ML20126K002
| ML20126K002 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 12/30/1992 |
| From: | Mike Franovich Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9301060363 | |
| Download: ML20126K002 (50) | |
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d/Ie7 e g.\ o UMTED STATES NUCLE AR REGULATORY COMMISSION p d.ic j [' WASHINOY ON, D. C. 20655 i y, ,I , k..,*./e December 30, 1992 j l Docket No. 52-002 *
. l APPLICANT: ABB-Combustion Engineering, Inc. (ADB-CE)
PROJECT: CE System 80+
SUBJECT:
PUBLIC MEETING OF DECEMBER 2 and 3, 1992, TO DISCUSS THE CE SYSTEM 80+ SEVERE ACCIDENT SUBMITTAL On December 2 and 3,1992, a public meeting was _ held at the ABB-CE facilities a in Windsor, Connecticut, between representatives of ABB-CE.and the U.S. Nuclear Regulatory Commission (NRC).' The purpose of the meeting was to l discuss ABB-CE's submittal entitled, " System 80+ Severe Accident Phenomenology ! and Containment Performance," dated August 1992. The meeting focused on l' questions resulting from the_ staff's review of the submittal. -Enclosure 1- - provides a list of attendees. Enclosure 2 is the staff's detailed meeting summary. Enclosure 3 lists the questions discussed that were originally included in the associated meeting notice. Enclosure.4 contains ABB-CE's . preliminary responses to the staff's severe accidents questions. The majority of the staff's questions (Enclosure 3) were either answered -- I during the meeting or will be addressed when the next--revision of the document is submitted on January 21, 1993. Fuel coolant interaction (FCI)= and-debris coolability were identified as the i two areas of greatest concern because an obvious path to resolution was not ! clearly defined. Because first-of-a-kind-engineering (F0AKE) has not been performed for the. System 80+ design, a detailed structural analysis will not take place until that time, which ABB-CE believes is a post design-certification action. Therefore, it was unclear as to whether.ABB-CE will be able to determine the reactor cavity's structural ability. to-accommodate _ ablation resulting-from core-concrete interaction-(CCI).and loads associated- . with a FCI. At- the conclusion of the meeting, ABB-CE agreed to evaluate what ' a could be done to define the forcing function and determine the ability of the reactor cavity to withstand ablation with static and dynamic loads resulting-- ; from these phenomena. The high ultimate strengths calculated for the-System 80+ containment did not consider the potential effects' of the containment penetrations. - Since the System 80+ does not have finalized penetration designs, ABB-CE has assumed for analytical purposes-that System 80+ will- have penetrationszthat have greater strength than the containment shell. ABB-CE stated that containment- , ' penetrations performance parameters 1will be addressed in:.the^ procurement , specifications that will require penetrationsuto be qualified to the ultimate capacity of the containment.- Thisls a different_' approach overiexisting= plants with steel containments. 'If the penetrations are as strong as the containment shell, catastrophic failure of the metal shell-.could become.a-credible event.
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December 30, 1992' . ABB-CE provided the staff with HAAP computer code input files on a floppy-disk during the meeting. By letter dated December 15, 1992 (LO-92-117), ABB-CE submitted documentation declaring this information as proprietary in accordance with the provisions of 10 CFR 2.790. In addition, it appeared that a significant number of MAAP computer code cases to supplement the probabilistic risk assessment have not yet been performed. t The staff commented that the aumber of hydrogen igniters appeared to be low. Representatives of the NRC's containment systems and severe accidents branch (SCSB) will meet with members of Duke Engineering & Services, Inc. (DES 1) to ! gain a better understanding of igniter placement and obtain the volume and-vent capacity for the compartments in containment. The following is a synopsis of connitments that were made and other concerns that were raised. A more complete and detailed summary is provided in Enclosure 2.
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(1) ABB-CE agreed to define and determine the forcing function and the reactor cavity's capability to withstand ablation and static and dynamic loads resulting from severe accident phenomena (e.g. CCI and FCI). . (2) ABB-CE committed to address the staff's concern of the unique susceptibility of the sumps (e.g. reactor cavity sump) to CCI. (3) ABB-CE committed to go through the various sequences and identify the equipment and instrumentation that are necessary to monitor and mitigate the consequences of a severe accident. ABB-CE's initial response to exactly what the System 80+ plant operators will be looking at to make this determination will not be provided to the staff until the January 21, 1993, submittal. (4) ABB-CE committed to meet with the staff after the January 21, 1993, submittal to discuss accident management, (5) ABB-CE committed to provide the staff with a list of station battery loads during station blackout and other severe accident conditions. (6) - ABB-CE committed to provide the staff with the calculation justifying that the containment can withstand 50-percent entrainment from a high-pressure melt ejection (HPHE) event. (7) ABB-CE committed to document their conclusion tnat the most probable rnode of reactor vessel failure is via the instrument tubes. ! (8) ABB-CE committed to orovide the hand calculation that determines the ! maximum pressure' spice resulting from adiabatic, isochronic complete combustion (AICC) of hydrogen. l u
i December 30,-1992 , (9) ABB-CE committed to evaluate the accident handling capability of the- ! System 80+ containment purge system. Use of the purge system lends the potential to reduce over 50 percent of the total risk if the system has the capability to purge under certain severe accident conditions. In-response to the staff's request, ABB-CE agreed--to revise the severe accidents mitigation design alternatives (SAMDAs) submittal accordingly. ABB-CE will also provide any previous assessments performed by DESI for , use of a filtered vent. (10) ABB-CE agreed to provide NRC's analysis group with the design-basis
- accident _ mass and' energy release rates as listed in the Combustion Engineering Standard Safety Analysis Report (CESSAR-DC) on.a floppy-disk.
Enclosure 3 contains the information requested by the analysis group to assist ' them in their performing MELCOR analyses for CE System 80+. 4 Sincerely, (Original signed-by) _ Michael X. Franovich, Project Manager Standardization Project Directorate - Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
As stated cc: See-next page DISTRIBUTION w/ enclosures: Docket File PDST R/F MFranovich MSnodderly PDR JKudrick SJLee RPerch RPalla JMonninger ADrozd DISTRIBUTION w/o enclosures:
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-.TWambach AThadani RBarrett .- WBeckner '
TKenyon 0FC:' LA:PDST:ADAR PM:Pd ADAR SC: T:ADAR NAME: PShea 08 MXFr ch:tz RBo ardt
-DATE: g'p/p/92
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i ABB-Combustion Engineering, Inc. Docket No. 52-002 cc: Mr. C. B. Brinkman, Acting Director (w/o encl.) Mr. R. Schneider (w/ encl.) Nuclear Systems Licensing ABB-Combustion Engineering, Inc. ABB-Combustion Engineering, In:, 1000 Prospect Hill Road 1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Windsor, Connecticut 06095-0500 Hr. C. B. Brinkman, Manager (w/o encl.) Washington Nuclear Operations ABB-Combustion Engineering, Inc. 12300 Twinbrook Parkway, suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch (w/ encl.) Nuclear Systems Licensing ABB-Combustion Engineering, Inc. > 1000 Prospect Hill Road Post Office Box 500 . Windsor, Connecticut 06095-0500 Mr. Daniel F. Giessing (w/o encl.) . U. S. Department of Energy NE-42 Washington, D.C. 20585 Mr. Steve Goldberg (w/o encl.) Budget Examiner 725 17th Street, N.W. Washington, D.C. 20503 Mr. Raymond Ng (w/o encl.) 1776 Eye Street, N.W. Suite 300 Washington, D.C. 20006 Joseph R. Egan, Esquire (w/o encl.) Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W. Washington, D.C. 20037-1128
o . , l HEETING ATTENDfES DECEMBER 2 and 3, 1992
!@i[ ORGANIZATION .- !
H. Franovich NRR/PDST R. Hitchell NRC A. El-Bassioni NRC ' J. Kudrick NRC
- H. Snodderly NRC -
J. Honninger NRC R. Palla NRC/NRR/SPSB R. Schneider ABB-CE-L. Gerdes ABB-CE S. Ritterbusch ABB-CE M. Jacob ABB-CE T. Oswald Duke Eng. & Sycs.- - e F Enclosure - . . . - . . - . - - . . -- . - .-. - .. : . . . .. .. -
r HEETING WITH ABB-CE ON THE CE SYSTEM 80+ DESIGN FOR SEVERE ACCIDENTS HEETING OBJECTIVE ,.- This meeting represented the first severe accidents working meeting between the staff and ABB-CE personnel. As a result, it is im)ortant that the staff obtain a better understanding of how ABB-CE has concluded tlat the design is able to adequately accommodate a significant and credible )ortion of the severe accident sequences. In addition, the current System 80+ su)mittal refers to various analyses which have led ABB-CE to substantiate design adequacy. The meeting iis intended to address, in more detail, these analyses. Finally, ABB-CE has applico the results of generic analyses throughout their severe accident evaluations. The staff needed to understand how ABB-CE concluded that the generic analyses were applicable to the System 80+ design. DISCUSSION . To focus the discussions, the staff generated a series of questions which were forwarded to ABB-CE. These questions were subsequently used as the meeting agenda. As a result, this meeting summary also used the questions as a basis for the meeting , findings. Computer Analyses The staff transmitted a series of questions concerning the MAAP runs which have been conducted in support of the design. In tddition, the staff also wanted to perform some HELCOR computer runs to aid in the staff's understanding of the System 80+ design. ABB-CE indicated that HAAP code was used to implement a series of computer runs. The results were then used as input for probabilistic risk assessment (PRA) analyses. Items of interest included timing for early core interactions, time to vessel failure,- fission product releases, and late containment failure modes (i.e., containment pressure, temperature, and non-condensible production). ABB-CE declared the System 80+ HAAP computer code input files as proprietary information. By letter dated December 15, 1992, the applicant has provided appropriate documentation attesting to the proprietary status of this information in accordance with the provisions of 10 CFR 2.790. In addition, ABB-CE agreed to send on a floppy disk the' design-basis accident (DBA) mass and energy release rates as listed in the CESSAR. As part of these' discussions, both parties decided that a meeting was needed to continue the NAAP discussions relative to how it was used in the PRA analyses. ABB-CE was interested in defining the specific HAAP runs which were needed in this regard. This level of detail, however, proved to be impossible to produce during ' the current meeting. It was agreed that a two-day meeting should be held January 4 and 5, 1993, in Windsor, Connecticut. Enclosure 2
i Severe Accident Chapter Ouestions Questions 1, 2, and 3 , The discussion focused on the methods used to develop the containment fragility I curve. The approach was to generate the curves assuming a plain sphere without i any penetrations. Although, the methodology has not been discussed with < members of our structural staff, ABB-CE indicated that the method is described in NUREG/CR-4870. The NUREG addresses penetration strains and the immediate area of the shell around the penetration. To account for each penetration, procurement criteria would be established in such a manner to ensure that the penetration would not be less robust than the shell. The staff indicated that , if ABB-CE were to adopt this approach, such performance parameters must be j specified in the associated ITAAC. ; The staff ex)ressed a concern that unless the staff would agree completely with this approac1, both the Service Level C and ultimate pressure values of the containment could change. As a result, the staff said that we would be in contact with our structural staff and set up a telecon with the appropriate .
- members of ABB-CE staff.
The use of expert solicitation in the development of the fragility curve was also discussed. f: pert solicitation from NVREG-ll50 indicated that a 50-percent failure probability would be reached at strains between 2 and 3 percent. ABB-CE reduced the strain to 0.5 percent to account for the presence of the penetrations as well as additional safety margin. This information would also be forwarded to our structural staff for their evalua-tion, In order to determine the sensitivity of Level 2 PRA results to the containment fragility characteristics, ABD-CE agreed to consider performing a sensitivity analysis assuming different containment fragility. The discussion also included how seismic loads should be considered in the calculation of the conditional containment failure probability (CCFP) value. The staff indicated that when one considers whether the design meets the 0.1 value specified in SECY-90-016, only internal events need be considered. As a result, seismic loads need not be considered, but seismic loads should still be discussed in the overall evaluation of the design to cope with various events. Question 4 Penetrations will be designed consistent with the fragility curve assumptions. This means that the penetrations would be stronger than the shell. It was noted that this is not consistent with existing containment designs. In fact, past analyses have taken credit for the fact that penetrations would-fail prior to shell failure. As a result, catastrophic failure of the metal shell need-not be considered as a credible event. ABB-CE noted the unique consideration for penetration designs and indicated they would get back to the staff on this matter. The question of what shell temperatures were considered in the development of the curves was raised. It was indicated that for early failures, a temperature
l I of 280 'f was used while 460 'f was used for later failures. HAAP results were used to compare the validity of these values. It was found that MAAP predic- , tions of late containment temperatures ranged between 380 'I and 430 'f. This information substantiated the selection of 460 'f temperature. The staff requested that ABB-CE reconfirm these temperature values, based on the revised MAAP 3.B analyses that will be performed in support of.the Level 2 PRA update. ABB-CE agreed to provide the reconfirmation results once the analyses are available. Question 5 The discussion concerning the question of what is needed for demonstrating equipment survivability was rnuch needed. The initial position of ABB-CE was that all equipment would be used early in the transient. As a result, no additional consideration was necessary since the environment would be no worse than DBA conditions. The staff discussed the sources from which survivability criteria are identi-fled. They are DBA, 50.44, 50.34(f), and SECY 90-016. It was acknowledged that neither 50.49 nor Appendix B of Part 50 were applicable in events which, were beyond DBA conditions. DBA requirements contain substantial margins to cover the unknown. However, beyond design conditions should consider best , estimate calculations without this added safety margin. It was suggested that ABB-CE look into the approach taken by both ice condenser and' Hark Ill designs to show that the-needed_ equipment would perform their specified function. ABB-CE indicated that they would go through the various sequences and identify the appropriate equipment and instrumentation and show why the equipment will function. Although they were unable to commit to a schedule, ABB-CE did indicate they would look to Duke Engineering & Services for the selected approach. ABB-CE will respond in the December 15, 1992, submittal with both an approach and commitments on when their future submittal will be provided to the staff. Accident management and how the System 80+ equipment could be used were also discussed. ABB-CE felt that identification of equipment and a discussion of how it would be used should be separated from accident management. The staff indicated that this issue and its resolution could be delayed until the more substantial PRA analyses are completed. As a result, agreement was reached that a separate meeting be scheduled after January 21, 1993, to further discuss this issue and other related items. Question 6 ABB-CE indicated that they intend to take credit for fission product removal by the annulus ventilation' system and secondary containment building for both DBA, as well as severe accidents. -It was aointed out that in spite of a severe event occurring within containment, tie secondary containment could well be unaffected. The approach to establishing the leakage value above design was also discussed. The staff agreed that penetration failure need not be consid-ered, especially in light of the design criteria discussed in response to
questions 1-4. However, increases due to the pressure should be considered. 1he staff suggested that a ratio of the actual pressure difference to the design pressure difference be used to increase leakage. ABB-CE had intended to .1 use the square root of the squares of the pressures. It is the intent of ABB-CE to select the more conservative approach, if at all possible. l Question 7 , The intent of the hold-up volume is to reduce the possibility of debris entering the in-containment refueling water storage tank (IRWST) during the post-loss-of-coolant-accident (LOCA) recirculation phase of operation, in ' addition, it was clarified that the containment sprays have the same initiation signals as more recently licensed CE plants. There is a scram permissive with a 2 out of 4 hi-hi containment pressure. Question 8 There were several clarifications of the intent of the words provided in the ABB-CE submittal. Question 9, 10, and 11 To determine the water levels throughout the containment during the course of the event, ABB-CE devel v ed a separate hydraulic computer program. The program is capable of computing the transient depth of water in all possible water pools within the containment during an event. A simplified version of this program was incorporated into the MAAP program via a contract with Advanced Reactor Severe Accident Program (ARSAP). ABB-CE indicated that the IRWST does not completely surround the containment, but only 300 degrees due to the presence of the refueling canal. The hold-up volume is rectangular in shape measuring about 37 feet by 11 feet. The maximum water capacity of the hold-up volume is about 60,000 gallons. , ABB-CE indicated that earlier PRA results indicate that a dry or partially flooded cavity is expected to occur less than 1.0 percent of the time, in addition, the design was established so that at no time will the water reach the bottom of the reactor vessel. The interconnecting valves which are required to open during the event to provide a flow path between the hold-up and other volumes are normally in a dry environment. However, it was noted that post-LOCA conditions will have the valves submerged. ABB-CE will check to ensure this condition is properly recognized in the valve specifications. The staff noted that these flood valves should be included as part of the pre-operational test program, as-well-as ITAAC for valves. These valves should undergo actual dynamic testing to verify their capability. 4-
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a . Question 12 NUREG-1465 is being used as the reference. ABB-CE will try to be consistent throughout the discussions since it was noted that:there may be some incorrect references in the current writeup.
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Question 13 F ABB-CE believes that the control room operators will know long before actual core damage that the event is going beyond DBA conditions. The initial response to exactly what plant information/ instrumentation the operators will use to make this determination will not be provided to the Ataff until the January 21, 1993,- submittal. But, ABB-CE said that this ir, formation would not be explicitly included in the emergency procedure guidelines (EPGs). The staff indicated that the NRC will evaluate MAAP runs to ensure flnat the response with respect to timing as well as the actions are fully supported. If not, the staff expects'the differences to be addressed in the response. Question 14 . ABB-CE acknowledged that it is possible that the operator could unnecessarily flood the reactor cavity. However, there does not appt:ar to be any serious downsides due to this operation. It would mean more cleanup. When questioned, ABB-CE stated that adequate net pop:tive suction head (NPSH) would exist-for-safety injection pump operation with the reactor cavity flooded. In addition, there is a dry-cavity analysis which shows acceptable results of core-concrete interaction (CCI), and therefore, there is no minimum water level is not an issue. An outside water source for the containment sprays has also been provided. An external tee on piping outside the containment has been provided to allow the , use of fire water. The tee would be normally blind flanged; however, a spool-
' piece would be provided to allow alignment with the fire water.
t L Current System 80+ documentation ind'.ates the hydrogen ignition system (HIS) l will not be initiated until an indication of core 1 damage. The staff. indicated-L that under this criterion, a significant quantity of hydrogen may be generated-prior to system activation. Based on the ensuing discussion, ABB-CE agreed that this may be too late. ABB-CE committed to review the HIS actuation criteria used in operating ice condenser and boiling water reactor Mark III plants, and modify their criterion accordingly. . Question 15 and 16 L L Either the question was already addressed, or the clarification was straight :
- forward.
Question 17 The threshold pressure value is-still under discussion at ABB-CE. ABB-CE believes 250 psi can be the final value. In fact,_it was stated that most, if not all the calculations to date, actually se this value. However, ABB-CE.is l^ m m y - ~ , - m.
not ready to commit to such a conservative value unless they are positive that
.the revised analyses show that this.value yields acceptable _results. .ABB-CE expects to respond to this question more. fully in the January 1993, submittal.
Question 18 ;
. 1 Depressurization of the reactor coolant system (RCS) using the SDS pathway will :
take about one hour while MAAP shows that it will take about 4 hours before you I
-get into trouble. ABB-CE also indicated that the shortest possible time for !
this operation is approximately 40 minutes, in light of the relatively short ! time needed for this operation, ABB-CE believes there is sufficient margin to c take this action. The event that ABB-CE uses for making this conclusion is a loss of feedwater. A depressurized RCS is necessary to initiate feed and bleed operation. The staff committed to identify those MAAP results that thes would like ABB-CE to submit in support of the PRA and severe accident closure analysis. This response will probably be provided during the scheduled January-4 and 5,-1993, meeting. . Question 19 ABB-CEindicatedthatthestationbatteriesaredesignedtoprovidepowerfoi only 2 hours. Four-hour capability per battery bank is achieved through use of a load management program. This was in accordance with the CESSAR-DC Chapter 8 description of the de system. The severe accident documentation also . indicated - that the design has a cumulative 8-hour battery capability. Discussions did not provide any clarification in this area as to what equipment is on the batteries. The staff will speak with electrical engineering reviewers to ascertain the battery adequacy for station blackout (S80) and accident capabil-ity. ABB-CE will provide loads for severe accidents including 580. Question 20 Results of MAAP runs were used to determine the timing of the various key events in the accident progression, such as time to core damage and time to. - reactor vessel failure. To allow everyone in the meeting to understand the models contained in MAAP, a brief discussion of the basic nodes in MAAP was provided. The RCS includes a crude reactor model, a detailed pressurizer, two steam generators, a PORV, ECCS systems,'and all critical pumps, The principle input parameters involve mass and energy. All the sumps as well as the containment are also modeled. The containment is divided into four nodes: an upper containment volume above the crane wall;.a volume above the reactor vessel and up to the crane wall; the reactor cavity; and the volume outside the cavity region and up to the crane wall. Input consists of typical containment parameters such as volumes, heat capacitance, etc. -
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In addition, heat removal systems such as sprays, IRWST, and RHR are also modelled as well. For the January 4 and 5, 1993, meeting, ABB-CE will discuss the HAAP base case and sensitivity analyses. This will include discussion of the Electric Power Research Institute's (EPRI's) recommendations for MAAP 3.B sensiti,vity analyses and how ABB-CE complied with these recommendations. - Question 21 Mainly straight forward clarifications. Question 22 ABB-CE's initial response was that the pool seal would eliminate the upward
, pathway thereby assuring the 90-percent assumption. However, further discus-sion called into question whether or not the seal could withstand the antici-pated pressures without rupturir.g. In addition, a second possible flow path around the reactor vessel hot and cold leg nozzles was acknowledged by ABB-CE.
ABB-CE indicated that calculations have shown that the design could accommodate up to 50 percent rather than the assumed 10-percent upward flow. Since the impact of these two additional debris flow paths could not be confirmed during the meeting, ABB-CE agreed to get back to us on their position on this issue'. Question 23 ABB-CE would get back to us the following morning. i Question 24 ABB-CE did not have a referencable document concerning the BETA last test evaluation. The staff indicated will check with the NRC staff, but the staff did not believe such a reference existed. In any-case, ABB-CE indicated that the writeup would be revised to use only acceptable references. Question 25 9 ' After a brief discussion, the staff agreed with the comment. Question 26 Straight forward clarification. Question 27 Detailed review will be delayed until the meeting scheduled for January 4 and 5, 1993.
. Question 28 The staff indicated that it was essential that an analysis be performed which would show the structural capability of the support structure. ABB-CE agreed
1 . -.. i with this assessment and planned to discuss what' could-be-provided to the . - staff. They indicated that internal meetings-were _necessary before they could- - have any meaningful discussions with the _ staff. As a result, ABB-CE cou'd only . indicate that they would get back to the staff. l Question 29, 30. and 31 ;- ABB-CE plans to get back to the staff on these items. Question 32 ABB-CE believes that they have performed the necessary analysis to adequately address this issue. ABB-l.E indicated that all that was necessary was'to locate the analyses and transmit the information to the staff. ABB-CE will get back - : to the staff when-this information-could be-provided, but they did.not-consider. this to be an issue. Question 33 ABB-CE indicated that the containment sprays can also _be powered by the combustion turbine. In-addition, if _ sprays are not available, the containment wiil fail eventually because operation of the sprays is the only way to remove energy from the containment via a heat' exchanger within the flow loop. The containment sprays are modelled in MAAP-as either on'or off. ~' Question'34 The selection of the one hour was based-on expert judgment. ABB-CE believes that whether the value is one or four hours, the categorization of events would not change. However, the amount of combustible gases present would be greater-if a 4-hour period were used. ABB-CE indicated they will reassess:the use of a 1-hour versus a _4-hour time period, and identify an. approach for dealing:with additional combustible. gases if.the one hour is retained.
-Question 35 ABB-CE indicated that they intend'to comply with the EPRI Utility Requirements Document (URD). However, it must ha re.lized that the EPRI requirements are in a sense a living document and are constantly-changing. By December _ 24, 1992,-
ABB-CE intends to submit an update to the. staff on the design comparison with' the EPRI requirements (URD). . Currently, they believe the: design is in general-- compliance with the EPRI-criteria with approximately 25 departures from the _ EPRI URD.
-Question 36 ABB-CE believes that the steam generator tubes are preserved when the reactor pressure falls'below 1000 psia. In addition, ABB-CE has concluded that below 200 psi, direct containment heating (DCH) is eliminated. It was noted that the current thinking of:our research staff is slightly more conservative. Below-
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N ! 150 psi, there-is complete agreement. Between 150 and 250 psi, it is felt that. it could be acceptable, but it is questionable. At 300 psi and above, the elimination of DCH would be unacceptable. In addition, the applicant had referenced an industry report from the Nuclear '
-Management and Resources Council on severe accidents. The industry report-suggests that creep in steam generator tubes is not credible with 9 feet-of water above the core. ABB-CE will attempt to provide the report or provide hand calculations to support this assumption.
Question 37 Discussion centered about a better understanding of the assumptions which went- - into the development of the figure in question. ABB-CE indicated that the underlying assumption was_ that everything reacts, ejects, and burns. They. believe that this is very conservative. They also considered no entrapment,- but no steel reaction was included in the calculations.. With respect to the general divisions in the figure, the following added information was provided, for the dry' cavity with no RCS water means that only , steam is considered. The values used in the figure were EPRI numbers. In addition, a dry cavity means that no hydrogen burns are considered with RCS water meaning only left over water. The region of full cavity with RCS water means that about 227,000 kg of water was assumed available. ABB-CE considers that DCH should only be considered for a dry cavity. This opinion is believed by ABB-CE to be consistent with HUREG-1150. However, it was again noted that if it could be shown that the' design could tolerate 50 percent involvement, this capability should resolve the issue. However. the staff would still question'the assumption of the 10-percent assumption, as presented by ABB-CE. Question 38 Current documentation references the EPRI Technical Basis Report (TBR) devel-oped~1n support of.the-industry accident management program report. This document is not available to the staff. However,'ABB-CE indicated that a simple calculation can be performed to prove the sume thing. ABB-CE agreed to-attempt to make the reference available or provide an acceptable basis.to resolve this issue (e.g., hand calculation). In ' addition, ABB-CE will identify the differences between System 80-and System 80+. For each feature that is provided in the. System 80+ design, there will be a discussion of the-risk improvement of the feature. ABB-CE believes that, for the most part, they have provided the staff with the necessary information. It is documented in the staff's draftisafety evaluation report in Section 19.1 starting on page 19-13'and continuing for about'12 pages. However, th'e staff indicated that this information is relevant only to severe accident prevention features and=their impact on core damage frequency. The staff recommended that ABB-CE consider developing comparable information- for _g_
the severe. accident mitigation design alternatives (SAMDAs) and existing features for System 80+. This would-include an assessment of the reduction in
. risk or CCFP for each feature._ For purposes of Chapter 19.2, the staff indicated that the section appeared to ae sufficient to describe why the design is an improvement over existing designs. It was further indicated that the detailed information would be nice. However, the staff indicated .that the staff would get back to ABB-CE if the added information would be*needed for design certification. The staff will discuss this matter further with ABB-CE as part of the January 4 and 5, 1993, PRA meeting.
Question 39 ABB-CE intends to review several key reports as well as performing some structural calculations. The applicant recognized the importance in establish-ing the structural capability of the structures. However, it was unclear during the meeting as to what was possible in this short period of time. In conjunction with the structural calculations, ABB-CE also agreed to perform related forcing function calculations. The staff recognized the large amount of work necessary to fully resolve this issue. As a result, they' agreed to provide guidance to the ABB-CE staff. In this spirit, the staff agreed to accept final documentation of this effort no later than March 21, 1993. If received by this date, the staff will evaluate the information in the System 80+- final safety evaluation report. However, it was noted that meetings of the type of this current meeting would be necessary to keep the staff informed of the progress in the severe accident analysis - area. Question 40 ABB-CE indicated that their calculations assumed a 20 kg mass participating with 100-percent efficiency for fuel-coolant interaction (FCI). In response to the staff's concerns, ABB-CE agreed to consider significant core involvement - (about 20-percent core) with lower conversion efficiency (greater than 2 percent). It was noted that this effort needed frequent communication between the structural staffs of both organizations. To this end, the. staff agreed to initiate these discussions so that an acceptable approach could be identified. The intent will be to establish criteria or considerations to be placed on the structures, so that when the actual calculations are performed, the necessary FCI information will be properly considered. Question 41 The structural capability of the reactor _ cavity:to withstand steam explosions reported in the System 80+ documentation was based or developed through design compar_isons-of existing plant structures and not System 80+ design-specific structural calculations. More importantly, ABB-CE did not expect to-perform calculations of this level of detail prior to design certification. _ The staff-indicated that this will require involvement of the NRC structural staff.
I However, the staff indicated that ABB-CE must specify and support performance goals for static and dynamic load capability of the reactor cavity structure. ABB-CE agreed to do this and attempt to define the forcing function. ABB-CE committed to provide an evaluation; however, ABB-CE stated that a detailed calculation for the structure would not be achievable at this time. This position was based on the premise that the finalized structual design of the reactor cavity would be accomplished through first-of-a-kind-engineering (F0AKE). The adequacy of both the number and location of the hydrogen igniters was identified as an area that needed to be discussed. This meeting will take place sometime in January of 1993, at the Duke Engineering & Services, Inc. (DESI) facilities in Charlotte, North Carolina. During this meeting, the staff will audit design information from the PACE system which yields 3-D computer visuals of the interior of the containment and nuclear annex. Question 42 For the case involving dynamic loads from ex-vessel steam explosion (EVSE),_the staff suggested that a rectangular load should be considered with the appropri-ate time of the applied load to be determined by ABB-CE. ABB-CE should perform a sensitivity study for timing in the case of dynamic load analysis for the reactor cavity. It was noted that per NUREG-2462, this profile represents the mcst conservative profile. As a result, ABB-CE would consider the staff's recommendation, but would not commit to using this profile without further study. Question 43 The staff expressed concern that recent TMI studies indicated that lower reactor vessel head failure / slump could occur prior to instrument tube failure. ABB-CE committed to document their basis for concluding that the instrument tube failure is the most probable, it was also noted that there are no other bottom penetrations in the reactor vessel. Question 44 Our previous discussion on FCI more or less addressed this issue. NUREG-4B96, page 27 further documents the basis of the design. However, it was noted that ABB-CE agreed to forgo the Zion probabilistic safety study as their basis and will undertake their own study. Question 45 The reference which was used to justify the selection of the upward heat flux was R. Henry. It was taken from the Westinghouse AP600 submittal. In the context of the System B0+ submittal, it represented a bounded number which is based on a combination of both ACE and SWISS experiments. However, this information was-not really used by ABB-CE, since a rate dependent load is not u ed.
- Il -
44,,b.6 ,d- -W 4 4l. -a 4 m / 4+ s-. /a - AAa+4s-s. .$
-Question 46-The pressure spikes for rapid steam generation event-were hand calculated. It was assumed that the spike would occur at the same time that the peak LOCA containment presvure occurred. This is a conservative assumption and not a mechanistically-c.erived conclusion.- The spike was based on 70 persent of core involvement with no added chemical reaction energy. -ABB-CE committed to
. provide the staff with a copy of the hand calculation.
Question 47 The reference for the calculation of the pressure rise as'sociated with a hydrogen burn was stated by ABB-CE to be NVREG/CR-5567, page 3B. ,The staff stated that a further check by the staff was needed to determine the accept-ability of this approach. It.was indicated that a_NUREG/CR-report _does not necessarily indicate staff approval, since it is a contract-produced document. Of particular note was the fact that ABB-CE indicated that their investigation has shown that caution should be taken by the. operator in turning on the sprays when steam inerting is suspected. ABB-CE indicated that hydrogen.instrumenta-tion should be checked prior to taking action to determine if hydrogen concen-trations would indicate a basis for a delay in operator-action. As an aside, ABB-CE verified that the design has safety-grade and redundant hydrogen , instrumentation which is capable of monitoring hydrogen concentrations up to-15-percent hydrogen concentration. The staff believes that this operator information on containment spray actuation during a potentially steam-inerted environment should be appropriately relayed to a combined license applicant through the EPGs or System 80+ accident management guidelines ~(AMGs). It was noted that in several instances, ABB-CE has used 100-percent metal-wate,' reaction to mean all zircoloy. clad interacts with water. This translates to 130 percent of the fuel clad surrounding active fuel. The staff clarified that 10 CFR 50.34(f) requires 100-percent fuel clad metal-water reaction of only - active fuel. ABB-CE used hydrogen control information from the Department of Energy (DOE) report, "Techn_ical Support for Hydrogen Control Requirements for EPRI Advanced PWR." The staff suggested that a common reference be used~to avoid any confusion. Since the majority'of-references. refer to active fuel, it was_ suggested to use this as the reference rather than the total amount of-zirconium within the core. ABB-CE agreed with.this suggestion and will modify all references to this common base. Question 48 ABB-CE supported their approach by comparing their' calculated values against experimental data. They concluded that their calculated values bounded all experiments and was therefore conservative. Figure 4.1-2 was.the focus of the-discussion. It is used by- ABB-CE as confirmation of the actual calculations
- performed by them. It was noted that-the-presence or steam seems'to indicate- -
that the peak pressures-would be about 15 psi higher than dry' air.
Question 49 Either previously answered by other question responses, or the clarification was straight forward. Question 50 ,
^
ABB-CE reported that the approach used the guidance of a advanced light water reactor document cited in Question 48. In any case, ABB-CE intends to redo-the MAAP calculations. The staff pointed out that local hydrogen concentrationi predicted by lumped parameter codes such as MAAP are a strong function of the volume of the nodalization scheme. While ABB-CE did not agree nor disagree with the limitation, they indicated that the maximum hydrogen concentration was a minor point since System 80+ relies on igniters to limit concentration. The staff reiterated the point that the 2-percent value is an artifact of the node chosen in the MAAP run. In any case, hydrogen igniters will be located immediately outside the IRWST. In-addition, ABB-CE will provide a discussion of why high concentrations'in the vicinity of the IRWST are not possible. As part of this, ABB-CE agreed to - review the mass and energy releases from the IRWST for selected sequences to identify potentially rich hydrogen concentrations. These hydrogen conditions include loss-of-feedwater, station blackout, and feed-and-bleed sequences. - This will be submitted as part of the January 21,199S, package. The staff raised the question of the number and location of the hydrogen ignitors in the design. It was noted that an ice condenser has about 100 igniters, while System 80+ has 42. This was puzzling since System 80+ is both larger in volume and core size and appears to be more compartmented. In addition, System 80+ has used the same placement criteria as an ice condenser. Clarification of quantity and location of System 80+ hydrogen igniters will < occur during the DESI meeting. ABB-CE should be prepared to discuss the differences between an ice condenser and the System 80+ relative to the number and location of hydrogen igniters during the DESI meeting. Question 51 The calculation of the pressure spike due to a hydrogen burn was performed by hand using the adiabatic, isochronic complete combustion (AICC) method. It is based on assuming adiabatic boundaries and uniform mixing within the selected volume. In addition, only global burns have been considered. ABB-CE has considered three different ranges of metal water reaction for their PRA studies: 50 percent for unrecovered events, 75 percent for recovered events, and 100 percent including CCI non-condensibles. The lowest value, which is assigned to unrecovered events, is based on the lack of water in the core. The initial containment pressure is assumed to be either 20 or 30 psia. These two values correspond to having sprays or without sprays, respectively. i 1
i w Using the Labove methods,. ABB-CE calculated an~ overpressure of 110 psiLdue to a -
. global burn:of 130-percent metal-water reaction, asLdiscussed in' question 47.
Therefore,-.forfa spray case, the peak-containment ~ pressure would be 120 psia. This compares with an overpressure of 140 pst_ reported in the NUREG/CR report., ABB-CE- believes the difference to be attributable to the variation of the specific heat at constant volume (Cv) with temperature. .. ABB-CE agreed to clarify the methods and approaches used to calculate. hydrogen combustion loads for early and late. deflagrations since different approaches have been referenced (e.g., AICC curve, Sandia report, other sources). Question 52 The containment pressure value of 30 psia represents the maximum steam concen-tration without entering into the steam inerted condition. Question 53 Combined with other responses. . Question 54 ABB-CE indicated that the steam generator (SG)_ represents only one of several bypass paths. In the spring of 1992, they submitted a response to;the staff's-questions regarding the three specific improvements of the SG design. Their response was consistent with the EPRI views as to why there is no valid reason to upgrade the:SG, interfacing system LOCA (ISLOCA), etc. The staff raised the question.as to why the containnient' bypass issue was not addressed in. the SAMDAs submittal. ABB-CE indicated'that a response to:this question was beyond the agenda of this meeting and should be deferred to a later time. However,.the staff noted that this' issue will teed to be appropri- ' ately addressed for SAMDAs. Other containment bypass pathways were discussed with respect to ISLOCA and pipe pressure capability.- High-pressure / low-pressure pathways, such as component cooling water-(CCW): interface with~the RCS through the high-pressure. seal cooler for RCP seal, and-the shutdown cooling-system:were mentioned. ABB-CE indicated that an ISLOCA submittal for SECY-90-016 purpose has recently been submitted to-the staff for review.
+
Question 55 The staff asked _why ABB CE did not consider venting of- the containment.: A look at the most significant risk contributors _shows that-over 50' percent of the-total risk .is associated with core melt sequences which are a result of containment failure due to over pressure, prior'to-vessel breach. It was acknowledged by- ABB-CE' that a containment vent'would l totally eliminate these sequences. In other words, total risk couldibe reduced in half if a vent' were available. Since the purge line dumps into the containment annulus,- this may be sufficient to completely address the venting issue. The annulus would
w QM qMy&% w serve as a large volume to attenuate the pressure to low enough values such that the secondary containment would not be damaged due to overpressure conditions. ABB-CE agreed to evalutte the accident handling capability of the containment purge system. ABB-CE's evaluation will include assessment of isolation valves operability and purge line piping capability under accident cond4tlons. The staff iterated that the System 80+ SAMDAs submittal must be revised accordingly on the issue of a filtered vent. In addition, ABB-CE committed DESI to provide any previous assessments for use of a filtered vent as a System 80+ severe accident design feature. ABB-CE will get back to the staff on this issue. Question 56 _ By letter dated May 8, 1992 (LD 92-042), ABB-CE indicated that two of the three features were addressed. The third feature was addressed in a November 18 or 24, 1992, su'mittal. o Their identifier to the letter is either LD 92-113 or ' 115. Question 57, 58, 59, 60 , ABB-CE indicated that from a PRA viewpoint, CCI and core debris coolability ' were being adequately covered by the ranges of values of the individual parameters. However, from a deterministic view, additional studies needed to be performed. The staff presented their views on the topic. First of all, it must be recognized that the experimental data is scant at best. Therefore, if ABB-CE does not want to be dependent upon future tests to close this issue, a strong effort is needed to determine the capability of the structures (reactor cavity) to withstand the EYSE. Structural analyses, as well as an effort to define the forcing function on the cavity, will be required to resolve this issue. ABB-CE indicated that their detailed structural analyses are not planned until F0AKE. However, ABB-CE understood and agreed with the approach to closure. ABB-CE needs some time to decide what could be done in the near future to address the structural capability issue. ABB-CE will get back to the staff by January 4, 1993. Another question raised was what constitutes containment failure relative to erosion of the base mat. For the reactor cavity area, ABB-CE presented the view that the containment shell is encased on both top and bottom with con-crete. Therefore, if the bottom portion of the shell~in the region of top and bottom concrete were to be penetrated due to CCI, it would not constitute failure. In fact, for PRA purposes they will continue to assume no failure under these conditions. However, for the deterministic approach,'ABB-CE will-reconsider their assessment. The staff pointed to containment integrity leak surveillance. tests (e.g., Appendix J) that have pulled the steel liner from the concrete, thercby
-creating a leak path due liner deformation and concrete fracture. . The staff also pointed out that it is hard to justify not failing the containment when the containment shell is breached. It was also pointed out that it did not appear that the placement of the sump in the reactor cavity was well thought-out with respect to a severe accident environment. The reactor cavity sump is located near the edge of the containment boundary where the top layer of concrete is of minimum thickness. Simply moving the sump toward the center should improve the ability to accommodate concrete erosion.
ABB-CE committed to reevaluate the location of the sump location and determine if suitable alternatives exist. The alternatives included: (1) evaluating an off center line location of sump; (2) reducing the size of the sump; and (3) adding a false floor in cavity (as stated in Chapter 5 of the EPRI URD). Questions on the Structural Analyses Due to an apparent miscommunication, ABB-CE had not received in advance the detailed structural questions generated by the NRC structural staff. As a result, ABB-CE had less than one day to look over the request. Based on this limited time, the following preliminary observations were provided to the , staff. Most, if not all, of the questions appeared to focus on the theoretical bases for the approach. Therefore, a response to these specific questions should not take much time nor put into question the basic. approach taken by ABB-CE. However, what will take a significant effort will be the response to structural issues raised during this meeting. ABB-CE will attempt to answer all our questions no later than March 15, 1993.
.____c_ __ _ _ _ _ _ _ _ . . .
. .' ' c ,
-i l
QUESTIONS RELATED TO CE'S SYSTEM 80+ SEVERE ACCIDENT PHENOMENO = AND CONTAINMENT PERFORMANCE SUBMITTAL DATED AUGUST 1992
!. Explain- the development of the containment fragility curve (ne the-accompanying calculations. /
- 2. 3-3 ASME Service Level C loading conditions allow material strains representative of incipient yield... 3-5 Foripeak' strains-typical of incipient yield conditions, the probability of containment 1f ailure' is less than 0.05. 3-7 ASME Service Level C failure probability _0.03.
Explain discrepancy.
- 3. 3-7 Basis _for assigning 5% failure probability to Ultimate. Capacity ASME pressure?
4 3-9 Penetrations designed to withstand ASME ultimate pressure and severe4 accident temperature. ASME ultimate pressure (169-147 psig) assigned failure probability of .05. Why not design.to 1.0% strain (220-204-psig) where failure? probability is equal to Ll? Credit-is taken for no- - failure of penetrations between ASME ultimate and 1.0% strain. t
- 5. 3-9 Discuss _the Equipment Survivability section of SECY-90-016 ini relation to-the penetrations, seals, any equipment,-and instrumentation relied upon:in severe accident prevention or mitigation. Provide a-list of all_ equipment, instrumentation, seals etc., that would be relied upon-for-severe accidents or be exposed to severe accident conditions,-and
- the accompanying parameters 1for which they are: designed to. tolerate.
Does the containment spray system, cavity; flooding-valves etc., have any special requirements for severe accident operation?_-
- 6. 3-10 What credit-for DBAs-and-severe accidents is.taken for'holdupiand filtration in the' containment-annuluslof'the shield building?_-is the annulus filtration system-included:in other CE-plants!'(Palo-Verde).?:
What was'the impedus for its-installation? AVS designed for design basis fission product loadings, what about severe-accidents? -Can the-AVS'be power'ed from the combustion turbine generator?.
- 7. :3-13 What is the inter-relationship between the.CFS and containment spray system for providing an-inexhaustible continuous supply;of; water?
- 8. 13: Provide preliminary indications of_the accident management guidance _
for whenito manually actuate the cavity floo' ding system._-Credit ~can not be taken for_ a pre-flooded: cavity unless there is assurance of the indications used for flooding.
- 9. 3-13 What-is the purpose-of the holdup volume tank? Why not flood '
directly from IRWST to reactor' cavity? .. .
- 10. 3-15 Provide timing for filling the HVT, reactor cavity, and final level in each. Do any accident sequences result _in a core mass in dry cavity or only par'tially filled cavity?
Enclosure 3 g g =em'q m - p ,.im 6.m. 4,--
4 2 3-15 Any equipment qualificationLrequirements for the CFS-valves?;
~
- 11. -
- 12. 3-15 What-features are discussed in "Section<.4.3.2"? ~IsLthis the correct reference?i -
..- )
- 13. 3-15 How will prolonged and irreversible core uncovery be determined?
Will the cavity ever be flooded when core damage is not known -to 'have ' ; occurred?
- 14. 3-15 What amount of water'is required in the cavity prior to introduction of the core melt,-for the assumptions made:regarding:
coolability and core concrete interaction to be true? 4 3-15 What is the impact on containment performance and core debris 1 15. coolability if the CFS works but:the containment' spray system.does not?- l
- 16. 3-16 Provide the 100% metal-water: reaction pressurization calculation. .
Does it assume one burn, intermittent burns, continuous' burn, etc. ;What pre-existing containment pressures are_ assumed and why?' 17, 3-19 Reactor coolant system pressure, where direct containment heatin.g is no. longer a concern (anticipated corium dispeisal-threshold value).
- 18. 3-20 RD: capability to depressurize the RCS-from 2500.to-250l psia prior-to reactor vessel melt-through. How long is this? How long.to core-melt-and vessel melt-through in representative sequences?
- 19. 3-20 Battery sized to power loads 'for 4 hours. What about SBO. coping-period of 8 hours? Sequences and timing for when depressurization~is needed.
- 20. 3-20 Provide results of the MAAP runs referenced.
- 21. 3-23 is the instrument shaft referenced, the'same'as the ICI: chase in Fig. 3.6-27
- 22. 3-24: 0CH steam exists.through louvered' vents under the refueling pool..
Are there.any paths where the pressures associated with a.HPME'would force any material (insulation).out- of the way creating a different flow _ path?'
- 23. 3-24 Provide reactor' cavity floor area and representative core ~ debris depths for various sequences-or amounts 1of core debris.
24, 3-27 Provide a copy of Reference 3.12 and- basis for: conclusion that. supports' coolability in the long term.
- 25. 3-27 How will cooling.of the upper layers of corium retard any~ concrete -
< attack?
- 26. 3-29 WhatEtype of concrete is used in the basemat?
ruw re as w -1=wi- p yt - - t er - ~ r & M-M = =g F 4 g -- y if '
. . - - - - _ _ _ ~ . - .. . . - - - . . . -. . . - - .
l9 -4 1 3
~ 27, 3-29' Provide analysts for 30 hours for corium to contact the containment-i liner.
- l
- 28. 3-29 Provide discussion on how the reactor vessel and RCS arp -supported - H and what structures are required for this.
- 29. 3-29 What in-cavity _ structures might be damaged by an ex-vessel steam explosion?
I
- 30. 3-29 Discuss the area above_the refueling pool-not prone to missiles. !
What amount of concrete is.below the core debr.is chamber and~HVT_ sump? Are any sumps located within the reactor cavity?-
- 31. 3-32 What is the external source of water for the containment spray.
system? (fire water, tee provided?)
- 32. 3-33 Are there dead-ended regions of the cavity'where-water from-the -
containment spray system would not be recirculated to the HVT and IRWST?
- 33. 3-33 Can the containment spray system be-powered from the combustion -
turbine generator? What is the impact on_ containment performance with no containment spray systems available? How is the containment spray _ system used in the MAAP analyses?
- 34. 4-1 What is the: significance of defining early containment f ailure as hour after the core debris penetrates the reactor. vessel?
35, 4-1ArethereanyareaswheretheSystem80+ design-departsfrom[the EPRI Requirements for. severe accidents?
- 36. 4-3 Discuss the SECY-90-016_ criteria for high pressure melt ejection,- ,
=
such that a depressurization system should pro. vide a rate of RCS depressurization to preclude molten-core ejection and to reduce RCS pressure suf ficiently to preclude creep rupture of steam generator tubes.
- 37. 4-6 Is this a generic graph or_ specific to the System 80+~ design?
- 38. What are the specific details of the cavity / containment design that will influence- the consequences of an EVSE event- and how does this relate- to' the statement on page 4-21 " Proper location of support: structures and.
cavity wall ~ design can effectively eliminate the containment threatening-potential of.- steam explosions,"? - (4.1.2 2.2, page 4-20)'-
- 39. What,is.the. basis for statement, "the actual mass of corium expected to be involved in my one explosion is small (under 20 kg)"? (4.1.2.2.2.3, page 4-21)
[ 40. Why was the applicability of the BETA V6.1 experiment not addressed?
' 41 '. Provide the analysis that establishes'the cavity design strength..to bt ~
approximately 225-psid. (4=.1.2.2.4, page'4-22)-
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'4
- 42. The cavity design strength mentioned above is for_a_ static load yet What..is.
dynamic loads _ are more likely in the case of a steam explosion. the cavity design strength for a dynamic load associated with a' pressure impulse lasting 5 ms? ,
- 43. Basis for stating the most likely M failure mechanism will be via instrument tube failure? (4.1.2.2.4, page 4-22) 44 The submittal states, "The energetics of this' type of an event (FCI) were estimated in (DOE /10-10271 " Prevention of Early Containment. Failure ,
due to High Pressure Melt Ejection and Direct Containment Heating for_ Advanced Light Water Reactors," March 1990.) to produce. localized cavity loads in the vicinity of 10 bar. 00E/ID-10271 refers to a part of the Zion Probabilistic Safety Study that determined this cavity load. How-is this analysis directly applicable to CE80+7 (4.1.2.2.4, page 4-22)
- 45. Whatexperimentaldataoncoriumquenchingindicatesjhatthequenching process exhibits maximum heat fluxes of up to 30 Mw/m for short time periods? (4.1.2.3.2.3, page 4-23)
- 46. We would like to review the 21 psi pressure spike calculation-and why it.
was assumed that the initial containment loading was_at design limits (49 psig) at the time of vessel breach. (4.1.2.3.3, page 4-24) ,
- 47. Are there any other means for compensating for loss of steam inerting besides the igniters? It is not-clear how the 80+ design will compensate for a loss of steam inertirg once the containment sprays are l
activated. (4.1.3.1.2.1.1, page 4-27)
- 48. Further discussion of Figure 4.1-2:-ALWR Combustion Potential.
- 49. We would like to review the analysis that establishes for 100% Metal Water Reaction complete AICC hydrogen burns result.in peak containment pressures of about 140 psia. (4.1.3.1.4, page 4-32).
- 50. We would like to review the analysis that the vented l_RWST hydrogen concentrations are only 2 v/o greater than the overall containment concentrations. (4.1.3.1.4, page 4-32)
- 51. What is the basis for the statement, " Igniter burns should produce p'
pressure spikes less than that associated with a 50% core wide oxidation"?_
- 52. What is the-basis for assuming the hydrogen burn will-be initiated from
[ ' a 30 psia base pressure? (4.1.3.1.5, page 4-34);
- 53. Further discussion of-Table 4.1-4: Summary of-PRA Assumptions for System L -80+ Hydrogen Deflagration Induced Loading and 4.1-5:. Summary of System L-
) 80+ Containment Failure Drobability Due to Hydrogen Deflagration. What
- 54. Section 4.1.4.6.1 describes the containment bypass phenomena. '
about f ailure of containment penetrations such as the personnel and _ l .- 5
;_ _ - _ , _ , .. _ _ -a
).
5 equipment hatches? (4.1.4.6.1, page 4-54)
- 55. Containment failure before core melt represents 62% of the containment failure frequency and 55% of the total risk for the CE System 80s.
Provide an analysis for inclusion of a filtered venting systeem and direct venting system for the CE System 80+. A direct venting system could be considered, if scrubbing though the IRWET is expected. Provide insights to scenarios were this scrubbing may be effective and the expected decontamination factors. A filtered venting system should be considered for sequences tnat release directly to the containmeit or if the IRWST is determined to be ineffective in scrubbing.
- 56. Containment bypass repr: m ts 28% of the containment failure frequency and 40% of t% total risk. Containment bypass consists of interfacing system LOCA and steam generator tube rupture with unisolable path to the atmosphere. Provide an indication of the benefits of inclusion of the SECY-90-016 criteria for addressing interfacing system LOCA and areas where the criteria has not been met. Provide an analysis for directing' the steam from secondary side relief valves back to the containment and through the IRWST.
- 57. Provide the analysis of basemat melt-through including assumptions of heat fluxes, amount of core, temperature, ablation rates etc.
- 58. Provide a best-estimate analysis of the impact on containment performance of continued core-concrete interaction for 24 hours.
- 59. What are the OCH assumptions ;; Figure 4.1-1 for the best estimate dry cavity with RCS water case?
E0. How much radial and axial ablation can the reactor cavity withstand - without failure?
$$QkWWWF Cf
$b a . . 6 f. _ i 4 RAls on CE System 80+ for Containment Performance
- 1. The statements under the item " Design Basis Pressure Capacity" section in page 3-3 of Reference I should be revised in accordance with the responses of RAls 220.45 through 220.48 and 270.44,
- 2. From the Response to RAI 270.42 for general membrane stress for 3-D finite element model for the steel containment vessel with openings, the maximum stress intensity calculated for the testing load condition (load combination is D + L + P + T,) at 53 psig is 24,614. psi and the .
allowable stress intensily for Level C Service Limit is 52,480 psi at design basis accident temperature of 290 'F. Since the stresses are at or below yield, a linear calculation for the allowable Level C Service Limit pressure can be determined as a check from the testing load condition as follows; 24,614 : 53 - 52,480 : X, X = 113 psig for 3-D The temperature is assumed to be 290*F. From Figure 3-.1-2 of model. Reference 1 for 2-D axisymmetric thin shell model, the level C Service Limit pressure is 145 psig. Since the internal pressure is dominant in the resulting stresses, the difference in the results (145 psig vs. 113 psig) needs to be explained. .
- 3. In page 3-5 of Reference 1, it states, "The material properties were represented by a bilinear stress-strain curve which was assumed to be essentially elastic-perfectly plastic in nature." which means the stress .
is maintained at yield, while the strain is increased. However, use is made of a 5% strain hardening modulus in SA-537 Cicss 2 Stress-Strain Curve (Ref. 2). Provide the reasons why a bilinear stress-strain curve with a 5% strain hardening is chosen.
- 4. In page 3-5 of Reference 1, it states, "the strain at the maximum pressure of 193 psig is approximately 0.003 in/in." Explain how this strain and pressure can be obtained beyond the yield point using von Mises theory when is valid up to the yield point.
- 5. In page 3-5 of Reference 1, it states, "The exact value varies depending _
upon element location and whether the midsurface or inner / outer surface is examined." Explain how the membrane strain can be varied with location and across the plate thickness?
- 6. Assuming a bilinear stress-strain curve, the stress calculations at o 3 0.003 o y in/in strain y
with 5% strain hardening can be performed as a+ ( o n values in following table. O r E Po,po3 P+ ao (psi) (psi) (psig) (psig) 110*F 59,500 29.00E6 177.5 193 290*F 52,480 28.35E6 157.8 169 350*F 51,100 27.80E6 153.7 161 450'F 48,800 27.30E6 147.1 147 Provide the reasons why System 80+ analyses produce higher pressures. ,
- 7. In page 3-6 of Reference 1, it states, "the 0.02 in/in actual tensile f ailure point of SA537 Class 2 material used in the containment shell construction." is the 0.02 in/in strain im lensile failure point for SA537 Class 2, or should it be 0.2 in/in i
_ _ _ _ _ . __ _ _ _ _ . _ _ _ _ . _ _ _ _ _ -- . _ _ - _ . - ~ _ - _ _ - _M
l
- 8. Explain how the extrapolation method in page 3-6 of Reference 1 be can used to get the ultimate pressures of 0.003 in/in strain at higher temperatures using a bilinear stress-strain curve?
- 9. Provide the bases for 3% and 5% of probabilities of failure at level C and ultimate pressures, respectively, in page 3-7 of Reference 1.
- 10. Provide the following information: ,-
- a. Material strength uncertainty, modelling uncertainty, and pressure distribution for containment fragility curve,
- b. Material characteristics for penetration seals to ensure minimal containment leakage at higher pressures and temperatures. It should be noted the containment should fulfill not only structural integrity function but also the leaktightness function. Structural integrity is necessary but not sufficient, because a 3/8"p hole in the containment may not fulfill its function of restricting the release of radioactive materials in case of a reactor severe accident even though it is structurally sound. Therefore, it is essential to establish the leakage through the seals of various .
penetrations.
- 11. Typographical errors:
- a. In Page 3-1 of Reference 1, the material is SA537, not SA357.
- b. In the column title of membrane strain in page 3-7 of Reference 1, the unit is in/in, not % in/in.
References:
- 1. ALWR-FS-DCTR-33, Rev. O, " System 80+ Severe Accident Phenomenology and t
Containment Performance," Combustion Engineering, Inc., August 1992,
- 2. Meeting handout, " System 80+" Steel Containment Vessel Code Design Activities," Combustion Engineering, Inc., April 22 and 23, 1992.
l 1
PRELIMINARY RESPONSE TO ITEM 1/2/ 3 3 The fragility curve is developed from shell membrane stress-strain curves- based on . structural analyses -performed: -by- DESI and information from FMREG-1150 (SAND-1309)'" Experts' Determination of Structural Responso Issues", Issues 5 and 6.~ . Based on Issue 5: EXPERT A I
- 1. Membrane failure' considered likely to occur between 1 and'.5
% strain.
- 2. Onset of general yield was taken to have a containment failure probability of .02.
EXPERT B
- 1. Onset of general yield taken to have a- cont. failure probability ofi.05
- 2. At 2%-strain containment failure has a failure probability of
.5 Based on Issue 6:
EXPERT C
- 1. Experiments show that general failure of steel containments will not occur until a global. strain of 2% has been reached.-
PRA Application ASME LEVEL C taken a onset of general yield with. a containment - failure probability of .03. (Between experts A and B of issue 5) ASME Ultimate calculation taken as a failure: probability of. 05 (strain = .003). Based on expert A (issue 1)- and assuming the failure probability in this range is linear ~ with strain, the failure probability for a .3% strain is between .03 and-. 15. A-value of .05 was selected as being consistent with the Sequoyah fragility curve. Median failure point conservatively selected based on -- Sequoyah fragility curve and expert judgement as 0.5% strain. Enclosure 4
s .. Figure 3.1-3: Comparison of Fragility Curves for System 80+ n n 0.9 - 0.8 - 0'7 - - ' System to+ @ 450 F System 80+ @ 290 F o6-b
)
.3 0.5 -
~
3 8 r 0.4 - DN 0.3 - 0.2 - 0.1 -
^
0.e y . , -, i i i 0 2 4 6 Ratio of Failure to Design Pressure 3-8
PRELIMINARY RESPONSE TO ITEM 4: Penetrations should be designed to be consistent with the PRA assumed fragility curve. Requirements in excess of this are considered out of scope of this effort. , 4
f PRELIMINARY-RESPONSE TO. ITEM'5:- See response to RAI 410.141. Also see page-41 of:NUREG/CR-5567 which discusses the excellent:-
- survivability'.'of hydrogen ignitors and other equipment in severe accident environments. .
9 e
-[
Q. 410.141 Your response to pal 440.20 lists, in part, the hydrogen mitigation system igniters and cabling,-as well as valves for the reactor cavity flooding system,_as equipment that is relied upon-to mitigate consequences of severe' accidents. SECY-90-016 requires that there be high confidence that this equipment will survive severe accident conditions for the period that-is needed to perform its intended function. However, SECY-90-016 has concluded that it is not necessary for redundant trains to be qualified to meet this goal. With this general background, there are several_ areas 'where information is missing in your response to RAI 440.20. Therefore,_ please provide the following:
- a. Provide the results of the calculations used to establish the environmental conditions for severe accident mitigative equipment. These conditions should include pressure, tem-perature, and radiation, as a function of time. In addition, provide the basis for concluding that the above conditions are bounding for the range of severe accidents,
- b. In addition to the environmental conditions, provide any .
further criteria that will be imposed on the mitigative equipment. Indicate if these added criteria are to justify that there is reasonable assurance that this equipment will-perform its function. Provide and justify the seismic design of this equipment,
- c. Describe the electric power supplies for post accident mitigative equipment, including train and bus configurations supplying clars IE and alternate power sources. Describe the provisions for switching between the power sources, if required in the course of a severe accident.
Response to 410.141 a: The equipment used in severe accident mitigation include: (1) hydrogen mitigation system igniters and cabling, (2) reactor cavity flooding system (CFS) valves, and (3) safety depressurization system (SDS) valves. The capability of igniters to function in harsh environment has been demonstrated via a number of NRC and EPRI sponsored test programs. For System 80+ application the igniters and associated cabling are expected to be available to perform their intended function if they survive the environment corresponding to the most limiting containment environment during a de' sign basis LOCA or Main Steam Line Break. Since hydrogen combustion is not a significant threat to System 80+, the primary intent of the igniters is to minimize potential containment combustion loadings. The design basis accident (DBA) qualification range is sufficiently restrictive to encompass most severe accident scenarios. Because the low likelihood of exceeding DBA limits a more restrictive qualification criteria is considered unnecessary. l
Response to 410.141 a (Cont'd)
- The CFS valves are intended for operation prior to a reactor vessel breach.
Therefore, these valves are not required to be qualified to extreme temperature, pressure, and radiation conditions representative of the later portions of a severe accident scenario. Thus, acceptable operation 'of the CFS valves is obtained during a severe accident scenario by qualifying them to design basis accident containment environmental conditions. The SOS valves are expected to be employed for severe accident mitigation prior to, or immediately following, core uncovery. Therefore, no additional qualification testing (other than that is required for design basis accident containment environment) is considered necessary. Response to 410.141 b: Cavity flooding System (CFS) and Safety Depressurization System (505). piping and components are designated in accordance with ASME Section Ill and ANSI /ANS S1.1. The CFS piping and components are Code Class 2 and Safety Class 2. The SDS piping and components that are part of the RCS pressure boundary are Code Class 1 and Safety Class 1. The remaining portions of_the SDS.are Code Class 2 and Safety Class 2. ASME Code and ANSI Safety Class designations for thes.e-piping and components are specified in CESSAR-DC Section 3.2 and Tables 6.7-2 and 6.8.2-1. As described in CESSAR-DC Section 3.2.1, all components in Safety Classes 1, 2, and 3 are Seismic Category I. Use of the specified classifications is intended to provide reasonable assurance that the CFS and SDS equipment will appropriately perform their functions. Response to 410.141 c: The major severe accident mitigative equipment that requires electric power supplies consist of (1) the hydrogen igniters, (2) the cavity flooding system, and (3) the safety depressurization system. The hydrogen igniters are powered from the Class lE 120V AC Vital Instrumentation and Control (l&C) Power system as described-in CESSAR DC. Section 8.3.2.1.2.1 and Table 8.3.2-3. This system normally receives power from offsite power sources, with the Diesel Generators, Alternate AC Source (combustion turbine generator), or the emergency batteries supplying power if offsite power is unavailable. As described in CESSAR DC Section 6.2.5.2.2, each igniter location consists of two igniters, one powered from each electrical division.. The cavity flooding system valves are powered from tha Class IE DC Vital-Power System. Each of the four holdup volume flooding valves are powered from-separate Class lE channels and-each of the two cavity flooding valves are
-powered from separate Class IE divisions as seen from CESSAR-DC Table 8.3.2-4.
The power to the Class _ IE buses.is normally supplied by either of two offsite _ power sources. Upon . loss of both offsite power sources, the Class IE Diesel Generators and the Class lE batteries supply power to the buses. The diverse Alternate AC source combustion turbine generator can power these buses if power from all other sources is lost.
Response to 410.141 c (Cont'd) ' The source of power for the rapid depressurization-valves of the safety depressurization system is the Class IE DC Vital Power System. The power to the Class IE buses is normally supplied by either of two offsite power sources. Upon loss of both offsite power sources, the class IE generators and the Class IE batteries supply power to these s. buse, Thedies'el diverse Alternate AC Source (combustion turbine generator) can power these buses if power from all other sources is lost. e 4 m.. sma,.
PRELIMINARY: RESPONSE TO ITEM 6: PART 1: YES. This system is'not considered necessary if realistic source terms are used. . PART 2: NO. Palo Verde is a reinfrced concrete containment. PART 3: The AVS can be helpful in removing fission-products following a severe accident where the containment remains intact'. This feature was not previously credited in the PRA. PART 4: Yes. (?) h
- - y
PRELIMINARI--RESPONSE TO ITEM-7
-The relationship is very close.
- 1. CFS is indirectly replenished by; sprays- .-
- 2. without sprays cont. - will fail even-if corium or coro is-cooled 0
4 O i l 1 1 1 i ci i
j s, . > l t PRELTHINARY REsponsg To 27gg a 1 SEE ATTACHED SilECT FOR HRA ISSUES ASSOCIATED WITH CFS ACTUATION i
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t i i ti;JTCFSMOVS: OPERATOR FAILS TO INITIATE CAVITY FLOODING SYSTEM A severe design basis event has occurred. After monitoring core level, temperature and pressure indicators and realizing that regardless of corrective actions taken, reactor vessel rupture is likely to occur, the operator must initiate the Cavity Flooding System (CFS). The CFS is used to
~
quench the debris beds in the reactor cavity by injecting water into the reactor cavity to flood the cavity. The source of this water is the In-Containment Refueling Water Storage Tank (IRWST), whleh delivers water first to the Holdup Volume Tank (HVT) and then to the reactor cavity. The operator must initiate this process. There are four pathways (called " spillways") from the IRWST to the HVT. Each spillway contains one manual isolation valve which is normally open, and one motor-operated valve which is normally closed. There are two spillwriys from the HVT to the reactor cavity. Each spillway contains one motur-operated valve which is normally closed. For this analysis, it is assumed that there is a procedure that requires that all the motor-operated valves be opened at the same time. In order to initiate the Cavity Flooding System, the' following actions must occur:
- 1. Operator must recognize the symptoms of the onset of core damage. These:
include: loss of water level above the core which is monitored by the Reactor Vessel Level Monitoring System (RVLMS), superheated steam temperatures at the core exit which is monitored by the Core Exit Thermocouples (CETs), and core voidage which can be determined by a review of the Self Powered Neutron Detectors (SPNDs).
- 2. Operator must open HVT spillway motor-operated valves SI 390, SI 391, SI 392, and SI 393. This is done from the control room.
- 3. Operator must open reactor cavity spillway motor-operated valves SI 394 and SI-395. This is done from the control room.
For the purpose of this analysis, it is assumed that the operator must initiate Cavity Flooding before vessel failure. This gives the operator approximately 1 hour from the time that the initial indications described in item #1 occur. The time required to open the valves is less than 5 minutes. The stress level is considered to be high. It is also assumed that at least 2 Senior Reactor Operators (SROs) and 1 Shift Technical Advisor (STA) are in the control room. The inclusion of this event in the fault tree model represents failure of the operator to open motor-operated valves SI 390, SI-391, SI-392, SI 393, SI 394, and SI 395 nom the control room upon recognizing the need to actuate the Cavity Flooding System. . - ~___,_ _ __ _ _ _ _ _ _ _ _ . _ . _ . _. .
.. _._.._ _. _ _ - , .._. _ _ _ . _ . . . . . . . . ...____ m.___ _ _ _ _ _ _ _ _ _ . _._. ._
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A a e i 1 e PRELIMINARY RESPONSE TO ITEMS 9,10 AND 11 l 6 h t
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System 80+ Cavity Flood System Cavity Height vs. Time 17.0 -- 16.0 - N 15.0 { 14.0 - , 13.0 , 12.0 : Holdup Volume T 11.0 -~
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l l 4 PRELIMINARY RESPONSE TO ITEM 12 I Section 4.3.2 discusses scrubbing capability of a water pool. An expanded discussion of the pool scrubbing feature can be added to ! the report.
. . . . _ . - _ _ . - . ~ _ _ - . --.-. . . ._. . - . - - ...-.. .- . _ . . . - . . . - _ , . . - - .
. 4 , PRELIMINARY RESPONSE TO ITEM 13 SEE RESPONSE TO ITEM 8: i There are no serious consequences of early CFS actuation, j O
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I PRELIMIHARY RESPONSE TO ITEM 14
~ Water in the reactor cavity prior to breach is desirable. However, ,
the system serves a similar purpose even if the CFS actuation is i delayed until after.VB.
- PRELIMINARY RESPONSE TO ITEM 15 .
If CPS steaming cannot be condensed the containment will ultimately. ; reach failure pressure. PRELIMINARY RESPONSE TO ITEM 16 - Discussion. i PRELIMINARY RESPONSE TO ITEMS 17/18 DCll threshold is about 250 to 350 psig. Soo DOE /ID-10271 . For a typical TLOFW event it will take about i hour to depressurize I to 250 psig ,
- RV failure time for this sequence will be about 4 hours.
Opening of SDS according to EOPs will afford sufficient time to establish low pressure conditions in the RCS prior to VD l y S m.. , . [ . m., ,. ,-,,r...,e'. - , , , , , -,, - - . , , , a m_ rwy , m W, te m ve-- v - M e- v- m W -"wMpm v t 1-v = -
- W etvi-------='w=--wW-F vy C - Wr " '*w---*
_._-_._._..__._.._.___.._._..___.__.___m ..- i e i f PRELIMINARY RESPONSE TO ITEM 19 i EXCERPT FROM CESSAR -DC PAGE 8.3 $ B.3.2.1.2.1.2 125V DC Vital Instrumentation and Control Power Batteries , 1 Each of the independent load group channels and divisions of-125 Volt DC Vital Instrumentation and Control Power is provided ' with a separate and independent 125 volt' battery. Each battery, is sized to supply the continuous emergency load : of its own load group for a period of 4 hours.- In addition,- the batteries provide a SBo coping capability which, assuming manual load shedding or the use of-load management programs, exceeds 4 hours and, as a minimum, permits . operating the 1 instrumentation and control loads associated with the-turbine-driven emergency feedwater-pumps for 8-hours.
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ITEM 20 DISCUSSIOli
-RESPollSE TO ITEM 21 YES.
PRELTHIliARY RESPollSE TO ITEM 27 Discuccion d
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PRELIMINARY RESPONSE TO ITEM 23 e SEE ATTACHED. ,_- 4
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PRELIMINARY RESPONSE TO ITEM 24 i t I e i SEE AT-IACHED. e ; a 1,
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t e ITEM 25 DISCUSSION b i d -
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- 4 I
e RESPONSE TO ITEM 26 SYSTEM 80+ IS DESIGNED FOR CONSTRUCTION WITH EITHER BASALTIC
- OR -!
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. g-PRELIMINARY RESPONSE TO ITEM 31 SEE ATTACllED FIGURE ,
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SCS TRA!N 1. i k PUMP 1 FIGURE : CONCEPTUAL DESIGN OF EMERGENCY - CONTAINMENT SPRAY BACKUP SYSTEM r ; r
. _ .- _. .. _ . . . . . . _ _ . - . . _ . _ . . _ . _ _ . . . . _ _ _ . . _ . . _ m_. . . . _ . . . . . _. . . .
h l PRELIMINARY RESPONSE TO ITEM'37 Ti1IS . CALCULATION IS APPLICABLE TO Tl!E SYSTEM 80+ DESIGN. SEE f ATTAcilED SHEET FOR DETAILG. WORK IS FROM APPENDIX A OF 1>0E/ID-10271. I g i t 4 1 9 S L i 9
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t A.0 DESCRIPTION OF BOUM0 LNG AMALYSES , The assumptions regarding the bounding analyses are described in Section 2.1 of the main report. This appendix provides a description.of t'he methodology and printouts of the results, which were obtained on a Lotus 123 spreadsheet. A.1 ligthodolooy A.I.1 Dry Case This case addressed the situation where no consideration was given to any liquid water in the RCS or the reactor cavity. (Steam in the RCS was assumed to be present). All the heat from the cooldown of the core debris materials l and from chemical reactions was assumed to be transferred to the containment atmosphere. The initial composition of the containment atmosphere was determined by l starting with the pressure and temperatur& obtained from the MAAP code 'for the station blackout sequence, just prior to reactor vessel meltthrough. ,The initial amounts of oxygen and nitrogen were detensined assuming a temoerature of 80F prior to the start of the accident sequence. The amount of water , vapor in the containment was estimated by taking the difference between the total pressure and the sum of the partial pressures of oxygen and nitrogen in containment at the point in the sequence just before meltthrough. The heat sources were computed by summing the sensible heat of cooling from the initial temperature of the melt to the final temperature'of the core materials, and adding the heats of chemical reactions. The sources of the-sensible, or stored, heat were the fuel, the zirconium oxide from the in vessel oxidation fraction, the unoxidized zirconius, and the steel. The l heats of reaction of zirconium oxidation in steam to form hydrogen ~and of L subsequent hydrogen burning in oxygen were considered. 07.idation of the , steel was not considered in this case because of the assumed absence of water . A-1
_ . . . - _ _ _ __ _ _ . _ _ _ _ _ . ~. _ _-__ _ i available to oxidize it to any great extent. It was assumed that the energy from the in core fraction of the zirconium oxidation contributes to the initial debris temperature, taken as 2533 K, and is not counted as an additional heat source. The heat sinks were considered to be the nitrogen ! and water vapor initially in the containment and the oxygen remaining.afte'r the hydrogen burn. The process was assumed to be a constant volume process in thermodynamic equilibrium. The final temperature was obtained by a trial and error solution as the - temperature where the heat sinks equal the heat sources. The final pressure was obtained by suming the partial pressures of the constituents in the gas , phase at the final temperature. A.I.2 Cases With Vater Present Two cases were analyzed where water was present: one where water and steam in the RCS were assumed to codisperse~with the core debris at the time of meltthrough, but where no water existed in the reactor cavity; and another identical case, except that 227,000 kg of water were assumed to exist in the reactor cavity at the time of meltthrough. In each case, the RCS water was assumed to be the liquid in the lower head and the steam in the remaining,RCS volume. Water in the pressurizer was not considered because it would folfow , later in the depressurization sequence. The heat sources were identical to those of the dry case. except that the oxidation of steel was allowed. The heat sinks were also identical, with the addition of the effect of the additional water. If the heat balance indicated that liquid water remained in the system at the equilibrium condition, saturation properties were used to determine the final temperature , and internal energies. If the heat balance indicated that no water remained, a separate area of the spreadsheet was used to calculate the final condition, using superheat properties that were calculated' starting with the saturation properties at the given pressure. The_ solution was obtained by assuming an- . initial temperature, which affected the pressure, which affected the steam A-2
.c y.-
, - . ,, +.3 .
_ . . , - _ . . , . . . -,y__ . _ , , , , - . . -
- _ _ _ _ . - . . - .. . - . . - . . . .... - - - - . . = - . - ._ .- . ..- -.. -.-
i saturation properties. These were used to compute the heat sources and heat sinks and interated until a convergent solut'9n was reached. A.2 Propertieg , i The thermodynamic properties were as follows: A.2.1 Properties . (Values at 1500K, unless otherwise stated, from MAAP User's Guide).Al o Cy 002 " 333 J/X9"K o Cy Zr - 356 J/Kg K o .Cy 02 - 660 J/Kg-K $ o Cy steam 1760 J/Kg-KA2 o Cy Zr02 - 645 J/Kg K o Cy N2 - 750 J/Kg K o Cy Steel - 663 J/kg - k o Heat of zirconium reaction with water - 5.51E8 J/Kg-mol Zrt2 (at 2500 K) o Heat of H oxidation'= 2.40E8 J/Kg-mol H 2O (at 2500 K) . , o Heat of iron oxidation - 4.113E8 J/kg mol Fe203 o Steam properties from Reference A2 and A4 A-3
- A.2.2 Physical Parameters o Initial containment temperature - 340 K
- o Mass 002 in core = 112,000 Kg ,
o Mass zirconium in core - 33,000 Kg o Mass steel in ejection - 10,000 Kg - o Containment volume - 3.3E6 ft.3 - 99,400 m3
/*
o Mass of water in reactor vessel bottom head 4 43,3 kg 9o . o Mass of wate$ vapor in RCS and}2003)st d 9,800 M . A.3 Calculations Performed 4
~
The spreadsheet program has the capability of varying the following: Core debris initial temperature , containment initial temperature Fraction of core ejected-
- Amount of steel ejected Fraction of ex vessel hydrogen reacting Fraction of steel oxidizing Initial amounts in containment .
Amount of water in the cavity. Amaunt of water in the reactor vessel Amount of steam in the reactor vessel. The term " fraction of core ejected" means the' fraction of the core wht'ch
.is participating in the energy transfer to the containment atmosphere. The
. core could be ejected from the vessel, but its participation-in'further A-4 l
.,,....,n.., , , , , , . . . . . , . . . . _ . . . . . . . . . . . _ _ . _ . . . . . . . . . . . . . . . . . .. . . . _ . . .
,,..,,,.,,,..,....._......--)
energy transfer could be inhibited by the cavity configuration, access to reactants, particle dynamics, reentrainment, and other factors identified in more depth in other parts of this re; ort. The three cases discussed in Section 2.1 of the main report were run to. varying fractions of the core, ejected to illustrate sensitivity to the perfo Mnce of the reactor cavity. The other parameters were set at best estimate values tonsistent with other analyses. 4 A.4 Result 1, Table Al shows the results for the dry case. Table A2 for the case where RCS water codisperses with the core debris during the high pressure melt ejection. Table A3 is similar to Table A2, except 227,000 Kg water was , assumed to be present in the reactor cavity. These results are sumarized in Figure 1 of the main report. ,
- s. .
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4 4 A5
,_ __ ~ _ . - - . _ . . - . . . _ _ _ _ -- _
Table A-1 Case 1. Contairment Prc Curo vs. Cora Fraction CDC10A3 ee depw sl P w MM304 COff.10000 823 STtfi N 79COtOOK. CDffMweJT Ptn5Urf CMCLLAl)ON. NO CAufY WAftR. IC5 ETEN4 CMY, VARY CCtf it.ACTOJ 2533r. 245fs2s. 8%H 8essimo Cel De est one . no 6 band Th D3 2$23 2133 2533 2333 2533 2333 1533 T aweem =v eind n M 150 350 350 350 350 350 350 Faraenelone o.J l 1 0.8 0.6 0.4 0.2 0 Am ww el m.J &gf 10000 10000 10000 10000 10000 10000 10000 f=suenelZbseammeJ h===J 0.245 0.24L 0.2 45 0.245 0.245 0.2 45 0.245 Es vs J 6.piroy.n. Genman el ames t 'l 1 1 1 1 .~ l Fammeele i =&ame 1 0 0 0 0 0 0
- 4T14 AMotetts ed CD4towa4f M Aan d O2 . glam42N/TT 8:9.nr4 719 719 719 719 719 719 719 MJe d N2 . pWrN.t#/rf Eg J 2877 2877 2877 2877 2877 2871 2877 Maine el H2O W. so wear 313 313 313 313 313 313 313 We u amer. Eg 217000 0 0 0 0 0 0 frecaum al ca.*r *** Parmapass=3 I l l I i 1 +1 WAftR N DC5 WLs=e = (4/31pi's3/21,=31 43.31982 43.31982 43.31982 43.31982 43.31982 43.31982 43.31982 Man u,ad Vd'1000rg/m*3 (Lc) 43319.42 .0' O O O O 'O c ess DC5 ==ser e 13.61 Mee UAd 1672701. 1672701, 1672701. 1672701. 1672701. 1672701. 1672701.
T ser DC3 waner e 2000 pne D3 60L8888 60L8888 608.8888 608 8888 608.8888 608.8888 608.8888 Mme ICS = aper Egl . 29800 29800 29800 29800 29800 29800 29800 Ce*r mener b.4ag pene K 435 447.7777 463.4917 457.3133 448.3032 435.9892 0 e se tG =ener a 2003 pne tungl 2337000 2257tXX3 2357000 2157000 2357000 2357000 2357000
- HEAT SOUDCE5 Ita.e dua's mandi AI es-eadJimes Irwa 6 ro eni Coulde d102 w 6e r se Ibd 7.8E e10 .$.0E+09 3.5E.09 9.4 09 1.21 10 9.2E +09 0 Ceni he v en.J Ic2 as le ==r so T I 1.5E+10 9.5E.08 6.6E .08 1AE.09 2.21 09 1.7E 09 0 Casi be ww.ao.J ZiO2 se 6e == se 11 4.5E+10 2.9E.07 2.0E.09 5.4 09 6 8E.09 5.3E.09 0
- Csul e.el . .m 1.4E+10 9.0E.08 6.3E.08 1.7E.09 2.1E.09 14E.09 0 Hee el e.momm. 6.in8y wr oo.J Zr 1.5E.1i ' .5E + 1.1 1.2E.11 9.0E + 10 6.0E+10 3.0E.10 0 Hee ede a man H W = > - 1.7t+11 G .7E.1M 1.5E+11 1.2E.11 9.5E+10 6.9E+10 4.3E+10 ftetniemenfe% Hem H2O 7.4E.10 0) 0 0 0 0 0 CmJ tOO be ince4o. shaly h O bl 6.5E.10 3.l E.09 2.3E.09 6.71 09 9.8E 09 1.lE+10 1.0E+10 *
.Asa d hees eawc e 6.lE +11 3.lE.l l 2.8E+11 2.4E.11 1.9E+11 1.3E+11 5.1E.10 . 255 fnal 6enp D10,pl e m. 7f 450 2668 2415 2114 1751 1306 747 .Ad c kmd ennJawe 80F 0/Kgl 111385.8 111315.8 11131L5.8 Ii1385.8 111315.8 111385.8 111385.8 ed.,mA es 6d om4 man K pue 677037.9 814278.4 796332.7 770464.0 732738.4 681179.6 591540.2 einer wped a le=d paa U/rg) 2364164. 2583290. 2582617. 1578548. 2372905. 2545004. 2544497 Ses ==pw e;6 w L e 6d p e bm3Agl 0.294103 0.142575 0.155437 0.178173 0.215537 0.287587 0.493994 Mme H2O besmay h wapar M M i865L8 1845L8 16692.05 14725.31 12758.56 10791.82 4825.076 Men r p 6= crivy 217000 0 . 0 0 0 0 0 Man =ee p 6 tC3 (nd ha vapi 73119.82 29800 29800 29800 29800 29800 29800 1end meer apras.J 290119.8 29800 29800 29800 29800 29800 29800 e ind e T.p tnal. ukg 2443125. 6147811. 5730450. 3232601. 4634116, 3901676. 2981354 U pas el weer vapur no find 5.4E.1I l.d+11 1.2E + 11 1,0E+11 L21 10 54E.10 2.3E.10 H3 MOtt CAV!TY WATEf, HO ENitY T 450 2668 2415 1114 1751 1306 747 MXIDCrut TEAT 5t*5 Heeme d O2 1.5E.09 3.5E 10 3.lE+10 2.7E.10 2.lE.10 1.5E.10 6.0E.09 Huaup el N2 6.0E+09 1.4E+11 1.2E + 11 1.lE+11 L5E.10 1.AE+10 2.4 10 Suasi k ma=4.Cr=1 U u H2Ol 5.5E.11 3.lE+11 2.81 11 2.4E +11 1.9E+11 1.3E+11 5.3E+10 Haaremwoessannaherses4: 4.4E+10 94813&58 1.1 E.C8 .44E+07 6170445213864909 21909463 Male O2 leli eher earm6 man 222.9117 357.2 411.8318 466.4636 521.0954 575.7272 630.359 a ,8 810 4802.4 4347 3805.2 3151.8 2350L8 1344.6
- f9et PEE 55Uti AT Waft 181.
P N2 EngIJi vnes p. 16.66&&1 98.82638 89.45492 78.30546 64.85944 48.37603 27.66990 7 021slrskerce4mnum 1.291494 12.26999 12.80513 12.69608 11J4763 9.680708 6.062556
* *CO sw. 99.38811 92.47688 80.31018 67.32659 53.30253 37.91896 2043786 L Pwm TOTAL P3 117.3442 203.5732 182.5702 15L3281 129.9096 95.97570 54.37032 Abw Had poem. <so r4 es te 179 G.== pasmsw. 0 0 0 0 0 0 Im/T Gw.s e,mr6s f e M 450 2648 2415 2114 1751 1306 747 Hee annow *=== 6e w4a L4E+10 94813&J8 .l.lE 08 4.8E.07 61708452 53864909.21909863 A-6 a
'DJH.I A-2 Case 2. Ccntalmeric Pressure vs. Caso PNtion ;
COPC6A3 ce aime met P *e itACTIOrJ CCat, 10000 823 Eftf1 D 7803tOOA. CorRAFMf4T Pet 15Uut CALCLLATCN. to CAWTY WAftf. AC5 Warts #40 $7EAM Vuy coat f8ACDON 2533K. 245hZr,8%H 8*=lne Cal De asi.e 1surelas had Th #3 1533 2333 2333 2133 2533 1533 2533 ie ,e cm.J fi p1 350 350 350 350 350 350 F,.: el cwe e,.oed 250 1 1 0S 06 0.4 0.2 0 Ame.v el i Agi 10000 10000 10000 10003 10003 10000 10000 e ae dtn esea.J h I 0.245 c 245 1145 0 245 0.245 0.245 0.24s E: I t ydroyen, Iraonen d man I i feao.m4an.d 1 i 1 1 1 -I t imo 1 1 1 1 1 1
#41W AMOUNTS N COr4A&MNT Molis el O2 . pbesK.2N/If r.p.el Fl9 719 719 719 719 719 719 Molen el N2 e p(awl (.8N/rf Kgmal 2877 2877 2877 2877 2877 2877 2877 Males J H2O bar en waper 313 313 313 313 313 313 313 Wow wca y,K 217000 0 0 0 0 0 .O freoum el cowy est =gparoupueng i i i i i i I Waft 8 N #C5 W&,,ne e (4/3)rd*t3/2 (m3s 4131982 43.31982 43.31982 43.31982 43.31982 43.31982 43.31982 Men ue d. Vel'1000Kg/ni*3 (lg! 43319.82 43319.82 43319.82 43319.82 43319.82 43319.82 43319.82 o ser RC5 = war a 13.6l MPs U/kg) 1672701. 1672701. 1672701, 1672701. 1672701, 1672701. 1672701, f ao RC5 weer a 2000 p=a S1 608.8888 60L8888 608 8888 608.8888 60L8888 608.8888 60L4888 Mom AC5 wepar Ad . 29800 29800 29800 29800 29800 29000 29800 Co.sy waner I,aleg p e K 435 467.7777 467.7777 462.8696 4521 848 437.9658 0 o ser IC5 waper a 2000 p=e (l/kg) 2357000 2357C00 2257000 2357000 2257000 2357000 2257000 HEAT SoueCES li si+n't mmch Al spr oM s Irorn i= , =J ,
CeeWe=si el U02 to Iwer meer w final 7.8t+l0 9.5E+09 1.5E+10 1.8t +10 1.7t+10 l.2f+10 0 Coal me reso d ZsO2 ee le ws to f f 1.5E+10 1.8E+09 19E+09 34 +09 3.3E +09 2.2f+09 0 Cool he ww.ooed ZsO2 w le wm no il 4.5E+10 15E.09 L9E+09 1.1E+l0 1.0E+10 6 4+09 0 Coal see.1 1.4E+10 1JE+09 21E+09 3.2E+09 11E+09 2.lE+09 0 . Hee d emeaum he dr eveeo d Zi 1.5E+11 1.5E+11 1.2E+11 9.0E+10 6.0E+10 3.0E+10 0 Heer el eeoces ,H nwn 1JEell I.7t+1I l.5E+l1 1.2E+11 9,5E+10 L9E+10 4.3E+10 t9 4 seccarm fe se H to H2O 74+10 74 +10 5.9E+10 44+10 2.9E+10 1.5E+10 0 Casi H2O 6 buedre, ear.edy n Q lan 4 6.5E+lo 10E.09 f.3E+10 1.7[+ l0 1.9E+10 1,7t+10 1.2E+10 ine el hear easue 6.lE+11 4.2f+11 2JE+11 3.lt+11 2.4E +11 1.5E+11 5.4E+10 VUE15 Inal e. nip 41 C,vne e es. 771 450 2279 2020 1720 1372 945 481 4
.A $ c hmai candeums tCF 0/Kgl 111315.8 111385.8 111385.8 111315.8 111315.8 111385.8 181345.8 edben a naal cadman x p.s 677037.9 814278.4 814278 4 793727.9 751503.0 689454 0 579060.0 4 eteer =aparl er Irel p.ae OAg) 2564164. 25E3290. 2583290. 2582155. 2576020. 236458L 2541284.
Ser waper op, sol, e Grad pne Im3/lg) 0.294100 0.142575 0.142575 O.157250 0.196300 0,274467 0J315I9 Man H2O hee 8y h vapor + b bwa 18&58.8 1845L8 1649105 1472131 12758J6 10791.82 8825.076 Mass =ueer e.up 6 eawr 217000 0 0 0 0 0 0 Men www ap Gem eC5 Ind ha vapi 73119.82 73119.82 73119.82 73119.82 73119.82 73119.82 73119.82 Tonal weer m d 290119.8 73119.82 73119.82 73119.82 73119.82 73119.82 73119.82 o baal c T.p Laal.1Ag 2443121 5496044 5062090. 4566583, 3994562. 3328478. 2537605. U gen eB =ew woow oo it =4 54+11 laE+11 14E+11 2.0E+ll I.6E+ll 1.lE+11 A4E+10 NO MCSE CAWTY WATER, PC ENTRY
- f. 450 2279 2020 1720 1372 945 ASI ACCrTICNAL TEAT 5NK1 Howup el 02 1.JE+09 2.9E+10 2.5E+10 2.1E 10 1.6E+10 9.3E+09 2.OE+09 Heeup d N2 ACE +09 1.2E+11 1.0E+11 L3E+10 4.2E+l0 3JE+10 7.9E+09 Swe d lem eds ised. U en tCol 3.5E+ll 4.2E+11 SJE+11 3.lE+11 2.4E.11 1.5E 11 5.4E+10 --
. H , ,e.e l e anne 6 4 +10 1.3E+08 1.9E+08 5.86+07 1.2E+08 13484289 19915827 Males 02 b& eke cenenamen 222.9117 222.9117 304.4011 3&$,8906 467.3803 548.8695 430.359 6.n, t 810 41012 3436 3096 2469.6 1737 845.8 fwl f9E55URE Al WATEt 87.
P N2 fngle inse gui 16.44861 84ei/29 74.82357 63.71116 50.82076 35.74492 17.81690 P On 1.6 che cwnbusean 1.291494 4.540703 7.916713 8.545548 L256035 6.819351 3.903734
'
- H2O man 99.38811 149,5099 1291656 108.0739 84.27771 57.91942 2L19302 W Pet 15URE fotAL PSI 117.3482 240.5679 212.5059 180.3306 143.3545 100.4837 49.91366 '
Almer ow.al guma. cJ eJ ee 8 179 Cw.m presas, 0 0 0 0 0 0 97UT C mapwhmsfan ,&3 450 2279 2020 1720 1372 965 481 Hee emme e-ma Iww ask 64+10 1.3E+08 1.9E+08 .$.8E +07 1.2E+08 13484289 19915827 A-7
i
. . Table A-3 Contalrument Pr:ssuro vs. Coro Fraction OPC7A4 en deperud. P w PtACTON Cott.10000 rc sitti Ci pvCosEKK ,
ONTAemeNT Mt.15Urf C4411ATO4, CAWY WAftt. DC5 WAftf #4 ETEA% VMY Coat PLA4.*fCN l 333K. 245FiZs. 9%H iws heel Th D3 2533 mm nore vend fl M 230
- wasse d eere e oed I 10000 sa.we d eeel ag) d h weal unoselTotroece 02L5 a womed 4.ydreyse, beaean ud seras I
- ocs e. el e.es -- &
- e t .
atW SMOUNT1 N CCHTMsD4f Moles el O2 . #d owK2N/IT KgW 719 M.A. d N2 . ,W e48u/rf Kgwiel 2a77 Ma&m al M20 ew eswepw 83 5 313 1 Weser b wy, r4 227003 fnoma el scr.ory weer p- , a 4:llt N EC1 < l elwas o 14/31se's3/2 (m31 43.31982 ' was (kom4. Val' 1000ry.e*3 Oual 43319.82
- m #C3 mese w 1341 MPe (l/kgl 1672701.
.ar #C5 ==er a 2(C0 pas #3 40&8888 ten GC5 waper Ag) . 29800 Cawy weer loo &ng prive K 45010Cr7 ,
em #C5 wepar a 2003 pao (J/he) 2257000
- EAT 50UPCES Iteeg don's M Al eelJe== 4 UO2 to hear weer w 7,6telo end ow c.oo.d ZrO2 se le =w w f f 1.4 10
- eul 6w ww.soed ZsO2 is le war se TI 44 10 eni seul 1,3E*10 ees ele echen, haie8y w d Zi 1.5E+1I
.er d e,ecnon H Inurn 1.7t+1i e el soucmen fe w H se H2O 7.d.10 '
eel H70 hose lawrvdNo, sir edy b O innd 0 Adelo iwa es hans eeweces AIE+1I UE15 lensi semp M fined a s w 77) SOS
~
.taqi o houd ecedawe SCF (uYg) Ii1383.8 -
w a I wl eaaa e x p.e 74a265.7 2573892.
'ser o ==per wepart an. =alceInal p=e ll/rg)(=3/Ld inal paw 0.207495 nam H2O ewiafy h vapor Gene lean 18654.8 '
Ines wave eveo Gorn cowy 227000 nous w ewp frorn RC5 lead inir wapi 73119.82-med muser ewepero.d 300119.8 Sad a f p Inal, J/kg 2544377. pen et www waper w fi ud A0E.Ii , o Mote CAttfY WATE1, to Et4 fry 505 DOfflONAL HEAT 58*5 4 eme d 02 2.4 +09 4=ene d N2 9.4 09 wa 4 k a e 4. Ind. U im H201 ale +11 - e onwon sar has enke -9568325 wins 021.Jr einer - ' - 222.9117 van E 909 FW PRE 150tt Af WATIt 8f, 2 ' N2 Engish w=e pas 1810589 , ' 8 02 leli e&w d _ _ l.449344 *
- 8 H2O mee . Ii1.1477 t#& PRE 15utt TGTAt P3 115.3029 kw kw at gumen, col ad u tow 179 kaas sn=enn. O M Queen m4.arhead Isenp % $O5 enew = h 4. .956a325 k
A .;
~ . - -.
Table A-3. (Continued) CDPC7A4 foC1 we 6 4 s. ' - ' P P84ccN Cott.10000 u3 511EL tt 79CDeDor. Go 10 K1 tot 8ESLLIS C3t4A#M4 79255URE CA40.A,ATON, CAWV WAltf. DC$ WATE8 NO $ TEAM DO iOf PMT COLUW4. *TIP COL 8 A5 SCEJKZ. DQ PCT *MCM* CCRIM4 g CAWY Dem.5 C JT, LLE LPEET K1 DOUf94. Case == seios se .2 pel J33E. 245fdt,swi f onei.e w .4 Th M 2333 2533 1533 1533 2533 1 areennwe . d is M 350 350 350 350 350 Fase elare epond as 04 Q4 0.2 0 A no el e J lbgl 10000 10003 10000 1CO30 10000 f=amm el Zet e.ao J 6i ===J 0.245 0.245 0.245 0.245 . 1.245 Em J 1.y4ws 6 :nas d ease 1 1 1 1 1 fmenna el e.d e 4rnig i I 1 I I e4ftat 4 Mourns N Ccur4sw.NT Make 4 O2 pdaad.2N/rf 9't.al 719 719 719 719 719 AWee al N2 . p4mzK aN/87 Kg d 2877 2L77 2877 2E77 2877 Mole el H2O he as ==,we 313 313 313 313 313 Wow m amy, K 227000 227000 227000 227000 227000 f=omm el ar.wr='ar *g '"" F - ~ as 1 I I I I waftstoPC3 vaLem . id/3)ps'e3/2 Irn3) 43.31982 43.31982 43.31982 43.38982 43.31982 A%e WM . Vd* 1000rg/m*3 (Lgl 43319.82 43319.82 43319.82 43319 82 43319.42 e w tcs w w 1361 MPs0/Lgl 1672701. 1672701. 1672701, 1672701. 1672101, f se PC3 =ow en 2000 p e N outA888 60L8888 60L4468 608 8888 60L88E8 Mass ICS wapse (kg) . 29400 29803 29000 29800 29800
- Co.vy .nsw in=1.ng prom K 4219 417 401.1 392 374 o we PCs waner as 2000 paa UAgl 2475 COO 2475000 2473000 2475000 2475000
- HEAT SOURCIS C. Ale.= d (C2 k low waar se flad . L3E 10 41E+10 3.2E+10 14 10 0 Cad W esao.J ZeO2 te le wee e I f 1.2f+l0 L9f 09 LOE.09 3.M.09 0 Cad W o.d ZrO2 6. le e as il 3JE+10 2.aE+10 1.aE+10 9.3E.09 0
- Ceil m.d 1.lf.10 84 09 5.6E.09 2.8E.09 0 Hasele m bumelv.ar oo JZr 1.2f+11 9.We10 6.0E+10 30E+10 0 Hear el ewsomm H lann 1.5E*1I l.*E+11 9.5E+10 6.9E.10 4.3E+10 Hrafe==manF seHse H2O 19E.10 4 4 +10 2.ml0 1.5E.10 0 3R+ 10 ARel0 34+10 ' 2.3E.10 1.21+10
( "ed H2O 6 kmen
== el beer ooweni 3.0E+lI 3.9E+11 2E+11 1.7t+11 3R.10 GUE15 l wd 6w, k3 (een a re- 771 4219 417 405.1 392 374 Emreweed lewd wpse pr= n= Mee 1503552 0.394157 0.282701 0.190476 0.107123 Em=ww.d I =d vap pris Ismil 1402222 58 41.15714 28 15J471.2 alkgl er wid wwk=ws SOF 0/Kg) I11381 8 111315 8 111345.8 111385.8 Ii1385 8 vf4adl e lewd and am a p.e 63893L2 601671.9 1516464 496996.9 421630.9 i 215684L 254711L 2535681, 2521639. 2500610.
wlms Lw wapar wpc.) op, wt a le=d p e U/rd(m3Agl at keul 0.368103 0.443712 E634534 a932654 1.735980 Mas vap. er I.=4 pm (Lg) p.e 270032.9 214354.9 156&50.3 106577.4 5725810 Mas H2O bwwdy n = aper + frwa lanii 20159.15 1762193 14692.31 1175849 8825.076 Man =ce e=ap 6 envey & r.y Anard 249473.3 196731.0 141957.9 94814.78 44433.62
%=a www e===w.g. aw + h.3 ( elryl ' SC646.45 1033681 158161J 201301.0 251686.8 Ce==manfarACs 14+11 ' 14+11 1 4+11 1.4E+11 1 A+11 Csmosan se acs amav +
U , a et weer (5 ' to few 3.0E+11 3.9E+11 2.aE+11 IJE+11 3R.10 f == N 4219 417 405.1 392 374 A00mCNAL HE.Af st*3 Heeup of 02 1.2E.09 1.0E,09 14 08 A4E.08 3.6E.08 Hwupd N2 44 09 4 m 09 32 09 2R.09 1.5E.09 Hean,d H2O w h u m_ a 7.5E+08 - A6t+08 3 2 ,08 4.2E+04 2.4E.08 5.= ad 6m aras (,wi U see H2Cl , 3.1E+11 (CE+ll 2.aE+l l 1JE*11 5.8E.10 H.oremn w 6.e sed. --Laf.08 -t m ot 43E.07 5.4E.09 3.1E.09 Mel.:02 IJi e/w credmanen 304 4011 385.8906 467.3800 544.8695 630.259 6.=p 8 76&o2 750A 729.18 705.6 673.2 ft4*t Pet 15Uri Af ft4A1 f P N2 Engluh wws psi 15.77592 15 44625 15D0146 14.52021 13.85347 P O2telielw caredunaa l.669172 2.071798 2.437694 1770144 3.035336
' H2O asa (Poi 7A02222 58 41.15714 28 1514712 rt4*t F1tiSut! TOTAL P2 91.44731 75.51805 39.00030 45.29036 32.63593 Origraf gw m 6.wp i*1 4 4219 417 4011 392 374 tt.seom,eo ww.e6.ceords .LaE.08 4 9E.09 42 07 5 4 09 3.lE.09 '
A.5 REFEREHCES t A1. Fauske and Associates, Inc., Technical Report 16.2 3,' HAAP (3.0) Modular Accident Analysis Program User's Manual - Vol !!, Atomic Industrial Forum, Bethesda, to, February, 1987. ' A2. E. A. Avallone, and T. Baumeister, Eds., Marks' Standard Handbook.for Mechanical Engineers, Ninth Edition, McGraw Hill Book Company, Rew York, 1978. A3. R. H. Pcrry, et al., Eds., Chemical Engineers' Handbook, Fourth Edition, McGraw Hill Book Company, New York,1963. A4. J. H. Keenan and F. G. Keyes, Thermodynamic Properties of Steam, John Wiley and Cons, New York, 1936. e [ t e 4 4 4 e t k A 10 s
...s
. 4 L PRELIMINARY RESWriSE TO ITEM ~39 S' c , explosions are millisecond phenomena and are typically
. triggered once-a mass of corium reaches a solid surface (floor).
nince the tV :al failure mode of the RV -is via f ailure of amall-
'nstrrt
. *ations, the loading associated with any given steam ar . ~ I' be that associated with the mass in the water
' ool fron i
.r" of corium with the diameter equal to the ICI ube diate.s - .3 . the depth of the pool. For a 1" diameter tube s
ad a 14 foot r' the volume of corium contributing to the EVSE
. auld be .0764 This le equivalent to a mass of 41.9 lbm or 19 kg.
RLSPONSE TO ITEi4 4 0 q BETA V6.1 information was unavailable. PRELIMINARY RESPONSE TO ITEM 41 The containmnet strength qm' ri in the document !( epresentative of cavity wall stronghts t) r n .1 of recent C-E (' ,aed PWRs. PLANT Cavity Wall Design Strength WSES 240 psid SCE 229 psid Millstone 2 247 psid
PRELIMINARY-RESPONSE TO ITEM 43
'The RV lower head failure map of NUREG/CR-5642 demonstrates that .
- the most likely mode - of : RV failure will be - caused by = either - 1 instrument tube ejection or tube rupture. These maps are gengrally 1 applicable to the System 80+ lower head. See attached sheet. l l
i
-)
I e 1 t s [ t - .. }
2,800 . . . - - - - - ' Lower h%d globd rupturo Oow2r 6md)
~
2 2,600 \ ----Tube e3ect,on power smit)
~
y 2,400 -
, Tube rupture Oower hmn)
Tube ejection (upper tenn) ( N 2,200 -
~
O [ 2,000 - ' e , g 1,800 - *' a 3 1,600 - 1,000 - 800 0 2 4 6 8 10 12 14 10 18 20 System pressure (MPa) . mn- a.w.cs Figure 4-28. Westinghouse instrument guide tube failure map at maximum radial gap (0.008 cm) 2,000 . , , , , ,
, (
2 2,600 - Lower head global rupture power hmn) ) w ----Tube ejection Cower Smit) e 2,400 - - -Tute rupture cower smit) .
$ * * * * * -Tube ejeeben (upper !anit) g 2,200 -
$. 2,000 .
E o 1,800 - 5 1,600 . 0 . 0 2 4 6 8 10 12 14 16 18 20 % System pressure (MPa) m u -- r.w .o. Figure 4-29. Westinghouse instrument guide tube failure map at minimum radial gap (0.002 cm). _.
~
~
pgp ,,uncjc e -rsy 2. t.ic ~ 1. - mA-65 ice . < s < l <x h<<c'< !- 1 A,, 'y,: i pu c }l,'b!!- W:
a 4 i i PRELIMINARY RESPONSE TO ITEM 44
= Calculations _ employed a water-filled cavity, and a PWR_with an-instrumented lower head. 1 l
1The CE design has a larger _ cavity, lower projected water depth and similarly instrumented lower head. .- 1 4 4
~ F
-r + - -p-e--- - - - e- --
T- T t f f
PRELIMINARY RESliONSE-TO A ITEM 45 See attached sheet. d W G l
. ums D
-'1 1 .1
A(A&oO S . e> n, N 2 One can provide a theoretical basis for heat fluxes in the ran of 10.4 x 10' Bru/h ft (30 MW/m') for a system with co dispersed debris and water as depicted
'in Figure 250. A steam velocity sufficient to levitate and se '
high temperature dense debris is given by . 4 y: , 3.7 /go(pf - p,) 4 i F where g is the acceleration of gravity, o is the steam water surface tension and p, and p,- ' represent the saturated water and steam densities respectively. .lf this is considered to be the maximum steam production rate which could exist without separation of th; water droplets from the co-disperse configuration, then the heat flux associated with the vapor production rate is given by glA =3.7hg [ Jgo(p, - p,) where hgis the latent heat of vaporization. Substituting 'the appropriate values for steam and' water at I atm into this expression results in a value of 10.4 x 10' Bru/h-ft' (30 MW/m 2); a value in agreement with those observed in the various experiments. Hence, the major ramification of
- an explosive interaction could be the co-dispersion of melt and water which then continues to transfer energy and vaporize water into the containment atmosphere at a rate limited by the ability of the water droplets to remain as part of the co-dispersed medium.
7
PRELIMINARY RESPONSE TO ITEM i8 This information can be found in DOE /ID Report " Technical Support-
- for:the Hydrogen Control Requirement for the EPRI Advanced PWR",
pages 37 to'40-(attached). 4 4
~
*e s , 9 F
. * . . .e .s i
2.2.4 Deflaoration Analytical Basis The above example can be generalized to find the maximum post cen.oustien pressure for an ALWR r.ontainsent. As steam is added to an atmosphere"of air and hydrogen, the' initial pressure and final post-combustion pressure both increase. However, eventually, the mixture becomes inert due to steam addition when the flammability limit is reached. An approximate method is derived in Appendix A. which can be used to determine final pressuret resulting from complete combustion of 13% H2 on a dry basis at various system steam pressures. A calculation using steam table values is presented here. The flammability limit of diagram of Figure 2-2 can be used to determine whether these mixtures are flammable. The maximum final pressure for a 13% dry basis H2 plus steam mixture which is possible in a containment under initially saturated conditionr. is defined for the flannability limit. Containment conditiens will not be superheated because ' such a condition is only possible when dry core debris exists in the containment, a situation precluded by the EPRI ALWR debris coolability recuirement (see Section 3.1). The results of this thermodynamic equilibrium calculation f:r possible containment atmospheres are shown in Table 2 5 and Figure 2 9. The final pressure increases as the ini'ial temperature increases because the initial pressure cust increase due to steam addition, but.the pressure ratio dccreases and is seen to be hignest for the dry case-(300*K). The taximum theoretical pressure in a containment ~r !'.owing a burn ;s 5.5 ata cased on the flammability limits. However, also shown in Figure 2 9 is the anticipated boundary between the complete and incomalete ccabustion regimes. The maxim e m probable pressure is thus about 6.4 atm.
- hen more steam is present. incomplete combustion would be expected and the final pressure would be le:s it.. this value.
37
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I Tabie 2 5 . PRESSURE R!$E AND FLANtABILITY RESUI.TS FOR 13% H2 IN ORY AIR VITH SATURATED STEAM ADDITION T(*K) 1 g9 3 Is b pc p d P /Prw Fla=' 300 0.127 0.,026 1.2 5.8 4.8 Y 325 0.117 0.097 1.4 5.9 4.2 Y 350 0.099 0.240 1.8 ;6.0 3.3 Y 375 0.074 0.430 2.5 '6.6* 2.6* Y 380 0.069 0.470 2.8 - - N 3 wet H2 mole fraction b H7 0 mole fraction C initial pressure (atmosphere) d final (postburn) pressure (atmosphere)
*Y flamable, N - not flamable
- overestimate due to incomplete combustion 38 m..
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.- L PRELIMINARY RESPONSE TO ITEM.49 -
CALCULATION - OF. 140 PSIA BASED ON NUREG/CR-5567'. PEAK PRESSURE CALCULATION SCALED TO A 100% BURN (PAGE-38) , Pb = 0.22 + -(1.42/.75) Mzr/V where Pb is pressure in Mpa gauge t For System 80+ Mzr=32653 kg and V=94650 m3 pb= 0.873 Mpa guage = 126.6-psig = 141.3 psia-Note: this estimate is generally conservative. More realistic AICC burns from sat. conditions will give pressures. closer to.110-psia-; t i s ,
h PRELIMINARY RESPONSE TO ITEM 50 See Attached Figures f b DOE /ID-O 1
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w' .. SYSTEM 80+ REQUEST TOR ADDITIONAL INFORMATION: SEVERE ACCIDENT PHENOMENOLOGY AND CONTAINMENT PERFORMANCE
~
NOTE: it may be helpful for the preparer to know, tha~t NRC 1 will be using CONTAIN and MELCOR codes.-
- 1. Please, provide sufficient information needed to perform multinode containment analysis:
- subcompartment volumes and elevations,
' - inter-compartment connections: junctions flow areas, elevations, flow resistance coefficients (forward and , reverse),
- heat structures data associated with-each compartment:
surface areas,- thicknesses, liners-and coating (if-any), materials type and physical properties .
- detailed description of the cavity-
- 2. Please, provide the following: .
- nominal core power,
- decay heat curve, if different from standard PWR ANS, I,.Te,
- initial fission product inventory (primarily Cs, La, Sr, Ru and Ce) ,. for a given burnup (e.g.. end of cycle, or 2/3 of the cycle)
- if possible, decay heat curve for CsI, zirconium and steel in the core
- total masses of the fuel, region (i.e. steel potentially melted during core degradation)
- 3. _Please, describe ti.s methodology used in the severe accicent evaluation. including:
- containment event tree with appropriate split fractions, ,
list of the risk dominant accidents i 4 For the risk-dominant (beyond DBA) accident (s)-please provide the following:
- mass end energy releasecrates.to the containment,
- fission product release rates to the. containment
- hydrogen generation: rate and total mass generated,
-melted core fractions,
- timing'of: melt start, vessel failure-
- assumed core melt temperature
~
- rate of core melt injection to the cavity
- composition of the core melt
-containment pressures and temperatures
- assumed or. calculated heat fluxes-("up" and "down") _during core-concrete interaction
- rate of-ablation (radial and axial)
Enclosure 5'
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