ML19296B828
| ML19296B828 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/16/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1664, NUDOCS 8002220151 | |
| Download: ML19296B828 (19) | |
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ISSUE DATE:
10/16/79 MINUTES OF THE AD HOC SUBCOMMITTEE
- [ogg MEETING ON THE THREE MILE ISLAND, UNIT 2 ACCIDENT IMPLICATIONS REGARDING NUCLEAR POWER PLANT DESIGN
% @ - [ b b 4,y THURSDAY, JULY 25, 1979
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A meeting of the Ad Hoc Subcommittee on the Three Mile Island 2 Accident Implications regarding nuclear power plant design was held in Washington, DC at 1717 H St., NW on
- July 26, 1979. The purpose of this meeting was to consider the long-tem implications of the THI-2 accident on nuclear safety design, policy, criteria, safety research, and plant operation. Notice of this meeting appeared in the Federal Register, Volume 44, No. 135. The schedule for discussion and a list of attendees at the meeting are attached to the minutes. No written statements were received from members of the public and no requests were received from members of the public to make oral statements. The Subcommittee did not issue, approve, or receive any written reports during the meeting. The Designated Federal Employee for this meeting was Mr.. Richard K. Major.
MEETING WITH THE NRC STAFF (Transcript pages 1-9)
Dr. Mattson stated that the NRC Lessons Learned Task Force will complete its review by early September. The Task Force is a retrospective review and has only briefly considered new Cps.
PROBABILISTIC TECHNIOUES TO PROVIDE INSIGHT INTO THE RELIABILITIES OF SYSTEMS AND COMPONENTS - H. Krugg, NRC Staff (Transcript pages 10-49)
Mr. Krugg began by noting that a small data base is available and if prob-abilistic methodology in an area having large uncertainty is employed, you may arrive at a conclusion with a large uncertainty, in which sig-nificant engineering judgment may be required. Design basis accidents are formulated using engineering judgment based upon experience and a detailed knowledge of the design, as those events which can reasonably be expected to provide an upper bound on limiting system states. The NRC Staff intends to use probabilistic methodology in licensing by de-fining design basis accidents to test the adequacy of the proposed design, I
t 8002220.;
The DBA is used to represent the character of potential challenges to those systems designed to provide a specific function important to safety. Definite reasonable acceptance criteria were not provided by the NRC Staff at this meeting. Although it is impossible to identify every failure path, probabilistic analysis helps provide the necessary assurance of adequate reliability.
Mr. Mattson commented that reactor safety is considered a defense-in-depth system. Because of this defense-in-depth philosophy, design bases have not always been coupled. A review of the hydrogen aspects and the degraded core consequences of Three Mile Island suggest that the containment design basis needs strengthening and perhaps decoupling from ECCS or other design bases.
Dr. Siess suggested (in a question) the uncoupling of containment design from the unrealistic siting criteria.
The Lessons Learned Task Force believes that it is appropriate to impose additional requirements on non-safety systems. Notable examples are the residual heat removal system and the reactor coolant makeup and letdown systems.
The NRC Staff recommended the continuation of the design basis approach such that it be used for specific systems to show adequate design and not across-the-board acceptability. The design basis includes the' design and operation of systems which are discrete functions. These functions should be addressed separately and discretely in instances such as the feedweter failure at TMI-2.
The Staff said that Three Mil'e Island showcd us that failure modes are scphisticated and that we cannot freeze the number of DBAs. Although a specific task force is not looking for new design basis accidents, ATWS, loss of all AC power, safe shutdown earthquake, have been indicated as poss}ble areas of, additional review.
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.a-2 Mr. Michelson asked about the analysis of pipe breaks outside of contain-ment. Dr. Mattson pointed out that the analyses of these systems are not perfonned with great rigor and that further review must be factored into the safety assurance aspect of the licensing process.
HYDROGEN DESIGN BASIS - H. Krugg, NRC Staff (Transcript pages 49-64)
The NRC Staff proposed that a hydrogen generation design basis for the con-tainment of about 50% would be more appropriate rather than the 1% metal-water reaction associated with core protection by the ECCS. When the final number is picked (5 to 100% metal-water reaction) for the containment design basis, careful attention must address the retrofit of existing designs or the rationale for not retrofitting.
ACRS members questioned the Staff on the amount of the metal-water reaction at THI-2. John Voglewede (NRC Core Performance Branch) said that the figure is about 40% based on the summation of the burn, bubble, and contaimnent volumes. A large uncertainty, however, still exists. Since the Staff approach has been that the possible amount of hydrogen generation must be considered in the containment design, the possibility of inerting PWR con-tainments must be addressed. The Subcommittee proposed that the full Committee should try to prepare alternative approaches and possibilities with regard to hydrogen generation and containment design.
CONCEPT CHANGES IN FUTURE LWRs - H. Krugg, NRC Staff (Transcript pages 64-75)
Examples of possible concept changes:
1.
shielding design review of certain systems such as makeup and letdown.
2.
high pressure bunkered RHR system capable of full system pressure inside of containment.
3.
concentration on degraded states vs. initiation of transients and accidents.
4.
energency procedures keyed to specific initiating events rather than mcderate frequency anticipated events.
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4 Generic Lessons Learned from TMI:
1.
increase attention to LERs 2.
get operations organizations to do a better job 3.
individual licensees evaluate historic experience on th" plant and from plants of like design.
Commissioner Gilinsky (in a cited quote) indicated that he thought the NRC would have to have a more detailed knowledge of the behavior of the reactors it's licensing. Dr. Mattson agreed that the Staff needs to build its practical understanding of day-in and day-out operations in nuclear power plants.
DEBRIS FROM PROCESS LINES - Jai Rajan, NRC Staff (Transcript pages 75-96)
The Staff described the extent to which attention was paid to the details of the seismic or blowdown capability of smaller components within the reactor coolant system of BWRs and PWRs. These smaller components include internal appurtenances and devices installed in the reactor coolant system which under blowdown condi-tions could possibly migrate to pumps, valves, and the reactor vessel and have potentially damaging effects. The ACRS members asked if the Staff and/or the Applicant perform analyses on these devices. The NRC Staff said that they accept the blowdown loads calculated by the Applicant.
The ACRS requested specific information about the San Onofre Plant (LER 78013).
The NRC Staff will discuss this LER at the next meeting in addition to providing a written reply. This report will address the basis for excluding secondary system appurtenances from blowdown load analysis.
Dr. Mattson stated that the NRC Staff had not done near the degree of audit calculations in the stress analysis area as has been done in the thermal hydraulics. Water hammer and two-phase flow instability were considered as generic items under the review of the NRC Staff.
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. STEAM GENERATOR OVERFILL AND THE INFLUENCE OF CONTROL SYSTEMS O John Olshinski, NRC Staff (Transcript pages96-123)
Steam generator overfill was first addressed in BSAR-205 as generic to all.
B&W once-through steam generators. Because of the complexity and difficulty in evaluating the secondary side effects, B&W has used an approach of pre-vention rather than mitigation. B&W has arrived at a conceptual design for overfill detection ano prevention and has gone out to seek bids on the design. Mr. Olshinski didn't think that the NRC Staff had undertaken an extensive evaluation of steam generator overfilling. The possibility of this accident, a s raised in BSAR-205, has not yet been answered by B&W. The probability of a steam generator overfill accident for Westinghouse U-tube steam generators has not been reviewed.
The adequacy of the current safety grade /non-safety grade classifications is being reviewed.
Is the list of safety systems complete? Should PORV, letdown system, and waste gas systems be included in the safety grade classification? While Section 7.7 of the Standard Review Plan considers the safety grade classification, there are no guidelines that generally apply to many other non ' safety systems.
The NRC Staff stated that the sequence of events in Chapter 15 of the SARs bears little similarity to the accident at TMI. 'The use of non-safety grade syst, ems should be reflected in these analyses.
Examples of the interaction between safety grade and non-safety systems:
1.
PORV sticking open 2.
auxiliary boiler unavailability (loss of vacuum and heat sink)
A great deal of work needs to be performed in the area of control system reliability. Transients where control systems fail to respond properly have not been analyzed. The NRC Staff does not have sufficient manpower to readily accomplish this task. The use of technical assistance contract' ors teamed with the Staff may alleviate the manpower problem.
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. BWR REVIEW - W. Hodges, NRC Staff (Transcript pages 133-190)
BWR Recommendations:
1.
relief and safety valve testing under design basis transients and accidents 2.
direct valve position indication 3.
instrumentation for detection of inadequate core cooling 4.
diverse containnent isolation 5.
nonessential systems review 6.
dedicated H2 control penetration for venting or for recombination 7.
inert containment for Vermont Yankee, Hatch 2 Mark I, and Mark II containments 8.
plant shielding review 9.
combustible gas control recombiner system
- 10. post-accident sampling of containnent atmosphere and reactor coolant in less than one hour and radiological spectrum analysis in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
- 11. ingrease the range of high range effluent monitors up to 100 rad per hour inside the containment
- 12. improve airborne iodine instrt, entation
- 13. transient and accident analysis for off-normal conditions (small break LOCA, inadequate core cooling, operator actic::s during accident conditions) 14.
feasibility of an on-line fuel failure detection capability
- 15. re. duction of the administrative duties of the shift supervisor
- 16. shift safety engineer or technical advisor to be on duty at all times
- 17. upgrade shift turnover procedure
- 18. limited access to the control room There was concern for overfilling transients for NWRs. A high water level in the reactor vessel occurred during the startup of Dresden 2 in the 70's. GE installed a non-safety grade high water level trip.
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The loss of feedwater transient for BWRs with a relief valve stuck open was discussed. If the relief valve is fully open pressure will decrease but if only partially open (or leaking) the pressure could remain constant, decrease initially and then increase. The indicators available to the operator will conflict depending upon the size of the break. The main indicator is the water level in the vessel, the operator cannot detect a stuck-open relief valve by suppression pool temperature or temperature of the discharge pipe since these temperatures are expected to remain higher than normal.
The rate of increase of these temperatures is the only real difference. A stuck 6 pen relief valve will increase the suppression pool temperature about 20F per minute. The failure of HPCI must be considered since an automatic actuation of the depressurization system may not occur.
(The depressurization system relies on concurrent signals of reactor vessel low water level, high containment pressure, and LPCI or core spray pump running). The relief valve discharging to the suppression pool will not cause high containment pressure so that depressuriza: ion must be actuated manually.
The water level in a BWR is generally measured in the downcomer region outs.ide the core shrouds. Level indication in the downcomer~ region generally gives good indication unless it is effectively isolat.ed from the core region. In the Oyster Creek event the discharge isolation valves in each of the five recirculation loops were closed. The downcomer was essential.ly isolated from the core region.
The utilities have been asked to address different types of transients:
1.
norma'. loss of feedwater 2.
loss of feedwater, offsite power, HPCI, and RCIC 3.
number 2 with stuck open relief. valves 4.
number 2 with one stuck-open safety valve 5.
loss of onsite AC and the failure of RCIC and HPCI 6.
loss of all AC
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8-The concern of isolating a pipe break LOCA and the ability of the ECCS to provide adequate core cooling after core repressurization was not considered serious by the NRC Staff.
BWR Items of Concarn 1.
identify various failure modes for the ADS (loss of DC power, loss of air pressure, etc.)
2.
stuck-open relief valve does not result in high containment pressure -- need other signals besides high containment pressure to isolate the containment 3.
identification of all lines which transport radioactive materials from containment 4.
recovery from an inadequate core cooling transient 5.
the possibility of radioactive material transport into the RCIC and HPCI steam systems during a LOCA since these steam lines are.not isolated The NRC Staff has requested the operating procedures for loss of feedwater, loss of coolant, loss of instrument air, stuck-open relief valves and loss of offsite power from Hatch 2 and Dresden in order to upgrade all procedures in general. A member of the NRC Staff felt that the casualty procedures are too specific event-oriented.
PRESSURIZER SURGE LINE DIFFUSER - G. Holahan, NRC Staff (' Transcript pages 190-196)
After discussion with various vendors, the NRC Staff has concluded that the pressurizer will not drain as long as the pressure is above 1400 psia and an opening exists at the top of the pressurizer.
It looks like there will always be water in the pressurizer under these conditions even if the core is com-pletely empty. This fact is being brought to the attention of the operators.
SINGLE FAILURE CRITERIA - C. Long, NRC Staff (Transcript pages 196-212)
A multiplicity of errors occurred at TMI. The category passive failure has not yet been defined even though it has been in the regulations since 1969'.
Passive failures, active failures, and operator action should be considered and should b'e incorporated f oto the single failure criterion philosophy.
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Regulatory Guide 1.33 addresses the need.for operating procedures for a safe shutdown during earthquakes.
The NRC Staff will probably look at the steam line break accident followed with a small primary break.
A new ANSI Standard (June 1979, Single Failure Criteria for Light Watar Reactors) has just been published for review. The NRC Staff had not voted on it as yet.
Administratively controlled components (e.g., locked open valves) and operating and maintenance procedures need to be considered in the single failure criterion.
A modification of the single failure criterion has been proposed which allows single failure in more than one system at a time.
AUXILIARY FEEDWATER SYSTEM REOUIREMENTS - M. Taylor, NRC Staff (Transcript pages 212-246)
Mr. Michelson asked why B&W maintained 30 inches of water in the steam generator while the auxiliary feedwater is running? If you controlled.at 6 feet (or some other height) it could buy you time in case of casualty.
Questioned referred to NRR.
Mr. Taylor stated that the unreliability of the auxiliary feedwater system was generally controlled by common faults in the two-train system. The reliability of the steam inlet valve to the turbine driven auxiliary feed pumps is addressed in an upcoming NRC report.
Several diesels in five plants have an AC dependency for lube oil cooling.
Several other diesels have dependencies on air operated valves failing shut after 30 minutes becar e of loss of air.
Some AC operated valves e
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failed open on loss of DC. No significant ventilation requirements were noted for the turbine electronic governors.
Mr. Taylor felt that a probability of 10-5 per demand was sufficient fcr auxiliary feed system startup unavailability.
The ACRS asked the NRC Staff to consider the uncertainty as to whether the HPI system would have enough capacity if feedwater were not available -- does the HPI system have enough flow capacity to remove decay heat? The Staff will discuss this in September or October.
Control Room Operational Aids:
1.
plant overview instrumentation 2.
systems and ccmponents data status monitoring 3.
service analysis systems (compares plant sensor signals with prestored patterns in order to detect anomalies in the plant)
The Subcormlittee considered touring the Singer-Link Co. located in Silver Spring, M3 which manufactures reactor control room simulators.
The Oak Ridge Surveillance system which attempts to recognize a disturbance analysis pattern from noise analysis was discussed. EPRI and J. C. Penny have perfomed studies on control system suitability for human operation.
The computer infomation system may require the use of a processor or sorter of information to be displayed to the operator.
Dr. Catton commented that the annunciator at TMI-2 had one button for both acknowledging and clearing events. The operators (at TML) did not push the button because they wanted to retain incoming infomation. The horn, because of this design, blared at them continuously.
CONTROL ROOM COMPUTERS - L. Beltracchi, NRC Staff (Transcript pages 246-282)
The ACRS Subcomittee asked the NRC Staff to discuss the Rancho Seco light bulb short-circuit incident. This item is to be reschsduled for a future meeting.
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i s Control room design has been evolutionary. The NRC does not have an overall philosophy or requirements with respect to control room design.
The post TMI-2 review has defined the man-machine Interface as a major operational problem in NUREG-0560 with the short-~ term reconnendations listed in N.UREG-0578.
Recommendations :
1.
direct indication of p0RV and safety valve position 2.
increased range of radiation monitoring 3.
monitoring inside the contaimnent 4.
the establishment of procedures and operator training for protection of inadequate core cooling with current instrumentation 5.
improved in-plant iodine instrwnentation 6.
installation of-primary system saturation meters LICENSEE EVENT REPORTS - T. Telford, NRC Staff (Transcript pages 282-310)
The Division of Operating Reactors has the responsibility for the evaluation of LERs. Because about 3,600 LERs are received annually and because its a part-time job (5 to 10% of the Staff's time), the review is not performed in a systematic manner.
The loss of pressurizer level indication at Oconee (March 6,1979) was discussed.
(A leaking valve packing in the primary system required that the plant be shut down. The reactor coolant makeup valve had failed shut and the-pressurizer level indicator had stuck. The plant was leaking and cooling down without makeup and pressurizer level indication.) Only one channel
-(of 3) of pressurizer level had been monitored during the evolution -- the failed channel.
Dr. Mattson stated that the lessons learned from this kind of thing are that the method of evaluation and determination for generic implications is inadequate; that simply assigning resources in dedicated fashion has not been done; and that the evaluation of LERs is not sufficient.
Fitzpatrick had been shut down with one diesel out for maintenance. Further maintenance caused a loss of offsite power leaving only one diesel available for power. The plant Technical Specifications were not violated. Additional review of maintenance procedures was recommended.
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. The failure of electric driven auxiliary feedwater pump bearings were discussed.as well as the resolution by Regulatory Staff of the possibility for a pair of reactors (Point Beach and Prairie Island), with swing diesels, having concurrent ECCS actuations.
The NRC Staff recommended that the number of LERs should be reduced and that the LERs should be more descriptive.
A transcript of this portion of the meeting has been sent to Dr. Moeller.
BACKFITTING - R. Cudlin, NRC Staff (Transcript pages 310-329)
The Regulatory Reauirements Review Comnittee is charged with reviewing backfit conversions. The authority and criteria are established in 10 CFR 50.9.
There are three categories of backfittir.g:
a.
front fit all new plants b.
case-by-case determination (14 instances) c.
backfit of all plants (9 instances)
The general topic of improving the process for detennining the need for new regulatory requirements is discussed in Commission Paper 79-8',
Comments from t
this paper include:
1.
lack of defined criteria for making changes in existing plants 2.
issues involved in improving the process are difficult 3.
the required level of safety is not well-defined 4.
criteria for determining when to backfit is unclear 5.
resource limitations in backfitting Dr. Okrent requested that all Subcommittee members draft comments for the full Comnittee concerning the important issues discussed at this Subcommittee meeting.
NOTES:
(1) For additional details, a complete transcript of the meeting is
. available in the NRC Public Document Room,1717 H St., NW, Washington, DC 20555, or from Ace-Federal Reporters, Inc'.,
444 North Capitol Street, NW, Washington, DC 20001.
(2) Materials provided to the Subcommittee at this meeting are on file in the ACRS Office.
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l Federal Registee / Vol. 44. No.141 / Friday l July 20, 1979 / Notices CS27 t,
I f
evaluate incidents involvirig loniting AI! other items regarding this meeting result in any sigmficant environmental radiation. The Agency continues to prepara remain the same as announced in the impact and that pursuant to 10 CFR for such response by providing the focowing.
Federal Register on July 12.1979 (44 FR Section 51.5(d;(4) an envirorunental (11 trained stali for advisement required to 40739).
impact statement, or negative 5
) trafnean eq ipped staif br Further information canhe obtained declaration and environmentalimpact l
emergency f'eid activiues:
by a prepaid telephone call to the appraisal need not be prepared h j
13] transportation by automobile to site of Designated Federal Employee for this connection with issuance of this inadent meeting. Mr. Richard K. Major, amendment.
(4) established liaison with appropriate (telephone 202/634-1414) between 215 For further details with respect to this NRC and DOE Operations Omces: and a.m. and 5.00 p.m, EDT.
action, see (1) the application for (5) training to iey personnel of other state /
amendment d6ted October 20.1978, as ICC'I'8'"C "-
DM bly 18.m supplemented January 24,1979. (2) g Radiolospcal assistance in the fo.rm of John C. Hoyle.
Amendment No.18 to Ucense No. DPR-monitanng. !aaison with appropriate Advisory Commitfee. Monogerwent Ofiern
- 7. and (3) the Commission's related
- M Safety Evaluation. All of these items are I
s ur y d cles paep id y
f Agency. The contamination guades used by
- E N4,
available for public inspection at the e
the Agency are in Table UI of the Protectave Commission's Public Document Room, Accon Gmdea contained in Appendix XL A!!
1717 H Street. NW., Washington. D.C.
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Agency personnel will be maintained at an Met No 50-tW and at the Humboldt County Ubrary.
operation. ready level of training Part of this traimns wM1 be provided through cooperation Pa fic Cas & Electrl@ hm of 636 F Street. Eureka. California. A copy y
of the NRCin Las Vegss. Nevada.
Amendment to Facility Oper:11rgi ofitems (2) and (3) may be obtained ne Annex C Nuclear Accident orIncident Ucense upon request addressed to the U.S.
i "" e he Nuclear Regula tory Commisalon.
De U.S. Nuclear Regulatory Wa shington, D.C. 20555. Attentica:
L ae se esA
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(DCPAlis included in Appendix XI.ha plan Commission (the Commission) has Director. Dtvision of Operating Reactors.
D addresses both transportation accidents and issued Amendment No.18 to Facility Dated at Be heeds, staryland. this 13th day C
off-soe releases from fiaed facilities. it Operating bcense No. OPR-7 rssued to j
reqmres that the State Police first notify Pacific Gas and Electric Company (the of b. ly trre.
y or the Nudur Reg.h@mh DCPA which in tum notifies the Ra6ation Incensee), which revised Techmcal
[
Control Agency. It is the responsibility of the Specifications for operation of the U *** AMI"'
Agency to adme the DCPA the extent of th*
Humboldt Bay Power Plant. Unit No. 3 Chief. Operetics Reactorr Smach No. :L hatard to the public health and safety and (the facility) located near Euraka.
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recommend protective actions as necessary, California. The amendment is effecuve F" D*' " N " ** *"8 All heensees wlll be given copies of the plan ggg an.uiso coca riso.eia and instructed ut proper reporting of 2
incidents which occur outside of their facCty The amendment revises the Tcc.r.sical Specificatiens to extend on a one-ti=e
[Doctet No. 50-2751 E
irn o m- % s.rs m e o..i basis the interval for containment p
" C# *'" " "
J integrated leak rate testing until Pacific Cas & Electric Co. (Diablo i
immediately prior to returning the Canyon Nuclear Power P! ant. Unit 1);
[]
Macility to power operation.ne Order Extending Construction Advisory Committee on Reactor 4
amendment also includes the following Completion Date
[
Safeguards; Ad Hoc Subcommittee on adminstrative changes to the Technical Pacific Cas and Electric Company is the Three Mile Island, Unit 2, Accident Specifications-the holder af Construction Permit Npa.
l Implications Re Nuclear Power Plant
- 1. Expression of an a!!cwable CPPR-39 and CPPR-49 issued by the Design; Addition to Agenda tolerance forperforming surveillance Atomic Energy Commissioo* on April The agenda of the July 25-:7.1979 intervals.
23,1968 and December 9.1974
.o meeting of the ACRS Ad Hoc
- 2. Addition of a membe'r tolhe Plant respectively, for construction of the W~--
^
Subcommittee on the Three Mile Island.
Staff Review Committee.
Diablo Canyon Nuclear Power P!act. -
i l
Unit : Accident-implications Re
- 3. Correction of a typographical error, Units 1 and 2. presently under Nuclear Puwer Plant Design, willinclude and construction at the Company's slteln
~. Clarification of GeneralOfSce San Luis Obispo Coimty. California.
the following.
Ndefear Power Plant Review and Audit On May 24.1979. Pacific Cas and 4
W I"E ##* #
Committee responsibilities.
Electric Company filed a request for The subcommattee will address the topic of De application for the amendmnt extension of the completion date for fcther ACRS review of pending apphcanons complies with the standards and Unit 1.
t I tala d.1.ru 2 ccid a requirements of the Atomic Energy Act On November 15,1978, the specifically, the Salem Nuclear Power of 1954. as amended (the Act), and the Commission s staff published Stanon. Urdt 2 review.
Commission's rules and regulations. De Supplement No. 6 to the Safety Background information regarding the Commission has made appropriate Evaluation Report for the Diablo Salem Nuclear Power Station. (Jnit 2 canbe findings as required by the Act and the Canyon Nuclear Power Plant found in documents on f'te and avaa!able for Commission's rules and regulations in 10 documenting the need to make certain,.
pubhc inspection at the NRC Public CFR Chapter I. which are set forthin the modifica tions. Conse~quent!y. additional.
Document Room. trir H St. N.W.
license amendment. Prior public notice time willbe requited to complete the i
Washington. DC :o555. and at the Salem Free
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bhc Lbrary. It2 % est Broadway. Salem, of this amendment was not required since the amendment does not involve a MecM lanuary a wt the Atende racur Co wamn becam ewa:Resutaaary.
signiCeant hazards consideration.
Ca umasics and Pernuta in thei on that day wre De Commission has determined that oo.iinued unda Be automy of 6e Nuclear the issuance of this amendment will not Ra stakry Co-a g-s
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Federal Register / Vol. 44. No.135 /hursday. July 12.197M Notices 40739 Endowment for the Arts. Washington, consultants. and Staff. Persons desiring Date July to.1s79.
D C. 20506. or cali (202) 634-6070 to make oral statements should notify John C. Hoyle.
lohn H. Clark.
the Designated Federal Employee as far Admory commmer Maragement O?iar.
Deector. 01lcr of CouncilandPwel in advance as practicable so that
- ni ows.n an vr.a r.nn naa nes Opent;ons. Nationa/ Endowmentfor the Arts.
appropnate arrangements can be made suo coot rsms p o. wi.es rua r.it., s es a.t to allow the necessary time during the eco com rur+.as meetmg for such statements.
The ager.da for subject meeting shall
[ Docket Nos. 50-522 and W523]
be as follows:
Federal Graphics Improvement Puget Sound Power & Light Co., et al Advisory Panel; Meetsng Thursday. July 26 cnd Friday. fuly 27.
(Skagit Nuclear Power Progect Units 1 Pursuant to section 10(a)(2) of the 1979. 8:30 a.m. untsi the conclusion of and 2); Order for Evicentiary Hearing Federal Advisory Committee Act (Public hsiness coch day.
and Related Matters Law 92-4i53). notice is hereby given that The Subcommittee may meet in
- 1. Pursuant to agmement among 6e a meeting of the Federal Graphics Executive Session, with any of its panin at a confennce m, Seattle.
Improvement Advisory Panel to the consultants who may be present, to Washmgton. on Apn! 24,1979. the i
National Endowment for the Arts will be explore and exchange their preliminary evidentiary hearing is scheduled to held on July 19.1979. from 9-30 a.m. to opinions regarding matters which should resume at 130 a.m Tuesday. July 17 4.30 p m.. in room 1125. Columbia Plaza Office Building. 2401 E Street. NW.,
be considered dunng the meeting and to heanng will be at Room 3038. New Washington. D C.
f mia e a mp n and me mmendation i
Federal Building. 915 Second Aven;e.
to the full Committee.
This meeting will be open to the Seattle. Washington.
pubhc on a space available basis.
AI the conclusion of the Executive
- 2. At the beginning of the heanng.
Accommodations are limited. Interested Sess.on, the Subcommittee will discuss limited appearances pursuant to 10 CFR persons may submit wntten statements with representatives of the NRC Staff.
2.715 will be permitted. Paragraph (a) to the panel the nuclear industry. various utilities, thereofis pertment.
The agenda for this meeting will and their consultants. state and local (al A person who is not a party may. in the include a discussion of the Graphic offic2als. and other mterested persons.
discretion of the pr.sidmg omccr. de matenal of the Consumer Product Safety the implicarica.s of the Three Mile permitted to make a hmned appearance by Comrmssion.
Island. Umt 2 Accident.
maiung oral or wntten statement of his j
Further mformation with re.~emnce to In addition. it may be necessary for pas um on the issues at any sesm of the this rneetmg can be ootained f:om Mr.
the Subcomrruttee to hold one or more
("j"['d o f'Q"
'P
'"8 Lance lay Brown. Coordmator of closed sessions for the purpose of by the presidmg officer. but he may'not ch con o a a e d
Federal Craphics. ynonal Endowment exploring matters involving proprietary otherwise parue:pate in the proceedea.
for the Arts. Washmgton, D.C. 20506, or information.1 have determined. in call (202) 6344286.
3.
n ns wWng to make a limited accordance with Subsection 10(d) of Dated. July e. It's.
Public Law 92-863, that should such 8ppearance at the heanng scheduled to John H. Clark.
sessions be required. it is necessary to gv advan not f ation to trie Detector OMa o/ Cow cif endPane/
close these sessions to protect Secretary. United States Nuclear Operor;ons NationalEsdonavent forthe.4rtr.
propnetary mformation (5 U.S.C.
Regula tory Commission. Washmston.
Irn on wisu rw v.iin us ami 5523(c)(4)).
D.C. 20555. Those planning to ma'ke a sea coes rstres Further mformation regarding topics hmited appearance at the heanng to be discussed, whether the meetmg begmnmg on luiy 17 are directed to has been cancelled or rescheduled. the check with the Brard's clerk at 9:00 a.m.
NUCLEAR REGULATORY Chairman's rulir:g on requests for the that day at the F eanng room.
COMMISS3ON opportunity to present oral statements Done this 5th day of July 194 at Advisory Committee on Reactor and the time allotted therefor can be Washmston. D C.
Safeguards, Ad Hoc Sut> committee on btamed by a prepaid telephone call to Atomic Safe'y and IJcensirig Board.
trie Three Mile Island. Unit 2 Accident, the Designated Federal Ernployee for y,;,ntic, s. o,,1.,
l Implications Re Nuclear Power Plant this meeting. Mr. Richard K. Major.
~
CAcirman.
Desagn; Meeting (telephone 202/634-1414) between 8:15 d
m., g,,n%3 The ACRS Ad Hoc Subcommittee on
- CC 7 5 '**'
the Three Mile Island. Unit 2 Accident-Background information concerning 1
implications Re Nuclear Power Plant this nuclear station can be found in
]
Design. will hold a meeting on July 28-documents on file and available for 27.1979 in Room 1187.1717 H St NW.
public inspection at the NRC Public NATIONAL TRANSPORTATION SAFETY BOARD Washington. CC 20555.
Documents Room.1717 H Street. N.W.,
in accordance with the procedures Washington DC 20555 and at the
[N.AR 79-28) outhned m the Federal Register on Government Publications Section. State l
October 4.1978 (43 l'rt 45926), oral or Library of Pennsylvania. Education Annual Report to Congress, Accident
'wntten statements may be presented by Building. Commonwealth and Walnut Reports, Safety Recommendations, members of the pubhc. recordings will Street. Harnsburg. PA 17126.
and Responses; Availacility be permitted only durm.g those portions AnnualReport to Congress.-The of the meeting when a transcnpt is being National Transportation Safety Board kept. and questions may be asked only i
by members of the Subcommittee.its released its report on June 29 and cep:es are now available to the pubhc. Sing!e
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ATTENDANCE LIST ACRS SUBCDMMITTEE ON TMI-2 ACCIDENT IMPLICATIONS JULY 26 & 27, 1979 WASHINGTON, DC ACRS NRC STAFF PUBLIC SERVICE ELECTRIC & GAS
- 0. Okrent, Chairman R. Tedesco E. Liden C. Siess R. Mattson R. Mitti J. C. Mark R. Frahm R. Burricelli H. Etherington J. Milhoan H. Heller M. Plesset J. Olshinski J. Zupko, Jr.
I. Catton, ACRS Consul tant R. Cudlin C. Veprek W. Lipinski, ACRS Consultant T. Tel ford T. Taylor C. Michelson, ACRS Consultant C. Long L. Reiter R. Major, Designated Federal Dnployee P. Mathews L. Lsitz E. Igne, ACRS Staff L. Beltracchi J. Wroblewski J. Gormly R. Patterson EPRI PASNY T. Martin F. Schneider R. Leyse G. Rangarao WESTINGHOUSE ELECTRIC CORP.
MCCRAW-HILL GENERAL ELECTRIC J. Scuss B. Adamson N. Shirley W. Gangloff J. Frent R. Burch J. Hard MORGAN, LEWIS & BOSKINS BABCOCK & WILCOX D. Maire K. Jordan E. Epner R. Borsum~
M. Hartley SOUTHERN COMPANY SERVICES INC.
T. Knight NORTHEAST UTILITIES D. Lambert D. Jaffe J. Raulston E. Foster R. Hernandez D. Wilson TEPC0 NUCLEAR PROJECTS INC.
H. Harmader BECHTEL POWER CORP.
J. Kelly, Jr.
W. House, II KEPC0 PCOTMI K. Ota VIRGINIA ELECTRIC & POWER CO.
L. Jaffe J. East
2-0FFSHORE POWER SYSTEMS LEHIGH UNIVERSITY OUKE POWER COMPANY D. Aabye J. Chen T. H11tman R. Sundaram PUBLIC T. Manjikian, States News Service J. Hebert, Associated Press H. Yocom, Gilbert Associates Richardson, Ace-Federal Reporters, Inc.
D. Hoffman, Ace-Federal Reporters, Inc.
J. Burns, Ace-Federal Reporters, Inc.
A. Riley, Ace-Federal Reporters, Inc.
K. Loryn, BRD m.
REVISION 1 7/19/79 TENTATIVE SCHEDULE ACRS AD ' HOC SUSCOMMITTEE MEETING ON THE TMI-2 ACCIDENT IMPLICATIONS 1717 H STREET, NW, WASHINGTON, DC, ROOM 1046 JJLY 26 & 27,1979 ITEMS FOR DISCUSSI0" ON JULY ES, 1979 - (8:30 am - 7:00 pm) 1.
What should be the design basis for hydrogen build-up for LWRs? Why?
2.
Examine events having the potential for overfilling the secondary side of steam generators (e.g., the March 20, 1978 Rancno Seco event).
3.
Examine events involving debris arising from structural failure of pieces within process lines.
4 How to resolve the question of the adequacy of the single failure criterion.
5.
How to examine the influence of control systems on safety.
6.
How to evaluate the merit of possible design changes for future LWRs:
a.
More RHR loops?'
b.
Full pressure RHR?
c.
Containment modi fications?
d.
Others
.7. How to review SWRs for TMI Implications.
8.
Generic Inplications of Recent LERs:
Loss of pressurizer level indication at Oconee - 3/6/79 a.
b.
Maintenance at Fitzpatrick c.
Others 9.
Auxiliary feedwater system Requirements:
What should be the quantitative reliability requirements? Why?
a.
b.
Others
- 10. NRR's use of probabilistic techniques in the licensing process.
- 11. Control Room Computer Systems for Data Processing. What is the state-of-affairs in other plants? What information is being given to operators?
Inexpensive off-the-shelf type equipcent is now an order of magnitude improvement over the T'il-2 systems. What is being asked of utilities in this area?
d
2
- 12. How are decisions nade regarding backfitting requiremants?
- 13. Discussion of diffuser screens in the surge line entrance to the pressurizer.
- postulating a break at the top of the p'ressurizer, will the pressurizer water dump?
14 Which operating procedures will be handled in the short-term? How will the adequacy of other operating procedures be detennined? Ex amples :
(1) Operating procedures for the SSE.
Rc,j, }
(2) Operating procedures for a steam line break coupled with a small primary side break.
- 15. Others 9
O 9
e*
P l
~
. ITEMS FOR DISCUSSION ON JULY 27, 1979 - (3:30 am - close of business) 1.
Discussion with the flRC Staff on how to apply the results of the Lessons Learned Task Force and the Bulletins and Orders Task Force to tne near-ter Westingnouse OL stage plants:
a.
design changes b.
operator actions and operating procedures The discussion should focus on the Salem 2 plant as the Westinghouse near-tenn prototype. The Subcommittee wishes to explore the procedures and requirements which will be placed on the near-term OLs by the Staff (e.g., how much of the review will be generic and how much plant specific?)
2.
Discussion with pSE&G on how they intend to implement the Staff's require-ments on the Salem 2 Station. The discussion is not intended to serve as a review of the Salem 2 Station, but rather to address Salem 2 as t pro-totype near-term 'dplant.
3.
The other near-term Westinghouse OL stage plants are invited to attend the meeting. Formal presentations are not required, however, they may be able to respond to Subcommittee questions arising out of design dif-ferences between the plants.
plants included are: 31cblo Canyon, !: orth Anna 2. Sequoyah, Farley 2, and McGuire.
So
.