ML20148K031
| ML20148K031 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/03/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1752, NUDOCS 8012020087 | |
| Download: ML20148K031 (50) | |
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f ACRS SUBCOMMITTEE MEETING ON SEQUOYAH NUCLEAR 6Y DATE ISSAD:
l h-)hp PLANT JUNE 2, 1980
- l M 3, mfd ll The ACRS Subcommittee on the Sequoyah Nuclear Plant, Units 1 & 2, held a one ,
The day meeting at' 1717 H Street, N. W., Washington, D. C. on June 2, 1980.
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purpose of this meeting was to develop information for the consideration of the ACRS in its review of the Tennessee Valley Authority (TVA) application .
l to operate the Sequoyah Nuclear Plant, Units 1 & 2. The notice of this meeting was published in the Federal Register on Thursday, May 8, 1980. A Copy of this notice is included as Attachment A. The schedule for this meeting is included The as Attachment B and a list of the attendees is included as Attachment C.
entire proceedings were held in open session. Portions of the material pro- ,
vided to the Subcomittee at this meeting are included as Attachment D. A complete set of the materials provided to the Subcomittee is in the ACRS files.
No oral statements were given by members of the public nor were there any requests for time to give oral or written statements submitted. Mr. W. Mathis, the Acting Chairman, and Mr. J. Ebersole were present. The ACRS consultants Dr. I. Catton and Dr. W. Lipinski, and Dr. Richard Savio of the ACRS staff were present. The Designated Federal Employee for this meeting was Dr. R' Savio.
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l INTRODUCTION Mr. Mathis called the meeting to order at 8:45 a.m. and made a bried intro-j ductory statement explaining the purpose of the meeting and called upon Mr.
Carl Stable of the NRC Staff to begin the presentations. l l
STATUS OF THE NRC FULL p0WER LICENSING - C. STAHLE, NRC Mr. Stahle semarized the course of the NRC review to date. He noted that the ACRS letter o? December 12, 1979 endorsed the low power operation and the performance of the special low test program at the Sequoyah Nuclear Unit 1 and that a list of post-TMI NTOL items had been developed in Supplement 1 to the .
Sequoyah SER (February 1980). The Commissioners on February 29,'1980 approved the 5% power license for the Sequoyah Nuclear Plant and the performance of the special low power test program, subject to the approval of the NRC Staff. -
Mr. William Cottle, NRC Resident Inpsector for the Sequoyah Nuclear, Unit 1 summarizedthestatusoftheplantoperation. Mr. Cottle stated that the fuel had been loaded in the core and that the turbine inspe'ction had been completed.
TVA is currently engaged in an augmented inspection program of the seismic restraints. This is expected to be competed by June 21, 1980 and will be followed approximately by a week of hot pre-operational testing. TVA's schedule currently estimates ~that the plant will be brought critical on July 4, 1980. This will be followed by.a week of low power testing and three weeks of the special low power test program. (TVA now estimates that the special low power test program will take three weeks rather than the previously estimated three to six weeks.) Mr. Stahle indicated that it was the NRC's expectation that the full power licensing review will have progressed to the stage where it can be discussed with the ACRS and a letter for fall power operation, requested at the July 10-12, 1980 Full Committee meeting. Mr. Stable noted that the SER documentation of the review will not have been issued by the July meeting.
Mr. Stable discussed the presently unresolved non-TMI issued on the Sequoyah j Nuclear U,it 1. These are listed on page 1.of Attachment D. Mr. Stahle indicated that the the path to , resolutions of these items appeared to be clear
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and that no problems were anticipated in resolving the items on this list j before the end of the July target date for full power licensing.
Mr. Walter Butler of the NRC Staff presented the Staff position on hydropn control measures for the full power licensing of Sequoyah Unit 1 and other ice condenser plants. Mr. Butler indicated that the NRC Staff review concluded that the existing Sequoyah design satisfied the current requirements of the 10 CFR 100, 50.44 (redundant recombiners, backup purge system, and a design basis for 1.5% metal-water reaction). The containment design pressure is 12 psig and the NRC's best estimate for failure pressure is 36 psig. It is esti-mated that the containment will accommodate the 25% metal-water reaction for the failure pressures reached. This conclusion is reached under the presump-tion that the hydrogen is generated and released and burned in the containment.
Complete mixing is assumed and no credit is taken for the presence of steam in the containment or initiation of containment sprays. Mr. Butler indicated that the Staff's positions were based on the presumption that recent changes in the licensing requirements have made the' likelihood'of severe accidents remote, ,
that the Sequoyah design has the capability to accommodate hydrogen. generation beyond the previously accepted design basis, and that studies addresting the. .
problem will be undertaken by the Staff and the Applicant and conducted on accelerated schedules. Mr. Butler also noted that the Staff feels that clearly beneficial medication systems have not yet been defined. A written statement submitted by Mr. Butler is include.d as Attachment E.
DISCUSSION OF THE LOW POWER TEST PROGRAM - R. BAER, NRC; J. WALKER 'TVA;
- R. TULEY - WESTINGHOUSE Mr. Baer summarized the overall objectives of the low power test program. He nc+ad that the Staff believes that these tests will::
- 1. Demonstrate the ability to achieve and maintain single-phase natural circulation under a variety of conditions
- 2. Provide further training of the operators under natural circulation conditions l
- 3. Demonstrate the ability of the plant and operators to remove decay heat under conditions simulating the loss of all AC power.
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Mr. Baer indicated that the saf'ety analysis of the special low power test l
program received by the NRC on April 9, 1980 and was currently still under review. He indicated that it was Staff's expectation that the review would f be completed by mid to late June 1980.
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Mr. Walker sumarized the test test objectives from TVA's standpoint and briefly '
described the proposed test. Mr. Walker noted that in addition to the goals identified by the NRC, TVA felt that the test will be valuable in that they would verify and improve the simulation models and training techniques used on the Sequoyah simulator. He noted that all five shifts would have experience in the performance of at least part of each of the proposed tests. The personnel -
responsible for the simulator training and the design engineers have had input into the development of the test procedures. Information obtained from these tests will be available for use in proving the presently empicyed simulator models and techniques. .
The special low power test, in order in which they will be performed, are:
- 1. The natural circulation test
- 2. The forced circulation cool-down test
- 3. The natural circulation with loss-of-pressurizer heaters test
- 4. The natural circulation at reduced pr, essure test
- 5. EffectofIteamgenerator.secondarysideisolationonnatural circulation tests
- 6. The natural circulation with simulated loss of offsite AC power test
- 7. The simulated loss of all onsite and offsite AC power test
- 8. The establishment of natural circulation from stagnant conditions test f
- 9. The cool-down capability of the charging and letdown system test
- 10. The boron mixing and cool-down under natural circulation conditions test.
A summary of the individual tests is given on pages 2 - 11 of Attachment D.
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The Subcomittee asked as to what consideration had been given to the inclusion l of reflux / condensation tests and feed / bleed tests in the special low power test program and why these tests were not to be included. Mr. Baer indicated that l such tests had been proposed by the NRC and discussed with TVA and W . TVA and i W felt that such tests could pose a hazard to the plant equipment and that the plant is not well instrumented to obtain meariingful information from these tests.
Mr. Baer indicated that the NRC felt that under these circumstances they could 1 not require such testing on a comercial power plant. The NRC feels that it would be more appropriate to perform such experiments in research facilities and to confin6 the testing in the comerical power plant to such tests as would be useful for training purposes and the check-out of plant equipment.
- Mr. Tuley presented a summary of the W safety evaluation for the special low power test. The' low power operation was evaluated for the FSAR spectrum of transients. The results were categorized into five classes. They are:
- 1. Bounded by FSAR analysis results
- 2. Transients were reanalyzed ahd fuel clad. integrity was demonstrated
- 3. Additional operator action would be required to assure protection
- 4. The probability of occurrence is reduced by restrictions on test conditions
- 5. The possibility of occurrence is reduced by the short period'of testing.
The evaluation results and the equipment and procedure modifications implemented are sumarized on pages 12 - 14 of Attachment D. Some highlights are:
- 1. DNB cannot be precluded if the evaluation modesi used in the FSAR analysis are used for the low power test analysis. Model improvements would preclude DNB and no significant fuel damage is predicted by the Original FSAR models.
- 2. Automatic actuation of safety injection will be blocked for all of these tests ,
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- 3. A number of bypasses will be used during the low power test series.
These are sumarized on page 15 of Attachment D.
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l Mr. Baer indicated that TVA is currently proposing that the test restrictions ;
0 require tripping the reactor if the subcooling margin falls below 15 F and initiating safety injection of the use of cooling margin is below 100F. The NRC is currently considering requiring larger subcooling margins. TVA has indicated a willingness to increase the margin for tripping the reactor but is reluctant to increase the margin for initiating _ safety injection ' Mr.
- Stable indicated that RES is intending to pre-predict the results of the special lowpowertestutilizingahailablebestestimatemodels. The Subcommittee suggested that much benefit might be obtained from the vendors performing such pre-predidctions of future NT0L tests and that RES should consider supporting
-additional instrumentation of future NTOL tests. Mr. Baer indicated that he
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' felt that the most practical application of additional instrumentation would be on Construction Permit (CP) applications. Opinions were expressed by the Subconinittee that additional instrumentation requirements would not be impracti-cal on NTOL applications.
FEED AND ' BLEED OPERATIONS ON THE SEQUOYAH PLANT - R. McKEEHAN, TVA; R. TULEY, W Mr. McKeehan gave a brief description of the systems which would be involved '
in feed and bleed at the Sequoyah Nuclear Plant. The schematic of these systems isgihenonpages15-16of-AttachmentD. The feed and bleed operation normally inholhesthePORVsandthesafe.tyinjectionpumps. The refueling water storage tank is the source fot the safety injection pumps. Two-train, low, intermediate, and high head pump configurations are used. Injection is into the cold leg of each loop. The safety injection system is fully safety grade. The PORV and block valhes are on three in'ch lines, jointed to the single six-inch line going into the pressurizer. The mechanical components of this system are safety grade up to the outlet of the PORV. The PORVs and block valves are three inchhalhes. The PORV will fail closed and the block valve will fail in the "as is" position. Motivepowerendcontr'o1powerfortheblockhalves'andthe control power for the PORV are class IE. This electrical equipment, however, l
is not qualified for LOCA conditions.
THERMAL HYDRAULIC ANALYSIS - R.'TULEY, W ,,
Mr. Tuley briefly discussed the thermal hydraulic analysis of feed and bleed f operation for two plant configurations. The firs was the analysis done for the two-loop 1,876 MWt plant documented in WCAP-9600. The assumptions of
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the analysis were the loss of all feedwater, zero break flow, and the PORV held full open. The results are sumarized on page 17 of Attachment D. For this case, the system rapidly depressurizes and experiences a slight core uncoverywithcladtemperaturesbeingheldbelow13bbF. Two additional analy-ses were performed for a four-loop 3411 MWt plant. The plant characteristics are quite similar to the Sequoyah Nuclear Power Plant. The analysis for this plant was perfonned with the PORV held open and feed accomplished by the safety injection and with the PORV in the normal pressure-activated mode and feed by safety injection. Results are summarized on page 18 of Attachment D. j W concluded that the decay heat could be removed without any core uncovery if l the PORVs were open prior to steam generator dryout. He noted that the quench l tank rupture disc would burst fairly early in the course of the transient .
(about 4-5 minutes after the initial opening of the.PORVs) and that the reactor coolant pump should be tripped to prevent a delay in repressurization of the l primary system. The analysis of the feed and bleed operation with the PORVs in their normal pressure-activated mode of operation indicated that decay heat couldberemohedandcoreuncoveryprehentedforsomelongperiodoftime(at
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least 10,000 seconds). Some core uncovery may be experienced after this point.
The results are sumarized on pages 19 - 20 of Attachment D. }l, concluded'the l feed and bleed operation (PORVS held in the open position) is an effective ' ,_
means of removing decay heat utilizing the existing hardware. They additionally noted that the feed and bleed operation (PORVs in the pressure-activated opera-tional mode) may b,e an acceptable, but not a preferred, mode of de~ cay heat removal. They indicated that procedures would be written for recognizing when this mode of operation should be used and for operating a reactor in this mode. ]
TVA indicated that they would review these procedures and utilize them at the Sequoyah plant if deemed appropriat'e. The Subcommittee noted that the existing hardware is not qualified to operate to the conditions that would exist in the containment af ter the bursti0g of the rup.ture disc on the quench tank. They questioned whether the equipment, although effective in removing decay heat,
- could be utilized in these conditions.
EFEECT OF UHI NITROGEN ON AN ISOLATED BREAK - N. LIPARULO , W Mr. Liparulo indicated that the UHI system was designed to perform in a-reliable fashion and insure the isolation of the $11trogen. Redundant isolation ,
valves are provided and W estimates that the probat,ility of failure to
- I isolate UHI nitrogen is less than 0.57.. A schematic of the UHI flow diagram isgihenonpage21ofAttachmentD. Mr. Liparulo . indicated that an analysis had been performed in which it was assumed that the failure of the UHI system to isolate the nitrogen whereby the nitrogen was injected into the system and that it accumulated in the U-tube vent in the steam generator. Heat transfer under these conditions were estimated. The conditions used in the analysis were:
- 1. An assumption of minimum safeguards
- 2. Best heat system transfer
- 3. Break size equihalent to the low rate of two stuck open PORVs.
Naturalconhectionisinterruptedandthedecayheatmustberemovedbya reflux / condensation process in the steam generator, with coolant being transported between the reactor core and the steam generator by counter-current flow through the hot leg. The blocked area of the steam gener& tor was estimated for various primary system pressures and the available heat transfer areas and required condensation transfer coefficients for the removal of decay were estimated.
The heat transfer coefficients for steam generator conditions were then esti-mated (see page 22 of Attachment D). The available heat transfer coefficient was found to be adequate to insure the effective removal of decay heat. Some discussion followed as to the adequacy of the analysis. Ther'e was . disagree-ment as to the adequacy of the analysis. There was, however, agreement that the decay heat could be effectively removed under these conditions.
DISCUSSION OF THE SUBCOMMITTEE QUESTIONS A number or questions that had been' put forth by the Subcocinittee were discussed. ,
The questions and the TVA responses are included as Attachment E. The high-lights of these discussions are as follows:
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- 1. Qualification of the main steam line check valve - The indicated that these valves had not been physically tested but had been analyzed. The NRC Staff was asked as to the extent which they hadbeenrehiewedbythisanalysis. The Staff indicated that they did not have the information at hand but would respond to the Committee on June 5th. .
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- 2. Vacuum breaker valve capacity - The Applicant indicated that the I valves had been analyzed and found to hahe, adequate capacity for the postulated transients involving the burning of'the hydrogen generated {i from~.a 25?.' metal-water reaction, the inadvertent actuation of the l containment sprays, and the inadvertent actuation of the containment ,
circulation fans. '
- 3. Plant hydrogen supply systems - The Applicant indicated that there was one hydrogen supply line in the plant. Theflowcontrolhalheislocated in close proximity to the supply tanks. Ventilation and hydrogen d&tec-tors are provided in this area. The supply line is presumed to have alowlikelihoodofd$nk,,ingleaksandadditionalhydrogendetectors arenotprohidedinotherareasoftheplant.
- 4. Containment circulation system - The Applicant indicated that intakes were provided at all high points in the containment and that the circulation flow was routed to the recombiner area.
- 5. Use of water for distinguishing fires The Applicant indicated that since Browns Ferry there has been extensive usage of automatic spray j
a systems in the critical areas and that all critical areas in the p.lant had been examined as to the effects of water hose operation. Water f can be used in all critical plant areas at the operator's discretion.
- 6. Qualification of equipment to be used in feed and bleed operations -
EPRI is currently conducting steam / water tests on relief valves. W !
iscompletingapreliminaryreportontheeffectihenessoffeedand bleed operation. The Subcomittee noted that:the PORV electical systems f are currently not qualified for the conditions which would be expected l 1
to exist in the feed and bleed mode.
Vulnerability of the ice condenser containment to negative pressure I 7.
loadings - The ice condenser con'tainment is more vulnerable to negatiie l pressure: loadings than to positive pressure loadings. Failure is likely to be in the buckling mode under negative pressure conditions.
All known events which could result in negative pressure loadings have been analyzed in the containment / vacuum release systems have been found to have adequate capacity.
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- 8. Qualificationofthecontainmentpurgevalhes-SimilarYalhesare installed on the D.C. Cook paint. These valves have been tested in the as-installed condition.
- 9. Seismically qualified auxiliary feedwater systems - The auxiliary feedwater system in the Sequoyah system is seismically qualified.
- 10. Water accumulation in the upper containment compartments - The question was raised to the loads that might be imposed on the containment structures if the upper drain plugs were left in place. The contain-ment is constructed such that 12 inches of water would be accumulatad before it would spill down other bypass paths. This would not result in significant structural loads.
- 11. Auxiliary Control Room - The question was raised as to the capability of the AQxiliary Control Room for overriding Control Room functions andthenatureofthefunctiontransferdehices. The transfer function devicesforthefourdihisionsofsafetyrelatedequipmentarehoused
' in four separate rooms. Each dihision can be separately isolated from the control room and transfereed to the Auxiliary Control Room thus obtaining a significant degree of protection against common mode failures. ,
- 12. Reliability class read /outindicating equipment in the control room -
All safety related instrumentation " control room" is seismically qualified and is Class 1E. About 90% of the instruments in the Control Room are Class 1E and remain available to tha operator even if offsite power is lost. Most of the balance of plant instrumentation and the plant computer will continue to operate for two hours on battery power.
The review of the plant instrumentation associated with I&E Bulletin 79-27 is still underway. - .
- 13. Vulnerability of Auxiliary Control Room to contamination penetration /
seal failures - The review associated with the Lessons Learned recommendations.is still underway. Critical areas in the Auxiliary Building have been identified and radiation maps are being developed.
Additional shielding will be provided to assure access to critical areas. Penetration failures have not been considered. It was.noted that electrical power to non-safety grade /non-Class 1E equipment is
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l being provided through containment penetrations. The possibility for the burn-out of a penetration following the non-Class 1E breaker failure was discussed. It was noted that all large non-Class 1E electrical equipment has backup breakers.
- 14. Environmental Qualification of pressurizer heaters and assorted equipment - The Sequoyah pressurizer heaters are powered by control- '
grade Class 1E circuits. The portions of the heaters outside the pressurizer are not environmentally qualified for accident conditions.
It noted that the PORVs and associated block valves are powered by emergency power in the event that offsite power is lost.
accident for the Sequoyah Nuclear Plant. However, design changes have been made to give the plant the enhanced capability for surviving such an event. The battery powered DC controi and instrument circuits will
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allow the plant to reach and maintain a hot shutdown of all AC power that was lost. The turbine driven auxiliary feedwater pump can run for at least two hours, using only battery power for control and a DC powered room fan to remove heat from the pump room.
- 16. DC battery redundancy - A question was raised as to whether or not '
systems in the plant might be served by only two batteries and if conditions might exist such a common fault could eliminate both batteries. The Applicant has looked at this possibility and has found that there' is no situation. in which the two-battery-lockout situation can arise.
- 17. Post-TMI review of decay heat removal during the design basis flood -
Decay heat removal during the design base flood has been reviewed in light of the post-TMI experience and has been found to be adequate.
- 18. Involvement of plant design engineers in the writing of the emergency procedures - The design engineers under present procedures interface !
both the personnel responsible for writing emergency procedures and the personnel responsible for simulator training. The engineers, however, are not directly involved in the simulator checkout of plant procedures. It was suggested that it might be worthwhile to institute such an involvement.
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32457 :
Fed:r:1 Register / Vd. G. No. 97 / Friday, May 16, 1980 / Notices Purther information regarding topics fe Exchange Floor. Transactions in ITS Washington, DC 20546. Telephone 202/ Pree List issues will be on a strict time to be discussed, whether the meeting 755-2243. has been cancelled or rescheduled, the priority basis as reflected by the
,,,tg gjica, Chaltman's ruling on requests for the timestamp which is affixed upon reCelpt aputy Assacrose AdaunalreforfbrErfarnc/ opportunity to present oral statements of the. order. All transactions in Free IJst .
Ae/obon, leaues will be considered effective and the time aDotted therefor can be my ,gggeo, obtained by a prepeld telephone call to avbject to the review of theloote
. % % ,, ,' the cognhant Designated Federal montage display.nerefore a Member
, , , , , , , . Employee. Dr. Richard Savio (telephone who is in the process of reviewing the quote display after having timestamped 202/634-326*) between 8:15 a.m. and his order may not be displaced by a/ 000 p m., EDT.
another Member who enters an order NUCLEAR REGULATORY V Background information concerning items to be discussed at this meeting whue'the process is in sHect."
CWWSSION dmd n o and Adytaory Committee on Reactor ,
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2.hdurws M WW Safeguards, Subcommittee on the P Organiection NRC Public Document Room.1717 H Sequoyah Nuclear Plant (Unita 1 and Street. N.W. Washington DC 20555 and De Board of Covernors approved the 2); Weeting
- at the Chattanooge. Hamilton County propowd amendments on February 28.
Bicentennial Library,2001 Broad Street. 1980.
%e ACRS Subcommittee on the Sequoyah Nuclear Plant will hold a Chettanooga,TN 37402. S. ne Exchange's Statament of the snetting on f une 2,1980 in Room 104C, 1717 H St., N.W., Washington. DC 20555
- P efe PosedCbses to continue its review of the Tennessee b C. To incorporate into the Rules the Advisory Cacmitsre Manageswnf Officar. g ggg 773 Valley Authority (TVA) applicstion for a license to operste Units 1 and 2. P' D" 8*" * *** ** *"I " free list" stocks, the best TIS SystJm ause coce name** bid or offer at the quoted she must be in accordance with the procedures ~ satisfied so as to avoid trade throughs of cutlined in the Federal Register on October 1,1979. (44 FR 56408), oral or superior markets which may be SECURmES AND EXCHANGE available on other market places.
written statements may be presented by COMWSSION esmbers of the public. recordings will 4. De Exchange a Statement o[the thet No. 34-16797;l'Be No. SR-OSE 40 41 hses of de Mposed Oarge '
be permitted on) during those portions cf the meeting w en a transcript is being Soston Stock Exchange,inc4 He bases under the Act for the kept, and questions may be asked only Propowd Rule Change proposed Rule changes is Section 6(b)(5) by members of the Subcommittee. Its consultants. and Stan. Persons desiring Pursuant to Section 19(b)(1) of the and 11(a)(1) since it calls for an order to Securities Exchange Act of1934,15 be executed in the best market to
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to make oral statements should notify the Designated Federal Employee as far U.S.C. 78a (b)(1), as amended by Pub.1.
No. 94-29,16 (June 4.1975), notice is peotect investors and the public Interests.
in advance as practicable so thst appropriate arrangements can be made hereby given that on May 2,1980, the E. caunenk Racehedfan Memb, to allow the necessary time during the above-mentioned self regulatory ClPants dere m 6posedRde meeting for such statements. organization filed with the Securities and Exnhange Commission a proposed O "#8'8 ,, ,
%e agenda for subject meeting shall ru!s change as follows: No comments were solleited or
' be as foDows: received.The proposed amendment is to Monday, June 2.1980, 8:30 a.m. Until g, g .s Statement of the Terms kic rporate into the Rules an axisting Ce Conclusion of Business. af Substecut the Propmd Rule practice.
%e Subcommittees may meet in Change Executive Session, with any of their
' 1. Textaf&poseddarge 8.#urd e m W h consultants who may be present, to No burden on competition is explore and exchange their pretiminarY An Interpretation of Section 4 of Perceived by adoption of the peoposed epinions regarding matters which should Article XVI of the Constitution to read amendment.The Exchange believes the be considered during the meeting. as follows: osed Rule will benefit the lovestors At the conclusion of the Executive trading ofITS Free List stocks .
Profthe public by executing orders in through supedor markets available in a
Session, the Subcommittee wi!! bear the best market in the TTS system.
presentations by and hold discussions other frS market p! aces will constitute On or before June 20.1980, or within with representatives of the NRC StafL conduct inconsistent with just and such longer period (1) as the Commission TVA their consultants, and other equitsble principles of trade and wiH be may designate up to 90 days of such interested persons. dealt with accordingly - date,if it finds euch longer period to be ,
in addition. lt may be necessary for New Section (b)(21) of Chapter XXXI ;
appropriate and published its reasons the Subcommittee to hold one or more of the Rules to reed as follows: for so finding, or (11) as to which the l closed sesalons for the purpose of " Prior to the time a transaction In an above-mentioned self-regulatory l issue within the Intermarket Treding '
exploring matters involving proprietary organization consenta, the Commfulon information. I have detennioed. in System in which no Dealer la registered by the Exchange is consummated. the will:
accordance with Subsection 10(d) of the (a) By order approve such proposed Federal Advisory Committee Act(Public best ITS System bid oc offer at the rule change. ce IJw 92-463), that, should such sessions quoted size, as shown on the montage display system, must be satisfled if . (b) Institute proceedings to determine be required,it la necessary to close whether the proposed rule changes these sessions to protect proprietary superior in price,by the issuance of an
/ TIS commitment to trade antered from should be disapproved.
Informatian. See 5 U.S.C. 55 b(c)(4).
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- W 4 M ' PROPOSED SCHEDULE
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.. . I 8:30 - 8:45 am 1. Executive Session 8:45 - 9:45 am 2. Status of the NRC full power licensing review 1 hr
- a. Sumary of review status 15 min Including actions on recomendations in the December 11, 1979 ACRS letter - NRC
- b. Outstanding Items and Schedule for 30 min Resolution - NRC
Program - NRC and TVA
- a. Test Program Objectives - NRC 15 min
- b. Schedule for and Description of Low Power 1 hr :
i Tests - TVA c.. NRC Safety Review of the Sequoyah Special I hr Low Power Tests - NRC (1) Summary of NRC Review and Conclusions
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(2)SystemBypassRequirements (3) Impact on FSAR Analysis and Technical Specifications
- d. General Discussion 45 min 12:45 - 1:45 pm LUNCH ,
1:45 - 3:45 pm 4. Discussion of the use of the feed and bleed 2' hrs process for decay heat removal in the Sequoyah )
plant I
- a. Description of hardware used on the 30 min Sequoyah plant (The discussion should include system capacity, safety class, post-TMI modifications, and expected reliability under accident conditions) - TVA
- b. Expected thermohydraulic behavior of the 45 min plant under feed and bleed conditions -
NRC and TVA
- c. Discussion of the capability of the system 30 min for decay heat removal under emergency conditions - TVA and NRC
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I hr 3:45 - 4145 pm 5. Responses to questions in Attachment A -
NRC and TVA ,
EXECtJTIVE SESSION .
4:45 - 5:00 pm 6. Discussion of ACRS review schedule - ACRS 15 min 8
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ADVISORY COMMITTEE ON REACTOR SAFEGUARD MEETING
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INCOMPLETE Nott-TMI ISSUES ON SEQUOY.Y.I UNIT NO.1
- 3. ATWS - REVIEW AND APPROVE
- 1. SEISMIC AUDIT PER ACRS LETTER PERATING PROCEDURES
- 2. POSITION REQUIRED REGARDING
- 9. C0fiPLIANCE OF IE 3ULLETIN FOUDATION MONITORif4G ON SETTLEMENT -
, OSS OF MON dLASS E
- 3. POSITION REQUIRED ON CONT;ItiMENT NSTRUMENTATION & CONTROL ROOM t
SUMP DEBRIS SYSTEM DURING 9PERAT10N
'!' 4. ECCS EVALUATION MODP CONCERNING 10. DIESEL GENERATOR 3ELIABILITY -
FUEL CLAD SWELLING S. POSITION REQUIRED REGARDING PROCESS AND IlORE6/C3-OSS9 l'
CONTROL PROGRAM p
- 6. E^UIP. QUALIFICATIONS COMPLY ';llTH 9230 AND 9236 3 ELATED TO i THE SUIDELINES OF flDREG-0539 %in STEAM & FEEDLINE BREAK ACCIDENTS
- . -. 7 . PAD 3-3 PERFORMANCE CODE - COMPLETE '
EVALUATION REGARLit4G RESTRICTION IN 12. 9-llST COMPLETE REVIEW OF THE USE OF THIS CODE "9-LIST" REQUIREMENTS
- 13. COMPLIANCE OF DIE BULLETit!
' C0-93 RELATED TO 3v-PASS.
OVERRIDE, RESET CIRCUITS
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_ _ _ _ _ _ _ _ . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ e _ -- _ - v, _ _ m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
SITQ4L ITE.T .T,1 N'.luP/L CIEL',TICN ES1 1
0:frTMS
- 2. IONSii%iE EAY HEAL PDU//L C#th1LITY Cf EWAL CIRCULATI0l.
- 2. EGSIW.iE PFESSLE NC ifEL ESP 0E TO A Lf6S OF F0i.3 CIROJLAT10', .
- 3. IBG5TAE FU3 iip FL'W COG (. ECCIPD TO PAlhTAll: AX0.%iE C00 Life LICE MIML CIRCLUT10, C00lT!0iC, L~LlHCGJlry,
- 1. TitlF P!Xi0i OJ.U"i f L iG FRZ T.. PEA 00R f G,1F.
- 2. PPfW.W. FESEJE RE LBR C0;TRCL lii ItTCI'/Tl'..
- 3. FEElT:R SALIE) D' AL7.lLIAP.Y FEIJiG ALT 0'ATi(ALLY C0iitus,
- 4. PAilML CIRCULA1101 ERif10140 FU,lT0FD, t
?
?
i SECIAL TEST M FOR S CIRW LATIQi COOL [G U .
RICTIVES 1, ETTillE N1 EXCDE DETtCTOR l!01CATED FUER CORECT10! FACT TIE SPECIAL TESTS IN ElCH SfSTEM TBTEPATUTS LE.U.S.
TEF TEf3Dlm
- 1. PEAC1Ci POYJ' AT IfWi!ELY ?T.
- 2. All RFf.CTOE. CLCU.9 PJOS IN QUATim.
3, ESTIGLISH A Star.' C00ilG'; VIA STE/E DUSS,
- 4. PEPTWi PRif93' SIE CAL %1tiEICS EV5N 10T DURifG TE C00LDE, 5, ECCC ll&E NO EXCCE PXR CALQJLAT10S ER 10 F DURif6 CD1D 6, STCP C0 CUM WHEI4 450E IS EADE NO EGIN A SLOW EAllF' TIE S#E MTA AS ECORDED IN TIE C00 LIM,
- 7. ALLOW SYSlH1 TEMPAlUE TO STABill2E AT iMAL SY B
FRP TlE ECOP!ED MTA GBEFATL AN EXECE ETEC70R If01C FACTOR AS A FLETIE T TE AVEPAGE GE LEG TDTEPATU
-gte amp ee w e w ymp , e.
- 4** 4Blee%._
f SPECIAL EST tO. 3 NATuffi. ClR01AT10; WITH LDSS T FRESSURIZEP. KATEPS OPJErTIVES
- 1. DEPDSTPAE ABILITY TO FAIKIAIN WEf( CIRCUL (EFTBS,
- 2. DEiEPHite RCS EPolSSURIZtTIO. FATE /FIER P
\
II !f'rE.
- 3. IUL: EF.^iTE S/E'G10 l&I!; Ctl, if CD~ ROUE USIIf; Pf,1!E. O!G!G f W NO SECUM SlU". FL'F, JIIT ItRMEE
- 1. TRIP RCP's NO PWSSURIZER EATEFS WITH REAC
- 2. EITOR PRit','O SYSTEI' DEPRESSUR12 Atla, PATE A'0 PARGIN
- 3. SWA.Y REDUCE REACTm fGE TO 1.5% IN W FIRST Lt. MG SATUPATION t%RGill UMlT (Fm TEST) IS TA0iE ,
CSElt0 RffM0 STEAM DlfF.
1
J .
- SPECIAL TE910, 5 MTiFAL CIRCLUTIG4 AT EDUED PESSUE crEcTIVES
- 1. ICETPATE TFE AElLITY TO PAlt?TAIN iMAL CIPILMTIG; AT REDUID PR SYSTB' PESSi1E, .
- 2. IEL5TPArt. tie tbE T llE SAltPAT10,' TETER TO KI;17.1, N%20: TO S4TiPAl
- 4. IBMTPATE Tlt EFECTIVDESS T 0%%I10 NC SECrWR( STE/f', FLN TO CGI SAILIATim PARGlii, TEST MS31PTIG,
- 1. TRIP EACTOR 0)Di.A!iT PafS WITH Em AT 3% POER,
- 2. IHESSLRIZE THE RCS EY RifiltE TF PESSURIZEF. KATEPS N0 FOSSIBLY m AUXILIARi SFFAYS,
- 3. SLCWLY FEUE EACTOR FGER TO 1.5% IN TIE FIRST KUR 7 T
- 4. KTilTOR SAMATim PARGl!i LSit0 SAUATim WiER ND PANiAL
- 5. ItN>SE 0%RGltG #0/0R STEAM FLOW TO llNEEE SARinAT!W UP.lT IS EAGG.
m
9 SPE.Cl/l TEST ND 4 Sl. b ik di 1
@JrTIVES
- 1. IETERMitE EFFECT T STET CBDATOR ISCOTICU ON tATLPAL C;ROAATil.
- 3. DEPDGITATE THE E-ESTAELISif D5 T tATUPAL CIROJtATlGi IN /;i ISCLATED STE#1(UEFATE.
ET TrfryJEIXfi
- 1. TRIP REACT 0'l C00LA!!i R.!R KITH PFACTOR I,T li FATO POE,.
- 2. ISCLATE SECGENN SIDE 7 UF TO TW STU!: CBDATORS ALL'El!6 EACP T ISOTEWAL Ca?ilTIUS.
- 3. WRIFY !ATUPAL CIRCULAT10t; IN ACTIE LOTS.
f4. PETUP,1 ISOLATE STEAM CODATORS TO SERVIE GE AT A TIE.
5, VERIFY E-ESTAFLISitBIT T iMTlPAL CIROJtATION IN TE UNISOLATED LOOPS.
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SPE.C] AL EST F3. '2 NATUPAL CIRCULAT104 WITH Situ.ATED LOSS OF 0FFSITE AC POWER ,
O'd FJilVES .
- 1. D0thSTRAIE NATUFR. CIRCULATICG CNi BE EST/6LISh3 #0 l'AINIAl!ED DURINS A LCES OF OFFSIE P38,
- 2. trG ESTOPAT10'l OF CFFSITE P3ER, EER3EiCY LOOS CRi EZ T%'rfEEED TO T: SITE PGER ND DIESE'_ GUEMTOR3 ESTORED TO ST# ruby STATU5.
IEST_IUT.s!EfiQi
- 1. TRIP PEACTOR CCG.NU PUMPS NC INIT! ATE A BLACK 0.ii 016.9-iN % S. 7D%' WITH TE FOLLGUNG IfilTl/L CGCITICiC:
A) REACTOR AT 3% P0r/ER.
B) ALL REACTOR COOLANT PLTPS 04.
C) AUXILIA'Y FEEDIATER SYSTEM (M7TCF.-DRIVEf4 PUMPS OILY) IN SERV CE (ON OFFSI PacR)
D) PRESSURIZER HEATERS IN SERVICE (CN OFFSITE POER) .
- 2. Vi.RIFY DIESELS START #0 ENERG12E 6.9-xV SWID0hii EMRDS.
- 3. VCRIFf DERGEEY EOU!WE.Ni SEGBiCED QiTO SWiID41 BOARCS.
- 4. VLRIFY ESTABLISHEfi 0F NATURA;_ ClkCULATIQi NO PAINTAIN STEADY STATE CUDITIC3.
._____._____________.__. yey e> ,,,- e e *
- e,m , , ,gggq.e e,.s
-* * %91a.6 eMbre WBDe up = 'N*4 N
b SKCIAL EST iD 7 SlFUtiiED LfE OF ALL QCITL NiD TFSITE /E PDG OrJFCTlWS
- 1. IUDGTFATE TTAl FOLIDil6 l.02 I ALL (tSITE MC TFSITE FGE DEGBEY DlESEL GCE/,~Cii, ME1 CIRCLUT101 CN! LE m!!JAl'8 TO PF06 /
l KCAY lEAT.
2, VERIFY 10i STNSY C00lT102 CM Et i AlfJAltE' FY l%'ili CM!RDL T Alt /lLIARf FEEBATER NO STFX FLO' l
- 4. VERIFY TMT llE 125-WLT VITAL FATiEFY Cfli SLFLY TIE BEiCBD LO JFSTIFSCRIPTid.
- 1. WilB llE RCACT0f AT A%0X1mrtLY 1% POER Sif0 LATE LDSS T TFSITE K. FGU.,
- 2. ASWE FNU.L CUiiROL T AUXILl/4,Y FIH@iFR LEWL CD~ 1 IfTER LML QUTG. .
V;1\E Fall CUE.
- 3. ION C03 POL AIP. StPFLY TO AlKSfk11C REllEF VliVES NO /S$ttE PNWL CC
- 4. PE-EST;dtlSl; PETRIP 5Fuii GBUATOR LEWLS /G iCS COLu Lf6 TUTUAliffS.
- 5. FAih7Alh PETRIP LEVEL.S FOR /KROX1i%TELY 2 KlTS, f;. [O13 l%lli STtN'i PES 9JRL C00EOL NO AUXILIARY FEDT.TER FLOW CallPbl TO ALITu% TIC,
- 7. RESTORE turit PQG TO ALL VITAL f(SES, l n.- J ~ d , & -g ;
ra% A w
l - - - ~ ~ ~ " ~
l
6 SPECIAL TEST 10. E ESTABLISit!' 7 T MlUP/L CIPCL'LATlU' FfE STAcirdiT CODITIG6 l
(MC11\ES l
VERlW llE ESTABLISflCR & MUAL CIRCULATIG1 FRf, STACEi 00 Fla] GOITIUS lh THE PfslhW S/SE.
tut EEGn1-1:
- 1. TRi? ALL TiriOE UUJF fitPS K!Tri PEACTM CRITIC /L AT HOT ZEF.0 2, IS0UV Id.L 9EA" CDFATOPS (SEC00/5 SIfD.
- 1. WFIR IS01.d!M. C'.lDlilG.E II: EACH Mll'RT LCCP, L:. UE 191. h Sib.'. IS'.UTid! VALES A$ PLAG. AEILIAD FE%TEP. CQJin Fl. TIC. -- C -<d<-y
- 5. EE.CIN A Sla.+CLliTFDL n ., MA . u eWITitiML
, . ~% GJ1L PEACTOR KE IS AP 2..... v +
6, REECCE STEAN CBEPAlm FfESSLRE IF EQUIRED TO ELP INDUCE inRIML CIR
- 7. VERIFY ESTAELISifE!E T MTURAL CIROJLATlG; Al0 PAlfRAIN FOR APPROX 15MitUTES, I
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s 1
i
\9 SPEC 1AL TEST t10. 5
/ '.., e ->
^ l C00LDOW'; CAPABILITY OF TE CHARGING NO LETIDr.
OB E TIVES IBOSTRATE TE. CAPABILITY OF 0%RGlNi AfD LETDO.d SYSTEM TO C00LD REA0 TOR CJJLA'iT SYSTEN ESIDESCRIPT10:1 1 TRIP THPE EACTOR C00LRTi PJM WITH PEACTCP SLBCRITICAL ND TE PR SYSTC1 AT HOT STNDEY CCtalT10;S.
- 3. IfUSSE CHARGIN3, LCFRi TO FAXIM' CAP 3. CITY, 14 . AhcR 30 MINJiES OF PAX 12', CHARGi'iG & LEIDE REDU;E TO MliilW; CKRGINS A'S LEFR1.
- 5. AFTt.R 30 MliUTES OF MINlE 0%R31NG & ETDO.R RETUP11 TIE STEM GENE TO SERVICE #D RESTART REACTOR C00LAKT F9TS.
4 e
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J SPECIAL TEST ID, 9B D0?.CN nlXI!5 KD CC:CCC U? DER IMTUPAL CTRCULAT10N CO?
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rcJECTIVES
- 1. DEFONSTPATE TPAT TE REACTCR CCOLaW SYSTE1 C#
MTURAL CIRCULATION, 2, DEF.3:STMTE TE CA? ABILITY TO C00LD241 TE PRlFARY SYSTEM Gl M USlis TE STEa?i EiEMTOPS, EST DESCRIPT10'i 1, IMTllRAL CIRCULATION IS ESTA3LISED WITH REACTOR POR AT APPROXIFA 2, VERIFY TE EACTOR CCCLK!T SYSTEM #D PPESSURIZER BORG) CONC 91 W1T}{lil20 PPM,
- 3. BIRGIZE PPESSURlIER HEATERS ND lillTIATE AUXILIARY S MIX 1tB EEPIEN TE PEACTOR C00L4fT SYSTEi ND 11iE PRESSURIZER, .
f4, lillTIATE SOSATION AT AFPROXIPATELY 500 PCM/HR, 5, WITHD3AW CGITROL M!KS TO FAINTAlil REACTOR POB CONSTNTT, 6 SMFLE REACTOR C0]LANT SYSTEM (HOT LEGS) #D PRESSUR VERIFYMIXIl5, 7, TEFMliMTE BCPATICN IFTER 103 PPM IllCPEASE IN REACTOR C00LN(T SYS CONCSITPATIC1, 1
8 STABILIZE PRit'ARY SYST31, 9, lillTIATE C00LECal BY li:CREASI:3 PATE OF STENi DLDP, 10 TEF?ilMATE CCOLC0d! hiEl FRif%RY SYSTEi g T ve EE WF, l
11, BRl:3 REACTCR SL2 CRITICAL KD EATUP TO NDP?AL tD-LOAD TBPERATURE, l
. . . . . . . . .-- .-----4.
oe .
. \* l I
SAR EVAUJATION RESULTS --
- 2) LDCA A) LOW PROEABILilY EVBU B) SUFFICIBH TIE AVAILABLE FOR OEPATOR ACTUATI0i 0F SAFETY IRJECTIQ1
- 3) STF#iINEBREAK A) LOW PROBABILITY EVBiT B) SUFFICIBE Tif E AVAILABLE FOR OPEPATOR ACT101 - ACTU SAETY INJECTIG1 NO LOOP ISOLAT101
- 4) 04CURROLLED RODlilT10PMAL A) LCW PROBABILITY EVBiT: TEST 8 #0 9B FDST SUSCEPTIBLE ROD FDGER E0JIPSETS i B) EVENT EVALUATED FAV0PABLY
- 1. DNB IS NOT EXECTED '
II, ADEQUATE HEAT TPMSER E01 ASStrilNG UB 111, NO SIGilFIC## FLEL DWAGE EXPECTED
- 5) 800 EJECT 101 ,
A) VERY Lai PFDBABILilY EVBU
~
B) 0FFSITE DOSE LOW EVEN ASSUMING thXIfUi C# ETIVITY R
EDJIPtB1T FDDIFICATIONS EESSARY TO EFFORM TESTS l l
l 1)
SEVEPAL EACTOR TRIPS,t0DIFIED TO PEVENT SPURIOUS TRIPS FKN INCOR INDICAT10tS OF SYSTEM STATUS DUE T0 l.0W Fl.0W ,
- 2) MITQ'ATIC ACTUATION OF SAFETY IRJECTim Bl.DCE TO PEVENT SIDUETOTESTCONDITIGS -
A) P%NUAL ACTUATION OF SI AVAllALE FOR OERATOR USE B) AUT@% TIC REACTOR TRIP OPEPABLE AT SETP0llf C) .Sl OW4NEL BlP STATUS INDICATION AVAll.ABLE FOR OEPATO
- 3) SI 01 STEAM FLOW IN B0 STE# LINES - HIGH COINClIDIT WITH STE PESSUE - LOW FDDIFIED TO ACTUATE AUTG'ATIC EACTOR TRIP O PESSUE - 1.0W OftY
- 4) TE0filCAL SPECIFIC Tim E)O1PT1006 EQUESTED TO REFl.ECT THE A P0DIFICAT10G
- 5) OPEPATOR ACT10f6 ARE EQUIE UPON EACHING SPEClflC VALLES O l
PAP #EERS TO EPLACE AUT@% TIC FUNCTICNS FDDIFIED ABOE w -- , , , ,
- M W4## W *
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g OPEFAT10f6 FDDIFICAT10tG ECESSAR( TO EFFORi TESTS
A) CLOSE PESSUR12ER POER OPEPATED RELIEF VALVE IS0lATim VALVES PELOW P-11 (i 1970 PSIG)
B) FEED FLOW THRCUGH FEEDWATER C0fER0L VALVE BWASS VALV O LIMIT CONTROL RCD INSERTION - w ~ ;-
' ( , , ,, ,, ,4
- 2) ESTABLISHED MINltlJM OR MAXIM VALUES FOR PPDCESS PAP #ETEP A) MINIM PRIPARY SYSTEM SUB-COOLING, MAX 1tui LOOP AT, RCS TAVGs AND CORE EXIT TEFFEPATUES TO EPLAE Q)RE LIMITS #0 DERl EPATUE AT EACTOR TRIP TO ASSUE PARGIN TO SATURATION B) MAXIM lt01CATED NBRRCN FLUX P%ER LEVEL TO EPLACE OE ATREACTORTRIP O MINIM STENi CBEPATOR MTER LEVB.TO ASSUE ADEQUATE SE SIDE EAT SINK D) MIN 1t.Ui PESSUR12ER MTER LEVEL TO ASSUE SUFFICIBR PE LEVEL TO PAltHAIN PESSUE CONTROL D MAXIM CONTROL RCD INSERTION TO ASSURE SLFFICIBE SHJ
- 0 PAINTAIN A ZEF0 MDEPATOR TEPPEPATUE COEFFICIDE
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' SEQUOYAH NUCLEAR PLANT j,
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UPPER COMPARTMENT !
A A CONI AINMENT 572 AT OtVtOER DECK 10 AttaOSPettt MAIN STE AM '
etttgp AND SAFETY I
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. .SEQUOYAH NUCLEAR PLANT PRESSURIZER RELIEF VALVES , .
' TO RELIEF TANK a .
I TVA I,,
CLASS 3"
G (NNS)
! CLASS A
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BLOCK VALVE TRAIN B BLOCK VALVE TRAIN A 3.. {3 l
8" SAFETY VALVE LINES
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4 2600
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- 1 i
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STAFF POSITI0tl Hydrogen Control Measures for Sequoyah -Me h jet TT*E.,
I INTRODUCTION In the staff's Safety Evaluation Report on the Sequoyah Nuclear Plant, Units 1 and 2, dated March 1979, we stated that the combustible gas In its letter, control systems for the Sequoyah station were acceptable.
" Report on TMI-2 Lessons Learned Task Force Final Report," dated Decem-ber 13,1979, the ACRS reconsnended with respect to hydrogen control measures that ".... special attention be given to making a timely deci-sion on possible interim measures for ice-condenser containments."
The staff has reviewed the matter of hydrogen control requirements in light of the TMI-2 experience. The staff's findings are reported in SECY 80-107 dated February 22, 1980. With respect to the Sequoyah and other ice condenser plants, the staff determined that the existing hydro-gen control measures that satisfy Section 50.44 of 10 CFR Pa'rt 50 trc sc-ceptable for full power operation, pending completion of certain studies ,
to be ' performed by the staff, the Sequoyah applicant and other ice con- .
denser owners.
II DISCUSSION In this section, the current status of the hydrogen issue, certain related study programs, a rulemaking proceeding, and TMI related safety improve-I ments will be discussed, l
A. CURRENT STATUS
- 1. SECY 80-107 In the staff's paper, " Proposed Interim Hydrogen Control Require-ments for Small Containments," SECY 80-107, dated February 22, 1980,
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scoping analyses were performed starting with the assumption that an accident involving a severely degraded core existed in each of the six classes of containments considered. These classes of containments include the Mark I, II, and III con-tainments for BWR's and the ice condenser, sub-atmospheric, and dry containments for PWR's. We concluded that inerting should be made a requirement for the Mark I and II classes of containments and that no ' additional requirements should be re-quired for the other classes of containments pending the up-coming rulemaking proceeding outlined in Task II.B.8 of the TMI Action Plan, NUREG-0660, dated May 1980.
In its risk-based studies, the NRC's Probabilistic Analysis Staff concluded that inerting the Mark I and II containments would not reduce overall risk. It was also their finding, how-ever, that overall risk would be reduced by inerting of the ice condenser plants.
Other elements of the NRC staff believe that although risk-based studies are worthwhile supportive studies, there remain substan-tial uncertainties in their ability to adequately treat actual accident sequences and operator intervention.
l l
The NRC staff concludes' on balance 'that the actions called for in the above cited SECY 80-107 relative to ice condensers, and particularly Sequoyah, Unit ~1 should proceed pending the out-come of the continuing studies in this area, i
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- 2. Hydrocen Each of the Sequoyah units is provided with a pair of re-dundant electrically heated thermal recombiners that satisfy the provisions of 10 Cr Part 50.44. Moreover, the purge sys-tem in these units can serve as backup systems should the re-dundant recombiners be unavailable. This combustible gas control system can accomodate up to abcut 1.5% metal-water reaction in the reactor core while maintainirg the hydrogen concentrations below the lower flarmbility limit of four per-cent.
- 3. Best Estimate of Existing Capability In the above cited SECY 80-107, we reported that the failure pressure for the Sequoyah containment was estimated to be 36 psig(thedesignpressureis12psig). We find that as much as
25% mr tal-water reaction can occur without exceeding the failure pressure of the Sequoyah containment, even assuming combustion . .
of the hydrogen. ,
- 8. PROPOSED STUDY PROGRAMS
- 1. Staff's Program The NRR staff is preparing a User's Request to have its Office of Nuclear Regulatory Research augment the existing programs on hydrogen control. This will be a substantial program of studies i
directed at developing an information base for use in the upcom-ing rulemaking proceeding, cited above. It will also call for l early treatment of those hydrogen mitigation measures suitable I
P' ' ""8 9' M '"7'- ~T-- W a ,, ,
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. . 4-for use in hydrogen control at ice condenser plants, with a completion milestone targeted for the end of 1981.
Among the mitigation measures that will be investigated in l the early phase for ice condenser containments are:
- a. Hydrogen combustion systems;
- b. Atmospheric fogging systems;
- c. Halon suppression systems;
- d. Inerting;
- e. Filtered-vent systems; and
- f. Other systems.
The advantages, disadvantages, and functional capabilities of each of these mitigation systems need to be determined in terms of their use in ice condenser containments. The inert-ing approach for example, which has been demonstrated to be a workable system for the Mark I/BWR containments, may not be a good choice for the ice condenser containments. The ice con-denser containment, being about four times 'f arger than the Mark I containment, has much more equipment located inside con-tainment. Containment entries need to be made several times a week for the ice condenser (maintenance purposes) versus about
. fise times a year for the Mark I containment. In our view, se-lection of the inerting approach or any of the other approaches at this time would be premature and inappropriate.
l C. RULEMAKING PROCEEDING In accordance with Task 11.B.8 of the TMI Action Plan (NUREG-0660),
'rulemaking proceedings will be conducted to determine whether and how the staff's existing design bases need to be changed to accom-modate those accidents involving severely degraded cores and melted cores. One of the principal items in this rulemakinp proceeding is the matter of hydrogen management for all classes of containments. ..
Although not yet established, the schedule for this proceeding is expected to range over two to four years.
D. TMI RELATED SAFETY IMPROVEMENTS As a result of the recorrnendations made by the staff's TMI Lessons
- Learned Task Force, and actions taken by the staff.'s Bulletins and Orders Task Force, a substantial number of safety improvements have already been implemented and will continue to be implemented at all ,
operating and new reactor plants. These improvements include changes in hardware, operating procedures, and operator training, which con ..
tribute to making more remote and acceptable the like'lihood of ac-cidents that involve severely degraded cores. Details of these im-provements are described in:
- 1) NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report l and Short-Term Recommendations," July 1979; f
- 2) Letter to All Operating Nuclear Power Plants from D. Eisenhut.
2 Acting ' Director, DOR, September 13,1979 (transmitted L re-quirements and clarification); and
- 3) Letter to All Operating Nuclear Power Plants from H. Denton, Director, NRR, October 30,1079 (further clarification of re-quirements). l l
- r. .. .. . .
i 1
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. 6-III CONCLUSION On the basis of the above discussion which indicates that the likeli- i hood of severely degraded accidents has been made acceptably remote, and that a substantial study program will be undertaken on an accelerated schedule by the NRC staff as well as by TVA and other owners of ice con-denser plants, the staff concludes that no additional requirements be-yond those of the currently effective 10 CFR Part 50.44 need be imple-mented for the Sequoyah plant and.other ice condenser plants, pending completion of the study programs identified above and possibly the rule-making proceeding also identified above.
Since the matter of full power licensing for the Sequoyah plant will have to be considered by the Commission, we request a statement of the ACRS views on the staff's position as outlined above.
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