliminary essessment is that the operator has amplo time faore than five minutes) in which to take action in order to n,eet the NRC established suppression pool temperature limit, This HRC pool limit is very conservative when cornparert in test data described in ilEDE-24222 which shows stable quencher parformance up to saturstion Leeperatures. It should be notr.d that the suppression pool temperature is not strongly dependent un SLC init!ation tirte and realistically the operator could t.ske much longer to initiate 51.C without incurring unocceptable suppre.ssion pool conditions. r, l th y p
~ i 452. Date and Subject list (2 pages) 453. Date and Subject List (2 pages) 454. Date and Subject List (3 pages) 455. Date and Subject List (4 pages) 456. May 26, 1976 letter to Robert E. Heineman from Kenneth E. Suhrke (11 pages) 457. March 8,1977 Note to S. Hanauer from A. Thadani
Subject:
DISCUSSION OF ATWS B&H ALTERNATIVE APPROACHES (22 pages) 458. March 24,1977 Letter to Dr. S. Hanauer from G. S. Lellouche (6 pages) 459. March 9, 1977 to E. G. Case from Thomas G. McCreless
Subject:
COMMENTS ON i ATWS, attaching November 26, 1976 letter to M. S. Plesset, October 28, 7974 l letter to W. R. Stratton (13 pages) 460. April 12,1977 Memorandum for P. S. Check from H. J. Richings
Subject:
PWR ATWS MTC (13 pages) i 4 61. April 29,1977 Note to Thomas M. Novak from Ashok Thadani
Subject:
TRANSIENT FREQUENCY (2 pages) 462. August 12, 1977 Note to S. Hanauer from D. F. Bunch
Subject:
ASSESSMENT OF l ATWS RADIOLOGICAL CONSEQUENCES (14 pages) 463. September 29, 1977 Memorandum for S. Hanauer from D. F. Bunch
Subject:
ASSESSMENT OF ATWS RADIOLOGICAL CONSEQUENCES (11 pages) l
n.. 24 464. November 23, 1977 Memorandum for Roger J. Mattson from Harold R. Denton
Subject:
COMMENTS ON ATWS REPORT (7 pages) 465. June 7, 1978 Routing and Transmittal Slip to A. Thadani from Del Bunch (5 pages) 466. June 13,1978 Routing and Transmittal Slip to A. Thadani from Del Bunch (5 pages) 467. June 26, 1978 Routing and Transmittal Slip to A. Thadant from Del Bunch (2 pages) 468. Date and Subject List (4 pages) 469. August 16, 1973 Note to J. M. Hendrie from A. Giambusso
Subject:
RP COMMENTS ON ATWS DRAFT (2 pages) 470. Connents on ATWS - Prepared by J. E. McEwen (2 pages) 471. November 10, 1977 Memorandum for T. M. Novak from D. F. Bunch
Subject:
ATWS DRAFT DRAFT PAPER (121 pages) 472. March 14,1978 Routing and Transmittal Slip to R. H. Vollmer from D. F. Bunch, attaching March 6,1978 Note to A. Thadani from D. F. Bunch
Subject:
HUANG COMMENTS ON ATWS, February 22, 1978 Memorandum for T. Thadani from Y. S. Huang
Subject:
COMMENTS ON TECHNICAL REPORT ON ATWS (3 pages) 473. April 20, 1978 Memorandum for H. Denton from W. Minners
Subject:
ATWS (3 pages) 474. September 15, 1978 Memorandum for Roger J. Mattson from Richard C. DeYoung
Subject:
REVIEW 0F SUPPLEMENT 10F NUREG-0460 (4 pages) 475. September 21, 1978 Memorandum for Roger J. Mattson from Anthony R. Buhl
Subject:
REVIEW 0F SUPPLEMENT 10F NUREG-0460 (2 pages) 476. October 20, 1978 Routing and Transmittal Slip to T. Novak from Gordon Chipman (5 pages)
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a-5. DATE SUBJECT 05-02-75 Info. on a Component Basis is Given in the Report Concern-Ing the Maximum Pressure that the Primary System Components of the 177 FA Plants can Withstand Without Exceeding the ASME Section III Emergency Condition Stress Limits. Provide this Info. for the 145 FA and 205 FA Plants. ~ 12-09-75 Status Report on Anticipated Transients Without Scram for Babock & Wilcox Reactors. 03-31-76 Item Section 3.1. 05-12-76 Item Section 3.2.1. 12-03-76 B&W Responses to the Evaluations of Enclosure (2) of the Agust 31 NRC Letter. (enclosure (1) Part I) 04-08-77 Preliminary Criteria for BWR ATWS Recirculation Pump Trip in the Areas of Electrical Instrumentation and Control Systems - Generic Request for Information. 05-04-77 To: S. Hanauer from H. Ornstein, B&W ATWS Meeting, March 31, 1977 w/ enclosure. 4 e i i
i DATE SUEJECl Undated T. Novak thru W. Minners from L. Olshan, Westinghouse.- Category "A" Plants ATWS Meeting with enclosures. 04-01-74 Meeting Agenda for Westinghouse ATWS Meeting. 08-02-74 D. Ross thru P. Norian from F. Odar, Attached Enclosure presuits details of loss of Teedwater.... 09-25-74 Letter to D. Vassallo from R. Salvatori, Received 9/30/74. 01-17-75 T. Novak from B. Grimes, Questions on WCAP-8330 & CENPD-158-P (ATWS) W/ Encl. 01-31-75 T. Ippolito from F. Rosa, Minutes of Westinghouse ATWS Peeting w/ encl. 02-06-75 Letter to D. Vassalo from Echeldinger, Received 2/18/75 02-12-75 B. Grimes from T. Novak, Westinghouse - ATWS 02-13-75 Applicant: Portland GE Co. (PGE), Facility: Trojan Nuclear Plant, Sumary of ATWS Meeting Held on Feb. 6,1975. 02-13-75 T. Novak from B. Grimes, Westinghouse ATWS. 02-20-75 T. Novak, from B. Grimes, Questions on Westinghouse ATWS (WCAP-9330) 02-21,75 Letter to D. Vassallo from C. Eicheldinger with Enclosures. 03-26-75 T. Novak from L. Olshan, Forthcoming Meeting with Westinghouse-ATWS "A" Category Plants. 04-22-75 Comments on Westinghouse's Letter of-Feb. 21, 1975. 04-30-75 NRC Distribution for Part 50 Docket Material from S. Burstein to B. Rusche, Ltr ref our 11-21-74 letter.... trans the following: Koshkonong 1 & 2. 05-01-75 D. Vassallo from C. Eicheldinger, WCAP-8330. 05-01-75 Letter to D. Vassallo from C. Eicheldinger w/ encl. 05-05-75 T. Novak from B. Grimes, Responses to ATWS Questions on WCAP-8330. 05-19-75 ATWS - Westinghouse, Coments on Their Answers to Ocr Questions. j 06-11-75 T. Novak from Grimes, Balance'of Responses to ATWS Questions on WCAP-833C 08-19-75 Letter to J. Bonnet from 0. Parr w/ encl. j 08-21-75 ATWS Status. 09-08-75 V. Stello from B. Grimes, Westinghouse ATWS Category B - Loss of Normal AC Power w/ encl. l i . ~.
o 3 3 ~ DATE SUBJECT 09-08-75 Letter to V. Stello from B. Grfres, Westinghouse ATWS Category 8 - Loss of Normal AC Power w/ encl. 09-08-75 ATWS - Westinghouse Station Blackout. 12-08-75 W. Kerr from R. Heineman, Status Report on Westinghouse Analyses of ATWS w/ encl. 02-02-76 R. Heineman from C. Eicheldinger, Re: Reviewed the Status Report on ATWS for Westinghouse Category B Plants. 02-13-76 T. Novak thru W. Minners from A. Thadani, Westinghouse ATWS Meeting Sumury w/encls., ATWS Meeting Sumary Distribution. 03-19-76
- 0. Ross from C. Eicheldinger, Contains P Material, Responses to Status Report Opend Items for WCAP-8330, " Anticipated Transients w/o Trip Analysis" 04-22-76 T. Novak from W. Minners, Westinghouse /ATWS Meeting Summary w/encls.
04-26-76
- 0. Ross from C. Eicheldinger, Responses to ATWS Status Report open Items for WCAP-8330.
05-21-76 Letter to A. Zechella from R. Boyd w/ enc 1. 07-06-76 Topical Report Distribution, Vendor: Combustion Engineering, "ATWS Model Modifications to STRIKIN-II" 12-30-76 NRC Distribution for Part 50 Docket Material to B. Rusche from W. Owen Letter re our 7/26/76 lt....concerning requested information related to ATWS. ~ 05-26-77 Memo for T. Novak thru G. Mazetis from A. Thadani, Westinghouse ATWS Meeting (Pittsburgh). i l I e +, _ _ _
i DATE SUEJECT 03-31-76 from Kenneth E. Suhrke to D. Ross, Partial Response to NRC Concerns Addressed in ATWS Status Report. 04-07-76 Letter to K. Duhrke from R. Heineman. 05-05-76 Letter to T. Novak from A. Thadani, Babcock & Wilcox/ATWS Meeting Summary w/enclosurds. ATWS Meeting Sumary Distribution. 05-12-76 To D. Ross from K. Suhrke, Further Respohses by B&W to Staff's 12/75 ATWS Status Report. 05-18-76 Letter to T. Novak from A. Thadani, Babcock & Wiocox ATWS Meeting Sumary, w/ enclosures. 05-26-76 Letter to R. Heineman from Power Generation Group. 05-26-76 To R. Heineman from K. Suhrke, Follow up of March 1 Technical Paper (ATWS Initail Conditions from Probabilities Analyses") Coments in April 7 Letter. 08-09-76 To T. Novak from A. Thadani (w/ encl.) ATWS Meeting Sumary Distribution. 12-30-76 To R. Boyd from J. Gilleland, Letter, re: Our 6/25/76 Letter and B&W's 12/3/76 Letter....Concerning Analyses of Anticipated Transients w/o Scram. 01-18-77 To B. Rusche from S. Jacobs, Cost Estimate for Balance of Plant Requirements in Connection with B&W NSS Changes Due to ATWS. 07-21-77 To R. Boyd from J. Taylor, Enclosed Revised B&W Positions Concerning l Items 6, 7, 8, and 12 in Section 1.6 of Report to ACRS by the Office of NRR in the Matter of SAR B-SAR-205, 7/8/77 (Draft). l l t
i ) 3 SUBJECT _ OATE Letter to T. Novak thru W. Minners from S. Salah, B&W ATWS Meeting, w/ 03-13-74 attachment. ATWS Meeting Summary,B&W, To T. Novak thru W. Minners from S. Salah w/ 03-22-74 attachment. To T. Novak from A: Thadani, Babcock and Wilcox - ATWS Meeting w/encls. 08-21-74 Letter to V. Stello from Babcock and Wilcox, Co. J. Mallay. 09-23-74 Letter to T. Novak, D. Ross, T. Ippolito, B. Grimes, J. Knight, and 11-20-74 G. Lainas from V. Stello, ATWS Schedule. 12-06-74 Letter to The Toledo Edison Company from A. Giambusso. Letter to T. Novak, D. Ross, T. Ippolito, B. Grimes, J. Knight, and 01-23-75 G. Laines from V. Stello w/ enclosure. Letter to T. Novak from B. Grimes, Questions on B&W-10099 (ATWS) w/ encl 02-20-75 02-26-75 ANSI-N661 on ATWS, Notes. Memo route slip to L. 01shan from H. Fontecilla, attached is list of 03-07-75 Input Data that we need for our analysis. 03-25-75 Summary of Meeting Held on February 27, 1975, to Discuss ATWS in TMI-2 03-28-75 V. Moore from V. Stello, Review of BAW-10099 (TAR 1338) w/ enclosure. 04-01-75 Letter to J. Mallay from A. Schwencer w/ enclosure. 04-21-75 Letter to J. Mallay from A. Schwencer.w/. enclosure. Memo route Slip to F. Schroeder, V. Stello, T. Novak, D. Ross, and 05-06-75 T. Ippolito from A. Schwencer, attached is Letter to Schwencer from K. Suhrke. Babcok and Wilcox Co. (B&W), Summary of Meeting with B&W Vendor: 06-30-75 to Discuss Topical Report No. BAW-10085 (P), " Reactor Protection System" from J. Giannelli. Letter to T. Novak from 0. 01shan, Babcock and Wilcox ATWS M u ting - 06-30-75 Category "A" Plants w/ enclosures. Undated Notes, Item Section 6.5.2. Letter to W. Kerr from R. Henineman, Strtus Report on Babcock & Wilcox 12-09-75 Analyses of Anticipated Transients without Scram. Letter to T. Novak from A. Thadani, Babcoxk & Wilcox - ATWS Meeting 01-26-76 w/ enclosure.
DATE SUBJECT 01-04-77 NRC Distribution for Part 50 Docket Material, to R. Boyd from R. Mitti Letter re our 6/24/76 ltr. and Sept.1973 report and Ge's 7/2/76 and 9/30/76 Itrs. and EPRI Aug. 1976 report.... concerning ATWS. 04-27-77 To S. Hanauer from H. Ornstein, GE ATWS Meeting March 24, 1977 withencl. 02-14-78 Memo for D. Ross from A. Thadant, GE ATWS Meeting w/ attached. 05-15-78 Letter to H. Krug from A. Malinauskas, w/ encl. 9 e Y 9 e h I I i i
DATE SUBJECT 08-01-75 From J. Embley to W. Butler, NED0-20626' Studies of BWR Designs for Mitigation of ATWS, Amend. 2. Undated Notes, GE-ATWS. Undated Computer sheets, Westinghouse ATWS Loss of Normal A. C. Power. 08-12-75 Memo Route Slip to A. Thadant from W. Pasedag, GE-ATWS. 1 06-18-75 From J. Larson to DL, Ltr re their 4-1-75 ltr.... requesting that an error in the report entitled "Anticipited Transients without Scram be corrected...... Pilgrim #1 10-16-75 Letter to A. Giambusso from G. Abrell, Dresden Station Units 2 and 3 Quad-Cities Station Units 1 and 2 Anticipated Transient Without Scram (ATWS) NRC Dkts. 50-237, 50-249, 50-254, and 50-265 with attachments. 10-16-75 NRC Distribution for Part 50 Docket Material from G. Abrell to A. Giambusso, Letter trans the following.... Dresden 2&3, Quad-Cities 1&2 11-04-75 GE-ATWS Radiological Consequences, Draft W. Pasedag. 12-08-75 Letter to T. Novak from A. Thadani, GE Trip Report with enclosures. 12-09-75 W. Kerr from R. Heineman, Status Re ort on GE Analyses of Anticipated Transients Without Scram. (w/ encl. 04-27-76 Letter to T. Novak from W. Minners, GE/ATWS Meeting Sumary, ATWS Meeting Summary Distribution. 05-04-76 Memo to Files from A. Thadant, GE ATWS Meeting Slides with enclosure. 07-02-76 To D. Ross from GE, GE Co. ATWS Progr^am. 09-07-76 Memo for T. Novak from A. Thadani thry G. Mazetis, RE/ Reliability Meetins. 09-27-76 From GE to D. Ross, GE Co. ATW3 Program Supplement I. 09-30-76 To D.Ross from E. Hughes, Appendix A: GE Report for BWR/5 Appendix B: GE Report for BWR/4. 12-15-76 Memo for T. Novak from R. Frahm thru G. Mazetis, Turbine Trip Without Bypass Reclassification Meeting Minutes with enclosures. 12-21-76 To Rusche from G. Sherwood, Letter re theri 6/30/76, 9/30/76, and 9/28/76 reports and our 12/9/76 report... concerning ATWS. 12-28-76 LettertoR.BoydfromJ.McGaughy,GrandGulfNuclearStationUnitsl&2l Docket Nos. 50-416 and50-417 File 0272/M-001. 0/6140 ATWS AECM-76/61. g l
DATE SUBJECT 04-30-71 Letter No. 183-193-71 to P. Morris from A. Bray, Received May 4, 1971. 04-27-73 Topical Report Review Bi-Monthly Status Summary Tar No. 472, Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor. Date Assigned: May 8, 1973. 07-31-74 Letter to T. Novak from A. Thadani, GE-ATWS Meeting with enclosures. 09-05-74 Letter to T. Novak from A. Thadani, GE - ATWS Meeting with enclosures. 09-30-74 Letter to W. Butler from W. Gilbert, General Electric Licensing Topical Report NED0-20626, " Studies of BWR Designs for Mitigation of Anticipated Transients Without Scram. Received October 30, 1974. 10-04-74 General Electric Topcial Report Distribution, Report No.:NED0-20626 " Studies of BWR Designs for Mitigation of Anticipated Transients Without Scram" from W. Butler. 12-12-74 Letter to E. Davis from A. Giambusso. Undated General Electric Agenda for ACRS Subcomittee Review of ATWS. 01-17-75 Letter to I. Stuart from R. Tedesco. 02-11-75 Letter to T. Novak from B. Grimes, Questions on NED0-20626 (ATWS). with enclosures. 03-04-75 Licensee: Consumers Power Company (CPCo), Facility: Big Rock Point, Summary of Meeting on ATWS from P. DiBenedetto with enclosure. 03-11-75 V. Stello frcm General Electric Company, Questions on NEGO 20626 Received on March 17, 1975. l I Undated R. DeYoung from V. Stello, Review of NED0-20626 (TAR-ll46), Studies of BWR Designs for Mitigation of Anticipated Transients Without Scram. with enclosure. 04-30-75 Letter to A. Giambusso from C. Aslaksen, Received May 3,1975. 04-30-75 to A. Gaimbusso from Jersey Central Power & Light Co., Ltr Trans the ___following-Oyster Creek _ Unit 1, with enclosures. NRC Distribution for l Part 50 Docket Material. l Letter to W. Butler from J. Embley, Amendment 1 to General Electric l 05-30-75 l Licensing Topical Report NED0-20626, " Studies of BWR Desings for l Mitigation of Anticipated Transients Without Scram" 06-09-75 Topical Report Distribution,. Report NED0-20626-1, Studies of BWR' Designs for Mitigation of' Anticipated Transients Without Scrams, l Amendment 1. l l l --}}