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PR-050 - 57FR44513 - Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations (ANPR)
	ML23151A498
Person / Time
Issue date:	09/28/1992
From:	Chilk S; NRC/SECY
To:	;
References
	PR-050, 57FR44513
	Download: ML23151A498 (1)
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TITLE:

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KEYWORD:

ADAMS Template: SECY-067 09/28/1992 PR-050 - 57FR44513 -ACCEPTABILITY OF PLANT PERFORMANCE FOR SEVERE ACCIDENTS; SCOPE OF CONSIDERATION IN SAFETY REGULATIONS (ANPR)

PR-050 57FR44513 RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

STATUS OF RULEMAKING PROPOSED RULE:

PR-050 OPEN ITEM (Y/N) N RULE NAME:

ACCEPTABILITY OF PLANT PERFORMANCE FOR SEVERE ACCIDENTS; SCOPE OF CONSIDERATION IN SAFETY REGULATIONS (ANPR)

PROPOSED RULE FED REG CITE:

57FR44513 PROPOSED RULE PUBLICATION DATE:

09/28/92 ORIGINAL DATE FOR COMMENTS: 12/28/92 NUMBER OF COMMENTS:

EXTENSION DATE:

I I

11 FINAL RULE FED. REG. CITE: 62FR53250 FINAL RULE PUBLICATION DATE: 10/14/97 NOTES ON:

SUMMARY

OF COMMENTS IN SECY-93-226.

NOTICE OF WITHDRAWAL ISSUED 0 STATUS N 10/7/97.

NOTICE PUBLISHED ON 10/14/97 AT 62FRS3250. SEE SECY 97 OF RULE: - 148 AND SRM DATED. 8/6/97 FOR BACKGROUND.

FILE ON Pl.

HISTORY OF THE RULE PART AFFECTED: PR-050 RULE TITLE:

ACCEPTABILITY OF PLANT PERFORMANCE FOR SEVERE ACCIDENTS; SCOPE OF CONSIDERATION IN SAFETY REGULATIONS (ANPR)

PROPOSED RULE PROPOSED RULE DATE PROPOSED RULE SECY PAPER: 92-292 SRM DATE:

09/17/92 SIGNED BY SECRETARY:

09/22/92 FINAL RULE FINAL RULE DATE FINAL RULE SECY PAPER: 97-148 SRM DATE:

08/06/97 SIGNED BY SECRETARY:

10/07/97 STAFF CONTACTS ON THE RULE CONTACT!: CHARLES E. ADER CONTACT2:

MAIL STOP: T-l0K-8 PHONE: 415-5622 MAIL STOP:

PHONE:

DOCKET NO. PR-050 (57FR44513)

DATE DOCKETED DATE OF DOCUMENT In the Matter of ACCEPTABILITY OF PLANT PERFORMANCE FOR SEVERE ACCIDENTS; SCOPE OF CONSIDERATION IN SAFETY REGULATIONS (ANPR)

TITLE OR DESCRIPTION OF DOCUMENT 10/22/92 11/09/92 12/21/92 12/24/92 12/28/92 12/28/92 12/29/ 92 12/29/92 12/29/92 12/30/92 01/04/93 01/27/93 10/07/97 09/22/92 11/02/92 12/16/92 12/21/92 12/22/92 12/22/92 12/23/92 12/24/92 12/27/92 12/28/92 12/29/92 01/21/93 10/07/97 ADVANCE NOTICE OF PROPOSED RULEMAKING COMMENT OF JOHNS. FUOTO (

1)

COMMENT OF ENTERGY OPERATIONS, INC (JAMES J. FISICARO, DIRECTOR) (

7)

COMMENT OF AECL TECHNOLOGIES (A.O. HINK, VICE PRESIDENT) (

2)

COMMENT OF WESTINGHOUSE ELECTRIC CORP (N.J. LIPARULO, MANAGER) (

3)

COMMENT OF NUCLEAR MANAGEMENT & RESOURCES COUNCIL (WILLIAM H. RASIN, VICE PRESIDENT) (

4)

COMMENT OF ADVANCED LIGHT WATER REACTOR UTILITY STEERING COMM (E.E. KINTNER, CHAIRMAN) (

5)

COMMENT OF DEPARTMENT OF ENERGY (E.C. BROLIN, ACTING DEP SECRETARY) (

6)

COMMENT OF OHIO CITIZENS FOR RESPONSIBLE ENERGY, INC (SUSAN L. HIATT, DIRECTOR OCRE) (

8)

COMMENT OF NIAGARA MOHAWK ET AL (MARK J. WETTERHAHN, ESQUIRE) (

COMMENT OF TENNESSEE VALLEY AUTHORITY (MARK J. BURZYNSKI) (

10)

9)

COMMENT OF FLORIDA POWER CORP (PAUL V. FLEMING) (

11)

NOTICE OF WITHDRAWAL OF ADVANCE NOTICE OF PROPOSED RULEMAKING. NOTICE PUBLISHED ON 10/14/97 AT 62FR53250.

DOCKET NUMBER PR c-A PROPOSED RULE

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( S ; F 'Y~ 'i Lf 5 J 3 NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AE38 DOCKETED US R.,

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\ I r'\i Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations AGENCY : Nuclear Regulatory Commission.

ACTION : Advance notice of proposed rulemaking : Withdrawal.

SUMMARY

The Nuclear Regulatory Commission (NRC or Commission) is withdrawing an advance notice of proposed rulemaking that outlined alternative approaches to generic regulation addressing the challenges from severe accidents for future light water reactors. The Commission has decided that a rule change to provide generic requirements for performance during postulated severe accidents is not warranted at this time. The basis for this decision is that a purpose for the rule was to provide guidance for future designs and to facilitate then ongoing design certification rulemaking. With all current design certification rulemaking either complete or nearing completion and future applicants not foreseen. expenditure of the resources to promulgate the rule is not warranted.

FOR FURTHER INFORMATION CONTACT: Charles E. Ader. Office of Nuclear Regulatory Research. U.S. Nuclear Regulatory Commission. Washington. DC 20555-0001, telephone (301) 415-5622.

SUPPLEMENTARY INFORMATION : On September 28. 1992. (57 FR 44513). the Commission published an advance notice of proposed rulemaking (ANPRM) to pJ. 0,,-, 10/14191,,;l 6J-Fv2SJ 'LSO

consider amending its regulations to provide generic requirements to address t he challenges from severe accidents for future light water reactors.

The advance notice of proposed rulemaking outlined three alternative approaches to the specification of requirements addressing severe accident performance.

The first alternative. described as a hardware oriented rule. would specify reasonable design features or design characteristics directed towards prevention or mitigation of explicitly identified risk significant phenomena.

The risk significant phenomena identified were:

hydrogen generation.

transport and combustion. high pressure melt ejection. core concrete interactions and basemat ablation. long term containment overpressurization.

steam explosions from fuel-coolant interactions. and containment bypass.

These phenomena represent the potential contributors to containment failure or bypass and thus the mechanisms for large offsite radioactive release.

Alternative 2. described as a phenomena oriented rule. was a modification of the first alternative wherein an overall containment performance goal would be specified along with the phenomena to be considered. as identified above.

The designer would then be required to perform analyses ' of the impact of those phenomena and develop and propose the design features to meet the goal.

Regulatory guides would address analytical methods. acceptance criteria and design criteria for hardware. This approach. similar to Alternative 1. would be an overlay on the existing design basis specified in 10 CFR Part 50 and justified on an enhanced safety basis.

The third alternative. described as a general design criteria (GDC) oriented rule. involved development of a set of new design requirements to address specific challenges and issued as changes to Appendix A. "General Design Criteria" to 10 CFR Part 50. Each new design criterion would describe the nature of the challenges as well as the success 2

criterion and involve the development of Regulatory Guides to provide additional guidance on analysis methods and assumption. This approach was similar to the other alternatives. especially Alternative 2. but differs in that the existing 10 CFR Part 50 design basis would be modified to include severe accidents.

A primary purpose for the generic severe accident rulemaking was to add consistency and standardization to the resolution of severe accident issues for future designs based on current technical information. While. in general.

consistency among many design reviews is best achieved through generic rules.

as a practical matter. since the number of new appl icants is l ikely to remain quite limited. it is more efficient to proceed with design-specific reviews.

In fact. the Commission is not aware of any new appl icants in the foreseeable future.

Another purpose of the generic severe accident rulemaki ng, i.e..

facilitation of design certification rulemaking. has been rendered moot by the experience gained in design certification rulemakings. The design certification rulemakings are completed for the General Electric Advanced Boiling Water Reactor and ABB-CE System 80+ and the only design currently under review is the Westinghouse AP600. The resolution of severe accident design specific requirements would be set forth in the AP600 design control document and approved in the AP600 design certification rulemaking.

While certain arguments in favor of generic rulemaking Ci.e.. promoting consistency and standardization in the resolution of severe accident issues 3

and providing guidance to future LWR designers and applicants) continue to apply in varying degrees. practical aspects limit the need for such an activity. At this point. given the lack of any new potential plant or design applicants. the Commission believes that the benefits of generic rulemaking do not justify the allocation of resources to proceed with the development of new regulations addressing severe accidents.

Upon consideration of the potential value of a generic rule. the status of the review and design certification of future reactors. and the potential resource requirements. the Commission believes that the value in pursuing generic severe accident rulemaking does not warrant the resource expenditure.

While the Commission does not perceive the need for generic rulemaking in the foreseeable future. should conditions change regarding potential applicants.

the Commission would reassess the merits of rulemaking at that time.

For the reasons discussed. the Commission is withdrawing the ANPRM.

-rt_

Dated at Rockville. Maryland this L_ day of October. 1997.

For the Nuclear Regulatory Commission John

. Hoyle.

Sec~ tary of the Commission.

4

January 21, 1993 f.GCK[1 [D US NHC The Secretary of the Commission, U.S. Nuclear Regulatory Commission Washington, DC 20555

93 JAN 27 P 3 : 1 3 ATTN:

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Subject:

Comments on Proposed Rulemaking -

"Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations"

GENERAL COMMENT

S:

The NRC is contemplating rev1s1ons to 10 CFR Part 50 to include design considerations for challenges from severe accident (SA) phenomena for future light water reactors ( LWRs).

As stated in the proposed regulation existing nuclear plants do not pose undue risk to public health and safety because present design and regulation is sufficient, although very complex and burdensome.

The NRC states that it recognizes the need to"... strike a balance between accident prevention and consequence mitigation in exploring the need for additional design features in the next generation of plants." However it is not clear how the NRC demonstrates that recognition in light of the proposed regulations.

In proposing the regulation the NRC identifies two fundamental areas of concern:

1) SA mitigation through improved design based on insights from IPE and resolution of several issues such as station blackout, anticipated transient without scram, hydrogen generation and control, and 2) Containment performance; establish new performance criteria.

The former concern s imp 1 y is not be expressed in terms that designers are able to use without skyrocketing the cost of building a nuclear plant. The latter concern attempts to encapsulate the SA in an impenetrable (energy/mechanical) device.

To aid in the determination and applicability of this proposed regulation three alternatives are provided. The first of these (Alternative 1) attempts to define hardware needs in terms of six (6)

"risk significant" SA phenomena.

Unfortunately the features that would be defined in the rule were not available for review now.

Although a statement is made to address "... these new regulations would not be considered to be traditional design basis requirements",

it does not differentiate what that means with respect to operability type evaluations. It is pointed out that a regulatory guide would be issued to"...

provide additional guidance on such design details such as redundancy, diversity, system capacity, power

supply, equipment survivability, and analytical assumptions." However it is not clear why this much detail is proposed for such a regulation if an understanding of the relationship between prevention and mitigation is really appreciated.

Finally, this alternative closes with a statement that would allow the licensee to justify design without performing "extensive" analysis to show compliance.

The problem is NRC still wants justification based on likely SA scenarios combined with deterministic analysis.

This is quite costly and may not be able to prove anything.

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Nuclear Regulatory Commission Page 2 of 4 Alternative 2 attempts to define containment performance requirements in terms of analytical methods, assumptions, acceptance criteria and guidance on design criteria for SA hardware. The NRC recognizes that this approach would probably (necessarily) limit the types of analytical tools and code to those specified by the staff due to the diversity, complexity, and uncertainty of available or yet to be developed codes. It would be very difficult, if not impossible, to agree on code selection and analytical methodologies.

Alternative 3 uses the General Design Criteria (GDC) approach. Because SA design basis would (probably) be different from traditional design basis, regulations would necessarily be more detailed and complex. This approach would rely on many of the methods used in Alternative 2 to define related basis which may or may not be acceptable to the NRC.

COMMENTS TO SPECIFIC CONSIDERATIONS:

1.

A rulemaking for severe accident space is not good business.

Rules and regulations should more closely correspond to issues that really challenge the health and safety of the public.

Integrating IPE information into design criteria for future reactors may be app 1 i cab 1 e for cases where significant safety improvement is derived.

2.

If source terms are significantly adjusted (with or without SA design modifications/codification) such that dose rate projections and significant benefits are derived from revising Emergency Planning Zone requirements then a basis for revision is indicated.

3.

Under worst case conditions containment pressure is projected to reach as high as 200 psig.

To prevent challenging the containment venting is indicated. It is not reasonable to design a containment to withstand such pressure and codify it. Although UPC is relatively high for many types of containment designs venting is a reasonable strategy to prevent exceeding it thus reducing the probability of an uncontrolled release.

If a hardness factor based on realistic failure probability can be applied to design criteria established for "traditional" design basis then there may be some safety benefit derived.

4.

This relates directly to earlier comments made to Alternative 2 & 3. Some codes and analytical methods may need to be developed.

It's not clear what standard(s) (if any) would be used to develop and evaluate such codes. With the wide variety of assumptions and existing codes, including revisions/versions it would be very difficult to get global buy in on any particular one, unless of course it is specified in regulations which may or may not be analytically correct.

5.

No.

Nuclear Regulatory Commission Page 3 of 4

6.

If SA design criteria "must" be codified then the only realistic approach would be high level general statements. Unfortunately this approach does lend itself to a host of other problems which typically begins with differences of interpretation leading to varied levels of compliance.

7.

SA space is independent of reactor design in that SAs begin with a degraded core which is "beyond" design basis (or should be).

Therefore applicability to passive designs should be the same.

8.

This is an open ended question and not really appropriate here.

9.

No.

There is no evidence that a HPME is even a reasonably possible (probably) event.

Even if it were the consequences of such an event are not worthy of design considerations in terms of safety benefits.

10.

Although this is an interesting concept surveillance activities and containment controls are sufficient.

Should a bypassed or impaired containment develop during a SA the usefulness of such instrumentation, provided it would be available, is minimal.

The cost of development to some arbitrary design criteria, installation, calibration, maintenance, and actions to take upon degraded performance would be very difficult to demonstrate safety benefits.

11.

This is a separate issue.

12.

If equipment is targeted for SA mitigation and "must" be regulated for that function then those regulations need to reflect a significantly lesser set of requirements.

What those would be and what they would be based on is not readily definable at this time.

13.

Existing design requirements with enhancements as identified by IPE and adjusted per comment 3 may be an appropriate means of establishing limits.

14.

The answer to this question should be directed to and obtained from NUMARC.

15.

If containment performance criteria includes SA phenomenal ogy then it would bound all other regulatory concerns.

Nuclear Regulatory Commission Page 4 of 4 CONCLUSIONS:

Although the concept of defense in depth may be applied to SA mitigation strategies as a part of overall accident management, it is not reasonable to generally include those in design criteria regardless of separation from "traditional" design criteria. If significant vulnerabilities exist which pose a threat the health and safety of the public (in SA space) then we are obliged to revise our design criteria (for the future) and operating practices. Based on opening NRC comments, existing nuclear power plants do not pose such a threat.

Therefore, consideration of regulatory change to include SA mitigation design criteria should be limited to reflect significant insights from IPE and clearly identified design feature improvements where safety benefits are meaningful and quatifiable.

Ir rnJ Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 December 29, 1992 Mr. Samuel J. Chilk Secretary of the commission ATTN:

Docketing and service Branch U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Chilk:

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NUCLEAR REGULATORY COMMISSION (NRC) -

REQUEST FOR COMMENT ON ADVANCED NOTICE OF PROPOSED RULEMAKING; ACCEPTABILITY OF PLANT PERFORMANCE FOR SEVERE ACCIDENTS: SCOPE OF CONSIDERATION IN SAFETY REGULATIONS The Tennessee Valley Authority (TVA) has reviewed the subject advanced notice of proposed rulemaking, which was noticed in the September 28, 1992 Federal Register (57 FR 44513-44518), and is pleased to provide the following comment.

TVA supports those comments submitted by Winston & Strawn on this proposed rulemaking.

Sincerely, Mark J. Burzynski Manager Nuclear Licensing and Regulatory Affairs cc:

U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. Thomas King Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555 JAN 2 1 1993 Acknowledged by card...............................,,,

DOCKET NUMBER PR 5 0 PROPOSED RULE~~ ----

( S 1 F fl L.J LJ 5 13 J WINSTON & S'rRAWN 1400 L STREET, N.W.

FREDERICK H. WINSTON (1853-1886)

SILAS H. STRAWN (1891-1946)

WASHINGTON, D.C. 20005-3502

92 DEC 30 (202) 371 -5700 FACSIMILE (202) 371-5950 WRITER ' S D IRECT D I A L NUMBER December 28, 1992 Mr. Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Attention:

Docketing and Service Branch f*'

CHICAGO O FFICE 35 WEST WACKER DRIVE p 3 '?, l~

GO, ILLI NOIS 60601 (312) 558-5600 NEW YORK OFFICE 175 WATER ST REET

\' NEW YORK, NY 10038-4981 (212) 269-2500 RE:

Response to Advance Notice of Proposed Rulemaking - -

Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations 57 Fed, Reg, 44,513 (September 28, 1992)

The Nuclear Regulatory Commission

("NRC" )

recently published an advanced notice of proposed rulemaking ( "ANPR" ) t hat would, for the first time, require new nuclear power facilities to be designed to withstand the effects of a severe accident.

57 Fed.

Reg. 44, 513 (September 28, 1992 ).

On behalf of the licensees identified below,Y we are filing the following comments in response to two aspects of the ANPR:

(1 ) the propriety of a rule amending 10 C. F. R. Part 50, as opposed to utilization of the existing design certification process codified in 10 C.F.R. Part 52 ;

and (2) reliance on severe accident miti gation design requirements as the basis for amending 10 C.F.R. Part 51 to preclude consideration of Severe Accident Mitigation Design Alternatives

( "SAMDAs").

These comments are timely filed in accordance with the September 28 Federal Register notice.

1.

The NRC Staff Should Refrain from Proposing Regulations that Specifically Address Severe Accident Mitigation, and Instead Pursue Such Issues In the Part 52 Design Certification Process The ANPR evidences a approach to severe accident risk.

limit consideration of i ssues fundamental shift i n the NRC In the past, the Staff chose to i nvolving severe accidents t o l'

These comments are submitted on behalf of t he following licensees of existing nuclear power facilit i es:

Niagara Mohawk Power Corp.,

Northeast Nuclear Energy Co.,

TU Electric, Tennessee Valley Authority, and Washington Publ i c Power Supply System.

JAN 21 1993 Acknowledged by card..................................

U.S. Nuclear Regulatory Commission December 28, 1992 Page 2 environmental evaluations and general safety goals, articulating this position in two separate policy statements.Y Even when the Staff recommended in a generic industry communication that existing nuclear facilities with Mark I containments consider beyond-design basis over-pressurization events, the Commission recognized that, absent voluntary commitment, the limitations of the backfitting rule applied)/

In the ANPR, however, the Staff proposes to require future facilities to specifically design against a severe accident,

~' an accident beyond current design bases, in accordance with discrete design requirements to be codified in 10 C.F.R. Part 50.

Not only is the Staff now focusing on an entirely new and previously undefined realm of design requirements (as opposed to the consideration of alternatives in an environmental review}

associated with severe accident mitigation, but also, at the same time, the Staff is proposing to implement this new philosophy in binding regulatory requirements to be codified in 10 C.F.R. Part 50, thereby casting doubt as to the appropriateness of the design basis for existing facilities.

This approach, if adopted, should not be implemented via a rulemaking that would amend 10 C.F.R. Part 50.

Rather, the NRC should implement design criteria of the posited nature by means of the design certification rulemaking process codified in 10 C.F.R. Part 52.

Because the ANPR would require new design criteria only in connection with the licensing of future light water reactors, and because, as a practical matter, the design of such plants will be certified by rulemaking pursuant to 10 C.F.R. Part 52, design criteria governing the performance of future LWRs under severe accident conditions should be £romulgated in the context of a design certification rulemaking.-'

~

50 Fed. Reg. 32,133 (Aug. 8, 1985}, "Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants;" 51 Fed. Reg. 30,028 (Aug. 21, 1986}, "Safety Goals for the Operations of Nuclear Power Plants; Policy Statement. 11

~

Generic Letter 89-16, "Installation of a Hardened Wetwell Vent," September 1, 1989.

Indeed, the Staff has proposed the form and content for a design certification rule. ~

SECY-92-287, "Form and Content For a Design Certification Rule, 11 August 18, 1992.

In pertinent

part, the proposed rule would require the consolidation of all design-related information into a single, stand-alone document called the Design Control Document

( "DCD"}

U.S. Nuclear Regulatory Commission December 28, 1992 Page 3

2.

NRC Should Not Revise The Part 51 "Remote and Speculative" Finding On the Basis of The Proposed Severe Accident Design Requirements The Staff specifically asked for comments on the advisability of using a

new containment performance design requirement as the basis for revising 10 C.F.R. Part 51 to "define a point of truncation" and to eliminate "the need for further review" of SAMDAs.V Presumably, the Staff would cite compliance with the proposed Part 50 severe accident design requirements as the basis for concluding that severe accident risk, as a result of implementation of the proposed severe accident design criteria, is "remote and speculative" within the context of the National Environmental Policy Act ("NEPA").

42 U.S.C. 4331.e.t..B.fill.

As a result, the benefit to be realized from the installation of SAMDAs would not have to be cost-justified under NEPA.

Under such a scheme, the proposed Part 50 severe accident design requirements would become a "floor" below which environmental alternatives regarding severe accident need not be considered further.

To proceed with such a plan would inappropriately precondition a Part 51 finding of "remote and speculative" risk on the inclusion of SAMDAs in plant design.

To date, the NRC has dismissed SAMDAs in its NEPA analyses without requiring that facilities specifically be designed to withstand the challenges of severe accidents.

For example, in an operating license proceeding that focused, in part, on the need to consider the mitigating effects of SAMDAs in an accompanying NEPA analysis (where the plant was not designed to withstand severe accidents),

the Staff

" 'discovered no substantial changes in the proposed action as previously evaluated in the FES [Final Environmental Statement]

that are relevant to environmental concerns nor significant new circumstances or information relevant to environmental concerns and bearing on the licensing of Limerick Generating Station, Units 1 and 2.' d' Moreover, the United States Court of Appeals for the Third Circuit has aptly observed that the Commission itself has noted that the impact of SAMDAs on the environment will differ with a particular plant's design, construction, and location.

Limerick 57 Fed. Reg. at 44,518 (Question 15).

Supplement to the Environmental Impact Statement, Limerick Generating

Station, Units 1

and 2,

at 1,

quoted in Philadelphia Electric co. (Limerick Generating Station, Units 1 and 2), LBP-89-24, 30 N.R.C. 152, 153 (1989).

U.S. Nuclear Regulatory Commission December 28, 1992 Page 4 Ecology Action. Inc, v, United States Nuclear Regulatory Comm'n, 869 F.2d 719, 738 (3d Cir. 1989).

Thus, the Court expressed doubt as to the feasibility and validity of according SAMDAs generic treatment under NEPA.

.I_d. ("it is unlikely that severe accident mitigation can be treated as a generic issue.")

Even if feasible, generic treatment of SAMDAs, as a practical matter, may deter the development and installation of cost-beneficial design changes at individual facilities.

In sum, we reiterate that Part 50 should not be amended because there is no need to require design changes to accommodate remote and speculative accidents.

In addition, in response to the question posed concerning Part 51, we believe that even if the Staff should proceed with rule changes to Part 50 to address severe accident risk, those changes should not be used as the basis for revision of Part 51.

Mark J. Wetterhahn Kathryn M. Kalowsky WINSTON & STRAWN

December 27, 1992 DOCKE:TcO U:,N1C t-:J.~

2-'1

92 DEC J1 A10 :03 (l)

COMMENTS OF OHIO CITIZENS FOR RESPONSIBLE Et{E"Rfl,'-; IN<}:~\ *,(OCRE")

ON ADVANCE NOT ICE OF PROPOSED RULEMAKING, "AdC.Et'T 1ABIL ITY OF PLANT PERFORMANCE FOR SEVERE ACCIDENTS; SCOPE OF CONSIDERATION IN SAFETY REGULATIONS,"

57 FED. REG. 44513 (SEPTEMBER 28, 1992)

In this notice the NRC proposes regulations which would require future nuclear power plants to be able to withstand severe acci-dent phenomena which present a challenge to maintaining contain-ment integrity.

OCRE generally supports this advance notice of proposed rulemaking.

The major flaw in the NRC's safety regulations for nuclear power plants is the exclusion of severe accidents from the design basis.

Instead of being designed to withstand all possible accidents, the current generation of nuclear reactors is designed to withstand only a limited set of events which are assumed to be mitigated by the operation of the ECCS, with minor radiological consequences to the public health and safety.

Severe accidents were deemed to be "incredible," at least until the TMI-2 accident proved otherwise. Due to this regulatory gap, severe accidents pose the greatest risk to the public health and safety from nuclear reactor operation, since current plants have limited capabilities to cope with severe accident phenomena.

E.g.,

NUREG-1150 found that even the strongest of current containments would not withstand severe accident loads.

It is OCRE's position that severe accident phenomena must be made part of the design basis for future nuclear power plants.

This advance notice of proposed rulemaking would accomplish precisely that.

Hence, OCRE supports this effort.

In addition to enhancing safety, this effort will also benefit the nuclear industry as it seeks a revival.

The public is un-likely to accept a new generation of reactors that cannot with-stand severe accident phenomena.

Of the three alternatives presented in the notice, OCRE prefers a combination of Alternative 1, the hardware-oriented rule, and Alternative 3,

GDC-oriented rule.

Due to its prescriptive nature, Alternative 1 minimizes reliance on the uncertain, imper-fect analytical methodologies which are required in the other two alternatives.

Howe ver, Alternative 1 does not make severe acci-dents part of the traditional design basis, which Alternative 3

does.

The disadvantage noted for Alternative 1, that it could 1

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discourage the development of other design approaches, does not appear to be that much of a problem; the requirements are not so restrictive as to preclude innovation, and in the event that they

are, the NRC could grant an exemption or modify the rule if necessary.
Innovative designs could also be evaluated in the SAMDA review process.
OCRE would suggest the following improvements to Alternative 1:
A.
The containment performance objective states:
"The design shall include a
containment system that provides a
barrier against the release of radioactive material for a
period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage under.
the rn likely severe accident challenges..
(emphasis added).
Any insertion of the words "likely" or "likelihood" into the rule will defeat its purpose by allowing licensees and plant designers to claim that they need not comply with its provisions because severe accident challenges are not likely.
This asser-tion will of course be buttressed by a PRA to which fudge factors will undoubtedly have been applied.
Even without the fudge
factors, PRA has enough limitations and uncertainties that it should not be used to evade compliance with the rule.
This is a
retreat to the "accidents oan't happen" mentality of the past.
The present wording of this Alternative creates a
substantial loophole which must be closed if the rule is to have any force.
B.
The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months timeframe for maintaining containment integrity is insufficient.
OCRE prefers the application of the present GDC 16, which requires the containment to provide "an essentially leak-tight barrier
.. for as long as postulated accident conditions require."
This is the approach suggested in the ACRS letter of May 17, 1991.
If containment venting is not to be prohibited, it should not be utilized until at least one week after the onset of core damage.
A one week period would provide more time for fission product removal mechanisms to work, would increase the likelihood that the accident would be recovered or stabilized, and would allow more time for offsite protective ac-tions.
The use of diverse containment heat removal systems should be required.
C.
The description of the severe accident phenomena should be more detailed and comprehensive.
The description of the phenome-na in the ACRS letter of May 17, 1991 is far more appropriate than the cursory descriptions in the Federal Register notice.
For example, the ACRS description of the hydrogen sources to be considered is: "in-vessel and ex-vessel oxidation of core materi-als, including (1) core degradation from overheating and melting, (2) steam explosions or high pressure melt ejection in the 2
presence of water, and (3) interaction between molten core mate-rials and concrete."
This is an accurate description of the actual sources of hydrogen gas.
The description in Alternative 1,
"a 100-percent metal-water reaction of the active fuel clad-ding,"
is not complete and accurate.
This description ignores many sources of hydrogen, such as the oxidation of metal other than active fuel cladding in the core, the reaction of boron carbide control rods with steam, and core-concrete interactions.
For another example, consider the ACRS-proposed criterion on core-concrete interaction:
The containment system would have the capacity to accommodate the following challenges resulting from the thermal decomposition of concrete by molten corium: (1) the degradation of containment cooling and of cleanup capability due to aerosol formation, (2) slow overpressurization resulting from the evolution of noncondensible gases, (3) functional degradation of structural concrete by erosion, including basemat penetration, and (4) combustion of carbon monoxide.
Challenges to the containment should not be sufficient to render inoperable that equipment required for containment cooling or atmospheric cleanup, nor to cause leakage in excess of the rate specified in Criterion 16 or to allow any release through the basemat within an appropriate time of the onset of the corium-concrete interaction sufficient to cause significant contamination of the groundwater.
This is a far more complete and accurate description and require-ment than is contained in Alternative 1.
OCRE recommends the incorporation of the description of severe accident phenomena and acceptance criteria from the ACRS letter into Alternative 1.
D.
The use of deliberate ignition to control combustible gases should be prohibited.
Reliance on ignition to control hydrogen requires that the igniters have a power supply, which is suscep-tible to failure.
If the power supply to the igniters fails and is restored at some later time, after the hydrogen has built up to higher concentrations, a severe deflagration or detonation could result.
The sudden condensation of steam in a
steam-inerted containment could likewise result in high hydrogen con-centrations which would then be ignited, with severe results.
Ignition of hydrogen adds the heat of combustion to the contain-ment atmosphere, can create pressure pulses which can damage the containment or equipment within it or resuspend fission products 3
which had plated out on structural surfaces, and creates the potential for accelerated flames or deflagration-to-detonation transition.
The NRC should require containment atmospheres to be inert.
This may be accomplished through full pre-accident inert-
ing, or through partial pre-inerting or post-accident inerting, or some combination of these two approaches, provided that the inerting mechanism is reliable. The potential for deinerting during the course of the accident must be eliminated.
E.
The calculation of containment stresses must be based on the as-built state of the containment.
The as-built containment will undoubtedly contain deviations from the perfect structure usually considered in such analyses.
For
example, the containment structure may be out-of-tolerance, out-of-round, or contain weld joints with indications dispositioned "use-as-is".
The effects of these imperfections, ~hich are probably acceptable for the loads resulting from the design basis LOCA, need to be evaluated for the higher pressures resulting from severe accidents.
The NRC should consider the potential for other failure mechanisms of steel containments, such as creep and low-cycle fatigue from multiple steam explosions or hydrogen deflagrations, should the NRC decide to allow deliberate ignition.
Also, for containments using suppression pools, the NRC should require evaluation of the effects of hydrodynamic loads induced by steam explosions and other rapid pressurization events, in-cluding hydrogen combustion pressure pulses, if the NRC decides to permit ignition.
F.
The NRC should require all future containments to minimum design pressure of 45 psig and a minimum net free of one million cubic feet.
have a
volume The NRC should prohibit the use of pressure-unseating equipment hatches and personnel airlocks in containments.
A pressure-unseating equipment hatch, with the hatch cover mounted on the outside of the containment, has such a large surface area that the closure bolt preload is overcome at relatively low pressures.
Containment integrity is then maintained by elastomeric seals, which may be degraded by the high temperature severe accident environment.
The NRC should establish standards for containment penetrations (including but not limited to electrical, mechanical, hatches and personnel locks, and purge/vent valves) which would require the use of penetrations types and materials which have been demonstrated to be the most resistant to severe accident condi-4
tions.
RESPONSES TO QUESTIONS POSED IN THE NOTICE OCRE is responding to selected questions posed in the Federal Register notice.
The numbering corresponds to the numbers of the questions in the notice.
1.
As stated above, OCRE believes that this rulemaking is necessary and desirable.
A rule will provide better coherence and predictability to the design review and certification proc-esses than an ad hoc approach.
A rule will also enhance the NRC's authority by requiring compliance with a rule rather than with non-enforceable guidance.
The ACRS, in its May 17, 1991 letter, provided a cogent explana-tion for why new containment criteria are needed.
The ACRS ad-vanced three reasons for the new requirements: to reduce risk and uncertainty, to clarify what is expected of applicants and to bring greater coherence to the regulatory process, and to in-crease the "robustness" of containments.
The ACRS explained "robustness" as follows: "A containment cleverly and narrowly designed to mitigate a set of accidents that has been precisely identified may not be able to cope with the unexpected.
A truly
'robust' containment would have improved capability to deal with the unexpected.
A containment that has been designed with ex-plicit consideration of a more extensive set of challenges is likely to be more robust than one designed with consideration of only a limited set."
OCRE agrees with this assessment complete-ly.
2.
The rule proposed herein does not provide a basis for revis-ing emergency planning standards.
OCRE presumes that the NRC is contemplating the elimination of offsite emergency planning, or shrinking the plume EPZ to the site boundary, or some equivalent
concept, as has been advanced by some in the nuclear industry.
The present emergency planning requirements should be retained and strengthened.
Emergency planning provides an important hedge against uncertainty.
Even with the additional requirements being
proposed, it is always possible that the containment will still fail.
We must have the emergency planning infrastructure in place if that happens.
Emergency planning also provides the last layer of defense in the event that plant safety features are deliberately defeated in acts of radiological sabotage or terror-ist attacks.
In addition, emergency planning yields significant collateral benefits to the community in that it is useful for a
spectrum of natural and manmade disasters, e.g, tornadoes, hurri-
canes, floods, chemical spills, nuclear attack, etc.
In fact, 5
many utilities use this as a selling point for emergency plan-ning.
Emergency planning ought to be in place for every communi-ty, regardless of whether it is host to a nuclear power plant.
3.
As noted above, OCRE believes that the containment should be designed to remain leak tight for the duration of the accident.
In the event that the NRC chooses to permit containment venting, it should only be permitted after one week from the onset of core damage.
As explained above, this is to allow sufficient time for fission product removal mechanisms to work, for recovery of the accident, and for offsite protective measures.
4.
The difficulties mentioned in this question are among the reasons OCRE does not favor Alternative 2.
In the event that the NRC adopts Alternative 2, OCRE recommends that it include a more detailed description of the phenomena and that acceptance crite-ria be codified.
5.
OCRE believes that future containments should have additional features and requirements.
Some of these have been addressed above, e.g., minimum design pressure and internal volume, prohi-bition of pressure-unseating hatches and airlocks, and require-ments for penetrations.
OCRE would also support the "supercontainment" concept.
That is, the containment must be designed to withstand any
accident, without consideration of probabilities.
If an accident or phe-nomenon is possible, the containment is designed to withstand it.
Germany is currently investigating supercontainments for future reactors.
Initial cost estimates suggest small impacts (less than 5%) on total reactor costs.
This is partially due to the fact that current German containments are conservatively designed to withstand extreme aircraft accidents.
See Forsberg and Reich, "Worldwide Advanced Nuclear Power Reactors with Passive and Inherent Safety:
What, Why, How, and Who," Oak Ridge National Laboratory, ORNL/TM-11907, September 1991.
OCRE believes that U.S. reactors should provide a level of pro-tection at least equivalent to that required in Germany and other countries.
The United States should be the world leader in protecting the public from reactor safety hazards, not the lag-gard.
OCRE supports the ACRS proposal that containments should against aircraft crashes, explosions, and other threats to the plant.
protect external The NRC should also require the use of a core catcher.
The 6
following characteristics appear promising (taken from Forsberg and Reich, supra, p. 59):
Under the reactor vessel, a portion of the concrete mat has a
specially controlled chemical composition.
The concrete con-tains a
mixture of different aggregates.
The aggregates are chosen so that when the various aggregates -cement, steel rebar, and core materials-melt, a waste glass that incorporates the core materials is created.
The glass contains one or more aggre-gates containing neutron poisons to prevent any possibility of a
criticality accident.
The glass chemical composition is chosen to have a very high affinity for volatile fission products.
The aggregates are chosen to minimize gas generation upon melting
and, hence, minimize aerosol formation.
The glass also has a
high surface tension to minimize aerosol generation.
The depth and width of the concrete mat with the special concrete aggregate is chosen to contain the reactor core.
A heat balance exists between radioactive decay heat and (1) heat needed to melt the concrete, and (2) heat conducted out or removed by other mechanisms from the molten core/concrete matrix.
Eventual-ly, heat conduction out of the waste matrix will exceed heat generation and the molten core/concrete matrix will begin to solidify.
The special aggregate concrete mat is sized to exceed the maximum volume of the molten core/concrete matrix, and the area is chosen to maximize cooling.
In particular, the top surface area is large enough to radiate sufficient decay heat so that it will cool and solidify the waste matrix over time, with-out meltthrough of the reactor basemat.
The concrete aggregate is a relatively low-melting aggregate (400 to 900 deg-C).
Low melting points are desirable for the following reasons:
(1).
A low melting waste matrix will quickly spread the molten core/concrete matrix over a wide area under the reactor.
This improves heat transfer and cools the matrix to quickly form a
solid.
(2).
A low melting waste matrix minimizes gas and aerosol gener-ation by two mechanisms.
First, the rate of release of semivola-tile radioactive gases is temperature dependent.
Lower tempera-tures imply less gas.
Second, the rate of release of semivola-tile radioactive gases is dependent on the concentration of those materials in the waste matrix.
Diluting the core material re-duces the fractional releases of radioactive materials.
7
6.
OCRE does not believe that the likelihood of accident scena-rios or phenomena should play a role in determining whether the containment should be designed to withstand them.
Our notions of "likelihood" are too uncertain, incomplete, and subject to manip-ulation to be sufficiently accurate for regulatory purposes.
OCRE agrees with the ACRS' assessment: "It is because quantita-tive risk estimates are not perfect that defense in depth is a
useful philosophy, and that separate containment performance guidelines make sense."
As stated above, OCRE believes that the challenges do need to be specified in more detail and with greater accuracy and complete-ness in the rule.
OCRE supports the incorporation of the de-scriptions of challenges and severe accident phenomena contained in the May 17, 1991 ACRS letter.
7.
OCRE sees no reason why the criteria proposed in the rule should not be fully applicable to passive LWRs.
9.
A reactor design should be required to have a reactor cavity design and/or a
reactor vessel support structure capable of mitigating and accommodating a high pressure melt ejection even if the design includes the capability to rapidly depressurize the primary system.
Why?
Because the depressurization system may fail or be deliberately defeated in an act of radiological sabo-tage.
10.
OCRE agrees with the ACRS that there should be a provision for on-line monitoring of containment isolation status.
However, if such a system is only capable of detecting "gross leakage" then it is not a sufficient replacement for the leak rate testing requirements of Appendix J.
11.
OCRE agrees with the ACRS that containments should signed to provide for ease of emergency closure during operation including station blackout conditions.
be de-shutdown
12.
Equipment provided only for severe accident prevention or mitigation should be subject to the same requirements as design basis equipment.
It hardly makes sense to require such equipment only to have it fail when needed because it was not subject to a
nuclear quality assurance program or was not environmentally qualified to withstand the very conditions in which it is re-quired to function.
Since severe accidents pose a greater risk to the public than design basis accidents, it makes little sense to have more stringent requirements for those accidents posing the lesser risk and less stringent requirements for those acci-dents posing the greater risk.
8
13.
OCRE believes that the ASME service level C stress limits for steel containments are appropriate for severe accident condi-tions.
Use of the level C limits is necessary to provide suffi-cient conservatism to account for all the uncertainties inherent in calculating such stresses.
For an excellent discussion of these uncertainties, see "On the Uncertainties Associated with Containment Analysis," Griemann and Fanous, NUREG/CP-0056, pp. 229-244.
15.
The codification of a containment performance objective does not provide a basis for the elimination of further review of SAMDAs for future LWRs under Part 51.
Consideration of SAMDAs is necessary because our knowledge of severe accident risks and phenomena will undoubtedly increase with the passage of time, and design innovations will undoubtedly be developed.
For
example, just 10 years shutdown risk was thought to be virtually nonexistent.
The SAMDA review process is essential for the rational evaluation of risk and mitigation measures that may be unique to a plant design.
Respectfully submitted, Susan L. Hiatt Director, OCRE 8275 Munson Road Mentor, OH 44060-2406 (216) 255-3158 9

Entergy Operations December 16, 1992 0CAN129203 Secretary of the Commission Attn: Docketing and Service Branch U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:
Proposed Rulemaking on Acceptability Entergy Operations, Inc.
Route 3. Box 137G Russellville. AR 72801 Tel 501-964-3100
92 DEC 21 P4 :J4 (j) of Plant Performance for Severe Accidents Gentlemen:
On September 28, 1992, the NRC published an advance notice of proposed rulemaking in the Federal Register (50FR57188) concerning acceptability of plant performance for severe accidents for future light water reactors. The NRC requested comments on the proposed regulations.
Please find our following two comments:
Specific Considerations, Item 5: Asks about future LWR design features beyond those described in Alternative 1. One of the aspects of Alternative 1 deals with controlled elevated venting (if provided in the design) after the initial 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.
Comment:
While there is no requirement for Containment Venting capability in this proposed rule, this feature should be encouraged due to its importance as a strategy in the severe accident environment.
JAN 2 1 1993 d
Acknowledged by car.............................
U.S. NRC December 16, 1992 0CAN129203 Page 2 Specific Considerations, Item 7: Asks for what reason would any of the criteria proposed in the three alternatives not be fully applicable.
Comment:
RCS depressurization capability. While the capability to rapidly depressurize the RCS may be applicable in some cases, it may not be the best option for all scenarios. MAAP runs have shown that RCS depressurization can cause earlier containment failure and at the same time not provide any benefit to accident mitigation.
Very truly yours, J:::l.1~
James J. Fisicaro Director, Licensing JJF/NBM/sjf cc:
U. S Nuclear Regulatory Commission Document Control Desk Mail Station Pl-137 Washington, DC 20555
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USN;~C December 24, 1992 Mr. Samuel Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Chilk:
92 ore 29 A10 :22 This letter provides Department of Energy (DOE) comments on the Nuclear Regulatory Commission's (NRC) advance notice of proposed rulemaking entitled, "Acceptability cf Plant Performance for Severe Accidents; Scope and Consideration in Safety Regulations." The proposed rulemaking has the potential for impact on the review schedules for design certification applications currently before the Commission, which DOE is sponsoring in cooperation with the nuclear industry.
We believe that the primary issue is whether generic rulemaking should be pursued at this time or whether the design-specific rulemakings that are currently scheduled should first be completed.
We recommend that the difficult issues related to severe accidents first be resolved in the context of specific designs through certification rulemaking.
If necessary, the certification rulemaking could be followed by a generic rulemaking.
The advanced status of the designs submitted for certification obviates one of the major benefits of generic rules, which is to provide uniform and consistent guidance to designers by identifying the requirements that must be met during the safety review.
The passive Advanced Light Water Reactors (ALWRs) have been designed and Standard Safety Analysis Reports have already been submitted to the Commission for review and approval.
Further, these designs are in accordance with the generic severe accident approaches incorporated in the ALWR requirements document.
Since there are no other U.S. ALWR designs anticipated for some time, the generic rulemaking can be conducted at a later date and still achieve this benefit of generic rulemaking for any future designs.
The lessons learned from the design approval process for current ALWRs can be incorporated in a future generic rulemaking.
In addition to being not necessary to the review of the two passive ALWR designs currently before the NRC, it is probable that the generic rulemaking would adversely impact the certification schedules for these plants.
Based on prior rulemaking experience, it is likely that the schedule for a generic rulemaking, which needs to address issues and situations that would be applicable for all future ALWR applications, would exceed the current plant-specific certification schedules. Thus, the likely outcome of pursuing a generic rulemaking at this time will be to delay the time when certified passive plants are available as an option in the United States.
JAN 21 1993 Acknowledged by card..................................
For these reasons, we request the Commission delay proceeding with generic rulemaking at this time and consider the need for a generic rule after the experience gained through the passive plant reviews is obtained.
Sincerely, E. C. Brolin, Acting Principal Deputy Assistant Secretary for Nuclear Energy 2
AL WrJ.
LIGHT WATER REACTOR 23 December 1992 Mr. Samuel J. Chilk, Secretary Office of the Secretary of the Commission U.S. Nuclear Regulatory Commission Mail Stop 16 G15 Washington, DC 20555 (i)
Subject:
Advanced Notice of Proposed Rulemaking (ANPR),
"Acceptability of Plant Performance for Severe Accidents"; Scope of Consideration in Safety Regulations" (57 Federal Register 44513 of September 28, 1992)
Dear Mr. Chilk:
The ALWR Utility Steering Committee assisted by the Electric Power Research Institute (EPRI) has reviewed the ANPR, "Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations," and has provided its input to NUMARC for incorporation in their letter forwarding overall industry comments. We agree with and support the NUMARC letter.
In addition, we wish to identify specific concerns regarding the ANPR that relate to the role of the ALWR Utility Requirements Document.
EPRI and NUMARC both commented in 9 January 1989 letters to the NRC on staff proposals to initiate severe accident rulemaking for advanced reactors.
The Utility Steering Committee (USC) reiterated our concerns in a letter to the Chairman on 4 May 1990. Key points included the following:
Rulemaking is neither necessary nor desirable to implement the NRC's Severe Accident Policy Statement.
The ALWR program is committed to meeting the letter and intent of the Severe Accident Policy.
JAN 2 1 1993 Acknowledged by card..................................
EPRI ALWR Utility Steering Committee 3412 Hillview Avenue, Palo Alto, California 94304
Telefax: (415) 812-2874
Mr. Samuel J. Chilk 21 December 1992 Page2 The ALWR Utility Requirements Document (URD) is the appropriate mechanism for translating industry and NRC policy into a specific set of baseline design requirements for ALWRs.
The ALWR URD with its attendant NRC Safety Evaluation Report followed by design certification will achieve the necessary resolution and codification of severe accident issues for future designs.
The steps taken in the Requirements Document toward reducing the frequency and severity of core damaging accidents will reduce the probability of accident measures being needed.
Nevertheless, all the severe accident issues and approaches to resolution were already identified and addressed by the Evolutionary Plant URD at the time generic rulemaking was first proposed.
No significant new severe accident issues have emerged, and no major changes have occurred in our utility requirements on how these severe accident issues should be addressed in the designs.
From what we have learned to date, no significant changes in the design of either the ABWR or System 80+ have been made and we do not expect any significant changes in Passive Plant requirements or designs.
Specifically, the "Purpose of the Rule" as stated in the ANPR can be achieved without an intervening generic Rule as follows:
"Codify the Commission's guidance on severe accident and containment issues that resulted from the review of advanced light water reactors"
{ achieved via the Design Certification rulemaking}
"Provide assurance that the performance of future L WRs under severe accident conditions is consistent with assumptions about severe accident conditions performance used in developing new source term information" {achieved via approval of the URD and the FDA process}
"Provide guidance to future LWR designers and potential applicants"
{ this is done via NRC input to and industry conformance to the URD}
"Add consistency and standardization to the resolution of severe accident issues based on the current technical information" {this is done via the conformance to the URD}
"Facilitate design certification rulemakings" {URD approval already accomplishes this; generic rulemaking will conclude too late to affect}
Intervening severe accident rulemaking cannot facilitate DC rulemaking for evolutionary plants because DC rulemaking will likely be completed well
Mr. Samuel J. Chilk 21 December 1992 Page3 before the generic rulemaking process could conclude.
Intervening rulemaking for passive plants will not be needed once agreement and approval of the Passive Plant URD are obtained.
Intervening severe accident rulemaking (already in development for over five years) will become a major source of unnecessary schedule delay in the ALWR design and approval process. This continuing regulatory instability would further erode investor confidence in the nuclear option and place a large drain on the already strained NRC resources.
We understand that the Commission gave tentative approval for proceeding with generic rulemaking "where appropriate" in its SRM on SECY-92-262. We believe the policy and schedule implications of the current path are so severe that a Commission review of the appropriateness of this path is needed.
Sincerely,
~~
E. E. Kintner, Chairman ALWR Utility Steering Committee 576L/TUM/ cdl c:
Chairman I van Selin Commissioner Kenneth C. Rogers Commissioner James R. Curtiss Commissioner Forrest J. Remick Commissioner E. Gail de Planque James M. Taylor, Executive Director of Operations, NRC Thomas E. Murley, Director, Nuclear Reactor Regulation, NRC Docketing and Service Branch, NRC John J. Taylor, Vice President, EPRI Joe F. Colvin, President and CEO, NUMARC
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1776 Eye Street, NW
Suite 300
Washington, DC 20006-2496 *02* Q[r 2B l :1 J (202) 872-1280 WIiiiam H. Rosin Vice President & Director Technical Division Mr. Samuel J. Chilk Secretary, Office of the Secretary of the Commission U.S. Nuclear Regulatory Commission Mail Stop 16 G15 Washington, DC 20555 December 22, 1992
SUBJECT:
Advance Notice of Proposed Rulemaking (ANPR), "Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations" (57 Federal Register 44513 of September 28, 1992)
Dear Mr. Chilk:
The Nuclear Management and Resources Council (NUMARC)1, on behalf of the nuclear power industry, has reviewed the ANPR, "Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations," and offers the following comments for consideration.
The industry has long been aware of NRC consideration of the need for generic rulemaking for advanced light water reactors (ALWRs) to supplement specific design certification rulemakings. In a January 9, 1989, letter to NRC Executive Director for Operations, Victor Stello, NUMARC provided its perspective, which we have maintained consistently since, that generic Part 50 rulemaking to address severe accidents, such as that contemplated by the subject ANPR, is not necessary and, in fact, may be counter-productive. Instead, the industry has supported generic resolution of severe accident 1NUMARC is the organization of the nuclear power industry that coordinates the combined efforts of all utilities licensed by the NRC to construct or operate a nuclear power plant, and of other nuclear industry organizations, in all matters involving generic regulatory policy and on the regulatory aspects of generic operational and technical issues that affect the nuclear power industry. Every utility responsible for constructing and operating a commercial nuclear facility is a member of NUMARC. In addition, NUMARC's members include major architect-engineering firms and all the major steam supply vendors.
JAN 2 1 1993 Acknowledged by card..................................
11.s. NUCL[J!.;i R[GL1t;-,*,2:w COMMlSSior~
OOCKET'.NG f., ~EfatV~Cf: S~CTION O::TICE r::.:: *r -i:: s:-=.cr.a,'\RY OF Tti~ cc-~~\1:S~)!ON
Mr. Samuel J. Chl1k December 22, 1992 Page 2 issues to the extent possible via the ALWR Utility Requirements Document (URD),
followed by review and codification of the design specific implementation as part of the design certification process. This position was explained to the Commission and NRC staff on several occasions, including our January 27, 1992, letter to the Office of the Secretary commenting on SECY-91-262, Resolution of Selected Technical and Severe Accident Issues for Evolutionary Light Water Reactor Designs. We continue to feel strongly in this regard, and this* letter and the enclosed responses to the fifteen ANPR questions reiterate our position and provide the basis for our conclusion.
We believe the URD/design certification process is the superior approach for the resolution of severe accident issues for AL WRs, both evolutionary and passive. A full discussion of this position is delineated in response to ANPR Question Number 1 contained in the enclosure, the key points of which are as follows:
severe accident issue resolution via the URD and design certifications is consistent with the approach embodied by the Commission's Severe Accident Policy Statement and NRC Generic Letter 88-20, Individual Plant Examinations for Severe Accident Vulnerabilities; the URD, together with design certification reviews and rulemakings, provide the optimal approach for comprehensive, integrated and design specific resolution of all safety issues, including severe accidents, associated with individual standard plant designs; severe accident resolution accomplished through the URD and design certification will avoid uncertainty associated with the impact of a "competing" generic rulemaking, thus promoting predictability and stability in the Part 52 licensing process; and there is minimal value added by one or more generic rulemakings that produces severe accident resolution "essentially the same" as that which will be accomplished via design certification.
A separate generic rulemaking on technical issues associated with AL WR designs, such as severe accidents, is not warranted. Rather than facilitating the design certification process, as indicated in the ANPR, generic severe accident rulemaking would more likely complicate the process and unnecessarily consume significant NRC and industry resources. Further, this approach would be inconsistent with the initiatives of the NRC and the industry to identify unnecessary, contradictory and/ or duplicative regulations and guidance.
Mr. SamuelJ. Chilk December 22, 1992 Page 3 We appreciate your careful consideration of these comments and urge the Commission to direct that severe accident issues be addressed via the URD / design certification process rather than move ahead with a proposed severe accident rule for future plants.
WHR\RJB\ljw Enclosure cc:
Chairman Ivan Selin
' Commissioner Kenneth C. Rogers Commissioner James R Curtiss Commissioner Forrest J. Remick Commissioner E. Gail de Planque Sincerely, jbf)Z~-l~
William H. Rasin James M. Taylor, Executive Director of Operations, NRC Eric S. Beckjord, Director, Office of Nuclear Reactor Research Thomas E. Murley, Director, Nuclear Reactor Regulation, NRC Docketing and Service Branch R~nses to ANPR Questions ANPR Ouestion Nnmher 1 Is ruleroaking addressing severe accident plant performance criteria desirable? If not, why? Would such a rule provide better coherence and predictability to the design certification review and certification processes for future reactor designs or is ruleroaking on these issues via individual design certification sufficient?
lndustty Response We believe that a generic Part 50 rulemaking to address ALWR technical and severe accident issues would riot be appropriate and is neither necessary nor desirable.
A substantial amount of severe accident research has been accomplished by the NRC as well as the industry since the Three Mile Island Unit 2 (TMI-2) accident in 1979. Indeed, the industry has incorporated into the ALWR development the lessons learned since TMI-2 through development by EPRI of the ALWR Utility Requirements Document (URD) and vendor development of individual ALWR designs, including design specific PRAs. AL WR severe accident issues are presently being addressed by the industry and the NRC staff via complementary URD and design certification interactions. Utility requirements have been developed for both evolutionary and passive AL WRs, including explicit consideration of a wide spectrum of severe accident phenomena and potential challenges to containment [l, 2]. These challenges encompass containment bypass scenarios, random system and equipment failures which could lead to breach of the containment boundary independent of any severe accident conditions, and potential phenomena that could challenge the structural integrity of the containment as a result of a core damage accident. The challenges considered include not only those referenced in the first two alternatives of the ANPR, but also the eight groups of challenges proposed by the ACRS for inclusion in the General Design Criteria [3].
References 1 and 2 identify the extensive list of challenges considered in the evolutionary and passive AL WRs and contain a summary of the design features available in each plant to limit the potential for or accommodate each of these challenges. Design specific PRAs are being performed which evaluate the potential for significant challenges given these design features and address plant response to those challenges that dominate risk.
Individual AL WR designs that implement the utility requirements and reflect the insights of a design specific PRA are being reviewed by the NRC staff and will become codified via design certification rulemakings. In sum, the industry and the NRC staff have pursued a course emphasizing generic resolution of severe accident issues to the extent possible via the AL WR URD followed by review and codification of the design specific implementation as part of the design certification process.
1
Consistent with this approach, the NRC staff argued persuasively in SECY-91-262 that individual design certification rulemak:ings provide the most efficient and effective mechanism for codifying the resolution of technical and severe accident issues. As noted in SECY-91-262, the design certification process provides for a thorough NRC staff, ACRS and Commission review of AL WR design requirements and 'criteria leading to their codification via the design certification rulemak:ings. Moreover, addressing severe accident issues in the context of the URD as well as within the overall design certification process provides greater confidence that these complex and interdependent technical issu~ will be coherently resolved. We concur in the essential tprust of SECY-91-262 that, considering the advanced state of the ALWR designs and associated reviews, the diversity among the standard plant designs, and the potential for delaying the design certification process, generic rulemak:ing to address AL WR technical and severe accident issues is not necessary or desirable.
Not only is resolution of severe accident issues via design certification amenable to evolutionary designs, as recommended in SECY-91-262, we believe severe accident resolution via design certification is the appropriate, superior and most technically coherent course for passive designs as well Indeed, an approach for passive plants that instead relies upon generic rulemak:ing would be contrary to the rationale behind the NRC staff issuance of Generic Letter 88-20, Individual Plant Examinations for Severe Accident Vulnerabilities, which recognized that severe accident challenges and resolutions are somewhat design specific. A generic rulemaking on severe accidents, especially one based principally upon evolutionary plant designs as suggested by the ANPR, may not address the particular performance characteristics of passive plant designs or allow for the realization of the full regulatory benefit in their regard. As a result, a generic rulemaking would likely not obviate the need to scrutinize severe accident issues in detail as part of a passive plant design certification. Therefore, we conclude that generic technical issues, such as severe accidents, can be most efficiently and coherently resolved for all ALWRs, including passive designs, via the URD and design certification reviews and rulemak:ings.
Furthermore, we are particularly concerned that promulgating a rule based upon this ANPR would likely generate uncertainty and could significantly disrupt the design certification process for both evolutionary and passive designs. Because evolutionary plant resolution of severe accident issues reflected in Part 52 design certifications and the rule envisioned by this ANPR are expected to be "essentially the same," there would seem to be minimal value added by such a generic rule for evolutionary designs. Indeed, we are concerned that issues relating to the consistency and applicability of the generic severe accident rule may be raised at a critical time in design certification proceedings with the potential for causing costly and unwarranted disruption of the certification process.
2
Regarding passive designs, parallel design certification reviews and severe accident generic rulemaking would add another dimension of uncertainty to an already complex undertaking. It is not clear what benefit a generic severe accident rulemaking based on evolutionary designs could add to the comprehensive and integrated resolution of all technical and severe accident issues via individual design certification rulemakings.
The ANPR essentially acknowledges this by noting that, "as detailed design information becomes available and review of the passive systems is completed, further rulemaking may be necessary." Presumably, the net result of such a series of generic rulemakings would be that generic resolution of severe accident issues for passive designs would be "essentially the same" as that reflected in passive plant design certifications. As in the case of the evolutionary designs, there would seem to be no value added by a largely duplicative generic rulemaking It is our yi¢w that the potential benefit of such generic ru1emaking(s) to a passive plant certification proceeding is outweighed by the greater likelihood that a rule to resolve expressed concerns would not be timely and could
generate uncertainty, rather than clarity, in the design certification process.
In the deliberations leading to publication of the Severe Accident Policy Statement, the Commission considered whether it was necessary to amend the NRC regulations. The Commission concluded that a statement of policy was more appropriate as both NRC and industry studies demonstrated that operating plants posed no undue risk to public safety and health. ALWR designs satisfy the Commission's stated expectation that future plants achieve a higher standard of severe accident safety performance. This enhanced performance is being achieved, in part, by integrating severe accident considerations throughout the utility requirements and design processes, thus providing for enhanced severe accident prevention and mitigation. While addressing severe accidents for future plants is appropriate and consistent with the Commission's Severe Accident Policy, inclusion of criteria for future plants related to severe accidents in the Commission's generic Part 50 regulations would be inconsistent with the Commission's previous conclusion that generic rulemaking was not necessary because current operating plants do not pose an undue level of risk to public health and safety. Because advanced plants achieve a higher standard of severe accident safety than the current operating plants, we would expect the Commission's previous conclusion regarding generic rulemaking to continue to be appropriate. Analogous to the plant specific approach to severe accident issue resolution for operating plants via Individual Plant Examinations, an integrated, design specific treatment of severe accident issues within design certification reviews and rulemakings is the best approach for future plants.
Finally, the industry and NRC staff are presently proceeding in efforts to take a hard look at eliminating or revising unnecessary, duplicative, or potentially confusing regulations and regulatory practices. The generic rulemaking envisioned by this ANPR would be inconsistent with that process in that it would promulgate duplicative severe accident regulation on top of comprehensive design *certification rulemakings, potentially introducing technical inconsistencies, schedular disruptions and licensing uncertainty. In addition to being the most practical and efficient approach to AL WR severe accident 3
issue resolution, the URD / design certification process will provide for severe accident issue resolution that is clear, tailored and codified for each standard design, and reflective -of integrated and coherent consideration by plant designers and NRC staff reviewers alike, of the full spectrum of safety issues associated with ALWR designs.
[1]
"Matrix Approach to Evolutionary Plant Containment Performance,"
E.E. Kintner letter to T.E. Murley, May 9, 1991.
[~]
DOE/ID-10291, "Passive ALWR Requirements to Prevent Containment Failure," December 1991.
[3]
"Proposed Criteria to Accommodate Severe Accidents in Containment Design," ACRS letter dated May 17, 1991.
ANPR Question Number 2; Would a new rule in 10 CFR Part 50 concerning plant performance for severe accidents, as discussed in the three alternatives, provide a basis for revising the
requirements on Emergency Planning Zones for future L WRs? H so, why? H not, why not?
Industn Remo;q~
Plant design to enhance performance under potential accident conditions and the improved state of knowledge regarding accident consequences provides an appropriate basis for reevaluating the requirements on Emergency Planning Zones and offsite emergency response programs. The traditional requirements should be reconsidered for AL WRs due to the enhanced plant design, including containment performance and other relevant technical criteria in the Utility Requirements Document Reconsideration of offsite. emergency planning requirements is also appropriate due to developments regarding improved understanding of, and anticipated changes to, the source term for AL WR designs. As emphasized throughout the industry response to this ANPR, the industry does not believe a generic severe accident rulemaldng is necessary. However, a generic rulemaking may be necessary to accomplish AL WR emergency planning and is being considered by the industry subject to finalizing the technical basis through review and issuance of the FSER on the passive plant URD.
ANPR Question Number 3; One option for an overall containment performance criterion that has been considered is that the conditional failure probability of the containment should be less than approximately one in ten. Two of the alternatives use a deterministic surrogate that states that the containment should remain leak tight for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage and after that time, remain a barrier against 4
the uncontrolled release of radioactivity when faced with challenges from the more likely severe accident phenomena. Is this criterion a suitable substitute for the conditional containment failure probability of one in ten? H so, explain why. Hnot, explain why not Is a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months an appropriate time frame? Is its degree of conservatism appropriate considering uncertainties and defense-in-depth? H not, what alternative would be appropriate? What other criteria [probabilistic or deterministic]
might be considered?
lndusta Remonse; Deterministic containment performance criteria are a suitable substitute to a quantitative conditional containment failure probability for demonstrating containment performance for AL WR designs.
The underlying purpose of containment performance criteria is to provide defense-in-depth in the design of containment and containment systems by assuring a balance between, severe accident prevention and mitigation. For a plant with highly reliable core damage prevention systems, it may be possible to show consistency with the safety goals simply by assuring an extremely low potential for core damage. The quantitative health objective, the large release guideline, and the mean core damage frequency guideline could all be achieved without the need for a containment. The intent of a containment performance criterion, either probabilistic or deterministic, is to assure that no matter how low the potential for core damage, there still will be a containment function that is available under postulated severe accident conditions. As provided for by the Commis.sion SRM in response to SECY-89-102, the industry has pursued a deterministic approach to demonstrating AL WR containment performance.
Indeed, the URD establishes design features and criteria which accomplish the containment performance objective of this ANPR, as well as that suggested by the NRC staff in SECY-90-016, thereby assuring adequate severe accident containment response for AL WRs. AL WR plant designs codified at the conclusion of the URD / design certification process will have reliable containments and containment systems capable of preventing or accommodating a wide spectrum of severe accident challenges as measured against the deterministic containment performance criteria of the URD. This deterministic approach to evaluating containment performance meets the intent of a quantitative conditional containment failure probability for demonstrating AL WR containment performance.
The time frame for containment remaining leak tight after the onset of core damage should be long enough to: (1) allow for fission product radioactive decay and aerosol settling so that if a release were to occur thereafter, the site boundary dose would be below the threshold for acute health effects when analyzed realistically; and (2) allow time for protective action as part of AL WR emergency planning. The URD establishes design features and containment performance criteria based on a 24-hour leak tight period to provide reasonable assurance that each of these objectives is 5
accomplished with significant margin. With these design features and using the best estimate methodology of the URD, AL WR containments can be shown to remain leak tight for at least 24
h~urs following core damage. In actuality, a period of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months or even less, depending upon the specific plant design and site, provides ample time for meeting both of the above objectives based on the half life of noble gases, the effectiveness of fission product mitigation systems and natural removal processes, and historical emergency evacuation experience, including ad hoc evacuations. Thus, the approximately 24-hour deterministic containment performance criterion proposed by the NRC staff in SECY-90-016 and this ANPR, in the absence of a common understanding regarding realistic evaluation methods, may be overly conservative.
This conservatism is acknowledged in Question 15 of this ANPR which notes that the 24-hour containment barrier criterion provides a level of safety that is three orders of magnitude more conservative than the ijiliiiititative health objective of the Safety Goal Policy statement While 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is an appropriately challenging URD requirement to ensure that AL WR designs contain features that provide a high level of severe accident protection, codifying this deterministic criterion would be inappropriate because doing so would establish a de facto higher safety goal. That result would be contrary to Commission guidance provided in response to SECY-89-102 concerning NRC staff imposition of industry design objectives as regulatory requirements.
Based on the preceding discussion, the criterion for protection against uncontrolled fission product release after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is similarly conservative and is inappropriate to codify. As noted above, it is estimated that a controlled release that would pose no undue risk to the public could occur significantly prior to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following core damage. Controlled release from the suppression pool airspace in the form of an overpressure protection device is being considered for the advanced BWRs.
Overpressure protection may also be provided by demonstrating that the size and strength of the containment allows meeting appropriate ASME limits for approximately three days, which provides adequate time for actions to bring the accident under control This latter method of protection from uncontrolled release is being implemented for large PWR containments.
In summary, design features for overpressure protection are specified by the URD and will be codified for individual AL WRs via the design certification rulemakings_
These features provide the basis for assuring substantial time for fission product removal and emergency response prior to any significant release, and their codification via design certification eliminates the need for generic severe accident rulemakiog The industry concurs in the appropriateness of deterministic performance criteria, and as discussed above, severe accident resolution for AL WRs achieved via the URD and individual design certification reviews and rulemakioes will reflect use of these criteria for demonstrating AL WR containment performance.
6
ANPR Question Npmher 4; Alternative 2 would require extensive reliance on analytical tools that calculate the effects of severe accident phenomena. Are there analytical tools that are sufficiently developed and adequate to allow effective implementation of such a phenomena-based rule? H so, what are they, and for what phenomena could they be used? How would Alternative 2 be implemented? For example, should the codes and input parameters be approved by NRC? Should acceptance criteria be codified or put in a regulatory guide?
Industcy Response; Specific response is not provided to this question relative to the implementation of Alternative 2 on the basis of the industry's overall position embodied by the response to ANPR Question 1 that generic severe accident rulemaking for AL WRs is neither necessary nor appropriate.
With respect to the URD / design certification process for implementing AL WR severe accident resolution, a combination of quantitative analyses and qualitative considerations has been used. The appropriateness of the implementation and criteria for a given AL WR will be indicated in the FSER for the design and codified via the certification rulemaking.
ANPR Question Npmher S:
Should future L WR containment designs include features beyond those described in Alternative 1 to prevent/mitigate severe accidents? If so, what are they?
Industn: Resoonse:
Specific response is not provided to this question relative to the implementation of Alternative 1 on the basis of the industry's overall position embodied by the response to ANPR Question 1 that generic sev~re accident rulemaking for AL WRs is neither necessary nor appropriate.
As discussed in response to Question 1, AL WR designs contain features required by the URD intended to limit the potential or accommodate a wide spectrum of severe accident phenomena and containment challenges. These features include those identified in Alternative 1 of the ANPR and, together with specific insights from the design PRA, provide the basis for severe accident issue resolution for AL WRs via the URD / design certification process.
7
ANPR Question Number 6; Alternatives 2 and 3 specify phenomenological severe accident challenges that should be considered in the design. Alternative 1 is based upon the same phenomena/ challenges. Are there other severe accident phenomena/ challenges that should be considered? Should the challenges be specified in more detail (for example, specifying the amount of hydrogen generation) or is a general statement of the challenge more desirable?
Industa Remonse; Specific response is not provided to this question relative to the severe accident phenomena/ challenges considered by Alternatives 1, 2 and 3 on the basis of the industry's overall position embodied by the response to ANPR Question 1 that generic severe accident ru.Jernaking for AL WRs is neither necessary nor appropriate.
As discussed in response to Question 1, the URD identifies an extensive list of AL WR severe accident phenomena/ containment challenges. These include containment bypass scenarios~ random system and equipment failures, and phenomena with the potential to challenge containment integrity. For each AL WR design, features are provided to limit the potential or accommodate each of the phenomena/ challenges.
These features, together with specific insights from the design PRA, provide the basis for severe accident issue resolution for ALWRs via the URD/design certification process.
ANPR Question Number 7; For what reason (e.g., not a risk significant phenomena, not a cost effective solution) would any of the criteria proposed in the three alternatives not be fully applicable to passive designed L WRs?
Industo: Remonse:
Specific response is not provided to this question relative to the implementation of Alternatives 1, 2 and 3 for passive plants on the basis of the industry's overall position embodied by the response to ANPR Question 1 that generic severe accident rulernaking for AL WRs is neither necessary nor appropriate.
As discussed in response to ANPR Question 1, severe accident issue resolution for passive plants is being coherently addressed via the passive plant URD and passive plant design certifications. This approach thus avoids the potential problem of incongruities between passive designs and a generic severe accident rule based on evolutionary designs.
8
ANPR Question Number 8; What features could an advanced L WR design include that would prevent or mitigate fuel-coolant interactions?
Industa Response; Fuel-coolant interactions (FCI) can be postulated to occur either in-vessel or ex-vessel. The potential for in-vessel FCis inducing containment failure (alpha-mode failure mechanism) was identified in the WASH-1400 Reactor Safety Study. ALWRs have depressurization system capabilities which could result in low enough reactor coolant system pressures such that energetic FCis could not be excluded based on system pressure. The reactor pressure vessel, and internals, head closure, and missile barriers
included in the AL WR design are similar to those employed in existing L WR plants.
Thus, the same mechanisms are operable in AL WR designs that led to past conclusions that the occurrence of an energetic FCI which could lead to alpha-mode containment failure is very remote in probability.
In the unlikely event of reactor pressure vessel lower head failure, ex-vessel FCls can result in both significant steam generation rates and shock waves induced by ex-vessel explosive interactions. In the PWR designs, any shock waves which occur will be within the thick walled reactor cavity, and the containment boundary integrity would not be challenged. Additionally, the thick walled reactor cavity provides a barrier against the generation of missiles which could impact the containment boundary and challenge its integrity. The lower drywell boundary provides the same type of barrier in evolutionary BWR designs.
In the SBWR, the lower drywell walls form the containment boundary. However, a thick walled shield is being provided to protect the lower drywell portion of the containment boundary from the effects of severe accidents. Thus, containment boundary integrity would not be challenged by shock waves or missiles.
For the AP600, the capability has been provided to externally flood the reactor pressure vessel and attached piping such that ex-vessel heat removal can be established.
This feature could protect the vessel lower head and prevent its failure. Maintenance of vessel integrity serves as a preventive means of precluding ex-vessel FCis since it avoids the discharge of molten core debris.
The above AL WR capabilities and features, which are being incorporated in the advanced L WR designs via the URD and individual design certification reviews and ruleroaki:ngs, provide adequate severe accident protection relative to both prevention and mitigation of FCis.
9
ANPR Questlon Nnmher 9;
H a design includes the capability to rapidly depressurize the primary system, should it also be required to have a reactor cavity design and/ or a reactor vessel support structure capable of mitigating and accommodating a high pressure melt ejection?
Industcy Response:
The capability to reliably depressurize the primary system provides sufficient means for limiting the potential for high pressure melt ejection (HPME) events.
AL WR designs include reliable means of primary system depressurization such that additional mechanisms and features, to address HPME are not required.
However, it should be noted that reactor cavity/support design has been extensively considered relative to assuring ex-vessel fuel/debris cooling under low pressure core melt scenarios. The resulting design of these features is also expected to limit the transport of fragmented core debris outside the reactor cavity region in the unlikely event of HPME.
ANPR Question Number 10; Should future L WR designs include an on-line instrumentation system that monitors containment atmosphere for gross leakage to reduce the risk from an inadvertent bypass of containment function? Would application of this system be sufficient basis to modify leak rate testing requirements under ~0 CFR 50, Appendix J?
Industa Response; The Utility Requirements Document for advanced LWRs for both evolutionary and passive designs requires that a means be provided to enable the operator to perform a periodic check for gross leakage of containment atmosphere during normal operation.
The intent is to provide the owner/operator of the plant with added assurance that penetrations connected to both the containment atmosphere and the environment are not inadvertently left open during normal operation. It was not the intent that such a system be maintained on-line continuously; instead, it would be used as a periodic check, e.g., following maintenance operations which involve opening of containment penetrations.
It is not practical to use such a system to modify the leakage rate testing requirements in 10 CFR 50, Appendix J. Testing conducted at the containment pressure which exists during normal operation would likely not be sufficiently accurate for detection of design basis leakage rates such as are detected by testing in accordance with Appendix J. Also, testing during normal operation would identify leakage in penetrations connecting the containment atmosphere to the environment, but not in 10
other penetrations which are covered by Appendix J testing ( e.g., penetrations of lines connected to the reactor coolant system). Thus, a system for periodic checking of gross conuµnment leakage during normal operation is a requirement of the URD. However, it is not practical or neces.sary to use this system to modify Appendix J leak rate testing requirements.
ANPR Question Number 11; What design criteria should be developed that provide asmrance that the containment's integrity could easily be established during certain shutdown conditions?
Industa Res;uonse; Like severe accident issues associated with power operation, shutdown risk issues are being appropriately addressed for AL WRs via the URD / design certification process descn"bed in response to Question Number 1.
The risk associated with shutdown operations is dominated by a few key plant configurations and operator actions initiating and responding to events occurring in the plant during shutdown conditions. These conclusions are variously reflected in NUREG-1449 [l], NUREG-1410 [2], the IAEA conference report on shutdown accident sequence modeling [3] and industry initiated studies associated with shutdown risk [4, 5, 6].
AL WR designs reflect past operating experience to minimize the* risks associated with the events identified in the referenced documents. The need to isolate containment during an event which might occur during shutdown conditions is dependent on the risk associated with the postulated event.
For example, for transient initiators which occur with the refueling cavity full, many hours are available prior to the beginning of bulk boiling even at the beginning of an outage, and days are required in order to boil off the inventory to the point that fuel damage would occur. This provides substantial time for operator actions to recover lost systems or mitigate the depletion by providing makeup from external sources. If the containment is open, the only risk during this period is that associated with the releases fro:qi the coolant activity. Such releases would be minima] (a small fraction of 10 CPR 100 limits), and therefore offsite consequences from this type of an event are expected to be very low. The URD specifies that the reactor designer demonstrate that the consequences of pool boiling with containment open are acceptable.
For those configurations when reactor inventory is low, such as head removal, the time available for recovery is short. Managing risk during these relatively brief and infrequent periods is best accomplished by plant system configuration control and by providing the operator with the capability to mitigate any loss. of inventory. Again, the 11
URD specifies that sufficient water inventory be available to provide makeup for losses of reactor coolant system inventory.
Because the risk associated with shutdown operations is very low and because residual risk is best addressed through configuration control and procedures, design features intended to limit releases by isolating containment are of limited incremental benefit and have not been established as design criteria. For purposes of establishing additional margin, the ALWR URD specifies th~t, where practical, capability shall be provided to restore containment closure through simple manual actions.
[1]
NUREG-1449, "NRC Staff Evaluation of Shutdown and Low Power Operation," February 1992.
[2]
NUREG-1410, "Loss of Vital AC Power and the RHR System During Midloop Operations at Vogtle Unit 1 on March 20, 1990," June 1990.
[3]
IAEA, "Modeling of Accident Sequences During Shutdown and Low Power Conditions," November 1991.
[4]
"Seabrook Station Probabilistic Safety Study, Shutdown (Modes 4, 5 and 6)," May 1988.
[5]
"Shutdown Risk, Prairie Island Dual Unit Shutdown," Northern States Power Co., ANS Executive Conference on IPEs, October 1992.
[6]
'The Diablo Canyon Shutdown Safety Assessment," Pacific Gas and Electric, ANS Executive Conference on IPEs, October 1992.
ANPR Question Number 12; Should equipment provided only for severe accident prevention or mitigation be subject to (a) the same requirements as design basis equipment (e.g.,
redundancy/diversity, power supply, environmental qualification, inclusion in plant Technical Specifications, maintenance priority, quality assurance; or (b) lesser standards
( e.g., reduced design margins or the regulatory guidance found in appendices A and B of Regulatory Guide 1.155, "Station Blackout?"). If lesser standards, what standards would be appropriate?
Industn Remonse; Equipment provided only for severe accident prevention or mitigation should not be subject to the same requirements as design basis equipment. This position regarding equipment survivability is co_nsistent with that expressed by the NRC staff in SECY 12
016 and subsequently accepted by the Commission via their SRM dated June 26, 1990.
The NRC staff recently reiterated this view in SECY-92-070, which observed that applying the requirements developed for safety-related equipment to equipment provided for severe accident would amount to inclusion of severe accidents in the design basis for the plant design. The industry concurs in the NRC staff position indicated by SECY 070 that, "[E]xisting requirements and tbe high degree of pedigree associated with them provide the design basis of the plant."
As stated in SECY-90-016, equipment provided only for severe accident prevention or mitigation should be designed so that there is reasonable assurance that it performs the function for which it is intended over its mission time. However, as indicated in the SECY, the equipment pedigree requirements need not be as stringent as those for equipment credited in design basis accident analyses. Specifically, features provided for severe accident protection only, need not not be subject to: (a) 10 CFR 50.49 equipment qualification requirements, (b) all aspects of 10 CFR Part 50, Appendix B quality assurance requirements nor (c) 10 CFR Part 50, Appendix A redundancy/ diversity requirements. Likewise, severe accident features would be included in Technical Specifications only to the extent that a design basis accident function is served by the same equipment.
The URD specifies that the design of equipment identified as useful for severe accident mitigation shall provide reasonable assurance that the equipment can perform its identified function during severe accident conditions. Design considerations for equipment under severe accident conditions include the circumstances of applicable initiating events and the environment ( e.g., pr~e, temperature, radiation) in which the equipment is expected to function. In addition, such equipment will be located and arranged, to the extent practical, to enhance its usefulness under severe accident
,conditions. Safety-related equipment that also serves for severe accident mitigation will be qualified and provided with quality assurance consistent with design basis requirements.
ANPR Question Number 13; Alternative 1 discusses not exceeding ASME service level C stress limits for steel containments under certain severe accident conditions. Are these limits appropriate for severe accident conditions? If not, what limits would be appropriate? Could these same stress limits also be used for loads generated by missiles? Hnot, what limits would be appropriate? What equivalent limits would be appropriate for concrete containments?
Industn Response:
Adherence to the ASME service level C stress limits for steel containments under severe accident conditions provides reasonable assurance that the containment will perform its intended function as an additional barrier against potential fission product 13
J release. The containment will perform satisfactorily if its integrity is not compromised and no unacceptable leakage develops during and after the severe accident. By limiting the stresses in the shell to the code yield, the service level C limits assure that the containment remains in the elastic domain by a margin of at least 20%. Within the stress rapge, the containment integrity is only minimally challenged since significant margin exists between service level C and ultimate failure; Tests and analytical modeling have shown the margin to be of the order of two or greater [1,2,3,4].
Limits less conservative than the service level C stress limits should be used for loads generated by missiles. Missiles have a local impact and do not threaten the overall integrity of the containment Criteria based on strain limits are appropriate for this type of loading. The energy is absorbed tlu:<>ugh plastic deformations limited by functionality requirements (i.e., no leakage and no zipper effect that would compromise the overall integrity of the structure).
For concrete containme:qts, the unity factored load combinations of Subsection CC of ASME Section ill are adequate requirements for severe accidents. They are based on the same philosophy as service level C stress limits applied to Class MC components.
The limits placed,on stresses and strains in the reinforced concrete load resisting elements, strain in the metallic liner plate and liner plate anchorage systems and other liner design details, provide significant margin against catastrophic failure and containment pressure boundary leakage.
The containment boundary strain levels associated with service level C will be within elastic limits such that any leakage that does occur can be expected to be via penetrations through the primary steel membrane. Tests have been performed on individual penetration types including airlocks, hatches, electrical penetration assemblies, and mechanical penetrations at pressures comparable to containment ultimate capacities.
Such tests have demonstrated that typical penetration designs can withstand pressures substantially greater than equivalent service level C containment pressure without loss of structural integrity or any significant degradation to their leakage barrier function.
The containment loading criteria discussed above are being incorporated into AL WR designs, as appropriate, based on utility requirements and will be codified via the design certifications. Therefore, as discussed in response to Question 1, generic rulemaldog such as that envisioned by this ANPR is unnecessary.
[1]
Keck, J., F. Thome, "Leak Behavior Through EPAs Under Severe Accident Conditions," Proceeding of the Third Workshop on Containment Integrity, NUREG/CP-0076.
[2]
Julien, J. T., S. W. Peters, "Leak Rate Test of a Containment Personnel Airlock," Fourth Workshop on Containment Integrity, NUREG/CP-0095.
14
[3]
Brinson, D. A, G. H. Graves, "Evaluation of Seals for Mechanical Penetrations of Containment Buildings," NUREG/CR-5096, August 1988.
[4]
Crapo, H. S., R. Steele, Jr., "Containment Penetration System (CPS) Tests Under Accident Conditions," NUREG/CR-5043, August 1988.
ANPR Question Number 14; What information is available regarding the costs ( capital and operational/
maintenance) of design features that wou1<! be required under these alternatives?
Industa R,emonse; Based on the industry position strongly preferring the URD/design certification process for accomplishing AL WR severe accident issue resolution/ codification, the industry does not have the appropriate data and has not addressed the comparative costs
of implementing the three 3;Iternatives outlined in this ANPR ANPR Question Number 15; The containment performance objective discussed in Alternatives 1 and 2 (i.e.,
containment shall provide a barrier against the release of radioactive material for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage) represents a level of safety for a 3800 MWt plant sited in accordance with 10 CFR Part 100 approximately three orders of magnitude below the Commission's quantitative health objective for prompt fatalities, as defined in the Commission's Safety Goal Policy Statement. It cou1d be argued that a future L WR design meeting this objective through analyses and the incorporation of design featur,es need not consider the addition of other features, since these other features wou1d be directed at even more highly unlikely severe accident phenomena and sequences which cou1d be considered "remote and specu1ative" under the National Environmental Policy Act (NEPA) and 10 CFR Part 51. Therefore, wou1d the codification and compliance with such a containment performance objective be sufficient to also define a point of truncation and serve as the basis for an amendment to 10 CFR Part 51 e1iroinating the need for further review of SAMDAs for future LWRs under 10 CFR Part 51?
Industcy ReSJ)Onse; The industry concurs in the underlying thrust of ANPR Question number 15, consistent with the intent of design certification under Part 52, that additional requirements shou1d not be considered for AL WR designs which will have a level of safety substantially exceeding that established by the quantitative health objectives of the Safety Goal Policy Statement. In keeping with the intent to resolve all design related issues at certification, the industry and NRC are addressing the requirements under 15
NEPA relative to consideration of severe accident mitigation design alternatives (SAMDAs) as part of the design certification process. The staff recommendation in this regard, later concurred in by the Cororois.sion in an SRM dated October 25, 1991, was contained in SECY-91-229.
While the concept of generically eHminating NEP A/SAMDA requirements for future plant designs demonstrating a high level of safety is sound, this benefit does not outweigh the industry's substantial concerns regarding the implications of conducting a generic severe accident ruleroaking that would be necessary to provide the basis for such a change. Further, as a practical matter, a generic rulemaking would likely not be timely to eHminate the NEPA/SAMOA review requirements for evolutionary plant design certifications.
16
Westinghouse Electric Corporation Mr. Samuel J. Chilk Energy Systems The Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555
92 DEC 28 p 4 :1 6 ATTENTION:
DOCKETING AND SERVICE BRANCH Box 355 Pittsburgh Pennsylvania 15230-0355 ET-NRC-92-3788 NSRA-APSL-92-0269 December 22, 1992
SUBJECT:
"Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations" (57 Fed. Reg. 44513, September 28, 1992) - Advanced Notice of Proposed Rulemaking (ANPR)
Dear Mr. Chilk:
The purpose of this letter is to provide Westinghouse Electric Corporation comments on the subject proposed rulemaking.
On June 26, 1992, Westinghouse submitted to the NRC an application for final design approval (FDA) under Appendix O to 10 CFR Part 52 and a Standard Design Certification (DC) under 10 CFR Part 52 for the AP600 plant design. As one of the three vendors with Advanced Light Water Reactor (ALWR) designs currently under review by the NRC, we have carefully reviewed the subject ANPR with particular focus on the impacts the proposed rulemaking would have on the pursuit of the FDA and DC for the AP600 design.
Westinghouse believes that generic rulemaking to address severe accident issues for future plants is neither necessary nor desirable. Contrary to a stated purpose of the proposed rule, we believe that the proposed rulemaking will not "facilitate design certification rulemaking" but would instead increase uncertainty with little or no added value. To pursue generic rulemaking for severe accidents at this point in the AP600 Design Certification effort would at best result in a duplication of effort, as well as a duplication of regulation and guidance. The intrusion of such rulemaking at this time introduces substantial uncertainty and confusion into the Part 52 licensing process, and could easily jeopardize design certification for the AP600 as well as other advanced plants.
AL WR severe accident issues are presently being addressed by the industry and NRC through the ALWR Utility Requirements Document (URD) and design certification interactions. The utility requirements developed for both evolutionary and passive plant designs address a wide spectrum of severe accident phenomena and containment challenges. The Westinghouse AP600 is designed to meet the utility requirements for passive ALWR designs. As such, we believe that NRC review of the 0732A
ET-NRC-92-3788 NSRA-APSL-92-0269 December 22, 1992 URD, as well as the design certification review of the AP600 design, provides the preferable method for resolution of severe accident technical issues. This process allows for generic resolution of severe accident issues to the extent possible via the ALWR URD and review and codification of design specific implementation as part of the AP600 design certification process.
We are particularly concerned that a generic rulemaking for severe accidents separate from the design certification process will significantly disrupt the design certification process being pursued for the AP600 with the potential for causing costly and unwarranted delays.
As noted in the ANPR, "this rule would be generally applicable to passive LWR designs. However, as detailed design information becomes available and review of the passive systems is completed, further rulemaking may be necessary."
The implication is that the net result of this potential series of generic rulemakings is a rule which would reflect the information obtained through the specific passive plant design certifications. Thus there is no need for a separate rulemaking at this time. We expect the Design Certification application for the AP600 to be docketed within the next month (a docket number has already been assigned). The review of the AP600 design by the NRC staff is already in process and we expect the AP600 FDA to be issued in mid to late 1994. This time frame does not lend itself to the process envisioned in the ANPR.
Westinghouse has participated in the preparation of industry comments on the proposed rulemaking and we support the comments being offered on this subject by NUMARC.
We appreciate this opportunity to comment on the proposed rulemaking and urge that the Commission not proceed with the proposed generic rulemaking for severe accidents. As discussed above, we believe the NRC staff should continue to work toward resolution of severe accident issues for advanced plants via the design certification process.
Very truly yours, L/~~tl',,,.
N. J. Liparulo, Manager Nuclear Safety and Regulatory Activities
/nja cc:
The Honorable Ivan Selin, Chairman Commissioner Kenneth C. Rogers Commissioner James R. Curtiss Commissioner Forrest J. Remick Commissioner E. Gail de Planque Mr. J. Taylor, EDO 0732A
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92 IJEC 24 A11 :06 AECL AECL Technologies 9210 Corporate Boulevard Suite 410 Rockville Maryland 20850 USA 1-800-USA-AECL (301) 417-0047 Fax(301)417-0746 Telex 403-442 File: 33-0002-122 December 21, 1992 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Docketing and Service Branch Re:
Advanced Notice of Proposed Rulemaking: Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations [57 Fed. Reg. 44513, (September 22, 1992)]
Dear Mr. Chilk:
AECL Technologies has reviewed the subject advance notice of proposed rulemaking. We appreciate the opportunity to review this document and to contribute our views for Staff and Commission consideration. AECL Technologies supports initiation of rulemaking to resolve severe accident issues generally prior to commencement of design certification proceedings to the extent practicable.
Sincerely,
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J) A. D. Hink
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ice President/General Manager AECL Technologies f:u\r\comanpr.
JAN 2 1 1993 Acknowledged by card..................................
A Division of AECL Inc.

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AECL TECHNOLOGIES (AECLT)
COMMENTS ON December 21, 1992 THE NUCLEAR REGULATORY COMMISSION'S (NRC)
ADVANCE NOTICE OF PROPOSED RULEMAKING (ANPR)
CONCERNING ACCEPTABILITY OF PLANT PERFORMANCE FOR SEVERE ACCIDENTS; SCOPE OF CONSIDERATION IN SAFETY REGULATIONS
[57 FED. REG. 44,513 (SEPTEMBER 28, 1992)]
I.
INTRODUCTION I.
The ANPR states that one purpose of a severe accident rule would be to "Provide assurance that the performance of future LWRs under severe accident conditions is consistent with assumptions about severe accident performance used in developing new source term information."
57 Fed. Reg. 44,514.
To achieve this purpose, AECLT believes that the rule should establish comprehensive requirements applicable to all reactors and that Regulatory Guides should be developed which provide specific guidance concerning ways of meeting the requirements for specific reactor systems or reactor *types.
These guides should become available in draft form at the time a proposed rule is issued.
Provided herein are AECLT's comments on the ANPR, as well as responses to those questions in the ANPR applicable to Pressurized Heavy Water Reactors (PHWR) such as the CANDU 3 reactor.
Since questions 8, 11, 13 and 14 are not applicable to PHWRs,, no responses to these questions are provided.
GENERAL COMMENT
S:
1.
The overall criteria for protecting the public f:r:om severe accidents should be the same for all water-cooled reactors.
A severe accident rule should specify these overall criteria.
AECLT believes the format of such a rule should be similar to the format described as Alternative 3 in the ANPR.
Adoption of this format would encourage designer flexibility and inventiveness in the incorporation of severe*
accident. prevention and mitigation features to
reduce the frequency and consequences of such accidents.
2.
The ANPR indicates that the criteria discussed in this ANPR would codify much of the Commission's guidance for general application to all future LWRs.
AECLT believes that this guidance would also be applicable to PHWRs.
Presently, the NRC is conducting a preapplication review of the CANDU 3 design.
In conjunction with NRC's review of CANDU 3 severe
accident prevention and mitigation design features, AECLT has prepared at NRC's request, a 1
December 21, 1992 comparison of the CANDU 3 features to the NRC Staff's recommended criteria in SECY-90-016, as modified by Commission
guidance, concerning severe accident prevention and mitigation in LWR designs.
Based on this comparison, AECLT concludes that the C.ANDU 3 design will conform with the SECY-90-016 recommendations and guidance.
3.
The ANPR discusses three potential -alternatives for design requirements related to prevention and mitigation of severe accidents.
Alternative 1 would prescribe hardware requirements to address risk-significant phenomena.
Alternative 2 would require designers to address risk-significant phenomena in the design, but would not prescribe specific hardware requirements.
Alternative 3 would specify General Design Criteria to describe the nature of the severe accident challenges as well as associated success criteria.
From the description of each alternative in the ANPR, AECLT cannot tell whether the alternatives are intended to be equally comprehensive in scope.
AECLT believes that, regardless of the format adopted for the severe accident rule, the rule and accompanying guidance concerning implementation of the rule should be comprehensive in scope and should address the following matters:
(a) criteria for establishing event sequence frequencies; (b) radiological consequence limits; (c) capacity and reliability of the design feature; and (d) criteria to establish load combinations and environmental conditions.
Additionally, implementing regulatory guidance should address redundancy, diversity, power supply, equipment survivability, analytical methods, and acceptance criteria.
4.
Specifically, in the rule and implementing guidance the following matters should be addressed:
A.
Selection Process for Severe Event Sequences Considered in the Design.
The selection process should be based on event frequency.
The process would establish the frequency limits to: (l)define the events requiring design changes to reduce their frequency, (2) define the events that require features to mitigate the event's consequences and (3) define events that need not be considered in the design.
B.
Consequence Limits: For each event sequence defined by A(l) and A(2) above (e.g. reactivity events, loss of heat sink at High/Low Pressure),
acceptable 2
5.
C.
D.
E.
December 21, 1992 consequences for the event frequency should be defined on an overall basis (e.g. containment stress and leakage, radiological consequence limits).
In addition, a phenomenon acceptance criterion should define the acceptable consequences for each individual phenomenon (e.g. hydrogen, molten fuel, non-condensable gas) associated with the event consistent with the overall acceptance criteria and the design features that produce the phenomenon.
Phenomenon Acceptance Criteria: For each phenomenon acceptance criterion, systems/features should be identified which provide the means to mitigate the consequences of the phenomenon.
System/Feature Design Criteria:
For each system/feature, design criteria should be established for
capacity, load combinations, environmental conditions vs time, and reliability.
The reliability criteria should include: redundancy, diversity, power supply, separation (from each other and from systems/features whose failures are involved in the severe accident event sequences),
and environmental qualifications.
System/Feature Demonstration Requirements: For each system/feature, the demonstration analysis/test requirements should be defined.
These should include assumptions, acceptance criteria, analytical methods, and test requirements.
For a criteria-oriented rule, similar which AECLT
favors, items A and B included in the rule; items C, D, and included in a Regulatory Guide.
to Alternative 3, above should be E above should be 6.
Because each reactor type may have some unique requirements, AECLT suggests that the rule be structured in two parts. The first part should present the overall requirements to be met by all reactors.
The second part would have separate sections for each reactor type (i.e.,
Pressurized Water, Boiling Water, Pressurized Heavy Water)
PLWR, BLWR, PHWR (i.e. CANDU) ).
Other reactor types (e.g. sodium and gas cooled reactors) could be added at a later date.
7.
A severe accident rule ideally should be of sufficiently comprehensive scope to permit severe accident closure determinations to be made for new designs.
AECLT believes that a rule and implementing guidance of the 3
December 21, 1992 scope described above in paragraph 4 would permit these determinations to be made.
Issues identified in 4 which are not addressed in a severe accident rule will have to be addressed in individual Standard Design Certification rulemakings or in COL proceedings.
8.
As discussed in 3 and 4 above, a severe accident rule should specify a cut-off event frequency such that events below this frequency need not be considered in the design and for which further analysis is not required.
NUREG/CR-5368, "Reactivity Accidents" reported the results of analyses of light water reactor reactivity events performed by Brookhaven National Laboratory.
For that effort, Brookhaven categorized potential event sequences as being worthy of further analysis, or not.
One of the screening criteria used to determine the importance of a sequence for further analysis was whether the sequence required too many low probability events to occur in combination. Brookhaven established a screening methodology with which low probability events could be eliminated from further consideration.
Event sequences with a frequency of less than lE-7 per reactor year were considered "incredible" and not recommended for further study.
AECLT believes that the generic severe accident rule should codify similar screening criteria.
4
II.
ANSWERS TO NRC'S SPECIFIC QUESTIONS Question 1 December 21, 1992 Is a
rulemaking addressing severe accident plant performance criteria desirable?
If so, why?
If not, why not? Would a rule provide better coherence and predictability to the design review and certification processes for future reactor designs or is rulemaking on these issues via individual design certification sufficient?
Response 1 AECLT believes a rule establishing generic severe accident criteria and plant performance criteria to prevent and to mitigate severe accidents is desirable.
AECLT prefers that the rule be in the format described by Alternative 3 in the ANPR.
As discussed in comment #4 above, the rule should establish the event frequency bounds, design criteria and radiological consequences associated with severe accidents.
These criteria should be independent of reactor type.
Such a rule would provide predictability in the certification process.
It would provide assurance to the public that the individual design certifications would be consistent with respect to degree of protection provided from severe accidents.
The rule should be supplemented by Regulatory Guide (s) which identify the phenomena identified to
date, the acceptable systems/features to cope with phenomena, and the acceptance criteria for such systems.
The supporting Regulatory Guide (s) should be issued at the same time as the rule.
Question 2 Would a new rule in 10 CFR part 50, concerning plant performance for severe accidents, as discussed in the three alternatives, provide a basis for revising the requirements on Emergency Planning Zones for future LWRs?
If so, why?
If not, why not?
Response 2 The three alternatives discussed in the ANPR do not address the offsite radiological consequences of a severe accident; therefore, they do not provide an adequate basis for revising the requirements on Emergency Planning Zones.
As discussed in our Answer to Question 1, AECLT believes that the severe accident rule should address such consequences.
If the rule does so, the rule would provide a basis for EPZ simplification for all future reactors (both LWRs and PHWRs) encompassed by the rule's scope.
5
December 21, 1992 Question 3 One option for an overall containment performance criterion that has been considered is that the conditional failure probability of the containment should be less than approximately one in ten.
Two of the alternatives use a deterministic surrogate that states that the containments should remain leak tight for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage and after that time remain a barrier against the uncontrolled release of radioactivity when faced with challenges from the more likely severe accident phenomena.
Is this criterion a suitable substitute for the conditional containment failure probability of one in the ten?
If so, explain why.
If not, explain why not.
Is a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months an appropriate time frame? Is its degree of conservatism appropriate considering uncertainties and defense-in-depth?
If not, what alternative would be appropriate?
What other criteria (probabilistic or deterministic) might be considered?
Response 3 Because of the wide range of the types of challenges to containment that may result from severe accident events, the NRC should not specify any specific criteria for evaluating these challenges in the rule.
Instead, general criteria concerning event frequency and radiological consequences should be in the rule.
Design specific criteria should be in the Regulatory Guides.
The applicant should provide the traditional justification for the analysis of containment performance during severe accident events.
Question 4 Alternative 2 would require extensive reliance on analytical tools that calculate the affects of severe accident phenomena.
Are there analytical tools that are sufficiently developed and adequate to allow effective implementation of such a phenomena-based rule?
If so, what are they, and for what phenomena could they be used?
How would alternative 2 be implemented? For example, should the codes and input parameters be approved by NRC? Should acceptance criteria be codified or put in a regulatory guide?
Response 4 Alternative 2 may dampen innovative approaches to the prevention and mitigation of severe accidents.
Alternative 3 would not be so dependent on the state of technology and so difficult to change to incorporate the results of ongoing research programs.
6
December 21, 1992 Question 5 Should future LWR containment designs include features beyond those described in alternative 1 to prevent/mitigate severe accidents?
If so, what are they?
Response 5 AECLT believes that Alternative 1 is unnecessarily restrictive. By codifying specific design requirements based on current knowledge, Alternative 1 does not allow for alternative designs.
This is an impediment to innovation based upon increased understanding of alternative technologies. Alternative 3 would codify the acceptance criteria and permit innovative designs to meet those criteria.
Question 6 Alternatives 2 and 3 specify phenomenological severe accident challenges that should be considered in the design. Alternative 1 is based upon the same phenomena/ challenges. Are there other severe accident phenomena/challenges that should be considered?
What should the criteria for deciding whether a severe accident phenomena or challenge is likely and should be considered?
Should the challenges be specified in more detail (for example, specifying the amount of hydrogen generation) or is a general statement of the challenge more desirable?
Response 6 As discussed above in General Comment 4, the criteria for deciding whether a phenomena or challenge should be considered in the design should be based on the event sequences to be considered in the design and the phenomena they produce.
This requires a systematic review of the plant for potential events and an analysis of their event frequency and their phenomena.
As discussed above in General Comment 8, AECLT believes that phenomena associated with event frequencies less that lE-7 should not have to be considered.
Question 7 For what reason (e.g. not a risk significant phenomena, not a cost effective solution) would any of the criteria proposed in the three alternatives not be fully applicable to passive designed LWRs?
Response 7 Alternative 1 may be design-dependent, as may Alternative 2.
Alternative 3 would be independent of specific reactor designs and, therefore, would be applicable to passive designs.
7
December 21, 1992 Question 9 If a design includes the capability to rapidly depressurize the primary system, should it also be required to have a reactor cavity design and/or a reactor vessel support structure capable of mitigating and accommodating a high pressure melt ejection?
Response 9 The need for either preventing or accommodating a high pressure melt ejection should be established on the frequency limit for the events, that should be considered in design. If the frequency limit can be met or exceeded with system(s) that prevent this event, it should not be necessary to accommodate the event.
It is more prudent to design to prevent this event rather than design to accommodate this event. The design of preventive systems, (i.e. depressurization systems, power supply, feedwater, etc.) is straightforward. The design of accommodation systems is speculative because the conditions of the high pressure melt ejection are nncertain.
Question 10 Should future LWR designs include an on-line instrumentation system that monitors containment atmosphere for gross leakage to reduce the risk from an inadvertent bypass of containment function?
Would application of this system be sufficient basis to modify leak rate testing requirements under 10 CFR part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."
Response 10 In the CANDU 3 design, containment air pressure and temperature, along with other data, are monitored while the plant is operating to provide a timely indication of any gross breach of containment.
The provision of a gross leakage monitoring system should be a sufficient basis to modify the requirements of Appendix J.
Question 12 Should equipment provided only for severe accident prevention or mitigation be subject to (a) the same requirements as design basis equipment (e.g. redundancy/diversity, power supply, environmental qualification, inclusion in plant Technical Specifications, 8
December 21, 1992 maintenance priority, quality assurance); or (b) lesser standards (e.g., reduced design margins or the regulatory guidance found in appendices A and B of Regulatory Guide 1. 155, 11 Station Blackout?")
If lesser standards, what standard would be appropriate?
Response 12 The question appears to suggest only two alternatives for requirements; however, there is a third alternative that considers the nature of the design feature, its safety function and the conditions under which it should operate.
The requirements for severe accident prevention or mitigation equipment/features should be appropriate for the specific feature, the time-history of the conditions associated with the event, and the desired reliability goal for the equipment/feature.
For example, the hydrogen igniter system, depressurization systems and heat removal systems would have different requirements from the reactor cavity and basemat.
Question 15 The containment performance objective discussed in Alternatives 1 and 2(i.e. containment shall provide a barrier against the release of radioactive material for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage) represents a level of safety for a 3800 Mwt plant cited in accordance with 10 CFR part 100 approximately three orders of magnitude below the Commission's Safety Goal Policy Statement. It could be argued that a future LWR design meeting this objective through analyses and the incorporation of design features need not consider the addition of other features, since these other features would be directed at even more highly unlikely severe accident phenomena and sequences which could be considered II remote and speculative II under the National Environmental Policy Act (NEPA) and 10 CFR part 51.
Therefore, would the codification and compliance with such a containment performance objective be sufficient to also define a point of truncation and serve as the basis for an amendment to 10 CFR part 51 eliminating the need for further review of SAMDA's for future LWRs under 10 CFR part 51?
Response 15 Regardless of the rule format (Alternative 1, 2 or 3), the rule should be sufficiently definitive to eliminate the need for further review of SAMDAs for future LWRs and PHWRs under 10 CFR Part 51.
The approach described in General Comment 4 is of sufficient scope to permit severe accident closure under NEPA for designs meeting the requirements.
The rule should include a determination to that 9
December 21, 1992 effect so that the issue cannot be raised successfully in a design certification or COL proceeding.
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DOCKET NUMBER PR (f)
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OQ(;K[l((J 1825 Prelude Drive USNHC The Secretary of the Commission U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Docketing and Service Branch Vienna, VA 22182-3345 November 2, 1992
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1.
Desirability of Severe Accident Performance Rulemaking A rulemaking on severe accident performance is appropriate and is highly desirable.
First, it will address for ALWRs many of the considerations that went into 10CFR50.44 that addresses a number of plants that will not be built. Second, a rulemaking will provide a consistent regulatory basis for certification and licensing of future plants.
Policy statements have their uses, but they do not provide a substantial basis for licensing and regulatory actions that may be required.
2.
Would a New Rule Provide a Basis for Revising EPZ Criteria for Future LWRs?
A properly crafted rule could provide a basis for revising EPZ criteria for future LWRs.
First, plants designed to explicitly meet severe accident performance criteria could have lower off-site dose consequences from core damage sequences. This can provide a technical basis for reducing the EPZ from a ten mile radius to something smaller.
Second, the prospect of reduction in the EPZ can likely give substantial incentive to even more robust containment designs, since there can be a measurable reduction in public risk.
Third, from a public acceptance point of view, how can one claim one has developed a safer reactor design when there is no change in the EPZ? Thus, coherent regulations in this area can demonstrate in a tangible way that the advanced and evolutionary LWR designs indeed do result in lower risk to the public.
3.
Containment Performance Criterion Options Obviously, the selection of a one-in-ten conditional probability of containment failure is based first on determining a suitably low core damage frequency probability. The use of a leak-tightness criterion option is based on the assumption that leak-tightness equals radionuclide retention, which is only strictly true in the case of noble gases. Existing NRG-sponsored studies show that substantial increases in containment leakage rates following severe accidents lead to only small increases in public risk. I believe that the suggested surrogates miss the point. It is how well the containment performs to reduce risk to the public that is the issue. Thus, neither of these options alone is sufficient to address the issue of minimizing public risk in a cost-effective way.
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U. S. Nuclear Regulatory Commission November 2, 1992 Page two I believe that a more appropriate containment performance criterion can be crafted.
would suggest multi-part criteria along the following lines:
a.
That the conditional probability of prompt (within about one hour after initiation of core damage) containment function failure be less than some appropriate value.
This probability would likely be less than 10% and probably on the order of not more than 1 %, and
b.
That the containment function be capable of retaining at least 95% to 99%
of radioiodines and particulate radionuclides following failure. This appears readily achievable based on NUREG-1465 and would give full credit to radionuclide retention features such as sprays, chemistry controls, controlled filtered venting, and in-plant radionuclide plate-out, and
c.
That off-site doses due to noble gas release be reduced on the order of at least 50% to 75% below that which would occur if the noble gases were released as soon as they were generated and transported to the containment breach(es). This is consistent with criterion a. above.
4.
Analysis Required for Alternative 2 I believe that Alternative 3, Severe Accident General Design Criteria, is the best choice.
Thus, I have no further comment on this point.
5.
Inclusion of Additional Features Beyond Those Addressed in Alternative 1 I believe that Alternative 3, Severe Accident General Design Criteria, is the best choice.
However, the severe accident phenomena and concerns addressed in Alternative 1 summarize the scientific, regulatory, and utility consensus on potential mechanisms for containment challenge. Based on my review of severe accident and PRA literature, I do not believe that any other mechanisms need to be considered.
6.
Phenomenological Severe Accident Challenges - Level of Specificity The phenomena to be considered should not be specified with some arbitrary level of challenge such as 100% clad-water reaction, etc. First, such specificity will not provide incentive for designers to avoid or minimize challenges.
If a designer can reduce hydrogen generation, surely there should be positive regulatory incentive to do so.
U. S. Nuclear Regulatory Commission November 2, 1992 Page three Second, arbitrarily determining level of challenge a priori will not promote cost-benefit tradeoffs to be made among protection features. Clearly, more weight should be given to prompt containment failure sequences than those that may occur later and to more probable failure mechanisms to less probable ones.
Third, a review should be included to assure that design features and procedure-related strategies do not introduce new failure mechanisms. Such a review does not have to be overly complex and should reflect best judgement. This would be consistent with the intent of 10CFR50.59, and v.Jould be a more complete treatment of severe accident containment challenges than would be likely under detailed rule-specified level of challenge.
7.
Basis for Non-Applicability to Passive Designed LWRs Given that the fundamental basis for passive designs is to rely on methods that will "always" work, determination of applicability based on risk-significance is most appropriate. This would give full credit for a passive design's ability to eliminate certain containment failure sequences or to reduce the public health risk consequences.
8.
Prevention or Mitigation of Fuel-Coolant Interaction Obviously, any feature which minimizes the probability of sustained core uncovery and prolonged inadequate core cooling goes a long way towards preventing fuel-coolant interaction. Given the phenomenological uncertainties, it is difficult to determine precisely what design and procedure strategies should be followed to minimize fuel-coolant interaction. However, I strongly recommend that the regulations presume that without clear evidence to the contrary, addition of coolant is always beneficial, no matter where in the core damage sequence it is finally (re)introduced.
9.
Mitigation Features for Mitigation/Prevention of High Pressure Melt Ejection This question concerns level of specificity required for containment feature performance assuming an Alternative 1 type rule. Please refer to my comments 3. and 6. above.
10.
Use of On-Line Instrumentation System to Determine Gross Containment Leakage See comment 3. above. Unless the designer chose a double-walled containment design, or a secondary containment completely enclosing the primary containment and all of its penetrations, such instrumentation would likely be inadequate when it would really be needed during containment bypass sequences and unnecessary for all other sequences.
U. S. Nuclear Regulatory Commission November 2, 1992 Page four 11.
Containment Integrity Design Criteria Applying to Shutdown Conditions During shutdown conditions, sensible and decay heat loads are far less than at power.
This substantially reduces the rate and the magnitude of radionuclides that can be generated. Additionally, the same features that are added to ALWRs to increase the ECCS reliability will be available to keep the core covered during shutdown event sequences.
As far as off-site public risk is concerned, shutdown events are a tempest in a teapot.
I have reviewed calculations on an existing PWR that would indicate that a core melt sequence starting from cold shutdown with the containment hatchway open would be required in order to have off-site releases corresponding to a General Emergency.
Improved Technical Specification requirements pertaining to available power sources, available makeup systems, and containment configuration should more than cover shutdown events.
12.
Standards Required for Severe Accident Equipment There appears to be little difference in the performance of well-designed and maintained "safety-related" and "nonsafety-related" equipment. Given the low probability of severe accident, "gold-plated" QA requirements are clearly inappropriate and will result in unnecessary cost with little or no risk benefit. The standards to be used should be performance-based, reflect the existence of the maintenance rule, and reflect importance as determined by probabilistic safety assessment.
The process for determining appropriate standards reflected in the Station Blackout Rule is probably a good model for severe accident equipment, since SBO is typically a significant contributor to core damage.
13.
Use of ASME Service Level C Limits As stated above, I believe that Alternative 3, Severe Accident General Design Criteria, is the best choice. However, use of level C stress limits is probably too limiting. Given that Level D stress limits are used for once in a plant lifetime design basis events such as LOCA, it certainly would be inconsistent to use level C stress limits for events that are supposed to be beyond the design basis and of even lower probability. Given that the real function of the containment is to mitigate off-site dose consequences of severe accidents, significant structural plastic deformation could occur without affecting the ability to retain radionuclides.
U.S. Nuclear Regulatory Commission November 2, 1992 Page five
14.
Information Available on Costs I am not qualified to comment on this point.
15.
Containment Performance Objective versus 10CFR51 SAMDAs See my comment 3. above. Use of multi-part criteria suggested above is consistent with the Safety Goal Policy Statement and can form a reasonable outer bound beyond which "remote and speculative" sequences can be removed from consideration based on explicit consideration of the end goal -- reasonable measures to reduce public risk.
By using surrogate conditions that may be more conservative than required in most cases, and that may not adequately address prompt failures against which no public protective actions can be effectively implemented, then the goal of coherent regulatory policy is missed.
By focusing severe accident mitigation features on reduction of public risk, rather than some intermediate expedient surrogate, the technical and safety basis is clearly defined, need for future backfits can be more readily determined, and most importantly, a determined intervenor cannot use apparent inconsistencies among regulations to use a non-technical court system to set aside regulations.
This has been done before, successfully, and to allow history to repeat itself would be unconscionable.
Sincerely,
~J.J~
John S. Fuoto, PE
NUCLEAR REGULATORY COMMISSION 10 CFR PART 50 RIN 3150 -
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Acceptability of Plant Performance for Severe Accidents; Scope of Consideration in Safety Regulations AGENCY:
Nuclear Regulatory Commission.
ACTION:
Advance notice of proposed rulemaking.
SUMMARY
The Nuclear Regulatory Commission (NRC) is considering an amendment to its regulations for future light water reactors (LWRs).
The amendment would add provisions for the design of the plant structures to withstand certain challenges from phenomena associated with severe core damage accidents beyond the current "design basis accidents."
The NRC is issuing this notice to invite advice and recommendations from interested parties on the proper scope and method to incorporate these provisions into safety regulations.
i._J~q 2-DATES

Comment period expires [90 days after date of FRNJ.
The NRC will consider comments received after this date only if it is practical to do so, but the Commission is able to assure consideration only for comments received on or before this date.
ADDRESSES:
Send written comments to:' The Secretary of the commission, u. s. Nuclear Regulatory Commission, Washington, DC 20555, Attention:
Docketing and Service Branch.
Deliver comments to:
11555 Rockville Pike, Rockville, MD, between 7:45 am and 4:15 pm on Federal workdays.
Copies of comments received may be examined at the NRC Public Document Room, 2120 L street NW (Lower Level), Washington, DC.
FOR FURTHER INFORMATION CONTACT:
Thomas King, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3980.
SUPPLEMENTARY INFORMATION:
Background
The Nuclear Regulatory Commission (NRC) is considering developing regulations under 10 CFR Part 50 for future LWRs to address the ability of the plant to withstand challenges from phenomena associated with severe core damage accidents.
S~vere core damage accidents are low probability events beyond the design basis established in 10 CFR Part 50 that can lead t6 significant core damage and radioactive material release from the reactor fuel pins.
The NRC believes that research and engineering on the significant severe accident phenomena, event sequences, and cost effective methods to mitigate them, coupled with its 2
understanding of the details of future plant designs, have sufficiently matured to allow the development of plant performance criteria for future LWRs.
This advance notice of proposed rulemaking (ANPRM) requests from interested parties advice and recommendations on the proper, scope and method to incorporate these considerations into the NRC's regulations.
This ANPRM reflects consideration of the extensive work accomplished in the severe accident area.
Specifically, the NRC has already taken various actions in response to severe accident concerns.
On October 2, 1980, the Commission issued (45 FR 65474) an ANPRM that invited advice and recommendations on the consideration of degraded or molten cores in safety regulation.
Based on recommendations received from that ANPRM, the commission developed a policy statement that addressed severe accident considerations and withdrew the ANPRM (August 8, 1985; 50 FR 32151).
In its "Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants" published August 8, 1985 (50 FR 32138), the Commission stated its intentions for rulemakings and other regulatory actions for resolving severe accident safety issues.
For existing plants, the Commission concluded that these plants posed no undue risk to public health and safety.
Therefore, it did not see a need for immediate action on generic rulemaking for these plants because of the low severe accident risk.
The Commission has continued to take all reasonable steps to further reduce the risk from severe accidents 3
at existing plants through its regulatory programs.
For example, the Commission completed rulemakings on several key issues related to severe accidents (i.e., station blackout, anticipated transients without scram, hydrogen generation and control), has implemented a containment performance improvement program based upon insights regarding containment performance under severe accident conditions, and has initiated a program for individual plant examination (IPE) for severe accident vulnerabilities.
For future plants, the Severe Accident Policy Statement established the criteria and procedural steps under which a new design for a nuclear power plant could be acceptable for meeting severe accident concerns.
The NRC recognized the need to strike a balance between accident prevention and consequence mitigation in exploring the need for additional design features in the next generation of plants.
Also, the NRC expected that these new plants would achieve a higher standard of severe accident safety performance than prior designs.
The Commission stated that a "clarification of containment performance will be made including a decision on whether to establish new performance criteria for containment systems and, if so, what these should be."
The NRC staff has been reviewing proposed criteria for future LWRs submitted by Electric Power Research Institute (EPRI) and several new LWR designs with respect to the Commission's severe accident policy and the design certification 4
aspects of 10 CFR Part 52.
In performing these reviews, the NRC staff has proposed criteria to address severe accident and containment issues that depart from the existing regulations.
For the evolutionary LWR designs, many of these proposed criteria are contained in a paper provided to the NRC Commissioners on January 12, 1990, SECY-90-016, "Evolutionary LWR Certification Issues and Their Relationship to current Regulatory Requirements."
The NRC staff has sought and received Commission guidance on the application of these proposed severe accident and containment criteria to the evolutionary LWR designs now under review.
Guidance from the Commission was provided in a staff Requirements Memorandum, s. Chilk to J. Taylor, dated June 26, 1990.
The criteria discussed in this ANPRM would codify much of the Commission's guidance for general application to all future LWRs.
Additionally, the NRC plans to improve its regulations for future plants by separating (decoupling) the acceptance criteria for a reactor site from the acceptance criteria for the design of various engineered safety features (ESF) via rulemaking changes to 10 CFR Parts 50 and 100.
The first phase of this decoupling of siting criteria from design criteria focuses on updating and revising siting criteria.
The* second phase of this process would focus on updating 10 CFR Part 50 for future LWRs to:
(1) implement new LWR source term information, 5
(2) specify performance criteria for plant design features based on improved knowledge of the release of radioactive material into containment (i.e., new source term), and (3) specify crite~ia for plant performance under severe accident conditions.
The criteria discussed in this ANPRM are associated with a portion of this second phase, namely item (3) above
Purpose of the Rule The NRC believes that adopting a rule to specify acceptable plant performance in response to severe accidents would accomplish the following:
1.
2.
Codify the Commission's guidance on severe accident and containment is~ues that resulted from the review of advanced light water reactors.
Provide assurance that the performance of future LWRs under severe accident conditions is consistent with assumptions about severe accident performance used in developing new source term information.
3.
Provide guidance to future LWR designers and potential applicants.
6
4.
Add consistency and standardization to the resolution of severe accident issues based on the current technical information.
5.
Facilitate design certification rulemakings.
This rule would then help assure that the risk to the public from severe accidents in future LWRs is maintained at very low levels in accordance with experience from existing plants, current insights from risk studies and research results, and the Commission's Safety Goal Policy (August 4, 1986; 51 FR 28044).
In addition, this rule could complement and support the review of severe Accident Design Mitigation Alternatives (SAMDAs) on future LWRs as part of the environmental review carried out under 10 CFR Part 51.
Basis for the Rule This rule would reflect the NRC's current understanding of severe accident issues from its research, experience with light water reactors now in operation, and review of future designs.
Accordingly, this rule would apply to light water reactor designs only, but could possibly provide guidance for establishing cri~eria for other reactor types.
The development of this rule relies on the major factors discussed below.
Since the accident at Three Mile Island in 1979, considerable 7
L research on severe accidents has been performed.
This research has explored the phenomena associated with in-vessel and ex-vessel severe accident processes; hydrogen generation and control; the form, quantity, and timing of radioactive material release into the containment; challenges to containment integrity; and the consequences to the public.
This research has led to the development of data and analytical tools to analyze severe accidents for current and future designs, assess severe accident risk, and evaluate potential risk-reduction improvements in design and operation.
Application of these research results has occurred in many areas.
One comprehensive application has been the development of NUREG-11501, "Severe Accident Risk:
An Assessment for Five U.S.
Nuclear Power Plants."
NUREG-1150 used probabilistic techniques to analyze five operating plants from a severe accident risk perspective.
This analysis provided the NRC staff with basic insights into the important event sequences that can lead to severe accidents and the mechanisms that can lead to a loss of containment function during severe accidents.
These basic insights identified challenges to co~tainment integrity that can be divided into two groups: energetic or rapid energy releases, Copies of NUREGS may be purchased from the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37,082, Washington, DC 20013-7082.
Copies are also available from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161.
A copy is also available for inspection and/or copying at the NRC Public Document Room, 2120 L Street, N.W.
(Lower Level), Washington, DC.
8
and slower, gradually evolving releases to the closed containment system.
Examples of containment loadings in the first group include high pressure core melt ejection with direct containment heating, hydrogen combustion, and the initial release of stored energy from the reactor coolant system.
Decay heat and noncondensible gas generation from core-concrete interactions typify the group of slow energy releases to the containment.
Further insights from this analysis identified major contributors to risk to the public and potential design solutions.
Also, the NRC has frequently interacted over the past several years with EPRI and various reactor designers concerning regulatory criteria for the future evolutionary and passive LWRs.
Some of these interactions have addressed both probabilistic and deterministic criteria associated with plant performance under severe accident conditions.
These potential criteria have been discussed in various correspondence with EPRI and the reactor designers, and are documented in draft safety evaluation reports on the EPRI Advanced Light Water Reactor Requirements Document.
Finally, in preparing this ANPRM, the NRC has benefitted from the insights provided by an independent study of containment design criteria made by the Advisory Committee on Reactor Safeguards
{ACRS).
In a letter to the Commission on May 17, 1991, the ACRS proposed a set of criteria addressing the specific challenges posed by severe accidents to the.containment design for future 9
light water reactor nuclear power plants.
In SECY-92-070, dated February 28, 1992, the NRC staff analyzed these ACRS criteria with respect to EPRI design requirements for evolutionary and passive LWRs, evolutionary vendor designs, passive vendor designs, and the existing Commission guidance for the NRC staff's review of severe accident issues for evolutionary LWRs.
This ANPRM reflects consideration of the results of this analysis and the ACRS proposed criteria~
Applicability of the Rule The NRC has accumulated an understanding of the evolutionary light water reactor designs to complement its understanding of severe accident issues from operating reactors.
Based on this understanding, it is expected that the criteria developed in this rule would be consistent and compatible with the criteria being developed and applied in evolutionary LWR reviews.
However, due to the advanced stage of the reviews of the current evolutionary designs (GE ABWR and ABB/CE System 80+),
it is likely that the resolution of severe accident issues for these designs will occur via the individual design certifications before completion of this rulemaking.
The NRC expects the resolution of severe accident issues for the evolutionary LWRs and the results of this rule to be essentially the same.
The NRC staff is also reviewing future LWR designs that use a 10
passive design concept.
In contrast to the evolutionary designs, the NRC staff has reviewed only conceptual design information from the passive plant vendors.
This preliminary review has not identified any unique features that would prevent the evaluation of these designs under the rule discussed in this ANPRM.
Therefore, this rule would be generally applicable to passive LWR designs.
However, as detailed design information becomes available and review of the passive systems is completed, further rulemaking may be necessary.
Proposed Changes to Part 50 Discussed below are three potential alternatives for incorporating plant performance criteria for severe accidents into the regulations.
Alternative 1:
Hardware Oriented Rule This alternative (as are the other alternatives discussed in this ANPRM) is based upon ensuring that the risk significant severe accident phenomena, which may cause a loss of containment function in an LWR, are considered in future LWR designs.
Based upon currently available information, including the results of risk studies and severe accident research programs, these risk significant severe accident phenomena are:
11
1.
Hydrogen generation and transport, including burning and/or detonation, resulting from metal-water and core-concrete reactions;
2.
High pressure ejection of molten core material from the reactor vessel;
3.
Interactions between molten core debris and reactor basernat material, containment wall and structural material;
4.
Containment overpressure and overtemperature from decay heat, non-condensible gas generation, metal-water reactions;
5.
Stearn explosions from fuel-coolant interactions; and
6.
Containment bypass.
Alternative 1 would specify reasonable design features or attributes of design features directed toward prevention or mitigation of the above phenomena.
Where design features cannot be precisely specified to prevent or mitigate a severe accident phenomenon, this alternative would require that the applicant provide an evaluation of the phenomenon with respect to the 12
overall containment performance objective specified in the rule.
This alternative is derived from the containment performance criteria developed as part of the Commission's advanced reactor reviews and essentially codifies those criteria.
In this approach, those features of the design needed for severe accident prevention and mitigation would be specified directly in the rule.
These requirements would be an "overlay" on the existing design basis requirements in 10 CFR Part 50 for nuclear power plants.
The requirements would be considered and justified on an enhanced safety basis (i.e. using safety goal, cost-benefit analysis and other appropriate considerations such as defense-in-depth and uncertainties) and would complement the existing design basis to enhance the level of safety.
However, because of the low likelihood of severe accidents, these new requirements would not be considered to be traditional design basis requirements.
For example, design features provided only for severe accident mitigation would not be subject to the same conservative analysis and design requirements that are necessary for systems developed to cope with design basis accidents.
A regulatory guide would provide additional guidance on such design details as redundancy, diversity, system capacity, power supply, equipment survivability and analytical assumptions.
An example of this alternative follows:
13
50.XX Prevention and mitigation of severe accidents.
(a)
Applicability.
The criteria of this section apply to the design of light water nuclear power reactors being considered for a construction permit or operating license under 10 CFR Part 50 or applications under 10 CFR Part 52 on or after the effective date of this rule.
The criteria of this section also may provide guidance in establishing the requirements for other types of reactor designs.
(b) Containment Performance Objective.
The design shall include a containment system that provides a barrier against the release of radioactive material for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage under the more likely severe accident challenges.
Following this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, the containment should continue to maintain a barrier against the uncontrolled release of large quantities of fission products.
This shall be accomplished by:
(1)
Including plant design features that:
(i)
Provide the reactor coolant system (RCS) with the capability to rapidly and reliably reduce RCS pressure.
(ii)
Provide a reactor cavity design that restricts as much as practical the amount of ejected core debris that reaches the upper containment or 14
impinges directly on the containment wall.
The cavity design, as a mitigating feature, should not unduly interfere with operations including refueling, maintenance, or surveillance activities.
(iii)
Provide a reactor vessel support structure sufficient to retain the reactor vessel in place under the loads generated by a high pressure core-melt ejection.
(iv)
Provide for containment-wide hydrogen control (e.g.,
igniters, large volume), that accommodates the hydrogen resulting from a 100-percent metal-water reaction of the active fuel cladding, and limits containment hydrogen concentration to no greater than 10 percent; or provides that the post-accident atmosphere will not support hydrogen combustion.
(v)
Reduce the potential for and effect of interactions with molten core debris by:
(A)
(B) providing reactor cavity floor space to promote core debris spreading and coolability; providing a means to flood the reactor cavity to assist in the cooling process and scrubbing of fission products; (C) protecting the containment liner and other structural members from direct contact by molten core debris; (D) employing basemat materials which reduce the production of non-condensible gases when in 15
contact with molten core debris: and (E) ensuring that containment temperature and pressure increases or the generation of missiles resulting from decay heat, fuel-coolant interactions, combustible gas generation and control, and core-basemat material interactions involving a range of event sequences which release core debris into the containment do not cause containment stresses to exceed ASME service level C limits for steel containments, or equivalent for concrete containments, or significant degradation of the containment design leak rate.
(vi)
Reduce the possibility of containment bypass and a loss of coolant accident outside containment by designing, to the extent practical, all elements of systems and subsystems (e.g. piping, instrument lines, pump seals, heat exchanger tubes, and valves) located outside containment and connected to the RCS to an ultimate rupture strength at least equal to the full RCS pressure.
(2)
Not crediting use of containment venting during the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period following the onset of core damage in evaluating the design for compliance with the containment performance objective in paragraph (b) above.
16
(c) Equipment Survivability.
Features provided for severe accident prevention or mitigation shall be designed to operate for the time period needed in the environment (e.g., pressure, temperature, radiation) in which the equipment is relied upon to function, including consideration of the circumstances of applicable initiating events (e.g., transients, loss of AC power, loss of coolant accidents).
Maintaining containment integrity for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage provides time for the remaining airborne activity in the containment (principally noble gases and iodine) to decay to a level that, when analyzed realistically, would be unlikely to cause prompt health effects if containment failure or controlled venting were to occur after that time.
In addition, it represents a level of safety significantly below the quantitative health objective for prompt fatalities defined in the Commission's Safety Goal Policy.
However, considering the uncertainties involved in analyzing the severe accident phenomena and progression and emphasis on defense-in-depth, it is not unreasonable to include some conservatism in the criteria.
This time period would also enhance the time available for offsite protective actions.
To the extent practical during this period, the passive capability of the containment and any related design features (e.g., suppression pool) should provide for containment 17
integrity.
Following this period, the containment should continue to provide a barrier against the uncontrolled release of fission products.
However, in keeping with the concept of allowing for intervention in coping with long-term or gradual energy release, controlled, elevated venting (if provided in the design) may be given credit in the design analysis after the initial 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period to reduce the chance of containment failure.
The intent of specifying in the design analysis no reliance on containment venting during the initial 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is to achieve the design objective of high containment integrity but not to constrain use of venting during operation if for some reason venting were the desired course of action.
Alternatively, a design may use diverse containment heat removal systems or rely on the restoration of normal containment heat removal capability if enough time is available for major recovery actions.
The advantages of this approach include prescribing those design features to reduce the risk from severe accidents, thus promoting a more standardized resolution to severe accident issues.
In effect, this alternative is also prescriptive regarding the severe accident phenomena that a future LWR design must address, since the design features specified are a direct result of the phenomena considered. The prescriptive nature of this alternative will also tend to facilitate the NRC review and design certification process by focusing the review on the severe accident phenomena which must be considered and the basic 18
features which the design must incorporate to address those phenomena, thus enhancing regulatory efficiency.
In addition, this approach essentially codifies Commission guidance on severe accident and containment issues from the advanced LWR reviews.
This approach does not require the applicant to perform extensive severe accident analysis to show compliance with the rule.
The applicant could primarily rely upon design features which, through previous analyses and research, have been shown to be effective in reducing the risk from the more likely severe accident scenarios, coupled with deterministic analysis to confirm that the containment performance objective is met.
The disadvantage to this option is that it could discourage designers of future LWRs from developing other design approaches that might be more cost-effective, innovative, or safer.
Alternative 2:
Phenomena Oriented Rule This alternative is a modified version of the first alternative.
It, like the first alternative, states an overall containment performance goal and is based upon preventing or mitigating the same severe accident phenomena as described in Alternative 1.
However, instead of specifying hardware requirements in the rule to meet the goal, this alternative specifies the severe accident phenomena that need to be addressed in the design.
Based upon analysis of these severe accident phenomena, the designer would 19
develop and propose the actual design features necessary to meet the goal.
Regulatory guides would address items such as analytical methods, assumptions, acceptance criteria and guidance on design criteria for severe accident hardware.
An example of this alternative follows:
50.XX Prevention and mitigation of severe accidents (a)
Applicability.
The criteria of this section apply to the design of light water nuclear power reactors being considered for a construction permit or operating license under 10 CFR Part 50 or applications under 10 CFR Part 52 on or after the effective date of this rule.
The criteria of this section also may provide guidance in establishing the requirements for other types of reactor designs.
(b) containment Performance Objective.
The design shall include a containment system that provides a barrier against the release of radioactive material for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage under the more likely severe accident challenges.
Following this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, the containment should continue to maintain a barrier against the uncontrolled release of large quantities of fission products.
This shall be accomplished by:
(1)
Minimizing the likelihood or effects on containment 20
integrity of the following severe accident phenomena:
(i)
Uncontrolled hydrogen burning and detonation; (ii)
Interactions between molten core debris and (iii)
(iv) the reactor basemat material, reactor vessel support structure, containment wall, and other structural materials; High pressure melt ejection; Containment bypass and loss of interfacing system integrity; (v)
Steam explosions due to fuel-coolant interaction; and (2)
Not crediting use of containment venting during the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period following the onset of core damage in evaluating the design for compliance with the containment performance objective in paragraph (b) above.
(c) Equipment Survivability. Features provided for severe accident prevention and mitigation shall be designed to operate for the time period needed in the environment (e.g., pressure, temperature, radiation) in which the equipment is relied upon to function, including consideration of the circumstances of applicable initiating events (e.g., transients, loss of AC power, loss of coolant accidents).
The approach in this phenomena-oriented alternative would be similar to the hardware-oriented alternative in that it is 21
prescriptive regarding the severe accident phenomena which must be addressed in the design; however, it does provide flexibility for the designer to propose solutions specific for the design.
This alternative could be made more prescriptive by specifying, for example, the amount of hydrogen or molten core debris which must be considered but, nevertheless, would provide designers with considerably more design flexibility to address severe accident issues than Alternative 1.
Applicants would be required to provide analyses showing that their proposed design meets the containment performance objective.
However, this alternative would place a heavy reliance on analytical codes to predict the likelihood of severe accidents and their behavior accurately.
Limitations of these analytical codes and gaps in knowledge of the phenomenological progression of severe accidents may make such a heavy reliance unacceptable, unless bounding parameters are used. Like Alternative 1, this alternative would facilitate the NRc' review and design certification process by focusing the review and limiting litigation on the severe accident phenomena which must be considered; however, it would leave open to review and litigation whether the designer has adequately addressed the severe accident phenomena.
Accordingly, this alternative could potentially require considerable NRC review effort prior to accepting an applicant's analytical results.
Similar to Alternative 1, this alternative would be in the form of an overlay on the existing design basis specified in 10 CFR Part 50 and justified on an enhanced safety basis.
22
Alternative 3:
General Design Criteria (GDC) Oriented Rule In this alternative, the NRC would develop a set of new design requirements that would include definition of specific challenges posed by severe accidents and issue them as changes or additions to Appendix A, "General Design Criteria" (GDC), to 10 CFR Part
50.
Each new GDC would describe the nature of the severe accident challenge or containment load as well as a success criterion.
Usually, success would be defined simply as maintenance of the containment function for an appropriate period following the particular challenge.
Regulatory Guides would be developed to provide additional guidance on items such as analysis methods and assumptions.
The ACRS outlined this approach in more detail in a letter to NRC Chairman K. Carr, dated May 17, 1991. 2 This alternative would differ from the other alternatives in that the existing 10 CFR Part 50 design basis would be modified to include severe accidents.
Accordingly, the design requirements for the severe accident equipment (e.g., quality assurance, equipment survivability, redundancy/diversity) would need to be determined in relation to those for traditional design basis equipment.
Different design requirements may be appropriate for 2The ACRS letter is available for inspection at the NRC Public Document Room, 2120 L street, N.W. (Lower Level), Washington, D.C.
single copies are available from Mr. Thomas King, Office of Nuclear Regulatory
Research, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555, telephone (301)492-3980.
23
severe accident equipment because of the low probability associated with severe accident scenarios.
This alternative would give the designer flexibility in devising proposed solutions to severe accident phenomena.
Like Alternative 2, this alternative would require applicants to submit analyses showing that the criteria are met.
Like Alternative 2, this would place a heavy reliance on analytical codes to predict severe accident behavior accurately and would leave open to review and litigation in a licensing hearing or design certification rulemaking whether the designer has adequately addressed the severe accident phenomena.
Plans and Schedules The plant performance requirements described in this ANPRM are part of the second phase of a program to decouple siting and design criteria.
In this phase, plant performance requirements for severe accidents, in combination with other necessary changes to 10 CFR Part 50, will result in a rule that would complete the decoupling of siting and design.
Currently, the Commission plans to publish the proposed rule for comment in mid-1993 and to publish the final rule in mid-1994.
Specific Considerations The NRC invites comments and recommendations from interested 24
persons on the three alternatives for the proposed rulemaking or additional alternatives, if desired.
Furthermore, the NRC requests comments and supporting legal and technical information on the following questions:
1.
Is a rulemaking addressing severe accident plant performance criteria desirable? If so, why?
If not, why not?
Would a rule provide better coherence and predictability to the design review and certification processes for future reactor designs or is rulemaking on these issues via individual design certification sufficient?
2.
Would a new rule in 10 CFR Part 50 concerning plant performance for severe accidents, as discussed in the three alternatives, provide a basis for revising the requirements on Emergency Planning Zones for future LWRs?
If so, why?
If not, why not?
3.
One option for an overall containment performance criterion that has been considered is that the conditional failure probability of the containment should be less than approximately one in ten.
Two of the alternatives use a deterministic surrogate that states that the containment should remain leak tight for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the 25
onset of core damage and, after that time, remain a barrier against the uncontrolled release of radioactivity when faced with challenges from the more likely severe accident phenomena.
Is this criterion a suitable substitute for the conditional containment failure probability of one in ten? If so, explain why.
If not, explain why not.
Is a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months an appropriate time frame?
Is its degree of conservatism appropriate considering uncertainties and defense-in-depth? If not, what alternative would be appropriate?
What other criteria (probabilistic or deterministic) might be considered?
4.
Alternative 2 would require extensive reliance on analytical tools.that calculate the effects of severe accident phenomena.
Are there analytical tools that are sufficiently developed and adequate to allow effective implementation of such a phenomena-based rule?
If so, what are they, and for what phenomena could they be used?
How would Alternative 2 be implemented?
For example, should the codes and input parameters be approved by NRC?
Should acceptance criteria be codified or put in a regulatory guide?
5.
Should future LWR containment designs include features beyond those described in Alternative 1 to 26
prevent/mitigate severe accidents? If so, what are they?
6.
Alternatives 2 and 3 specify phenomenological severe accident challenges that should be considered in the design.
Alternative 1 is based upon the same phenomena/challenges.
Are there other severe accident phenomena/challenges that should be considered?
What should be the criteria for deciding whether a severe accident phenomena or challenge is likely and should be considered?
Should the challenges be specified in more detail (for example, specifying the amount of hydrogen generation) or is a general statement of the challenge more desirable?
7.
For what reason (e.g., not a risk significant phenomena, not a cost effective solution) would any of the criteria proposed in the three alternatives not be fully applicable to passive designed LWRs?
8.
What features could an advanced LWR design include that would prevent or mitigate fuel-coolant interactions?
9.
If a design includes the capability to rapidly depressurize the primary system, should it also be required to have a reactor cavity design and/or a 27
reactor vessel support structure capable of mitigating and accommodating a high pressure melt ejection?
10.
Should future LWR designs include an on-line instrumentation system that monitors containment atmosphere for gross leakage to reduce the risk from an inadvertent bypass of containment function?
Would application of this system be sufficient basis to modify leak rate testing requirements under 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."
11.
What design criteria should be developed that provide assurance that the containment's integrity could easily be established during certain shutdown conditions?
12.
Should equipment provided only for severe accident prevention or mitigation be subject to (a) the same requirements as design basis equipment (e.g.,
redundancy/diversity, power supply, environmental qualification, inclusion in plant Technical Specifications, maintenance priority, quality assurance); or (b) lesser standards (e.g., reduced design margins or the regulatory guidance found in Appendices A and B of Regulatory Guide 1.155, "Station Blackout?).
If lesser standards, what standards would 28
be appropriate?
13.
Alternative 1 discusses not exceeding ASME service level c stress limits for steel containments under certain severe accident conditions.
Are these limits appropriate for severe accident conditions? If not, what limits would be appropriate?
Could these same stress limits also be used for loads generated by missiles? If not, what limits would be appropriate?
What equivalent limits would be appropriate for concrete containments?
14.
What information is available regarding the costs (capital and operational/maintenance) of design features that would be required under these alternatives?
15.
The containment performance objective discussed in Alternatives 1 and 2 (i.e., containment shall provide a barrier against the release of radioactive material for a period of approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the onset of core damage) represents a level of safety for a 3800 Mwt plant sited in accordance with 10 CFR Part 100 approximately three orders of magnitude below the Commission's quantitative health objective for prompt fatalities, as defined in the Commission's Safety Goal 29
Policy Statement.
It could be argued that a future LWR design meeting this objective through analyses and the incorporation of design features need not consider the addition of other features, since these other features would be directed at even more highly unlikely severe accident phenomena and sequences which could be considered "remote and speculative" under the National Environmental Policy Act (NEPA) and 10 CFR Part 51.
Therefore, would the codification and compliance with such a containment performance objective be sufficient to also define a point of truncation and serve as the basis for an amendment to 10 CFR Part 51 eliminating the need for further review of SAMDAs for future LWRs under 10 CFR Part 51?
The preliminary views expressed in this ANPRM may change after considering the comments received.
In any case, the NRC-will provide an opportunity for additional public comment on any proposed rule developed as a result of this notice.
30
LIST OF SUBJECTS IN 10 CFR PART 50 Antitrust, Classified information, Criminal penalties, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.
The authority citation for this document is:
Sec. 161, Pub.
L.83-703, 68 stat. 948, as amended (42 u.s.c. 2201); Sec. 201, Pub. L.93-438 88 Stat. 1242, as amended (42 u.s.c. 5841).
Dated at Rockville, Maryland, this"')..)J day ofM-,k.,, 1992.
For the Nuclear Regulatory Commission.
k, of e Commission 31

ML23151A498

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Subject:

GENERAL COMMENT

Dear Mr. Chilk:

Subject:

Dear Mr. Chilk:

Subject:

Dear Mr. Chilk:

SUBJECT:

Dear Mr. Chilk:

SUBJECT:

Dear Mr. Chilk:

Dear Mr. Chilk:

GENERAL COMMENT

SUMMARY

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ML23151A498

Text

REFERENCE:

SUMMARY

SUMMARY

Subject:

GENERAL COMMENT

Dear Mr. Chilk:

Subject:

Dear Mr. Chilk:

Subject:

Dear Mr. Chilk:

SUBJECT:

Dear Mr. Chilk:

SUBJECT:

Dear Mr. Chilk:

Dear Mr. Chilk:

GENERAL COMMENT

SUMMARY

Background

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