ML19309D493
| ML19309D493 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/07/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19309D487 | List: |
| References | |
| ACRS-1677, NUDOCS 8004100381 | |
| Download: ML19309D493 (19) | |
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E ISSUE DATE:
12/07/79 I
[t il F0IA EXEMPTION (b)5 TES OF THE ACRS AD HOC SUBCOMMITTEE MEETING 27 D THREE MILE ISLAND IDENT IMPLICATIONS REGARDING NUCLEAR POWER PLANT DESIGN OCTOBER 3,1979 WASHINGTON, DC The ACRS Ad Hoc Subcommittee on the Three Mile Island 2 Accident Implications regarding nuclear power plant design held an open meeting on October 3,1979 in Room 1046, 1717 H St., NW, Washington, D.C.
The purpose of this meeting was to discuss the implications of the March 28, 1979 incident at the Three Mile Island Unit 2 station on nuclear station design, policy, and criteria.
Implica-tions of the Three Mile Island Unit 2 accident as they relate to reactors similar to the Diablo Canyon Nuclear Generating Station and the boiling water reactors that expect to receive licenses in the near term, as well as reactors similar to Westinghouse ice condenser / upper head injection plants that are expected to receive operating licenses in the near term were also discussed. Notice of this meeting was published in the Federal Register on September 18, 1979. A copy of this notice is included as Attachment A.
A list of attendees for this meeting is included as Attachment B, and a schedule for this meeting is included as Attach-ment C.
Selected portions of the meeting handouts are included as Attachment D.
A complete set of handouts has been included in the ACRS Files. There were no written statements or requests for time to make oral statements received from members of the public. The Designated Federal Employees for this meeting were, Mr. R. Major and Dr. R. Savio of the ACRS Staff.
SUMMARY
OF STAFF PLANS FOR CONTENT OF DIABLO CANYON REVIEW FOR THI IMPLICATIONS -
D. B. Vassallo Mr. Vassallo gave a status report concerning plans over the next several months for the Diablo Canyon review. He noted that presently the Staff is working on a number of outstanding issues that are not TMI related.
Following the TMI incident and the establishment of the Lessons Learned Task Force, the Applicant for Diablo Canyon submitted material regarding TMI implications, on their own volition, and then added more information to their response following the Lessons Learned Task Force report (NUREG-0578). Currently the Staff is having difficulty appropriating resources to review this material.
80041003S1
TMI-2 Accident Implications OcGoser V Wz/s Mr. Vassallo noted that until the preceding week, the Applicant has never fomally b:en asked to respond to the Lessons Learned Task Force recommenda:.10ns. The letter sent to those applicants who have pending OL applications indicated that the utility was to implement NUREG-0578 as modified by a number of issued in the letter.
In addition, the utility was to implement some additional requirements on emergency preparedness. Mr. Vassallo noted that Pacific Gas and Electric g
would have to modify its previous responses in accordance with the current requirements given in the letter. Mr. Vassallo noted that within the next few weeks members of the Lessons Learned Task Force will be made available to review implementation of NUREG-0578 requirements on North Anna, Salem, and then Diablo Canyon. Mr. Vassallo noted that during the preceding week the Atomic Safety and Licensing Board on Diablo Canyon rendered a favorable decision on seismic matters. The Staff is at the point of trying to resolve all remaining matters so that a final recommendation on issuance of an OL can be made and brought before the Commission. Mr. Vassallo noted that prior to the issuance of any OL, the Commission is to be briefed on it and make a final decision.
Mr. Vassallo noted that the Staff could be responsive to some of the Committee's concerns but that the Staff would be better prepared to come before the Com-mittee in November and address additional Committee concerns. Dr. Mattson noted that the Lessons Learned Task Force has very nearly completed its assignment and plans to issue a final report within the next couple of weeks. He noted that several members of the Lessons Learned Task Force are being reassigned this week to the implementation of the short term lessons learned from Three Mile Island for operating reactors. Over the course of the next several weeks more and more people from the Task Force will be reassigned to OL cases. Dr. Mattson noted that the Lessons Learned Task Force could be back in November to talk in more detail about Diablo Canyon's conformance to the 0578 short term requirements to accommodate events similar to TMI-2,
'IMI-2 Accident Implications October 3,1979 RESPONSE TO SUBCOMMITTEE QUESTIONS ON SEISMIC CLASSIFICATION OF EQUIPMENT - J. Stolz/
B. Buckley Mr. Stolz noted that the September 14, 1979 ACRS letter written in connection with the Three Mile Island review of both Salem and North Anna indicated that there was special interest in the seismic implications of Diablo Canyon with respect to the Three Mile Island review. He also noted that the Staff will be prepared to discuss two aspects of the thirteen specific questions noted in a preliminary meeting agenda. The Staff discussed what they believed to be the two most important aspects of these questions. The first and what was characterized as probably the most important aspect of the Diablo Canyon design with respect to seismic qualifications was the assurance that Diablo Canyon could come to cold shutdown from the control room assuming a single failure using seismically-qualified equipment. A second item is the criteria and assumptions that the Staff uses in connection with non-seismic piping.
It was proposed to cover these two areas during this meeting. Mr. Stolz also noted that PG&E has prepared a report which addresses the Committee's concerns expressed in the September 14th letter and also addresses the thirteen items on the preliminary agenda.
Dr. Okrent noted that the basic question is No.1, what are the seismic impli-cations of Three Mile Island, if any? He noted that the items in the tenta-tive agenda represent some possible examples. He noted the concern was are there items that go beyond the short term lessons learned that should be con-sidered. Dr. Okrent noted that he wanted to have benefit of the Staff's thinking.
Dr. Mattson noted that with respect to the Staff's thinking there is one qualifier, which is, there is more to come. He noted there would be a need to have a dialogue between and the Staff and the ACRS about the necessity and sufficiency of the final report when it comes out, and recommendations it might make. He noted that in the course of that dialogue discussions will be held concerning recommendations from other quarters such as the Kemeny Commission. He also noted that short-term recommendations were developed from a generic perspective. It did not accommodate unique aspects of individual plant designs. Dr. Mattson noted that such aspects as OL applications in areas of high seismicity and how non-safety grade equipment can influence the course of an accident or how might non-safety grade equipment be relied on to mitigate an accident will have to be studied.
'TMI-2 Accident Implications October 3, 1979 In response to questions from the Subcommittee, the Staff noted that as of yet they have not been reviewing operating procedures, the Staff is just beginning to review this subject. Dr. Mattson noted that in reviewing procedures, the NRC will have to bring on board a lot of human factors expertise. A decision from the human engineering, human factors point of view will have w be made as to the right way to proceed. He noted the learning experience in proceo es reviews for the near-term OLs will feed into the process. Dr. Siess notest that not just written procedures but operator training and the use of those procedures should form a part of the review. Dr. Mattson noted that until this point the Staff has reviewed the design of Diablo Canyon for achieving cold shutdown following an earthquake. The Staff has not reviewed the emergency procedure but noted that the Staff will. How the Staff will judge its adequacy is unknown at this point. Mr. Buckley of the Staff responded to item No. 3 in the pre-liminary meeting agenda: What anomalies in system behavior during an earthquake should operators be trained to handle?" He noted that the Staff believes that the anomalies that would result from an earthquake would not be much different from anomalies that would result from any other transient. The Staff did not know of any special anomalies that would require special attention as a result of an earthquake.
Dr. Okrent expressed surprise at this answer. He noted that the plant might be in a very different situation than would be anticipated for almost any other transient situation as a result of an earthquake resembling the current design basis. Dr. Mattson noted that there were two principal aspects, one being the performance of non-safety grade equipment, and the other being procedures that the Staff would want to look at for TMI-2 implications.
He explained that the review of non-safety grade aspects would irclude what would happen to the equipment in the plant that wasn't designed for earthquake forces should it fail.
Dr. Mattson briefly explained the Salem non-safety grade equipment problem.
He noted that Westinghouse, in responding to an earlier Staff interest in a thorough culling of environmental qualifications of equipment inside contain-ment, went through in a systematic way the traditionally non-safety grade equipment inside the containment to see how it would behave in adverse en-vironments from a much expanded perspective compared to the past.
In the
TMI-2, Accident Implications October 3, 1979 past the staff noted that non-safety equipment cannot be relied on to perform in the event of a transient or accident. Now the question being asked is:
If it performs anomalously, can it aggravate the course of the transient or the accident? Three pieces of equipment which were felt could exhibit anomalous behavior under adverse environmental circumstances were discovered during the Salem 1 review: A bulletin has been issued and requires a reply within twenty The bulletin asks licensees to look at the environmental qualifications days.
of non-safety grade equipment or the performance of non-safety grade equipment in an adverse environment. Mr. Bender noted the importance of operator education 50 that operators will be prepared to deal with upset control systens which may result from earthquake motions. Mr. Buckley addressed item No. 5 on the preliminary agenda:
"What are the assumptions concerning failure of non-seismic Class 1 piping? To what extent can the failure of such piping be tolerated?
Is its design based on the single failure criterion?"
He noted that there is no credit taken for the non-seismic Class 1 piping following an earthquake. None of the components in the non-seismic Class 1 systems were considered functional following an earthquake. Each of the non-seismic Class 1 pipes was assumed to have failed. Flooding was examined. The single failure criterion was applied to the seismic Class I systems that remained intact. The plant was still capable of being shut down safely following an earthquake.
Dr. Siess asked if anybody had made a failure modes ?nd effects analysis to indicate that assuming all the non-seismic Class 1 equipment does not work is the most conservative assumption. Mr. Stolz noted that the Staff has not. Dr. Mattson noted that only one non-seismically qualified pipe at cne location at any given time is postulated to break as a result of an earth-quake. After postulating a pipe failure, a review is made around the break location to determine that the consequence of the break cannot degrade a Class 1 system.
Mr. Knight responded to a question from Mr. Michelson on assumptions used in the review of non-seismically qualified pipe concerning how many of those pipes break during an earthquake. He said the basic assumption is that one non-seismically designed pipe is assumed to fail.
TMI-2 Accident Implicatio s October 3, 1979 Dr. Okrent asked if this assumption has been looked at probabilistically to see if it was justified. The Staff could not recall. However, the Staff would check into pipe break assumptions used in WASH-1400 concerning seismically related pipe breaks.
Dr. Mattson noted the final lessons learned report would address the topic of nonsafety grade equipment. Dr. Mattson also mentioned the concern of a
" hassle" factor affecting the operators caused by the loss of nonsafety grade equipment, and indicated the importance of adequate operator training.
Mr. Bender suggested to the Staff that they write down the definition of " safety grade" equipment.
He noted he would like to see a philosophical set of criteria describing safety grade.
With regards to the seismic qualifications of the connections to the refueling water storage tank at Diablo Canyon, it was noted that all related piping is seismically qualified with the exception of a single cleanup recirculation line which is used intermittently (approximately 7 days out of a year). The connection to the RWST (isolation valve)is seismically qualified.
Dr. Okrent noted that attention should be paid to the subject of stagnant borated water lines where cracking has been observed. This is especially true in Class 1 piping systems. The f4plicant should be confident that such cracks could not degrade the seismic design of the plant.
Mr. Buckley of the Staff noted that the seismic class of the PORV, the block valve, and equipment related to the operability of these devices are all seismic Category I.
Dr. Okrent suggested the Staff should ask themselves whether other cquipment associated with the pressurizer or other plant components in general should be reexamined for seismic capability in the light of TMI.
Mr. Hoch of PG&E stated his opinion that many of the concerns being raised on Diablo Canyon were generic in nature and had only a slight dependence on the seismic environment of Diablo Canyon.
Mr. Etherington mentioned that it should be recognized that the probability of failure of non-Category I equipment is much greater in a region of high seismicity.
There is also a bigger chance of multiple failures of non-Category I equipment where you have high earthquake probability.
TMI-2 Accident Implications October 3,1979 Dr. Okrent noted the original intent of the questions in a preliminary agenda o
was for the Staff to report to the Subcommittee on special considerations re-quired for a plant i.n a high seismic area.
Dr. Mattson said he expects the Lessons Learned Task Force final report to be issued in about two weeks. The members of the Task Force would be reassigned by the first of November although they would be available to discuss the report with the ACRS after that date. Any reorganization with in NRR would probably not take place until after the Kemeny and Regovin reports are issued.
BWR REVIEWS The schedule for OL applications is on a first-due, first-served basis as far as assigning Staff resources. The first BWR OL application is Zimmer which is preceded by five PWR applications. The Staff will address this application with an open-minded attitude with respect to the implications of TMI and expects to have the benefit of the reports prepared by the Kemeny Commission and the Rogovin NRC/TMI Special Inquiry Group. Presently all action on the Zimmer docket is unrelated to the TMI incident.
FINAL LESSONS LEARNED REPORT CHANGES IN THE NRR LICENSING PROCESS - R. Cudlin, NRR Mr. Cudlin discussed final recommendations to be made by the Task Force for changes in the NRR review organization and procedures for reviews.
It was mentioned that recommendations will include a suggestion for establishing a technical review board to maintain quality and unifonnity across case reviews. Also to be recommended is a technical specialist pool to provide immediate response capability for emergencies, implement on a trial basis a modified review sequence that eliminates rounds of questions, consoMdate technical review and project manage-ment activities for licensing reactors ana coerating reactors, and to establish an accident evaluation function. Three other recommendations included establishing on a trial basis, an interdisciplinary review process to conduct reviews of selected cases, consolidate activities relating to reactor operations, and streamline the SRP to differentiate between required criteria and guidance.
l
TM1-2 Accident Implications October 3,19F9 Mr. Cudlin discussed backfitting regulatory requirements. He noted there seems O
to be some lack of clarity and lack of consistency in t'ie way backfitting is accomplished. Backfitting of regulatory requirements should be based on a cri-terion of what is required for safety. Required for !afety should be equivalent to the aggregate of requirements, practices, and poli:ies set forth in the regula tions. Mr. Cudlin also noted that regulations should be amended to reflect multiple levels of safety. This would mean some minimum level of safety re-quired of succeeding generations of plants. A final recommendation made by j
Mr. Cudlin was that improvements in safety should be applied prospectively to new plants. These improvements should be inventoried in three-year increments
,rior to implementation.
Mr. Cudlin moved on to discuss emergency preparedness.
In the area of information requirements, he noted that information required to assess plant status following off normal operation at the plant's "onsite technical support center", will be transmitted to the NRC's Incident Response Center (IRC). The minimum set of specific parameters needed by NRR for the evaluation of reactor status and radio-logical consequences will be identified by a Sandia Study and a Standards Development Task Force for the identification of instrumentation required to follow the course of an accident.
Mr. Cudlin discussed NRR's role in emergencies. An NRR Emergency Response Team (ERT) should be designated and on immediate call in the event of emergencies.
The ERT will provide on-site information gathering e.apability, and advise the established line authority at the NRC's IRC of appropriate action. The ERT's first task will be to identify resources requirements, procedures, training, and facilities to enable effective emergency response by NRR.
Mr. Cudlin commented on operational surveillance.
In addtion to the ED0's Operational Data Analysis and Evaluation Group, it will be recommended to establish within NRR, a dedicated staff organization for the review and evalua-tion of operating experience. The NRR organization will screen significant operating experience (LERs) and be the NRR focal point in the NRC/ Industry network of operations evaluation.
TMI.-2 Accident Implications October 3,1979 Dr. Mattson noted that it would be helpful if the Staff could get more guidance from the Committee on how to handle recome. dations from the ACRS as a whole, its Subcommittees, and individual members. A cethod of placing priorities on recommendations was requested.
Mr. Bender gave a personal observation that the Staff should return to the Committee and describe how they interpret the Committee's recommendations.
He noted not all ACRS questions asked required direct answers.
UTILITY TECHNICAL SUPPORT CAPABILITY Mr. Haass of the Staff presented a description of plans for development and imple-mentation of criteria and the status of an NRR survey of present utility technical and management support capability at operating nuclear power plants.
Infonnation now being submittsd by utilities will be given to an outside contractor to con-sider and to identify weak points of a utility in handling an accident situation.
The project is expected to be completed by March. A utilities present capability will be compared to acceptance criteria to be established.
Mr. Michelson raised a concern dealing with the low-level cutoff on the pressurizer heaters. He noted the cutoff is non-safety grade in many cases presently. He noted that for some operating transients, it appears the water in the pressurizer can empty out below the level of the Calrod heaters. For those transients, if the non-safety grade heater cutoff should fail to function, a failure modes and effects analysis on the Calrod heaters would be of interest. Mr. Michelson noted the failure of Calrod sheatts and remaining backup insulation, which represent the primary system pretsure boundary, represent the major concern.
l l
,.,PROBABILISTIC ASSESSMair OF HYDROGEN C0tfIROL MEASURES - M. Trylor & T. Cyb We Mr. Taylor sumarized the history of the NRC criteria for hydrogen control.
NRC hydrogen control criteria was derived from post IDCA hydrogen generation con-sideration. Safety Guide 1.7 specifies a 5% metal-water reaction as the design criteria. D1I experience indicated that there is a range of accidents for which he amunt of metal-larger concentrations of hydrogen need to be considered.
I water reaction which has occurred in DII-2 is still uncertain but it appears clear that it was in excess of 25%. Present hydrogen control philosophy includes the use of hydrogen recombiners and purging of contairment and is generally tied to the criteria of Safety Guide 1.7.
Recocbiners are believed to have sace mar-gin for controlling hydrogen above the 5% value. W e margin available is a matter of the specific plant design. Recently, partly as a result of some of the conclu-sions arrived at in WASH-1400, soce consideration has been given to a controlled filtered vented contairrent.
Were was some discussion as to the analysis of the hydrogen release in DII-2.
he NRC Staff indicated that estimates of the anount of metal-water ranged between he nature of the pressure pulse observed in the DII-2 event was discussed.
25-50%.
Mr. Taylor indicated that his analysis indicated that a compart: tent concentration of about 7% with some locally higher concentrations existed. h e lack of a quasi-static " pressure tail" on the observed pulse would indicate that the high concen-trations of hydrogen were local rather than uniformly distributed throughout the More uniformly distributed hydrogen concentrations would have led containment.
to higher containment temperatures with a quasi-static " pressure tail" being ob-served. Mr. Taylor estimated that approximately 550 pounds of hydrogen was burned. E is corresponds to about a 30% metal-water reaction. W e hydrogen burn It was noted that occurred about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the zirc-vater reactions started.
the contairment circulation times are on the order of minutes.
h e criteria for net positive suction heads for containment sunp pumps was discussed.
It was noted that during a large 1DCA the liquid in the sunps could be above 2120F and containment failure with a depressurization to atmospheric pressure would result in the flashing of some or all of the liquid into steam. h e sump pumps could then be deprived of a source of liquid.
It was noted that open purge valves or airlocks would also result in lower than anticipated pressures under IDCA conditions.
The principal eccident :~pences leading to release of radioactive materials was discussed within the WASH-1400 context. These are shown on page 1 of Attachment D and The are illustrated for a typical RR and BWR on pages 2 and 3 of Attachment D.
risk contributions of core melt accidents as calculated in WASH-1400 are given on page 4 of Attachment D.
It was noted that containa nt failure caused by over-pressure dominates the risk. The relative importance of the various postulated containment failure nodes is sunmarized on page 5 of Attachmnt D.
The relative probabilities of containment failures by various mechanists is sumarized on page 6 of Attachment D.
Various alternate containment concepts have been studied which nny significantly reduce the risk for these types of nuclear accidents.
He results of a Sandia study are shown on page 7 of Attachmnt D.
B e shaded area on this figure depicts the uncertainty for the WASH-1400 calculations. The use of filtered atrospheric venting, compartment venting, and deep un:lerground venting resulted in comparable reductions in risk. The use of stronger contain-nents/ increased containnent volume, thickened mats, double containment, evacuated containnent, and shallow underground siting were also cor.sidered and were shown to result in a lower reduction in risk.
The scenario during which the failure of a check valve between a low pressure /high pressure system would result in a failure of the low pressure system was discussed.
Questions were raised as to whether or not the probability of such an accident scenario had been made acceptably low. The NRC Staff was asked to give this infor-mation to the Subecruittee at or before the next Subcocntittee teeting.
Probabilistic analysis work which compared the WASH-1400 RR with the ice condenser RR was discussed. The results are sucrrarized on page 8 of Attachment D.
The level of risk associated with the ice condenser containment was predicted to be somewhat higher than the WASH-1400 R R.
Opinions were expressed that the uncer-tainties in the risk assesscent for the ice condenser would be higher because of the relatively fewer years of operating experience obtained with this type of design. It was also noted that the ice condenser design had the benefit of core years of conventional RR operating experience than the WASH-1400 type work and would be expected to show lower risk under this type of analysis as the level of risk associated with accidents identified in WASH-1400 had been reduced (for in-stance, the check valve accident scenario). It is noted that these studies had identified an accident sequence in uhich the drain between the upper and lower l
l
ccxrpartments was left blocked through human error and would prevent the return of water to the ECS sumps. Page 9 of Attachrtent D sumarizes the risk associated with hydrogen burns for the WASH-1400 ER and the ice condenser PWR.
Page 10 summarizes scrae of the accident sequence predictions for the WASH-1400 ER design (a ER 4, MARK I inerted containment).
It was noted that the inerting i
of the reactor safety study ER design had little effect on the risks associated with the contaiment overpressure failure scenarios. Controlled filtered vent-ing had a nuch larger effect.
It was noted that the DE-2 vent event had demon-strated that even releases that were small enough to have little health and safety inpact were large enough to be objectionable to a significant portion of l
the public. On that basis, it was suggested that devices that would significantly l
reduce small risks might be assigned a larger benefit than had ordinarily been li ascribed to them in past studies.
l Mr. Taylor indicated that he had concluded that:
1.
Inerting appears to have small value in reducing the overall risks and i
that in some cases reducing the accident sequence probability appears to have an equal value or greater benefit.
l 2.
We larger high pressure containments are less sensitive to the effects of hydrogen, f
i 3.
Hydrogen control measures that may be adopted after DR-2 should have the benefit of insights based on overall risk analysis.
4.
W e WASH-1400 study enphasized the core melt accidents,. W e risk reduction benefits of current licensing H2 control measures for such accidents appears to be small.
5.
Research on improved IRR safety concepts such as controlled filtered l
venting should be continued and given priority.
i Mr. Cybulskis discussed estimated containment failure pressures. A 60 psi dry AR containment by his estimates would be somewhere between 85 and 125 psi, whereas 12 psi ice condenser containment was estimated to fail at about 27 psi.
Results are sumarii.ed on page 11 of Attachment D.
B ere was some discussion as to the basis of these estimates. Only the containment shells were considered and the ultimate stress was taken as the failure criteria. Penetrations were not considered. Mr. Cybulskis estimates of the potential contributors to contaiment overpressure are sumarized on page 12 of Attachment D.
W e Westinghouse RESAR i
6 cu. ft.,
cor'e was used as the reference core and dry contaiments with 1.2 x 10 1.8 x 106 cu. ft., and 2.1 x 106 cu. ft. volumes were considered. L e basemat was assumed to be constructed of limestone concrete.
i he effectiveness of distributed ignition sources as a means to control hydrogen l
i was discussed. Opinions were expressed that the effectiveness of these devices j
could be determined experimentally but that it would have to be done in a large I
scale experimnt to be convincing. L e question was raised as to if an ice condenser plant was equipped with controlled filtered venting whether or not it would be advantageous to also inert the containment. It was noted that the principles of using ignition sources for other types of hydrogen reccubiners the large dry were the same for large dry containments and ice condesners in that Recombiners are dependent containmnt had more margin against overpressure.
upon an elcctrical power supply to function. Recombiners with very large capa-cities are being considered as a design possibility.
It was noted that a repre-sentative spectrum of accident sequences should be considered in evaluating the effectiveness of these various hydrogen controlled devices. h e NRC Staff was asked to report on programs that they would use to evaluate hydrogen control measures in ice condenser containments and their preliminary thoughts on how the problem might be handled at the next 'IMI Subcomittee meeting (tentatively scheduled for November 5, 1979).
REPORT OF TEE 1ESSONS IEARNED TASK FORCE - R. Mattson & W. Minners Dr. Mattson and Mr. Minners reported on the final recomendation of the lassons learned Task Force regarding hydrogen control. Dr. Mattson indicated that the lassons Imarned Task Force was recomending that measures for controlling hydrogen in events involving degraded core cooling be provided for all reactor designs.
He noted that the design basis for metal-water reaction was difficult to specify Criteria would be but that they envisioned it as being greater than 20-25%.
established in rulemaking procedures. Le second recomendation was that the Comission undertake rulemaking to establish design requirements for controlled filtered venting of containments as to provide a method of interdiction of the gaseous pathway to the environment for the core melt scenario.
Dr. Mattson indicated that he agreed with Mr. Taylor's conclusions regarding the effective-ness of contaiment inerting for the core melt scenarios but thought that con-tainment inerting would be desirable for smaller accidents involving degraded He noted that the lessons learned Task Force had come to the cooling events.
itigation were required for proth
- onclusion that both prevention an mDr. Mattson indicated that a plan fo d
against the degraded cooling event.
d to develop these concepts would be pre-It was suggested that r: search program that might be requireNovember meeting.
i d sented to the 1MI subcomittee at the tor cavity might be further exam ne.
cooling of the dispersed core in the reac reactor cavity's capability for F
h further The use of refractory materials to augment t eted as an area which would retaining the core debris was also sugges study.
k-USE OF A SINGIE STANDARDIZED UE IESig d views on the advisability of the idered The Subcomittee and the NRC Staff exchangeDr. Mattson indicated that single UE standardized design.
reliminary views on it.
the matter in depth and only had very pd at the same level of compet presuming the design work was performe h NRC could review the concepts in bably did not h ~e recently proposed standardized designs, t eViews A stm-the inhouse expertise necessary to per or d and it was concluded that the two years.
dardized gas cooled reactor design was discusse difficult than an U a pl d
design and review of such a plant would be moreVulnerability of plant i d in plant design were identified a because of experience factors.
i the narrowing of the focus of think ng vantages of pursuing a single standardloosely specified configuration design.
design could take any fom ranging from a verys' equipment and con to one which named specific manufacturer i
at 6:00 pm.
The meeting was adjourned after this sess on ipt of the meeting is 1717 H St., NW, Washington, For additional details, a complete transcravail NOTE:
DC 20555 or from Ace-Federal Reporters, NW, Washington, DC.
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.All worLets of Stnde Rite Manutecturing ya om mams med su es "I Corporetion's Haatt Shoe Drvision,locsted m ye on. saurin ru.e e-it.ra se -1 sumus cons an*sHs lawrence. Mauechusetts, who became suasc coca eswe.a totally or partially separated from eenployment on or after November 10.1978 (TA W-37141 are eligible to apply for edrustment Stride Rite Manufacturing Caep., Histt NY73$4 Title 11. Chapter 2 of the See! Tenning Co., Manchester, N.H ;
Shoe Dhrision. Lawrence. Mass.;
Signed at Wuhmston D.C. this 7th day of I
R: vised Determination on Certification Regarding Eligit>ility To september 1979 APP y for Wortler Adjustment knee F.Teylor.
l Reconalderation On July 23:1979 (44 a 44304). the D"'C'#' Ol e ofMonesement in accordance with Section 223 d the Adansnistmtion anmnium Department of Labor granted cdministrative reconsideration of the Trade Act of W4 (19 U.S.C. 2273) the sen. Saarira.o.39.7 an ni Department of Labor herein presents the ausso ooon as+m.=
Nrgative Determination Regarding meu a an gation ngardmg Ehgibihty To Apply for Worker certification of eligibihty to apply for
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Adjastment Assistance which it had worker adjustment assistance.
NUCLEAR REGULATORY
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1 made on June 19,1979 (44 FR 36516)
In order to make an affirmalive COMMISSION pursuant to Section 223 of the Trade Act determination and issue a certification cf 1974 for all workers at the of eligibility to apply for adjustment Advisory Committee on the Three Mlle Manchester. New Hampshire. plant of suistance each of the group eligibility toland, Unit 2 Accident, implications re the Seal Tanning Company.
requirements of Section 222 of the Act Nuclear Reach Safeguants, Ad Hoc In its rec 6nsideration. the Department must be met.
Sut> committee on the Three Power reviewed its file on the Seal Tanning De investigation was initiated on July Plant DeeJen; Meeting Compary.The review and an additional 5.1979 in response to a worker petition
%e ACRS Ad Hoc Submmmittee on customer survey conducted by the
.nceived on July 2.1979 which was filed the Dree Mile Island. Unit 2 Accident-Department revealed that a substantial on behalf of workers and forme-Implications Re Nur: lear Power Plant
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part of Seal Tannmg's decline in workus producing children's fotwear Danign. will hold a meeting on October finished leather sales was accounted for at Stride Rite Manufacturing
- 3. W9 in Room 1046.1717 H St.. NW.
by customers who reduced purchases Corporation's Hiatt Shoe Division.
Washington. DC 20555. Notice of this located in Lawnnce. Massachusetts. It meeting was published on August 23.
l from Seal Tannmg and increased their I
import purchases of fmished leather.
is concluded that all of the requirements 1979.
have been met.
In accordance with the procedures Further. the Manchester. New gg gg.s nonrubber outlined in the Federal Register on Hampshire plant of the SealTanning footwear, except athletic. inemased October 4.1978. (43 m 45926). or written Company ceased operations on June 30.
relative to domestic production in the statements may be presented by W9-first quarter of1979 compand to the members of the public. recordings will U.S. imports of tanned and finished same period in 1978.The ratio ofimports be permitted only during those portions cattlehides increased from 94.0 million to domestic production increased from of the meeting when a transcript is being square feet in 1977 to 123.7 million 80.8 percent in the first quarter of1978 to kept. and questions may be asked only square feet in 19'8 and from 25.1 million 100.0 percent in the first quarter of1979 by members of the 9ubcommittee,its equare feet in the first quarter of1978 to A Labor Department survey of consultants, and Staff. Penons desiring 35.9 million square feet in the first chtomers who bought children's shoes to make oral statements should notify
~
quarter of 1979 The ratio of imports to and sandals from Stride Rite the Designated Federal Employee as far domestic production increased from 8.7 Manufactursng Corporation revealed in advance as practicable so that percent in 19 7 to 12.3 percent in 1978.
that obme of these customers nduced appropriate arrangements can be made purchases from Stride Rite and to allow the necessary time during the Cooclusion increased purchases ofimported meeting for such statements.
Based on additional evidence, a children's shoes and sandals in 1978 he agende for subject meeting shall review of the entire record and in compared to 1977 and in the first quarter be as follows:
l cecordance with the provisions of the of me compared to the same penod in Wednesdoy, Octo5er 3. 2979; a:30 a.m.
Act,1 make the followmg revised untilthe Conclusion ofbusiness.
determination:
Conchsalon ne Subcommittee may meet in All workers at the Manchester.New After carefal review of the facts Executive Session, with any of its Hampshire, plant of the Seal Tanning abtained in the investigation. I conclude consultants who may be present, to
~
Company who became totally or partiaDy that increases of imports of articles like explore and exchange their peeliminary separated from employment on er after April or chrectly competitive with the opinions regarding matters which should
- 30. tira, are ehgible to apply for =-?"'
children's shoes and sandals prnAad be considered during the meeting and to assistand under Tttle R. Chaptar a of the at Stride Rite Manufacturing formu!ste a report and recommendation Ttade Act of1974.
Corporation's Histt Shoe Division.
to the full Committee.
located in Lawrence. Massachusetts.
At the conclusion of the Executive contributed importantly to the decline in Session, the Subcommittee will hear j
sales ce production and to the total or presentations by and hold discussions i
f3 MO j
r aan==ga-eatia ta,sta and Hamilton Cosaty Bunaisenaa! Ubenry,
'emeials, and cther tsarested paressa.
2001 Broad Street. Mattaanopa.TN tega.dag the imp!scanoes cf the Does 37402 regarthag Segnoyah; and the Mde Island. Unit 2 Actedsat 43 they Clerunoel County Ubrary.Thsrd and relate to the fenowetas Broadway Streata. Batavia. Ohio 45133.
~
g a g, y,
togarthag Z.unser.
i Reactore sindar to the Diablo Cneyan
. Dated September it.154 Nucles' Geores% Staben W 6e I'I' C O bolhng-water reactors that are erpected AcMsory Cemsvese Mpuresuret OfFeer to receive operating bonnaea in te near.
Pt D=. we 'W *-e 8a as =l term (Zunmer and I4Salle) amassa ones ruso.e64 Mr. Richard K Major is the Designated Federal Earpioyee fue ein g
porton of the meetm6 2.100 p m until the conclusica of business i
Ren: tors similar to Westinghouse Ice Condeneer/ Upper Head inrect>as (UHI)
Plants that are expected to racerve operaung beenses in the near term (Seqcnah and McGuire)
Vol. 44. No.18e / Monday. September 24 Dr Richard P. Savio is the Designated Fed re Employee for th.is porboc of the gfe Wns by the Dmene.NSE an snectinst In adicon. It tony be acceesary for the SA-rettee to hold one or more joyce F. taplaats.
closed semons for the purpose of Acrag ce,,jeg,e Aramvement Coortfmotor explormg matters involymg peopnetary September 1s. tem informataco. I be ve deternned, in 7 accordance vnth Subesetaoc 10ld) af '
ire o =,w m,,...i Pubbe la = 92-4a3. that. abould aud
/
aan.se esse m sessioes be required. n is necaseary to close these sessions to protect NUCLEAR NEGULATORy prepnetary miormanoc (5 U.S.C.
CCMMtSSeON 552b:c)(44 Further informabac regarthag topics A% ComwWttee oct Rosetor et be d;scuseed. whether the meetirqg
$sfogmtte Ad Hoc Subcommittee on has been canceled or rest.bedu!ed, the the Three Mlle taland Unit 2 Accident-Chassan's ruhng oc requests fcr the impecations re Nuedear Powerplant opportunity to present oral stataments Desh Meeting and the tune allotted therefor can be cof,,eff,3 obtained by a prepaid telephone cad to the cogmzant Dessnated Fadeca!
In FR Doc. 758tNn:la the f aeue of Employae.Mr Rackard K. Mast:
Tuesday. September te 1779. on page (telepbooe E/8M-14141 oe Dr. mchard 54139. the heading should be corrected P.Smo.(telephone El&M 32E*)
to read as set forth above.
bmeen t15 en and 5.00 p.m EITT.
same come me.e Background aformation concaming Items to be discassed at this meeting can be found in docuecents on Es and as allable for pubbe inspectico at the NRC pubhc Document Room.1717 H Street. N W Weshington. DC 30655; 16 Gov emment Pubbcations Section. Stet Ubrary of Pennsylvania.Educatnae Buildmg. Commonwealth and Wahnet Stree1. Hamabarg PA 1712L regardag Three blale taland. Unit 2 the Documenti and Maps Department.
Ca!&rnia Polytechnic State Unieurat IJbesry. San 1.ais Obispo.CA SMo7 regarthag Diablo Canyoe; the Bhnois Val ley Cosamunity College Ruts! Roo
- 1. Ogienby. E. 81Ma, regarchas la l
Saus;the Pubhc ubesry of Chadoes 7Y 3 D)AlW(
D 9TD eh wI
ATTENDACE LIST ACRS AD HOC SUBCOMMITTEE MEETING ON THE TMI-2 ACCIDENT IMPLICATIONS WASHINGTON, DC s
OCTOBER 3,1979 ACRS NRC STAFF D. Okrent, Chairman G. Bagchi M. Carbon R. DiSalvo W. Kerr J. Burns W. Mathis C. Thomas M. Plesset
- 0. Parr C. Siess J. Martore S. Lawroski J. Milhoan H. Etherington P. Stoddart M. Bender H. Krug C. Mark W. Minners I. Catton, ACRS Consultant W. Haass W. Lipinski, ACRS Consultant R. Cudlin C. Michelson, ACRS Consultant P-Y. Chen T. Theofanous, ACRS Consultant J. Knight R. Major, Designated Federal Employee S. Varga R. Savio, ACRS Staff D. Vassallo J. Stolz B. Buckley PACIFIC GAS & ELECTRIC R. Tedesco L. Rib B. Norton W. Milstead M. Furbush J. Long G. Blanc M. Silberberg J. Schuyler J. Curry P. Crane W. Butler R. Young F. Rowsome R. Laverty N. Merriweather T. Crawford R. Birkel R. Patterson H. Holz A. Marchese D. Skovholt TENNESSEE VALLEY AUTHORITY L. Mills ELECTRIC POWER RESEARCH INST.
J. Ballentine D. Lambert R. Leyse C. Morgan W. Lau WESTINGHOUSE ELECTRIC CORP.
GILBERT ASSOCIATES INC.
W. Luce W. Brown H. VanKessel J. Hord M. Barnisin S. Palusamy ATTACHMENT B
..o THE CLEVELAND ELECTRIC ILLUMINATING CO.
0AK RIDGE NATIONAL LAB.
G. Flanagan J. Szwejkowski GENERAL ELECTRIC CO.
BBR K. Laver g
N. Shirley OFFSHORE POWER SYSTEMS EDS NUCLEAR R. Reymers D. Aabye R. Bruce BATTELLE'S COLUMBUS LABS.
TOKYO ELECTRIC POWER CO.
H. Hamada P. Cybulskis
I REVISED TENTATIVE SCHEDULE FOR 1HE OC1VBER 3,19791MI-2 ACCIDE!C IMPLICATIONS AD HOC SUBCOMMITTEE MEETIN3 MORNING SESSION 8:30 A.M. - 12:00 NOON 1.
Discus'sion of Seismic Design of Plants.
(1/2 hr.).
- Briefing on Seismic Classifications Failure Assumptions 2.
Discussion of 1MI Implications - Lessons Learned as they apply to BWRs.
(15 mins.)
3.
Discussion of a single standardized reactor plant concept.
(1 hr.)
4.
Report on selected aspects of the long-term recommendations from the Lessons Learned Task Force.
( l 3/4 hrs.)
- LUNCH ***********
AFTERN E SESSION 1:00 P.M. ~
1.
Discussions with RES-PAS.
(1 br.)
Response to Dr. Okrent's questions at the September 12, 1979 meeting of the Reliability and Probabilistic Assessment Sub-mittee regarding the probabilistic assessment of hydrogen contral criteria.
Technical status report on PAS-sponsored work on accident delineation.
- Technical status report on Sandia's work on filtered / vented containments.
2.
Status of NRC review of hydrogen control measures for various contalment designs.
(1 hr.)
3.
Options for modifying containment to cope with accidents beyond the design basis.
(1 hr.)
4.
'INI implications for ENP -- Status of NRC response to ACRS letter of July 25,1979 (attached).
(1 hr.)
b L_
99 cs%q'c, UNITED ST ATES NUCLE AR REGULATORY COMMISSION
. ;. 7 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS s, e e
w ash tN GT ON, o. c. 20555 Q ""
f
- ..c [ p July 25, 1979
=
Harold R. Denton Director, Office of Nuclear Regulatory Regulations ACRS REVIEW OE THE FLOATIN3 NUCLEAR PIANT CORE LADLE DESIG
SUBJECT:
ACRS Subcomittee Meeting on the Floating Nuclear At the June 27, 1979 Plant, menbers of your staff requested that the ACRS meet at an early Mr. Gossick con, enting on that preliminary design prior to the
'lhe Acting ACRS Subcomittee Chairman issuance of its safety evaluation.
informed your staff and representatives of Offshore Power Systems that the suggestion to held an early ACRS meetire would be considered at the July 1979 ACRS meeting.
'Ite proposal to hold an early ACRS review of the conceptual design of It was the M4P core ladie was discussed at the July 1979 ACRS meeting.
dreided that additional information, as indicated below, is necessary be-fore the Comittee can proceed with its review of the FNP.
I Ite.s Related to the Impact that the Core Ladie Will Have on Other c.
Contain ent Structures Calculate the fraction of decay heat radiated from the pool 1.
for the proposed design.
Calculate the effects'of heat radiation in Item 1 on the rate of:
2.
disintegration and collapse of exposed concrete (a) disintegration and collapse or melting of concrete behind (b) the 6 inch magnesite brick wall collapse of steel from the reactor cavity.
(c) l Discuss the consequences of Item 2 with respect to:
I 3.
(a) loss of integrity of superstructures
, b) loss of hearth capacity
(
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