ML19250C830
| ML19250C830 | |
| Person / Time | |
|---|---|
| Site: | 05000515, 05000471, Allens Creek, Atlantic Nuclear Power Plant, Perkins, Skagit, Black Fox  |
| Issue date: | 02/11/1981 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML19250C831 | List: |
| References | |
| REF-10CFR9.7, TASK-RICM, TASK-SE SECY-81-020A, SECY-81-20A, NUDOCS 8103030450 | |
| Download: ML19250C830 (12) | |
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SECY-February 11, 1951
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RULEMAKING ISSUE gE n(
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For:
The Cc Tission W
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Executive Director for Operations
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POLICY ON PROCEEDING lTH 'ENDIN3 CONSTRUCTION CERMIT f
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AND iG..UFACTURING LICtNSL AFDLICATION3 4
7 Purocse:
The Comission consicered this subject at : meeting
~2 January 13, 1980.
As a result of that nesting. the
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staff was recuested to:
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(1) Prepare a final rule; w
E (2)
Study further the staff recommendation for re';uire ents 3
re'ated to degraded core rulemaking (Action Item II :. 8); and E.
j (3) Provide staff vius tn how the construction per-it 2
process can be acdified *o ersure that major desian 3
chuges are subject to prior NRC apprnvel, a$
This ceper resocads to that reuuest.
a-Discussion:
1.
Final Rule Attached as Enclosure 1 is a draf t of the orcocsed final 3,
h rule.
NUREG 0718, reieranced in the ule, centains a statement of recuirnent for Action Plan Item II.B.S
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censistent with the discussinn anc recorrendations in the following section.
Contact:
j R. A. Furple X279BO
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SECY NOTE:
This paper, which is identical to ADVANCE COPIES which were distributed to Comissioners en February 11, 1931, is currently scredueld for
-l discussion at a Comission meeting on Thursday.__ february 12.
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2.
Decraded Core Rulemakino In January, the staff position included a requirement for (a) a probabilistic ris< assessment; (b) a dedicated containment penetration for possible future venting, and (c) a strengthened containment.
As a result of further staff consideration, the results of studies and evaluatiens by utilities and ven6 s, and ACRS coments, the staff has deveicoed a proocsed final position.
A detailed statement of the reuuirements for pendiag CP applicatiens is attached (Ercicsure 2).
This statement, when approved, will be incorporated in NUREG-0718 under Item II.B.S.
In brief, the present oosition is as follcws:
a.
We have retained the recuirements for probabilistic risk assessments and dedicated cenetrations as creviously des cri bed.
We believe that there should be conside r argin against detonation (rapid combustion) of c.jarogen that could threaten containment integrity.
To provide this margin, we have required that the containment and associated systems be designed so as to ensure that uniformly distributed hydrogen concentrations do not reach a level greater than about ene-half of a typical detonation concentration.
This recuirement, we believe, can be met only if some means of hydrogen control are included in the design.
We have specifically recuireo that additional hydrogen control r easures be in. iuded in the facility design.
The two most likely options for hydrogen control in the near term are post-accident containment inerting and distributed ignition systems.
Given r2 nydrocen release into coi.tainment, either of these systems would result in an increased internal containment pressure.
Neither system has been evaluated enough to clearly demonstrate that one is superior to the other.
Therefore, we have structured the containment strengtn provisions in the recuirement statement so that either cotion is allcwed and can be accortnodated.
To ensure that sone containment strengthening is, in fact, acc eplished for the small containment designs, se propose recuiring that all containments ce capable of w-thstanding, without exceeding CE code yield stress values, an internal static oressure of 45 osig' (modest deviations would be considered).
Containment strencthening in the ice concenser and EWP Mark III containments will be required to meet the stated requirements. We believe that this strangthening can be achived without major redesign or significant delay.
In:ustry presentatiens generally support this belief.
We believe that the attached requirement statement achieves or criginal goal - i.e., to permit the Cp review process to resume withcut foreclosing cssible mitigative or preventive features (except fer core retentien features) that may result from the degraded core rulemaking, without requiring major facility redesign, and in a manner that minimizes uncertainty.
3.
ACRS Recomendations The ACRS, in its consideration of the staff cosition regarding II.S.8, has suggested that, for these pending CP/ML applications:
1.
The Comission state an aim of seeking such improve-ments in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant.
Acplicants should take steps that are in harmony with that aim.
2.
Modest deviatiers from the specific containment strength criteria in cosition 4a of Encicsure 2 should be censidered on their merits, if prc:csed by applicants.
3.
Applicants should be required to perform design studies of possible hydrogen control measures and filtered venting for containment which have the potential for mitigating consequences of large scale core damage or core melting and thct the choice of hydrogen car. trol measures at each plant be made with the benefit of the broader study.
We agree with these recommendations and have included them, with one exception,in the II.B.3 recuirement statement (Enc ~csure 2).
Tne erception related to Item 3, above, on design studies.
In cur resconse to similar ACRS mcomendatiens (December 13,1979 and Sc;tember 8,1980), we advised the Commission (by mero dated Septercer 25, 1980) that a clan for implementing this recomendation would be submitted by the staff pursuant to Action Plan Item II.S.S.
~his submittal will include consideration of wnetner to require, during c:nstruction, tna tne recomenced studies be conducted by the pending C?
applicants.
_4_
4 Post-CP Desien Chances An advanced nr.tice of proposed rulemaking en this issue has been published in the Federal Register.
The coment ceriod ends February 9.
The Comission has recuested that the staff develop a croccsed rule 50 days after the and of the ccrent ceriod, i' pcssible.
Assuming this schedule can be met, a final rule acpears ocssible within about one year.
We consider that it is unlikely that a new CP ;ill te issued prior to about tid-1982.
It is possible, therefore, that a regulation deeling with post-CP design changes will Le in place before the next CP is issued.
If that should prove not to be the case, we would still have the option to include acprcprir.tc conditiers in the constructinn permit.
We recommend, therefore, that we defer attemots to specify such conditions at this time.
By the time such conditions are needed, our ability to articulate raasonable and practicable conditions sill be greatly imoroved, because of the considerations that will have gene into the rulemaking effort.
Recommendations:
That the Cc=ission:
1.
Accrove for finalization the draft crocosed final rule in Enclosure 1.
2.
Accrove the orccosed statement of recuirement related to cegraced core rulemaking in Enclosure 2 for incorporation into NURC 3718.
3.
Defer further consideration of controlling post-CP design or crocedural changes until the ongoing rulemaking en this subject is finalized or the first new CP is ready for issuance, whichever cccurs first.
DISTRIBUTION f) k Comissioners f
Comission Staff Offices 4ill M, m.4 t, f Exec Dir for Operations J. Dircks CRS 7
Executive Director for 0:eraticns 45LEP Secret y g, ares-1.
Draft of Proposed Final Rule 2.
Staff Dositien Re: CD Recuirement with Fesrect to Degraded Core Rulemaking
ENCLOSURE 1 NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 Licensing Requirements for Pending Construction Permit and Manufacturing License Applicaticns AliENCY :
Nuclear Regulatory Comission ACTION:
Final rule
SUMMARY
The Nuclear Regulatory Comission is adding to ita power reactor safety regulations a set of licensing requirements applicable only to construction permit and manufacturing license applications pending at the effective date of this rule. This rule contains the necessary and suffi-cient set of requirements derHed from the lessons learned from tne accident at Three Mile Island as applicable to these constructions oermit and manufacturing license applications. Each applicant for such a permit or license coverec by this rule must s.eet these requiremnts.
EFFECTIVE CATE:
[ Insert date 30 days after publication in the FEDEPAL REGISTER.]
FOR FURTHER INFORMATION CONTACT:
Albert Scnwencer, Office of Nutlear Reacto-Regulation, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555. Telephone: (301) 492-7011.
SUPPLEMENTARY INFORMATION:
Background of the Rulemaking The events leading up to promulgation of this rule were discussed in detail in the Notice of Procosed Ruiemaki:;g. This Notice aopeared in the FEDERAL REGISTE? on October 2,1950, at pages EE247-552tS.
2 Among othe r things, the noti ce requested public co :nent on the requirements Contained in NUREG-0713. The notice stated that "Following receipt of public coments,
ne Commission will finali:e its position and *-ke appropriate action, including the possible issuance of final rules on sore er all of inese matters."
At a public meeting on January 13, 1951, tne Cc=ission instructec its staff to prepare a final rule codifying the require ents cf NUREG-0718. On February
, the Commission voted approval of this rule, and instructed that it be published in the FEDERAL REGISTER.
Analysis of Public Comrents
[ Insert Enclosure 2 to SECY-51-20]
Service of tbc Rule on Affected Parties This rule, together with the final draft of NUREG-0718 published in February,1981, and other supporting documentation, is being served directly upon all affected parties. Service will be made upon alt persons, organizations, and governrental entities appearing on the docket service lists for the following licens,ing croceedings:
P rocee di ne NRC Docket No.
Allens Creek 50 456 Black Fox, Units 1 and 2 50-556, 557 Skagit, Units 1 and 2 50-527.
523 Pilgrim, Uni: 2 50 7.71 Perkins, Units 1, 2, and 3 50 455, 459, 490 Pebble Springs, Units 1 and 2 50-515 Offshore Power Systems 53 427
3 Availability of NUREG-0718 and Other Documents Nt' REG-0713 can be obt;'ned bywritten request sent to either of the following addresses:
GPO Sales Program, Division of Technical Information and Document Centrol, U.S. Nuclear Regulatory Commission Washington, D.C.
20555; National Technical Information Servi ce, 5255 P:rt Royal Road, Springfield, Vi rginia 22151. Cther NRC publications and documen:- referencen in NUREG-0715 are available as follows.
Document Clas:
Source NUREG GPO Sales Program, NTIS Standard Review Pitn NTIS Regulatory Guide GPO Sales Program I&E Sulletin 5eth Matosko Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Commission Orders NRC Public Documettt Room 1717 H Street N.W.
Washington, D.C. 20555 Requests should identify the specific document or documents needed, by full title and date of publication if possible. All of the above-listed documents are available for public inspection and copying at the Concission's Public Document Room,1717 H Street N.W., Washi ngton, D.C.
4 Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorgani:ation Act of 1974, as amended, and sections 552 and 553 of Title 5 of the United States Code, the folicwing amendment to Pa rt 50 of Chapter I, Title 10 of the Code of Federal Regulations is published as a document subject to codification.
Part 50 - Donestic Licensing of Production and Utilization Facilit'es A new paragraph (e) is added to @ 50.34 to read as follows:
(e) Additional THI-rel ated recuirements.
In addition to the requirements of paracrach (a) of this sect on, each acplicant for a construction permit or manufacturing license whose application was pending as of [ insert effective date of rule] shall meet the requirements set out in NUF.EG-0715 as published in February,1931.
[ Secs.161b,161i, Pub. L.83-703, 68 Stat. 943, 42 U.S.C. 2201; Secs. 201, 204(b)(1), Pub. L.93-438, SS Stat.
1., 1243, 1245, 42 U.S.C. 534!, 5 844.]
Dated at Washington. D.C., this day of Feoruary,1931.
For the Nuclear Regulatory Commission, Samuel J. Crnik Secretary of tne Comissicn
tnclosure z-STAFF POSITION RE. CP REQUIRE"ENT WITH er.c o. r r.. 4 0 nrc n> r..e n CCm. e. nL, rv..a I o.iar s-
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1.
Commit to performing a site /clant-specific crobabilistic risk assessment and incor: orating the results of the assessment into the design of the facili ty.
The commitment must include a program plan, acceotatle to tne staff, that dencnstrates hcw the risk assessrent program will be scheduled so as
- influence syste-designs as they are bein: cevelcrec.
The assessment shall be completed and sutritted tc 'GC within two years of issuance of the construction cermit.
The outccre of :nis study and the NRC review of it will be a determinatien of s ecific creventive and mitigative actions to be implemented to reduce these risks.
A creventicn feature that must be considered is an additional decay heat removal system whose functional recuirerents and criteria would be derived from the FRA study.
It is the air of the Conmissicn thrcugh tnese assessments to seek sucn im rovements in the reliability of core and c ntainment heat removal systems as are significant and cractical and do not intact excessively On the plant.
A clicants are encouraged to take stens that are in harmcny with this aim.
2.
In order not to preclude the installation of systems to orevent centainment failure, such a fil tered vented containment systen, the containment design shall include Orovisions for one or more dedicated :enetrations, enuivalent in size to a single 3-fcet diar.eter c:ening.
. 3.
Hydrogen control measures shall be Droviced.
4 Arolicants snali provice preliminary cesign informa :icn at a level consistent with that normally recuired at the consti ucticn oermit stage of review sufficient to demonstrate that:
a.
Containment integrity will be maintained (i.e., for s teel containments, ASME Service Level C based on ASME, Division 1 Code specified minimum yield values and considering pressure and dead load alone.
For concrete containments, membrane tensile strain of the liner plate not to exceed 0.003 in/in, consider pressure and dead load alone, based on ASME Division 2 Code) during an accident that releases hycrogen generated from 100% fuel cicd metal-water reacticn accompanied by eitner hydrogen burning or the added pressure from post-accident inerting assuming caroon dioxide is the inerting agent depending upon which option is chosen for control of hydrogen.
As a minimum, for steel containments ASME Service Level C (based on ASME, Division 1 Code, specified minimum yield values and consicering pressure and dead load) will not be exceeded at an internal pressure of 45 psig.
For concrete containment s tructures, membrane tensile strain of the liner less than 0.003 in/in cased on ASME, Division 2 Coda, is satisfied at tne same internal press ure.
Modes t deviations from these criteria will be considered by the staff, if good cause is shown by an applicant.
-3 Systems necessary to ensure containment integrity shall also be demonstrated to cerform their functicn under these conditions.
b.
The containnent and asscciated systems will crovide reascnable assurance that uniformiv - distributed hydrcten ccccentraticns cc nct excesc IC= asscciated witn ar accicent that releases hvdrcgen cenerated frcm ICJT fuel clac etal-water reacticn, or that the cost-accident atmos;here will not succort hycr gen combustion.
c.
The facility design will provide reascnable assurar.ce that, based on a 1005 fuel clad metal-water reaction, ccmoustible concentraticns of hydrogen will not collect in areas whern unintended comoustion or detonation could cause less of containment integrity or loss of accropriate mitigating features.
d.
If the cotion chosen for hydrogen control is cost-accident inerting:
(1)
Containment structure loadings produced by an inadvertent full inerting (assuming carbon dioxice) b.: not including seismic or design basis accident loadings, will not produce stresses in excess of the acceptable maximum for Service Level A specified in ASME Division 1 Code (for concrete containments membrane tensile strain of tne lirer not to exceed 0.002 in/in based on ASME, Division 2 Coce).
~
(2)
A pressure test of tne containment a 1.10 and 1.15 times the pressure calculated for steel and concrete containments rcsec.tively, to result frca carbon dioxide inerting can be safely concucted.
(3)
Inadvertent full inerting c# :ne contair ent can te safely acccc ccate: during D' int creratirn and de~:Onst rate: Oy test.