ML20248M126

Discusses Evaluation of Rulemaking Language Proposals Re 10CFR50.59 (Changes,Tests & Experiments).Comments on Concept Offered by Nuclear Energy Inst in for Decoupling Scope of 10CFR50.59 from SAR Encl

ML20248M126
Person / Time
Issue date:	05/27/1998
From:	Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Diaz N, Dicus G, Shirley Ann Jackson, Mcgaffigan E, The Chairman NRC COMMISSION (OCM)
References
NUDOCS 9806150037
Download: ML20248M126 (7)

Text

_ _ __ . _ _ _ _ - .

~~r k' f-f2 9 fu f - .

m t//

p/.g, /,g; ' g

!* ,; UNtTED STATES

f. COMSECY-98-013 l j _ ( -. ' I };, NUCLEAR REGULAf0RY COMMISSION mam rc u a e nn m

.m f g**co...............,,,,

• • ..* May 27, 1998 e RELEASED TO THE PDR .

MEMORANDUM TO: Chairman Jackson o data kf[ bk o Commissioner Dieus ****.****oesce . g....,M Commissioner Diaz Commissioner McGaffigan FROM: L. Joseph Callan Executive Director erations

SUBJECT:

+

EVALUATION OF RULEMAKING LANGUAGE PROPOSALS CONCERNING 10 CFR 50.59 (CHANGES, TESTS AND EXPERIMENTS)

In an SRM dated March 24,1998, the Commission requested that the staff evaluate, for Commission consideration, the advisability of allowing (without prior NRC approval) licensee changes to a facility that result in the creation of an accident or malfunction of a different type than previously evaluated and that has " minimal" safety impact. As discussed in more detail below, we recommend that the Commission not make such a revision to 10 CFR 50.59. An alternative proposal is offered which we believe is responsive to the Commission's intent. If the Commission agrees, the staff would plan to include this alternative proposal as part of the '

L proposed rulemaking package for 10 CFR 50.59.

The SRM also asked the staff to reassess its position on acceptance limits on consequences and margin of safety, and report to the Commission on this matter. In addition, the SRM I directed the staff to allow " minimal" reductions in margin of safety as part of the 10 CFR 50.59 rulemaking. This memorandum also responds to these issues.

Finally, this memorandum provides staff comments on the concept offered by the Nuclear Energy Institute in a letter dated April 16,1998, for decoupling the scope of 10 CFR 50.59 from the safety analysis report ACCIDENT OF A DIFFERENT TYPE in accordance with 10 CFR 50.59, a change to the facility or procedures as described in the safety analysis report, or conduct of a test or experiment not described req'uires prior NRC approval (in the form of a license amendment) if, among other reasons, a possibility for an i

accident of a different type than any evaluated previously in the safety analysis report may be created. In determining whether an amendment to the license may be issued without a prior }

g-hearing, the NRC uses the requirements of 10 CFR 50.92. In particular, the NRC may make a prD ill i

NOTE: TO BE MADE PUBLICLY AVAILABLE

Contact:

Eileen M. McKenna, NRR WHEN THE FINAL SRM IS MADE AVAILABLE 415-2189

, Q - \ \$ Y_ ' O(

l 9906150037 990527 1 l PDR COMMS NRCC  !

CORRESPONDENCE PDR ,

_ _ _ _ _ _ _ __ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . A

The Commission 2 determination that an amendment involves a "no significant hazard consideration" (and thus may be issued without a prior hearing)', if operation of the facility in accordance with the proposed amendment would not, inter alia, create the possibility of a new or different kind of accident from any accident previously evaluated.

The staff recognizes that the Commission wishes to adjust the 10 CFR 50.59 process so that l

minor changes to the facility do not require approvalin the form of a license amendment, as reflected in the March 24,1998, SRM. One of the specific proposals the staff was asked to consider was changing the process to allow changes that create accidents of a different type with minimal safety impact. The Commission did not define the extent of what was contemplated as being of minimal safety impact or how this was to be determined.

The staff believes that it can be successful in modifying the rule to provide greater flexibility to licensees to make changes without prior NRC review and approval through changes to other criteria in 10 CFR 50.59, as discussed below and in GECY-97-205. However, the staff also believes that creation of a different type of accident is one area in which such latitude should not be permitted. The staff believes that any plant change that creates a new accident that was not previously evaluated should be submitted for staff review. Accidents as typically considered by the staff entai! challenges to the barriers for fission product release (fuel, pressure boundary, containment) and thus would have an impact on safety. In theory, a new accident will challenge existing plant safety s) stems or integrity of these barriers and, therefore, it is imperative for the NRC to have a firm understanding of the, causes of the new accident and the impact on the plant if this new accident should occur. Only through staff review can the NRC establish what the safety impacts will be and have assurance that existing plant systems will respond and mitigate the event. It would be extremely difficult to develop a meaningful definition of minimal safety impact that wou!d be effective in providing these assurances when dealing with new types of accidents. In addition, the staff notes the following concerning the regulatory and legislative history of the regulations about accidents of a different type.

For instance, a standard of minimal safety impact is not compatible with the stated intention of past Commissions, when they adopted both the unreviewed safety question (USQ) test and the no significant hazards consideration (NSHC) criteria, that 10 CFR 50.59 and 10 CFR 50.92 establish procedural star,Jards that do not require a determination on the merits of +he propoead changer . Commenters on the 10 CFR 50.91 and 10 CFR 50.92 rulemaking offered proposals to l

1

'The vast majority of license amendments that are issued involve changes for which l

there is a no significant hazards determination.

2 There is also some legislative history that suggests that Congress, when adopting the "Sholly" amendments to the Atomic Energy Act of 1954, as amended (AEA), with respect to NSHC, understood the Commission's intention that the determination whether a prior hearing is ,

j required before the NRC can finally determine an amendment request was not intended to be a j

i safety determination, and did not object to this approach.

i

! i E- - _- -- -- - )

The Commission 3 establish a threshold for creation of accidents of a different type, which were rejected by the Commission in 1986 (see 51 FR 7748). The statement of considerations noted:

In regard to the second criterion in the proposed rule, a number of commenters

} recommended that the Commission establish a threshold level for accident consequences (for example, the limits in 10 CFR Part 100) to eliminate prior notice for insignificant types of accidents. This comment was not accepted. The Commission stated that setting a threshold level for accident consequences could eliminate a group of amendments with respect to accidents which have not been pt eviously evaluated or which, if previously evaluated, may turn out after further evaluation to have more severe consequences than previously evaluated (48 FR 14868).

Further, the standards in Sections 50.59 and 50.92 are very similar, and there is no threshold established in 10 CFR 50.92 for significance of the accident, only that it be of a different type from any previously evaluated, if the Commission were to revise the rules such that some changes resulting in accidents of a different type were allowed to be made under 10 CFR 50.59, it would also need to revise 10 CFR 50.92 to limit the accidents of a different type that involve significant hazards considerations to exclude those that can be made under 10 CFR 50.59.

Otherwise, there would be a conflict in the rule provisions in which one section would say the change can be done without NRC approval, and the other section saying the same change requires a prior hearing (if one is held) on a required license amendment.

Finally, on the basis of information gained from the public comments and from reviews and inspections of licensee 10 CFR 50.59 evaluations, the staff concludes that the criterion related to accidents of a different type is not causing implementation problems. Rather, it is the other parts of the 10 CFR 50.59 criteria that are of concern (which are being addressed through other contemplated rule changes). Therefore, the staff believes that the specific proposal suggested by the Commission should not be implemented.

As an alternative, the staff could revise the language for accident of a different type from may be created to is created. This has the effect of requiring an affirmative conclusion on the part of a licensee that a new acci6nt has, in fact, been created. With the change to rule language in 10 CFR 50.59 presented above, the criteria in $50.59 would align with the criteria in @50.92.

Therefore, for those changes that require a license amendment because an accident of a different type is created, a prior hearing would be required if one is to be held at all.

MALFUNCTION OF EQUIPMENT OF A DIFFERENT TYPE in accordance with 10 CFR 50.59, a change to the facility or procedures as described in the safety analysis report, or conduct of a test or experiment not described requires prior NRC approval (in the form of a license amendment) if, among other reasons, a possibility for a

' The change in the standard of USQ with respect to accidents of a different type may also be inconsistent with Congress' understanding of the definition of USQ when it amended the hearing requirements in Section 184 of the AEA.

1

l The Commission 4 malfunction of equipment of a different type than any evaluated previously in the safety analysis report may be created.

The regulatory concerns expressed above about Sections 50.59 and 50.92 as related to

" creating the possibility of an accident of a different type" do not apply to the criterion established by the phrase " creating the possibility of a malfunction of equipment of a different type" because Section 50.92 does not have any criteria on malfunctions. Thus, the staff believes that it could provide rule language responsive to the Commission cirection to allow some flexibility for this criterior.. In view of the use of 10 CFR 50.59 as a procedural standard, rather than as a safety standard, the staff would agt propose language of minimal safety impact. Rather, the staff recommends adoption of the NEl proposal for use of the phrase " possibility of a malfunction with a different result is created." The staff concludes that this rule change would accomplish the intended purpose so that creation of a malfunction whose effects were already considered in the safety analysis does not require approval. Such changes would clearly be of minimal safety impact Those changes that give rise to the possibility of malfunctions with a different result from what has been previously evaluated would require review to determine whether there is a safety impact. The determination as to whether there is a different result would need to be assessed at the same level (i.e., component, train, or system) that the equipment being changed was previously evaluated.

ACCEPTANCE LIMITS ON CONSEQUENCES Section 50.59 states, in part, that an unreviewed safety question is involved if the probability of occurrence or the consequences of an accidere or malfunction of equipment important to safety previously evaluated in the safety analysis repon may be increased. In its guidance document, NEl 96-07, NEl stated that increases in consequeraes can be evaluated by determining whether acceptance limits (which for some accidentJ would be the regulatory guidelines in Part 100) continue to be met, despite the wording of Sectbn 50.59(a)(2)(i). The staff has stated that the determination as to whether consequences "may t.s increased" is made against the previous evaluations in the safety analysis report. In the March 24,1998 SRM, the Commission asked the staff to reassess its position on use of acceptance linits for consequences. The Office of the General Counsel has advised the staff that the interpretation in NEl 96-07 is not consistent with the rule and, therefore, enutd not t,a accepted without rulemaking. For the reasons discussed below, the staff does not recommend that such rulemaking he undertaken.

The current industry guidance, NEl 96-07, would permit, in some instances, ir. creases in consequences up to the regulatory thresholds (such as those contained in Part 100) without review. Allowing increases up to the acceptance limit, without review, could be inconsistent with a " minimal" increase standard since increases up to an acceptance limit may be more than minimal. Similarly, if the change results in a significant increase in consequences (still meeting acceptance limits), such an amendment would trip the criteria in 10 CFR 50.92 and require the l

l 4 It is interesting to note that the guidance also includes language that "where increases in consequences are so small that it cannot be determined that an increase has actually occurred, this is not an increase in consequences." This guidance appears inconsistent with guidance allowing increases up to acceptance limits.

l 1

. _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______ J

The Commission 5 amendment to be issued as a significant hazards consideration amendment. (Increases beyond the acceptance limits might not be found acceptable even if reviewed, or might require an exemption). Thus, limiting the need for prior approval to changes in which established ,

acceptance limits for consequences are exceeded would also appear to be in conflict with 10 l CFR 50.92(c)(1), namely, if operation of the facility in accordance with the proposed amendment would involve a significant increase in the probability or consequences of an accident previously evaluated.

NEl proposed in its November 14,1997, letter that the rule be revised so that a change would only require approvalif the consequences of an accident or malfunction previously evaluated j

exceed the established acceptance limit. As NEl discussed further in its letter, the established acceptance limit would be the value that was previously reviewed and approved by the NRC.

Attempting to use values from the staffs safety evaluation report (SER) as acceptance limits for I

consequences would oe difficult since SERs were not written for the purpose of establishing such limits5 . It is the staffs view that the reference to the SAR ve'"e as the baseline for l l

comparison is the most effective way to implement the regulation consistently for all plants.

The staff typically performs independent evaluations of radiological consequences of accidents, rather than an in-depth review of the licensee's calculations, during the licensing process for the pisnt. As a result, the degree of conservatism in the licensee calculations may differ from that used in the staffs assessments (typically the staff would conclude that the licensee is already closer to the " acceptance limits" than the licensee did). Although the staff would not be concerned about minimal increases in consequences or in cases that are not near the regulatory guidelines, the staffis concerned about allowing licensee changes without prior review which, when evaluated with licensee assumptions and methods, result in doses at or very close to the regulatory guidelines (e.g., Part 100). In these instances, the staff is concerned that these changes, if reviewed using the staffs assumptions (or starting from the staffs estimation of the accident dose), would result in the regulatory guidelines not being met. The staff would also have a concern about cumulative effectf, particularly when a licensee changes its analysis assumptions as well as its facility design or operation.

Finally, the staff notes that when considering acceptance limits with respect to margins, these are generally evaluated with respect to one of the three fission barriers for prevention of 5 in a literal sense, neither the SAR nor the SER set an "acceptancejimit." Rather, the SAR documents an applicant's/ licensee's analytically derived conclusion that a given event has a certain consequence that is within the regulatory bounds set by NRC regulations. The SER is intended only to reflect the staffs confirmation or modification of that conclusion. The applicant's/ licensee's SAR value, as modified through the staff review and approval of the SAR, then becomes the baseline for future analyses.

l

' The staff believes that the issue of cumulative effects will need to be considered if revised 10 CFR 50.59 criteria are adopted. Under consideration are revised reporting l l

requirements and possible changes to the level and method of NRC oversight. The staff believes such consideration is consistent with RG 1.174, which also discusses tracking of l l

cumulative effects.

l l

l l

l .

The Commission 6 radioactive release. When considering increases in radiological consequences, it is important to remember that all the barriers have been breached or degraded and, therefore, the calculations have a direct bearing on establishing or confirming the adequate protection of public health and safety (i.e., limiting release of radiation off site).

In summary, the staff concludes that proposed rule language that would allow " minimal' increases in consequences above those previously evaluated in the safety analysis report is a better standard than one based upon regulatory acceptance limits for consequences.

USE OF ACCEPTANCE LIMITS TO DETERMINE REDUCTIONS IN MARGIN OF SAFETY Section 50.59 states in part that an unreviewad dety question is involved if the margin of safety as defined in the basis for any technical specificat.on is reduced. Guidance has been provided by both NEl and NRC that licensees evaluate whether the margin of safety (defined by the TS bases) has been reduced by conside^g whether acceptance limits established during the license review are exceeded. In the staff guidance, acceptance limits are defined as specific values within which the licensee has proposed to operate the facility and which the NRC has accepted during its review of a license application. The acceptance limits in some cases are the calculated values reported in the safety analysis report, and in other cases are the acceptance criteria established by the Standard Review Plan (or other guidance). If the acceptance limits continue to be met, the margin of safety that has been established is not reduced. In contrast to the criteria based upon not increasing the probability or consequences of accidents / malfunctions previously evaluated in the safety analysis report the margin of safety criterion focuses on preserving margins that exist within established assumptions, methodologies and analyses used by the licensee to meet acceptance limits. As long as these limits continue to be satisfied, the

" margin of safety"is maintained.

For this reason, the staff does not recommend revision of 10 CFR 50.59(a)(2)(iii) to explicitly refer to " minimal" reduction in margin of safety, but rather plans to continue the current approach with acceptance limits'. The staff concludes that this approach will allow at least the degree of flexibility that would be afforded by a " minimal" reduction in margin criterion for changes when the existing acceptance limits are still met, but no reduction if the acceptance limit (which might also be a regula'ory limit as established by rule or license) would be exceeded.

NEl CONCEPT FOR REVISION TO SCOPE OF 10 CFR 50.59 in a letter dated April 16,1998, NEl stated that the industry and the NRC should take the opportunity of the forthcoming rule changes to 10 CFR 50.59 to not only clarify and simplify the criteria for requiring approval, but to also improve 10 CFR 50.59 by clarifying its scope of i

7 in SECY-97-205, the staff recommended rule changes on increases in probability or consequences (so that negligible increases could be allowed). Further, the staff proposed clarification on " bases for any technical specification" for margin of safety determinations. It was never the staff's intention to modify this criterion with a term such as " minimal" for these reasons noted.

l L --- - - - - - - -- - - . - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

~

\

The Commission 7 applicability. Specifically, they believe that the determination of the need for an evaluation should be whether the proposed change affects the safety analyses (that NRC approved in a safety evaluation report). NEl met with the staff on April 23,1998, to discuss this concept (see attached meeting summary).

The concept has not been sufficiently developed for the staff to understand how such a change could be implemented or to comment upon its merits. The staff recognizes the need to make a recommendation to the Commission regarding the scope of 10 CFR 50.59 in February,1999, as stated in the March 24,1998, SRM. However, as indicated at the April 23,1998, meeting, the the staff would not be able to assess such a concept for possible inclusion in the rulemaking package and still meet the schedule established by the Commission (July 1998) for providing a proposed rulemaking package addressing the other changes to 10 CFR 50.59 discussed in the March 24,1998, SRM. Therefore, the staff does not currently plan to pursue this concept as part of the proposed rulemaking package. However, the staff will continue to interact with NEl to obsn a better understanding of their proposal, and will report tc the Csmmission as to the desirability for adoption of a change to scope once this understanding is reached. Additionally, this subject will be a point of discussion at the June 4,1998. Commission meeting requested by NEl.

Conclusion in summary, the staff does not recommend that the Commission revise 10 CFR 50.59 as suggested in the March 24,1998, SRM concerning accidents of a different type. Further, the staff continues to believe that the criteria relating to consequences should not be revised to allow increases up to the acceptance limits without review (although the staff does plan to revise the rule to allow minimalincreases in consequences as directed by the Commission). The staff also does not recommend revising the rule language to refer to " minimal reductions in the l

margin of safety." Finally, barring a change in direction from the Commission following the l

upcoming Commission meeting, the staff is not planning rule changes that would revise the scope of 10 CFR 50.59 from the current language at this time.

The Office of the Geneial Counsel has no legal objection to this memorandum.

Unless otherwise directed by the Commission, the staff plans to prepare the proposed rulemaking package consistent with the above staff proposed alternatives (and the other direction provided in the March 24,1998, SRM).

SECY, please track.

4

Attachment:

Meeting Summary cc: SECY OGC OCA OPA OlG CFO CIO

wo oq y '-  % UNITED STATES E

)* j NUCLEAR REGULATORY COMMISSION W ASHINGTON. D.C. 20%M001

,o

• • • • • May 4, 1998 MEMORANDUM TO: Thomas H. Essig, Acting Chief Generic issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation FROM: Peter C. Wen, Project Manager +C/

Generic Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation

SUMMARY

OF APRIL 23,1998, MEETING WITH THE NUCLEAR ENERGY INSi TUTE (NEI) REGARDING 10 CFR 50.59 SCOPE ISSUE On April 23,1998, a public meeting was held at the U.S. Nuclear Regulatory Commission's (NRC's) offices in Rockville, Maryland, between representatives of the NRC, NEl and other interested parties. Attachment 1 provides a list of attendees at the meeting. Attachment 2 includes the agenda that was used for the meeting and the presentation material provided by NEl for the meeting. Attachments 3 and 4 are supplemental information provided by NEl for the discussion of safety analysis."

Before beginning discussion on the main agenda topic (scope of 10 CFR 50.S9), NEl first raised three other topics related to the Commission SRM of March 24,1998. Concerning guidance for  ;

J updating the final safety analysis report (FSAR), they indicated that they wanted to work with the staff to reconcile their draft guidance document (NEl 98-03) with the staff's draft generic letter guidance as soon as possible. The NRR staff members stated that such discussions would occur once we received Commission response to the Commission paper forwarding the

. draft GL (sent to the Commission on April 20,1998).

The second topic was the use of acceptance limits with respect to changes that result in increases in radiological consequences. NEl continues to believe that a change should not be an unreviewed safety question (USO)if the acceptance limit (such as the Part 100 guidelines),

used by the staff to judge acceptability, are still met with the change. The staff responded that it typically performed independent calculations of consequences, rather than specifically reviewing and approving the analyses (methods and assumptions) performed by the licensee.

As long as the staff's calculations confirmed that the limits were met, the staff would approve the facility design and operation. However, the degree of margin remaining to the limits might be less as viewed by the staff than by the licensee. Therefore, if a licensee subsequently made l

changes that would have the effect of increasing calculated doses up to the limits, it is possible '

that the staff conclusion would be that the limits were actually exceeded. NEl stated that this was an area that they wished to explore further such that increases above the SAR calculated values could be allowed without always requiring NRC review. They noted a specific example I

Acan kmW" n m  !

3M L

e T. Essig of a recent enforcement action where the et ange was from 22 Rem to 23 Rem (the limit was 30 Rem), as a case that should not have requi ed prior review (and thus which should not have been a violation because it did not).

The next topic was the issue of enforcement discretion. NEl stated their conclusion that enforcement policy changes should be made immediately such that no enforcement action is taken for circumstances that are clearly not safety significant, in order to achieve stability. The staff indicated its plans to continue to exercise discretion with respect to severity levels or issuance of civil penalties (under existing enforcement policy), pending further interaction with the Commission on enforcement policy changes.

Finally, NEl stated that as part of upcoming rulemaking on 10 CFR 50.59 criteria, the NRC should address the scope of changes that require evaluation directly in the regulation, rather thun indirecuy through the FSAR. Specifically, they would redefine the changes requiring evaluation against the USO criteria to be those that affect safety analyses. They would propose to include in the rule a functional definition of " safety analyses" (see preliminary thoughts in Attachment 3), referring to analyses performed pursuant to Commission requirement, or requested to validate conformance with requirements, or other analyses that are approved by NRC (by issuance of safety evaluation reports). They would supplement the definition with lists of such analyses in a guidance document. A draft outline of how such safety analyses and changes affecting them might be characterized was distributed at the meeting (see Attachment 4). The staff stated that it would consider this proposal but noted that this could not be done on the July 1998 schedule for the proposed rule established by the SRM. Further, the staff emphasized that even if such a rule change were pursued, there is still the need for licensees to update their FSARs to be complete and accurate in accordance with 10 CFR 50.71(e). Further meetings with NEl are anticipated to discuss FSAR update guidance and other issues.

Attachments: As stated cc w/atts: See next page I

NRCINEI MEETING ON 10 CFR 50.59 ISSUE LIST OF ATTENDEES April 23,1998 NAME ORGANIZATION David Matthews NRC/NRR/DRPM Tom Essig NRC/NRR/DRPM i

Frank Akstulewicz NRC/NRR/DRPM Eileen Mckenna NRC/NRR/DRPM Peter Wen NRC/NRR/DRPM Geary Mizuro NRC/OGC Cornelius Holden NRC/OCM/GID Briari Holian NRC/OCM/SAJ Tony Hsia NRC/OCM/NJD Ken Hart NRC/SECY Tony Pietrangelo NEl Steve Floyd NEl Doug Walters NEl Russ Bell NEl Bechtel j Nancy Chapman '

Herb Fontecilla VAP/APS Charlie Brinkman ABB-CE Jerry Dosier NUS Info Services Jenny Weil McGraw Hill Roben Vondrasek PSE&G l' Sam Crowley Winston & Strawn

)

1 Attachment 1 {

l 1

NEI Licensing Issues Meeting with NRC April 23,1998 V'

Agenda .

= FSAR Update Guidance

= Acceptance Limits on Consequences

= Enforcement Discretion related to USQ Determinations a Scope of10 CFR 50.59 PgE I Attachment 2 2

D FSAR Update Guidance Objective: Mutually acceptable guidance for utilities ASAP

. Most effective to interact now to

. reconcile industry and NRC draft guidance, per SRM

+ then publish result (revised NEl 98-03) for public comment V'

Status of NEl 98-03 .

Distributed for industry comment last November

= No major comments received

= NEI is ready to work with NRC staff now to reconcile with draft GL 7'

2

I Acceptance Limits

= NRC position in Jan. 9 letter to NEI 1

l = Example of the problem

= SPl.1 requests staff to reassess position l

7' Enforcement Discretion .

. No enforcenient action should be

.taken during the period prior to the '

rule change in circumstances that are clearly not safety significant

= Enforcement policy change should be instituted before July 10 QEI 3

l Purpose of 50.59 1

= Require licensee review of proposed changes

= Determine if change exceeds previously approved design or operational lines

= Require prior NRC approval if any authorized limit is exceeded QEl T.

Clarifying the Scope of 9 50.59 Principles

= 50.59 is just one part of a hierarchy of plant change processes

= FSAR is neither appropriate or efficient as the scope of e 50.59 TE3 l 4

d REGULATORY OVERSIGHT OF PLANT CHANGE CONTROL PROCESS Proposed Change S**' *.nemption '

y, _

per 10 CFR 50.12 Meets 2 Regulations or

Stop Yes Amend License No per 10 CFR 50.90 Meets -

or Operating License? -

Stop Yes 1P Seek Amendmentto -

No Order per 10 CFR 2.202 Meets -

or Orders?

- Stop Yes ,,

I i

Affects Yes  :

Apply Safety Analysis? 10 CFR 50.59 I

J No Change to QA, EP. Process per Security Plan?

10 CFR 50.54 (a),(p),

or(q)

No ,

Apply NEl y,,

Commitment Change to Management Commitments?

Guideline Proceed With Change Update hSAR NO <

No Regulatory ~ per 10 CFR 50.71(e)

Interaction Required 1.

L_________.___.---_______- - - - - - - - -- -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - -

i l

Why change Q 50.59('aX1)?

= Too many safety evaluations oflittle or no safety / regulatory value

= Address scope of 50.59 directly in the regulation, not indirectly via the FSAR

= Improve consistency between rule and implementation v'

I What are the benefits? .

= Clarify the appropriate role and focus of 50.59

= Avoid the need for extensive changes to FSARs, including removal or reformatting of information

= Avoid assigning roles to the FSAR and 50.71(e) for which they are not well suited e Address concerns about small vs. big FSARs i e Facilitate use of acceptance limits criterion for  ;

l evaluating the effect of changes on consequences V'

l l

5

v

,g Why now?

= Convergence of 50.59 and FSAR update issues

= Scope issue recognized by industry, NRC staff and Commission

= Include with Q 50.59 rule changes -- the first in 30 years - - planned for 1998

= More efficient and coherent to address Section a(1) changes in conjunction with other 50.59 changes and FSAR update guidance

't" '

Why Safety Analyses?

= Final exam of NRC safety review - -

principal basis for NRC safety approval l

= Provide a nexus to protection of public health and safety

= Encompass design bases

= Only context that makes sense for (a)(2) criteria byI 6

How would it work?

= Identify safety analyses

. from NRC requirements

. other analyses approved by SER

= Idt ?tify explicit inputs, assumptions, etc.

s Identify mitigating equipment and operator actions credited

= Changes that do not affect analyses would screen out

't* '

i .-

I .

Summary i

= 50.59 enforcement discretion ASAP i

= Work with NRC staff on

. reconciling draft FSAR update guidance ,

f . { 50.59 scope issue

. reconciling staff comments on NEI 96-07 QEI 7

Proposed Changes to NEl 96-07 Include a definition of Safety Analysis SAFETY ANALYSIS A safety analysis is an analysis that is performed pursuant to Commission requirements or requested by NRC to validate compliance with existing requirements, and is necessary to demonstrate the integrity of the reactor coolant pressure boundary, the capability to shutdown the reactor and maintain it in a- safe shutdown condition, and the capability to prevent or mitigate accidents sat could n sult in potential offsite exposures.

Safety analyses include:

analyses included in the FSAR and approved by the Commission as part ofinitial licensing analyses performed pursuant to new or amended Commission regulations subsequent to initial licensing analyses performed in response to a generic or plant specific issue to validate compliance with existing requirements analyses specifically approved by the NRC via SER Note: When a new analysis or change to plant or procedures "affects"one or more safety analyses, the safety analyses should be updated to reflect the change to maintain an accurate baseline for evaluation offuture changes.

Safety analyses do not include:

detailed calculations ana other non-docketed analyses performed in support of safety analyses environmental, firancial and other analyses unrelated to nuclear safety docketed information controlled by other regulations (QA, EP, Security) analyses submitted to the NRC in response to generic communications that do not affect analyses required to support initial licensing or demonstr ate compliance with new or amended regulations (Note: required analyses should be updated to reflect the effects of other changes, analyses or issues.)

analyses provided in LER or NOV responses except as required to demonstrate compliance with NRC regulations; the effects of such analyscs should be incorporated in the UFSAR in a subsequent update Attachment 3

DRAFT Identification of Safety Analyses Safet) Analysis Basis for NRC Safet) Analysis SER or other NRC Approsal Requirement Reference

1. General GDC I 2. Decrease in FW Temperature GDC 10.15.26

3. Increase in FW Flow GDC 10,15,26

4. Increase in Steam Flow GDC 10,15,26

5. Inadvertent Steam Generator Safety or GDC 10,15,26 Relief Valve Opening (PWR)

6. Steam System Pipmg Failure Inside and GDC 27,28,31, Outside of Containment (PWR) 35,10 CFR 100

7. Loss of External Load GDC 10,15,26

8. Turbine Trip GDC 10.15,26

9. Loss of Condenser Vacuum GDC 10,15,26

10. Loss of Non-emergency AC Power to GDC 10,15,26 the Station Auxiliaries

11. Loss of NorTnal FW Flow GDC 10,15,26

12. FW System Pipe Breaks inside and GDC 27,28,31,

. Outside Containment (PWR) 35,10 CFR 100 13, Loss of Coolant Flow including Pump GDC 10,15,26 Trip

14. Reactor Coolant Pump Rotor Seizure GDC 27,28,31, 10 CFR 100

15. Reactor Coolant Pump Shaft Break GDC 27,28,31, 10 CFR 100

16. Uncontrolled Rod Withdrawal from a GDC 10,20,25 Suberitical or Low Power Condition

17. Uncontrolled Rod Withdrawal at Pow er GDC 10,20,25 1 A. Control Rod Misoperation (System GDC 10,20,25 Malfunction or Operator Error)

19. Startup of an inactive or Recirculation GDC 10,15,20.

Loop at an incorrect Temperature 26,28

20. CVCS Malfunction that Results in a GDC 10,15,26 Decrease in the Boron Concentration in the Reactor Coolant (PWR) 21, Inadvertent Loading and Operation of a GDC 13, Fuel Assembly in a improper Position 10 CFR 100 Spectrum of Rod Ejection Accidents GDC 28, 22.

(PWR) 10 CFR 100 Inadvertent Operation of ECCS GDC 10,15,26 23.

24. CVCS Malfunction that increases GDC 10,15,26 Reactor Coolant Inventory (PWR)

25. Inadvertent Opening of a PWR Pa. GDC 10,15,26 Relief Valve or a BWR Relief Valve

26. Radiological Consequences of the GDC 55, Failure of Small Lines CarTying PWR 10 CFR 100 Primary Coolant Outside Conta'mment

27. Radiological Consequences of a Steam 10 CFR 100 Generator Tube Failure (PWR) l l

Attachment 4

l

28. LOCAs Resulttng from Spectrum of 10 CFR 50.46.

Postulated Piping Breaks within the App. K, GDC 35.

Reactor Coolant Pressure Boundary 10 CFR 100

( 29. Radioactive Liquid Waste System Leak

or Failure (Release to the Atmosphere) l- 30. Radioactive Gas Waste System Leak or Failure i 31. Postulated Radioactive Release due to GDC 60,

Liquid Containing Tank Failures . 10 CFR 20

32. Radiological Consequences of Fuel GDC 61, l

l Handling Accidents 10 CFR 100

33. Spent Fuel Cask Drop Accidents GDC 61, 10 CFR 100

34. Containment Analysis GDC 50

! 35. Power Uprote Analysis NA

36. Temperature Effects on PWR Level IEB 79-21 Measurements

37. Analysis of a PWR MSL break with IEB 80-04 Continued Feedwater Addition L

38. MOV CMFs during Transients due to IEB 85-03

! Improper Switch Settings

39. Pressurizer Surge Line Thermal IEB 8811 Stratification in PWRs

40. Seismic Qualification Of Auxiliary GL 81 14 Feedwater Systems

41. Resolution of Gl A-30, Adequacy of S- GL 91-06 R DC Power Supplies,10 CFR 50.54(O

42. Reactor Vessel Structural Integrity GL 92 01 l 43. WEC Rod Control System Failure and GL 93-04 Withdrawal of RCCAs,10 CFR l 50.54(D l 44. Equipment Operabihty/ Containment GL 96-06 Integrity under DBA Conditions

45. Assurance of Sufficient NPSH for ECC GL 97-04 and Containment Heat Removal Pumps 46, Anticipated Transients Without Scram 10 CFR 50.62, GDC 10,15,26, 27,29

47. Pressurized Thermal Shock . 10 CFR 50.61 l 10 CFR 50.63 l 48. Station Blackout

! 49. Fire Protection Appendix R i 50. Environmental Qualification 10 CFR 50.49

! 51. TMI Items 10CFR 50.34(0 l

L _ ._ --- -__-- -- ---__-- _ -_-_

4 DRAFT Generic Communications That May Have Led To New Analyses Bulletins Bulletta

• Title Comment

1. IEB 96 03 Potential Plugging of Emergency Core Coohng Suction Stratners by BWR Debris in BWRs

2. IEB 96 02 Movement of Heavy leads Over Spent Fuel. Over Fuelin the Reactor ALL Core. or Over Safety Related Equipment

3. IEB 96 01 Control Rod Insertion Problems WESTINGHOUSE

4. IEB 93 02 Debns Plugging of ECCS Suetion Strainers ALL

5. IEB 90 02 Loss Of Thermal Martin Caused By Channel Box Bow BWR

6. IEB 89 03 Potential Loss Of Required Shutdown Margin Dunng Refuehng PWR Operations

7. IEB 8811 Pressunzer Surge Line Thermat Stratt6 cation PWR

8. IEB 88 08 Thermal Stresses In Piping Connected To Resetor Coolant Systems ALL 9 IEB 88 07 Power Oset11ations In Bothng Water Reactors (BWR) BWR

10. IEB 8e-04 Potential Safetv Related F-p loss ALL

11. IEB 88-02 Rapidly Propagating Fatigue Cracks In Steam Generator Tubes WESTINGHOUSE

12. IEB 85 03 Motor Operated Valve Common Mode Failures Dunng Plant ALL Transients Due To lmproper Switch Settings

13. IEB 84-03 Refuehnr Cavity Water Seal ALL

14. IEB 83 07 Apparently Fraudulent Products Sold By Ray Miller. Inc. ALL

15. IEB 8102 Failure Of Gate Type Valves To Close Against Differential Pressure ALL

16. IEB 80 23 Fatlures Of Solenoid Valves Manufactured By Valcor Ergineenng ALL Corporation

17. IEB 8018 Maintenance Of Adequate Mtnimum Flow Through Centnfugal PWR Charging Pumps Following Secondary Side High Enerrv Ltne Rupture

18. IEB 8017 - Failure Of Control Rode To insert Dunne A Scram At A BWR BWR

19. IEB 8016 Potential Misappheation Of Rosemount Inc., Models 1151 And 1152 ALL Pressure Transmitters With Either "A" Or "B" Output Codes

20. IEB 8011 Masonrv Wall Design ALL IEB 80-07 BWR Jet Pump Assembly Fatlure BWR 21

22. IEB 80-04 Analysts Of A PWR Main Steam Line Break With Continued PWR Feedwater Addition

23. IEB 79 27 Loss Of Non Class IE Instrumentation And Control Power Systems ALL Bus Dunnr Operation 24 IEB 79-21 Temperature Effects On Level Measurements PWR

25. IEB 7914 Seismic Analysis For As Built Safety Related Piping Systems PWR

26. IEB 7912 Short Penod Scrams At BWR Faethties BWR

27. IEB 79 07 Seismic Stress Analysis Of Safety Related Piptnr ALL ,

I

28. IEB 79 02 Pipe Support Base Plate Designs Ustng Concrete Expansion Anchor ALL Bolts 29 IEB 79-01 Environmental Quah6 cation Of Class le Equipment ALL

DRAFT Generic Communications That May Have Led To New Analyses Generic Letters Comment j Genene Ltr Title I #

GL 97 04 Assurance Of SufEcient Net Positive Suetiora Head For Etnergency ALL 1.

Core Coohng And Containment Heat Removal Pumps GL 96 06 Assurance Of Equipn.ent Operabthty And Containment Integnty ALL 2.

Dunng Design Basis Accident Conditions 3 GL 96 04 Loraflex Degradation In Spent Fuel Pool Storage Rach ALL

4. GL 95-07 Pressure kcking And Thermal Binding Of Safety. Related Power. ALL Operated Gate Valves

5. GL 95 03 Circumferential Cracking Of Steam Generator Tubes PWR Intergranular Stress Corrosion Cracking Of Core Shrouds in Bothog BWR

6. GL 94 03 Water Reactors GL 93 04 Rod Control System Failure And Withdrawal Of Rod Control Cluster WESTINGHOUSE 7.

Assembbes.10 CFR 50 54(F)

8. GL 92 04 Resolution Of The issues Related To Reactor Vessel Water Level BWR Instrumentation in BWRs Pursuant To 10 CFR 50 54(F) 9 GL D2 01 Reactor Vessel Structural Interntv ALL

10. GL 9106 Resolution Of Geuenc lasue A 30," Adequacy Of Safety Related DC ALL (No direct Power Supphes." Pursuant To 10 CFR 50 54(F) response requiredi

11. GL 89 21 Request For Information Concerning Status Ofimplementation Of ALL Unresolved Safety Issue (USI) Requirements 12 GL 8910 Saferv Related (1) Motor Operated Valve Testing And Survedlance ALL

13. GL 88 20 Individual Plant Examination Of External Events For Severe Accident ALL Vulnerabihties

14. GL 8814 Instrument Air Supply System Problems Affecting Safety Related ALL Equipment 15 GL 88 01 NRC Position On IGSCC In BWR Austenitic Stainless Steel Piping BWR Nureg 0737, item li.K.3.44, Evaluation Of Anticipated Transients BWR (Referencing

16. GL 8132 Combined With Single Fatlure BWROG response to NUREG 0737 ilk.3 44

17. GL 8120 Safety Concerns Associated With Pipe Breaks In The BWR Scram BWR System

18. GL 81 14 Seistoie QuahEcation Of Aux 1harv Feedwater Systems PWR

19. GL 81 12 Fire Protection Rule (45 F/R 76602, November 19,1980) ALL (Licensed pnor :o 1/1/79) 20 GL 81-07 Control Of Heavy bads ALL

21. GL 78 09 Multiple Subsequent Actuations Of Safety /RebetValves Following An BWR Isolation Event

lIll)l jl ;ji1 l iI e e g

_ , g

. a a r

_ v r e b

5 i

t s

n le s

e e ies i

s - d ic m m s

s C A um e e

r a leS s

T eWoa ickn r dknn oa i

u q I

/

iM d, n h

i DRBT4cT R

e I I oo oi

. e cl t e e e n iF a e9 g m.r s ep a e r

T1 m50 n is e

DESFS O Qireie h t a5R w

y o Er np a

r dt x

• e

• e

• uF rE oS e ta d i

t t cn e wa gt s sC $md I n e n ns iei g

ie n

ie t e r a sp W10 ip yaF g t t knxaouhna e

SCACan r Rn e i

g t u i

x x Mfo s r i q

e T h11 lu i

r u

l e e e e e *e ME e t a wv er re l oe g

h n r a e puF s R w mst A l

i k r pto e r r Pl we e e

er aS n ow S P x x h P(ohv OPCI noo s R n

+ * * * . i o

t p

e p t r

ut pr n nme m ns e n e taioe T u mio r

u s p r

ePlT s ni t x s

e mw1 5 S s iadn x

r ek! W R A t s

PTt I n o s

a d C oC M g *

• e *

• n r

oi n pg a

e/

a g V lu t s r o

r sesudqui Pe u

p mta a r x i b e e a u r i.

P1 l n t S e n

a inh m i G

gr t s

aw r

MIn

• =

e  !

C, n ny s

y r

e wit- l s

',d oi a ei c t c r s

m. l;nh tuae n oy x x MfSR A ChP

• .

• e . t y

~ ";,, e

, er p '" f s e ul ws e n e ['l r

a t i

osr v P e e. T S i m

x e r S

,Pt L x x -

cPPRZ oinC S C

R

. . = . *,

e r

t u

s a r

e s e s y p i l

a mw e o s

y n Tl F la A . . . s - -

n y e A t

e W

F FW s

ly t n

f n a e a i in n m S e s e A a in e es r a A t a

'X n c o hrn I I X

I C 8

l . 2 2

M

~

E!! ' l

Nuclear Energy Institute Project No 689 cc: Mr. Ralph Beedle Ms Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy institute Nuclear Energy Institute Suite 400 Suite 400 17761 Street, NW 1776 i Street, NW Washington, DC 20006-3708 ,

Washington, DC 20006-3708 l Mr. Alex Marion, Director I

Programs Nuclear Energy institute Suite 400 4 1776 i Street. NW  !

l Washington, DC 20006-3708 Mr. David Modeen, Director Engineering )

j Nuclear Energy Institute Suite 400 1776 i Street, NW Washington, DC 20006-3708 ,

l i

Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400 1776 i Street, NW >

Washington, DC 20006-3708 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities )

Nuclear and Advanced Technology Division l Westinghouse Elec'*ic Corporation P.O. Box 355 j Pittsburgh, Pennsylvania 15230 {

Mr. Jim Davis, Director Operations i

Nuclear Energy Institute Suite 400 1776 i Street, NW  ;

Washington, DC 20006-3708 l

l 1

_ _ - _ _ _ _ _ _ _ _ _ _ - _ _