ML19338E237
"Draft Withdrawal" is not in the list (Request, Draft Request, Supplement, Acceptance Review, Meeting, Withholding Request, Withholding Request Acceptance, RAI, Draft RAI, Draft Response to RAI, ...) of allowed values for the "Project stage" property.
| ML19338E237 | |
| Person / Time | |
|---|---|
| Issue date: | 09/04/1980 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML19338E238 | List: |
| References | |
| FOIA-80-485, REF-10CFR9.7 SECY-80-364, NUDOCS 8009250276 | |
| Download: ML19338E237 (20) | |
Text
i I
UNITED STATES OF AMERICA r$?
=
2 NUCLEAR REGULATORY COMMISSION 3
.:e
- 35 4
PUBLIC MEETING e
5 3n j
6 AFFIRMATION SESSION 80-39 R
7
=]
8 d
ci 9
- i h
10 Nuclear Regulatory Commission i
Room 1130 II 1717 H Street, N.W.
8 Washington, D.C.
12 E
i c
Thursday, September 4, 1980 13 g.g.
3
~
The Commission met, pursuant to notice, at 4 :05 p.m.
g 14 BEFORE:
1 2
15 JOHN F. AHEARNE, Chairman y
16 d
VICTOR GILINSKY, Commissioner 6
17 JOSEPH HENDRIE, Commissioner 18 E
PETER A. BRADFORD, Commissioner 19 5
NRC STAFF PRESENT:
20 LEONARD. BICKWIT, General Counsel 21 M. MALSCH b_
22
~~
J.
HOYLE 23,
24 e.p 25i
%Q9% sob %
ALDERSON REPORTING COMPANY,INC.
O y rD' AEf
}bd lTj Ifi Y Y [' ~
~
JkF 2
=
- ==
=
DISCLL.MIR
.. ~ _ ~
"E=
This is a= u=cificial ::a= scrip: of a. =ee:1:8 of the Uni:ed Sca:es Nuclear Regulat=ry Co-dssic= held on Sect 0 4, 1980 i= the Coi::c:issio='s offices a: 1717 3 5::eet, N. W., Washi=gton, D. C.
The meating.as ope = :o public at:anda=ce and abservation.
This tra=se:1p has =o: been reviewed, cc ec:ed, or edi:ed, and i: =ay co= ai= i= accuracies.
The transcript is i=re=ded solely for gn=eral d #c =a donal purposes.
A.s p cvided by 10 Cn 9.103,1: is not par. of the for..a1 or d #c:=al record of decision of.he =ar:ers discussed.
Expressio=s of opd-d en 1: - #s ::a= script da co: secessar 17 reflec: fi=al data. h.tions or b=1 d eis.
No pleading or other paper =ay be filed " h the C.
'ssion i= a=7 p cceedi=g as the resul: of or addressed to a=y s a:e=e== or argu=e== ce=:ai=ed here1=, excep: as the Co-dssio= =ay autherfen.
==
E
. :5
=
.y*J 4
2 y
_P _R _O _C _E _E _D _I _N _G _S
=
'-~
1 CHAIRMAN AHEARNE:
We are on the record.
3 MR. HOYLE:
The first affirmation item, Mr. Chairman,
==
==
EF 4
is SECY-80-364, the subject is Fees for Withdrawn License Applica-5 tions for power reactor construction permits, operating licenses a
3 6
and other reviews.
=
R R
7 The staff has recommended a rule change to authorize N
8 8
the General Counsel to take action if and when necessary to a
dd 9
g collect fees for licensing reviews, for denied, withdrawn and oH 10 y
suspended or postponed applications.
You've all approved the E
11 g
text of the amendment to CFR Part 170, and you've all now agreed d
12 that it should be made effective upon publication, rather than z
E=.
- d 13 yr:
g a proposed rule.
May I have your affirmative votes?
E "8
g (There was a chorus of Ayes.)
e C
15 h
MR. HOYLE:
I will speak to the other two for a moment.
m
?
16 g
SECY-80-373, the General Counsel has provided a memorandum which y[
17 asks that that be deferred for a few days.
You've all agreed to x
18 that.
And 80-365 the Chairman has circulated a memorandum on P"
19 8
which you agree.
20 COMMISSIONER HENDRIE:
Good.
21 CHAIRMAN AHEARNE:
Very well, thank you.
==
22
- QF (Whereupon,- at 4
- 10 p.m.,
the Affirmation Session 80-39 23 lhwas adjourned.)
l db 24
}
l
E.
25 l ALDERSON REPORTING COMPANY, INC.
5 NUCLEAR RIGULATORY CO.WCSSION This is Oc ce.-:ify that the attachec pecceecing: bercre -he
.~.
, c5 in the =atte.- ef: AFFIRMATION SESSION 80-39 Date cf ?rcesecing:
September 4, 1980 Decket Nu=ber:
? lace cf ?receecing:
Washington, D. C.
..e... _, a
....<...e.s.eg,
.u - -
..._a, 3
.a.-a.._-
a
-- =-
... a.
e
.....s w.
thereof for :he file of the Cc==issica.
Suzanne Babineau Official.ieperte.- (Tfpec) s
/
/):
/
ner,$=EE.~. e h C.# ~ee.
, D - h $ $ ~ *w* r e,'
we
'g e
e"~;}
k o-W y
a w
w w
U
[7590-01]
(r,
- p. 3a)
NUCLEAR REGULATORY COMMISSION
[10 CFR Part 50]
DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Consideration of Degraded or Melted Cores in Safety Regulation AGENCY:
U.S. Nuclear Regulatory Commission ACTION:
Advance notice of proposed rulemaking
SUMMARY
The U.S. Nuclear Regulatory Commission is considering amending its regulations to determine to what extent [ if rd commercial nuclear power plants should be designed fer : cope wi%
- reactor accidents to broaWe-o Ehid kvehe-demege-to-fue7-.and-r-cicasc cf-rsditad vity,
cluding-& sign f w acto w cider.t3 beyond those considered in the current " design basis accident" approach.
In particular, this rulemaking would consider the need for nuclear power plant designs to be evaluated over a range of degraded core cooling events with resulting core damage and the need for design
' improvements to cope with such events.
This advance notice of proposed rulemaking is being issued to invite advice and recommendations on several questions help--the-NRC depe W pelic4e]s concerning design and operational improvements for dealing with degraded core cooling.
Therefore, the preliminary views expressed in this notice may change in light of comments received.
In any case, there will be an opportunity later for additional public comment in con-nection with any proposed rule that may be developed by the Commission.
re,.i ne... 1
n-t l,
- [7590-01]
DATES:
The comment period expires [90 days after notice in the Federal Register].
ADDRESSES:
Interested persons are invited to submit written comments and suggestions to the Secretary of the Commission, U.S. Nuclear Regu-latory Commission, Washington, D.C. 20555, Attention:
Docketing and Service Branch.
Copies of comments received by the Commission may be examined in the Commission's Public Document Room at 1717 H Street, N.W.,
Washington, D.C.
Comments may also be delivered to Room 1121, 1717 H Street, N.W., Washington, D.C., between 8:15 a.m. and 5:00 p.m.
~
FOR FURTHER INFORMATION CONTACT:
M. S. Medeiros, Jr., Office of Standards Development U.S. Nuclear Regulatory Commission Washington, D.C.
20555 or telephone (301) 443-5913.
~
SUPPLEMENTARY INFORMATION:
HISTORICAL BACKGROUND The Nuclear Regulatory Commission (NRC) is responsible for licensing and regulating nuclear power plants.
Before a nuclear power plant can be built at a.particular site, a construction permit must be obtained from the NRC.
As a major part of the application for a construction permit,
'th.1 applicant files a Safe,ty Analysis Report.
This report presents the design criteria and preliminary design information for the proposed nuclear power plant and provides information on the proposed site.
The report also discusses various abnormal conditions and accident situations and describes safety features to be provided to prevent accidents or, if they l_
should occur, to mitigate their effects on public health and safety.
l l
I l
l i
u Enclosure 1 2
_ =. - -
f
[7590-01]
In nuclear power J ants,~1arge' amounts of radioactive material are ;
il generated during fission of nuclear reactor fuel.
Although this radio-active material generally remains in the fuel pellets, significant amounts 15 t4,s. m.ctor c. cola.h
~
can be released during accident conditions.
For appreciable amounts of g
radioactive material to'be released from the fuel, it must experience damage from one or more of several possible causes.
For example, a hydraulic-mechanical accident at. normal fuel temperatures can burst fuel cladding resulting in release of radioactive material normally retained in thegapbetweenthefuelpelletsandti)efuelclad.
A more serious type of accident-involving higher fuel temperatures might, in addition to rupturing fuel cladding, cause oxidation of the cladding.
This, in turn, would
.n cause hydroge~n to be generated and 'releas.ed which would compound the severity of the accide t.'
A still more serious accident might involve-very high fuel' temperatures and oxidation of a large fraction of the core's zirconium.
In this case, not onl e, A< conhent-6u.%o.y would large amounts of hydrogen be release {but other thermal reactions could result in the release of radio-active material normally held captive in the fuel pellets.
Finally, an accident so severe that core melting occurs could release large amounts of' radioactive material to the environment if reactor containment integrity a.h a to be -
l wereg ost.
Based on these considerations, a broad range of nuclear power plant abnormal ccnditions and accidents with the potential to cause fuel clad damage and release of radioactive material to the environment has been identified and categorized for analysis.
Attempting to prevent abnormal conditions and accidents and mitigating their potential conse-quences have been the primary objectives of nuclear power plant safety design.
The Safety Analysis Report is a key analysis document supporting the adequacy of this aspect of nuclear power plant design.
3
=
[7590-01]
As discussed in Title 10, Chapter 1, Code of Federal Regulations, section 50.34 (a), in the Safety Analysis Report the applicant is required to determine margins of safety for both normal and abnormal operations and to determine "the adequacy of structures, systems, and components provided for prevention of accidents and the mitigation of the consequences of accidents." To assist the applicant in complying with this regulation, the NRC has published Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Powar Plants, which describes 1
the information to be provided in the Safety Analysis Report.
In partic-ular, section 15 of Regulatory Guide 1.70, provid#s guidance to an appli-cant concerning " design basis assumptions acceptable to the NRC for purposes of determining adequacy of the plant design to meet 10 CFR Part 100 criteria."
Regulatory Guide 1.70 explains that these design -
~
basis assumptions can, for the most part, be found in regulatory guides that deal with radiological releases and suggests use of Regulatory Guides 1.3 and 1.4, Assumptions Used for Evaluation of the Potential Radiological Consequences of a Loss-of-Coolant Accident.
Regulatory 1
Guide 1.70 further states that "This analysis should be referred to as the ' design basis analysis'." Operating events corresponding to design basis assumptions are termed " design basis accidents'j and satisfactory concerdd* them analysis conclusions allow a judgement that the facility can be operated 4
without undue risk to the health and safety of the public.
It should be nc+=d that these events are analyzed primarily for the purpose of estab-l lish? 7 the adequacy of engineered safety features, such features being those structures, systems, and components, designed into a plant to l
2Available from the U.S. Nuclear Regulatory Commission, Washington, DC 20555 4
=.: - -
3 -.. --
[7590-01]
mitigate the consequence's of ' postulated design basis ' accidents, and which supplement plant features designed to meet performance specifications for normal operations and anticipated ' abnormal conditions.
In the Safety Analysis Report the app % desiGm %M5t3 e,cegden4E.licant is n ssedents more seme. 6sn to. explicitly analyzegth: =:t :tri::: type ef pa::ib h-a;+ Ments, : :n, m- -
p th: ;dhereareotherdesignrequirementswhichwould. pres.upposeevent.s where significant core damage and release of radioactivity,has oc:urred.
For example, radioactive source terms of Technical Information Document.
TID-14844, Calculation of Distance Factors for Power,and Test; Reactor, Sites,2 which imply a major reactor accident, are used_to judge d.e. sign adequacy of various engineered. safety features and certain.other plant
\\
~
\\
systems and components, uikte ic:.lly, the beliefi as b;;r. that the Emst-4eMev:- p ::%,Jn A assmeti.n%E M Tds w reac ibl:4 accidents are of sufficiently low probaMi;ty that.auaka-
.F rkise censequences is nei necertary con pWe'. Sa* Wy.
W. :nd mitigation ir et axplicitly requir+d.
This low probability g
wo A acM-t-Ge.
resulty from y " defense in depth" approach that requires conservative g
design, multiple physical barriers, quality assurance for de.ign, manufac-ture and operation, and. continued surve.illance and testing.to prevent such accidents.
Furthermore, in reviewing reactor plant designs using the " design basis accident" approach, the NRC does not review all structures, systems, va:tko-and components but reviews, in varying levels of detail, only those 4
considered " safety grade".by the applicant submitting a Safety Analysis Report.
Items considered by the applicant to be outside the scope of design basis accident analyses are generally not considered to'be " safety 3Available from National Technical-Information Service, U.S. Department of Commerce, Springfield, Virginia 22151.
I'
.L s '. --
m o,
a
~
~
[7590-01]
grade" and are not reviewed by the NRC to see whether they will perform
~
as intended or meet various dependability criteria.
This method of classi-fication is based on the notion that things credited in the analysis of
- lesign basis event or specified in the regulations are important to -
safety and thus are " safety grade" while all else is "non-safety grade."
Non-safety grade items do not receive continuing regulatory supervision or surveillance to see that they are properly maintained or that their design is not changed in some way that might interact negatively with other systems.
Instead, these items simply receive what attention may' be dictated by routine industrial codes and by desires to enhance plant availability.
I a.rsamph h Historically, a further perumptie in design review and licensing was
\\
ta:- bee.mtsssume that if reactor plant systems can handle large-scale design basis accidents, ie-.most-cases. they can also handle a spectrum of smaller accidents that are regarded as being "within the design envelope."
resdEit The accident at Three Mile Islandfkh invoked-e-sequenc+of-event-s-cariag core damace more severe than that considered in current design basis events {ut ^^+ ""sg-e-release of fission products from Tim u.u s A/o g
the core more severe than that presumed in 10 CFR Part 100 or TID-14844 %
- hasshowntheneedtore-examinethesehistoricalapproachestoanalyzing reactor plant design and plant accidents. hep ',
- he October 1979 Report of the President's Commission on the Accident at Three Mile Island 8 i
recommended that in-depth studies be initiated on the probabilities and consequences (onsite and offsite) of nuclear power plant accidents, 4"The Need for Change:
The Legacy of TMI," available from the U.S.
Government Printing Office, Washington, D.C. 20402.
~
/ k "#-
6 Enc losure 1- - - - -
[7590-01) c o e es including the consequences'of meltdown. -This report recommended that these studies include a variety of small-break, loss-of-coolant accidents and multiple-failure accidents, with particular attention to human fail-ures.
The report stated that "from these studies may emerge desirable modifications.in the detign of plants that will help prevent accidents and mitigate their consequences.
For example, consideration should be given to equipment that.would facilitate the controlled safe venting of hydrogen gas from the reactor cooling system," and " consideration should be given to overall gas-tight enclosure of the let-down/make up system with the option of returning gasss to the containment building.
Similarly, the January 1980 report, Three Mile Island, A Report to the Commissioners and to the Public, states, "...we have come far beyond 4
~
the point at which the existing, stylized design basis accident review approach is sufficient.
The process is not good enough to pinpoint many important design weaknesses or to address all the relevant design issues.
Some important accidents are outside or are not adequately assessed within the ' design envelope'; key systems are not ' safety related'; and integra-tion of human factors into the design review is grossly inadequate."
COMMISSION'S INTENTIONS Accordingly, it is the Commission's intent to determine 4e. what c.ka.wqcs, W
are nee (<4 to pt^^t, if any,greactor plant designs and safety analysesghcuM take into account reactor accidents beyond those considered in the current design basis leerden ts uder e owsWe sA&<
accidentapproachgi<:ludp.g.arangeoflossof-core-cooling,c'eredamage,
' Copies may be obtained from the GPO Sales Program, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555.
.:..==
[7590-01]
and core melting events both inside and outside historical design envelopes.
Furthermore, the Commission will consider whether to require more coherent consideration of this range of core damage events in the design of both' normal operating systems and engineered safety features.
Therefore, this advance notice of proposed rulemaking is being published to provide the,
erptd< Ms<. vi<f neo-oAtw % %e c-ecsa+i on regulated industry and the public an opportunity togii:e er the ::a--
At sWd be. the.
ge.atent of a regulation requiring improvements to cope with degraded core ooling and te cepe with accidents not covered adequately by traditional design envelopes.
The rulemak.ing proceeding will address the objectives of such'a regulation, the design and operational improvements being con-sidered, and the costs of such design improvements compared to expected benefits.
Recognizing the need for prompt action to correct specific deficiencies -
identified during the Three Mile Island accident and subsequent investiga-tions, the Commission is publishing, h-parclld with-tM4 Iedarm1 Dagister-Netsca, a proposed rule that would require certain interim improveinents TLd-later;m to better cope with degraded reactor cores. 3% e proposed rule should,
(
not be viewed as prejudging the final action concerning this advance notice of proposed rulemaking, and comments should be framed accordingly.
In addition to this FEDERAL REGISTER N,otice, the Commission's Office of Standards Development is making a direct mailing to affected licensees and other known interested persons to ensure that they are aware of this advance notice of proposed rulemaking.
~
[7590-01]
SUMMARY
OF FEATURES BEING CONSIDERED FOR PROPOSED RULE The Commission is considering initiating rulemaking'that w
-have-to iviivwing icote ;.;.
(
equire that a broad,hange of accidents of both lesser and greater severity than the design basis accidents e considered in plant design, plant operation, and reactor D /' D '-
a alyses.*decigr and :n 3y4i.s right be required -for a r=ge=cf less-of-cccc coeWg ~~+-
nd--
ren tant :;rc d=:ge, including a fully melted core), se t%t cartqia
~
ipredidcR cewsegcacc:- &ght-be prevented-er-substenticil; -iti -
pt-i
[
Require mer-c-cchcrent-cons 4deration vf cv e -damage-and r:1 case-.of
-r-adioac-t4ve-mater 4a1 #n design _ nf plet etwc4eres ;-sys-tW5r-end
.capenent+ and e1 % ate-uneven-treat.nent vi acridentm:-htstes-by-di.f.ferent parts nf +homA44ons Commcd ir-a l: o I-ife/ om
( tk uhf-hkiet, q
~
' 4 h e-( m%c A(4 IFIC CONSIDERATIONS u Qe f,-Wd.J Advice and recommendations on a proposed rule reflectin the foregoing featuresandonanyothelrpointsconsideredpertinentgarein 5
all interested persons.
Comments and supporting reasons are particularly requested on the following questions:
1.
If loss of core cooling and resultant core damage occur in a nuclear i
power plant, there are certain predictable consequences.
Can these consequences be mitigated substantially, and the risk of severe public health danger thereby reduced substantially, by practical design improvements?
If not, why not, or, if so, what design improvements can be made and at what estimated cost? How would your recommendations impact on other safety considerations?
f
{
- -. ~ _..
s
~.
[7590-01]
2.
The Three Mile Island accident was terminated after the core was damaged severely but before substantial melting occurred, a condition beyond the current design-basis accident events considered in the safety analysis.
Should the.NRC require that events of this type be considered in future safety analyses?
If not, why not, or, if so, what criteria would you impose to judge design acceptability?
3.
Although the consequences of core-melt accidents have been considered to some extent in assessing nuclear power plant safety, such as in requirements for siting, emergency response plans, and certain engineered safety features, explicit consideration of the capability of current designs and casualty procedures to cope with core-melt accidents has not been a part of safety analysis scrutiny by the NRC.
Should core-melt accidents be specificCly evaluated in safety analysis reviews, and, if so, to what extent, or, if not, why noti 4.
Recognizing that there can never be complete assurance that only analyzed events as delineated in a Safety Analysis Report will occur, a
what additional analys)(s, procedures, or design features would you propose to mitigate fuel damage accidents in the range from exten-sive clad perforation without oxidation, through a few percent clad oxidation, through extensive oxidation to full core meltdown? Would you recommend different and perhaps overlapping design features depending on t'he severity of core damage to be coped with?
5.
To-what extent should reactor design and reactor safety analysis account for engineered safety features not working at all, not working well, or being defeated by the operator resulting in severe core damage? What limits should be placed on multiple failure and operator 10
[7590-01) error assumptions made in safety analyses and how probabilistic risk assessment be used to determine suci: limits?
Mu.M MRC. necdr4-cc4hwe. fib 3 6.
Arc j= # - f=cr cf a n:w requirement t: conct =:t, at each nuclear oF reactor plant site, a new structure for controlled filtered venting of the reactor containment structure? Would you limit the function of such a new structure to filtering particulates, elemental iodine, and inorganic iodine or would you atead swh ca-:ppendage +e include adsorption bed systems using charccal or other processes so that she organic iodine and noble gases could,be trappef? What quantities and r~elease rates of gases and particulates would you design such a structure to handle and at what removal efficiency and cost? Do
& teractwrc.
the potential reductions in risk expected from such cr appendags-offset potential increases in risk that may materialize from incidents such as inadvertent operation or the concentration of hydrogen in the filtering apparatus?
Shou {d pitc. Yea &n lucerfcecNA 7.
Are-you-in-feve r-of--requkement: t: in :rpor+te-into containment design, systems for controlling combustion of hydrogen? Do you favor methods of control that suppress combustion or do you favor controlled burning?
If you favor suppression of combustion, what techniques would you recommend and should they vary as a function of the design capability of current containments?
If you favor controlled burning, do you recommend open flames, spark plugs, catalytic combustors, or odde h%,k A cert, some other means? What percent of a :n c's zirconium being exidi :d s
would you design for and at what rate? Would you respond differently for different reactor or containment types?
If so, what differences would you recommend?
e 1,
m-
= - -
[7590-01]
8.
Would you recommend that all nuclear power plants cperate with a nitrogen-enriched containment atmosphere as some BWR plants currently do do? Why or why not,and, if not,ywhich types of' containment, if any, n.wwd would you limit nitrogen enrichmentlte?-
skd pec nastM iucorpomb 9.
A m =u-in favor cf : nc.: requiremcM-L xW into containment m......, a core retention system to mitigate the consequences of core meltdown by, for example, increasing resistance to molten core debris penetration and thereby substantially reducing gas, vapor and aerosol generation to less than that which occurs when core ~ debris is allowed to interact with concrete? Assuming a core retention system is required, do you favor a device that delays melt-through of the eR containment basemat verce: a device that permanently retains core f
debris within the containment building?
If you favor delay of core melt-through, do you recommend refractory materials (such as Mg0, Zr0 ) to ' protect the containment cor. crete basemat, or do you recommend 2
some other means?
If you favor permanent retention of core debris, do you recommend using refractory materials in combination with cooling systems that rely either on natural convective cooling or forced pumping of coolant around the extremities of the refractory material, j
or do you recommend some other concept? Would you and differently for different containment types?
If so, what difft wces would you recommend? How do your recommendations impact on other safety considerations? _
h l4 Alt?c_ Ye q Ire de N M e k gg Yo accomife
- 10. M et-des 47-chm;er, 4"-anypnuM-vou u-exf-t; asunt-fer increased radioactive material that may be transported during an accident by systems normally funct.oning with much lower levels of 19 Cn,-1 n e e ma__1
[7590-01]
radioactive material such as the steam and residual heat removal systems and the containment drainage system?
5(ould. A)R. C re ade e 11.
he A ir f ver cf-ac444ns-ac-h-a:- equiring-more extensive operator training, +eg i ? g strict literal compliance with new and improved detailed operating procedures, increased reliability of emergency el cooling or decay heat removal capability, and expandmg control room minimum manning ~as alternatives or supplements to degraded cooling design improvements?
5' m (d A)(tC W M t M.
12.
Art ycs in favor-cf : rege+ cement--to-r % an alternate, add-on, self-cont'ained decay heat removal system to prevent degradation of the core or to cool a degraded core,in contrast to Stev4dieg the previously discussed schemes which are aimed toward mitigating the consequences of degraded core. cooling? How would such a decay heat removal system affect other safety considerations?
Are&$ 9t(Jltc. Yc.wtba S
13.
-youmn-f+voMf-a requirement-t+-4+o-+4 systems such as the make-
% 6c WM up and purification systems in a leak-tight building? Would such a requirement add to or detract from overall plant safety?
14.
What design, quality and seismic criteria would you recommend for any additional systems to prevent the potential breeching of contain-ment such as systems for controlled filtered venting, hydrogen com-bustion control, and core retention mentioned in previous questions?
Do you favor evaluating designs of such systems on a realistic basis, as opposed to the conservative method used to evaluate engineered safety features? Do you favor establishing design criteria for such systems that are equally stringent, less stringent, or more strin-n Fnt1ncuro 1
[7590-01]
gent than those applied to engineered safety features?
Please explain your response in terms of criteria you would recommend, including consideration of redundancy, diversity, testability, inspectability, and structural design limits (including seismic requirements).
15.
Can probabilistic analysis be used both as an aid in determining i
and comparing the adequacy and usefulness of the several futures mentioned in previous questions and as an aid in determining the design criteria and reliability requirements for these features?
How do you view the utility of quantitative risk analysis in better understanding the safety advantages and disadvantages of the several features mentioned in previous questions?
tk 16.
In weighingscosts of design and operational improvements to cope
.ge, og acir Jithb, with degraded core cooling against benefits what quantitative methods g
g a % e q c r G.u. b e e._
j eu.c of tAumb would you suggest to facilitate preparation of a 1
useful value-impact assessment? Would you consider useful or appro-priate comparisons between nuclear power plant risks and other risks to which people are exposed?
17.
What aspects of degraded cooling or melted-core accidents are suffi-i ciently unknown or uncertain se-as to impede mM4satsg-system design
,F w,W h %t.rs #c.*s and analysis and thus requir[Edditional research or experimentation?
s TheNRChasunderway[separaterulemakingproceeding/concerning 18.
-emer-gency--phnning 2nd-reactor siting If you are familiar with btheseseparateactivities,howwouldyoumodifypresentandproposed requirements for emergency planning and reactor siting if accidents t.mMency pknn Lc-g le M a
mutsy been uppnad.
Y
[7590-01]
beyond the present design basis were to be considered in nuclear power plant safety analyses?
Dated at Washington, D.C., this day of 1980.
For the Nuclear Regulatory Commission.
Samuel J. Chilk Secretary of the Commission e
i
^
w
_