Text

PR-050 and 072 - 64FR36291 - Reporting Requirements for Nuclear Power Reactors
	ML23152A136
Person / Time
Issue date:	07/06/1999
From:	Annette Vietti-Cook; NRC/SECY
To:	;
References
	PR-050, PR-072, 64FR36291
	Download: ML23152A136 (1)
	v • d • e

ADAMS Template: SECY-067 DOCUMENT DATE: 07/06/1999 TITLE: ,PR-050 AND 072 - 64FR36291 - REPORTING REQUIREMENTS FOR NUCLEAR POWER REACTORS CASE

REFERENCE:

PR-050 AND 072 64FR36~91 ,

KEYWORD: ' RULEMAKING COMMENTS I

~ I Document Sensitivity: Non-sensitive - SUNSI Review Complete

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Docket No.: PR-050 and 072

10/30/2000 FR Cite: 64FR36291 In the Matter of Reporting Requirements for Nuclear Power Reactors Comment Comment Docket Document Miscellaneoll8 Accession Number Submitted by Representing Date Date Description Number

- 07/01/1999 06/25/1999 Federal Register Notice -

Proposed Rule M.A. Krupa Entergy Operationa, Inc. 08/24/1999 08/20/1999 Director- Nnclear Safety and Llc.

2 R. John Oianfranccsco, Jr Florida Power & Ught 09/14/1999 09/07/1999 Company

3 Otto L. Maynard Wolf Creek Nnclear Ope.rating Corporation 09/17/1999 09/13/1999 President & CEO 4 James W. Davis Nnclear Energy Institute 09/20/1999 09/17/1999 Duector 1

Docket No.: PR-050 and 072 10/30/20(){)

FR Ciu: 64FR36291 In the Matter of Reporting Requirements for Nuclear Power Reactors Comment Comment Docket Document Miscellaneous Accesmon Number Submitted by Representing Date Date Description Number John H. Mueller Niagar Mohawk Power 09t.m'l999 09/17/1999 Sr. VP&CNO Corporation 6 S. K. Grunbhir Omaha Public Power 09/20/1999 09/17/1999 District 7 Ted C Feigenbaum North Atlantic Energy 09/2CV1999 09/17/1999 Execullve VP & CNO Service Corporation

8 Sheny L Berohoft Flonda Power Corporation 09/20/1999 09/14/1999 Dtrcctor 9 I. F Aloxande.r Ptlgrim Nuclear Power 09!.m'l 999 09/17/1999 Director Statioo/Entergy Nuclear Gcnemtion Company 2

Docket No.: PR-050 and 072 10/30/2000 FR Cite: 64FR36291 In the Matter of Reporting Requirements for Nuclear Power Reactors Comment Comment Docket Document Miscellaneous Accession Number Submitted by Representing Date Date Description Number H L Sumner Southern Nuclear 09/20/1999 09/20/1999 Vice President Ope.rating Company 11 Steven A Toelle Uruted States Ennchment 09/20/1999 09/20/1999 Corporation 12 A Edward Schera- Southern Califorrua 09/21/1999 09/20/1999 Edison 13 Mark J Burzynski Tennessee Valley 09/22/1999 09/17/1999 Authority 14 John R. Caves Carolma Power & l.J.ght 09/22/1999 09/20/1999 Company 3

Docket No.: PR-050 and 072 10/30/2000 FR Cite: 64FR36291 In the Matter of Reporting Requirements for Nuclear Power Reactors Comment Comment Docket Document Miscellaneous Accession Number Submitted by Representing Date Date Description Number

-15 Larry Newman Nebraska Public Power 0912'2/l 999 09120/1999 District 16 James Knubel New Y ode Power 0912'2/l 999 09120/1999 Sr. VP&CNO Authonty 17 James H McCarthy Virgima Power 0912'2/1999 09121/1999 18 Nutban L Ho.slcell Consumers F.nergy 09/23/1999 09/17/1999 Director, l.Jcensing Company 19 Warren A Witt Union Electnc 09/23/1999 09f2'2/1999 4

DlJCket No.: PR-050 and 072 10/30/2()()(}

FR Cite: 64FR36291 In the Matter of Reporting Requirements for Nuclear Power Reactors Comment Comment Docket Document Miscellaneous Acce88ion Number Submitted by Representing Date Date Description Number

.20 R. P. Necct Vice President Millstone Nuclear Power 09/'24/1999 09/'23/1999 Station/Nortbeast Nuclear Energy Company 21 LawrenceF. Womack Pacific Gas and Electnc 09m/I999 09/'20/1999 Vtce President Company 22 C Lance Torry TXU Electric 09/'27/1999 09/20/1999 Sr. VP&PNO 23 Kenneth E. Pevelec lES Utilities, Inc. 09/27/1999 09/'20/1999 24 Charles H Cruse Baltlmore Gas and 09f27/1999 09/22/1999 Vtce President Electric Company 5

Docket No.: PR-050 and 072 10/30/2000 FR Ciu: 64FR36291 In the Matter of Reporting Requirements for Nuclear Power Reactors Comment Comment Docket Document Miscellaneous Accession Number Submitted by Representing Date Date Description Number

-25 Mork V McKeown Northern States Powec 09/29/1999 09/20/1999 Company 26 W Glenn Warren B011ing Water Reactor 10/04/1999 09/17/1999 Chairman Owners' Group (BWROG) 27 Richard M Fry North Carolina Division 10/21/1999 10/12/1999 of Radiation Protection 10/20/2000 10/18/2000 fRN-Fmal Rule 6

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[ lolfFR.3~;L qI) I ,-1 [7S9d-01-P]

NUCLEAR REGULA TORY COMMISSION 10 CFR Parts 50 and 72 RIN 3150-AF98

AD Reporting Requirements for Nuclear Power Reactors and Independent Spent Fuel Storage Installations at Power Reactor Sites AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending the event reporting requirements for nuclear power reactors to reduce or eliminate the unnecessary reporting burden associated with events of little or no safety significance. This final rule continues to I

- provide the Commission with reporting of significant events where Commission action may be needed to maintain or improve reactor safety or to respond to heightened public concern. This final rule also better aligns event reporting requirements with the type of information NRC needs to carry out its safety mission, including revising reporting requirements based on importance to risk and extending the required reporting times consistent with the time that information is needed for prompt NRC action. Also, NUREG-1022, Revision 2, "Event Reporting Guidelines, 10 CFR 50.72 and 50.73," is being made available concurrently with the amendments.

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f DATES: The final rule is effective [INSER'f DA'l'E 9@ DAYS A ~ ~ ~ ~ I I <

FEDERAL REGISTER].

ADDRESSES: Documents related to this action may be examined, and/or copied for a fee, at the NRC's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Documents created or received at the NRC after November 1, 1999 are also available electronically* at the NRC's Public Electronic Reading Room on the Internet at http://www.nrc.gov/NRC/ADAMS/index.html. From this site, the public can gain entry into the NRC's Agencywide Document Access and Management System (ADAMS),

which provides text and image files of NRC's public documents. For further information contact the PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Dennis P. Allison, Office of Nuclear Reactor Regulation, Washington, DC 20555-0001 , telephone (301 ) 415-1178, e-mail dpa@nrc.gov.

SUPPLEMENTARY INFORMATION:

Contents I. Background II. Analysis of Comments Ill. Discussion

1. Objectives

2. Section by Section Discussion of Final Amendments 2

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3. Revisions to Event Reporting Guidelines in NUREG-1022

4. Reactor Oversight

5. Enforcement

6. Electronic Reporting

7. State Input

8. Plain Language IV. Environmental Impact: Categorical Exclusion V. Backfit Analysis VI. Regulatory Analysis VII. Paperwork Reduction Act Statement VIII. Regulatory Flexibility Act Certification IX. Small Business Regulatory Enforcement Fairness Act X. National Technology Transfer and Advancement Act XI. Final Amendments I. BACKGROUND

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The reporting requirements in Sections 50. 72 and 50. 73 have been in effect, with minor modifications, since 1983. Experience has shown a need for change in several areas. On July 23, 1998 (63 FR 39522), the NRC published in the Federal Register an advance notice of proposed rulemaking (ANPR) to announce a contemplated rulemaking that would modify reporting requirements for nuclear power reactors. Among other things, the ANPR requested public comments on several concrete proposals for modification of the event reporting rules.

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l, Public meetings were held to discuss the ANPR at NRC Headquarters on August 21, 1998, in Rosemont. Illinois on September 1, 1998, and at NRC He'adquarters on November 13, 1998.

A proposed rule was published in the Federal Register on July 6, 1999 (64 FR 36291),

including a conforming change to Section 72.216. Concurrently, a draft revision to the associated event reporting guidelines was made available for public comment (NUREG-1022, Draft Revision 2). A public meeting was held at NRC Headquarters on August 3, 1999, to discuss the proposed rule and draft guidelines. Public comments were due on September 20, 1999. Additional public meetings were held on February 25, and March 22, 2000, to discuss public comments.

II. ANALYSIS OF COMMENTS The comment period for the proposed rule expired September 20, 1999. Twenty-seven comment letters were received, representing comments from 24 nuclear power plant licensees (utilities), two organizations of utilities, and one State agency.

I In addition to the written comments received, the proposed rule was the subject of a public meeting on August 3, 1999, as discussed above under the heading "Background," and

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comments made at that meeting have also been considered.

Most commenters expressed support for amending the rules in accordance with the objectives discussed in the proposed rule. However, they objected to some of the specific provisions. Many comments also provided specific recommendations for changes to the proposed rules. The resolution of comments is summarized below. This summary addresses the principal comments o.e., comments other than those that are: minor or editorial in nature; 4

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supportive of the approach described in the proposed rules; or applicable to another area or activity outside the scope of sections 50.72 and 50.73}.

Comment A (Do not require reporting of degraded components): The proposed rule Included a new component reporting criterion. It would have required reporting "Any event or condition that resulted in a component being in a degraded or non-conforming condition such that the ability of the component to perform its specified safety function Is significantly degraded and the condition could reasonably be expected to affect other similar components in the plant" The term "significantly degraded" was defined by providing several examples of' 4t reportable and non-reportable events. The stated purpose was to ensure that design basis or other discrepancies would continue to be reported if the capability to perform a specified safety function is significantly degraded and the condition has generic implications.

Most commenters strongly objected to the proposed component reporting criterion.

Among other things, they indicated:

(1) The proposed component reporting criterion is not needed because, after deleting the requirement to report a-condition that is outside the design basis of the plant, any significant events would still be captured by the other existing criteria.

(2) The proposed component reporting criterion would be unclear and subject to widely varying Interpretation with regard to the meaning of the term "significantly degraded" and the term "could reasonably be expected to apply to other similar components."

(3) The proposed component reporting criterion would be overly burdensome. For example, it would become necessary to screen all single component failures for reportability.

(4) The proposed component reporting criterion would be contrary to the stated objectives of the rulemaking. For example, it would result in many additional reports for events with little or no safety- or risk-significance.

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Response: In the final rule, the proposed reporting criterion has been retained, but modified to address the concerns about unnecessary burden and clarity expressed in the comments. It requires reporting any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdown condition; (2} Remove residuaf heat;

{3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident.

Events covered by this criterion may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to this criterion if the, event results from:

(1) A shared'dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.

Subject to the two exclusions stated above, this criterion, as modified, is needed to capture those events with enough generic significance that a single cause could have prevented the fulfillment of the safety function of multiple trains or channels, but the event

{1) Would not be captured by§§ 50.73(a)(2}(v) and 50.72(b)(3)(v} [event or condition that could have prevented fulfillment of the safety function of structures and systems needed to ... ] because the affected trains or channels are in different systems; and (2) Would not be captured by§ 50.73(a)(2)(vii) [common cause inoperability of independent trains or channels] because the affected trains or channels are either:

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0) Not assumed to be independent in the plant's safety analysis; or (Ii) Not both considered to be inoperable.

The criterion, as modified, would not be unclear because it uses the term "could have prevented fulfillment of the safety function," which is already used in a previously existing criterion.

The criterion, as modified, is not considered over1y burdensome because It is estima_ted to result In fewer reports than the previous requirement to report a condition outside the design basis of the plant It is not necessary to screen all single component failures for reportability.

The criterion, as modified, is considered consistent with the objectives of the rulemaking for the same reasons.

Comment B (Do not change the term "any engineered safety feature [ESF] ... j:, In the proposed rule, the term "any engineered safety feature (ESF), including the reactor protection r"'

system {RPS)," which defines the systems for which actuation must be reported, would be replaced by a specific list of systems. It was recognized that this proposal to list the systems in the rule was controversial and public comment was specifically invited in this area. In particular, three principal alternatives to the proposed rule were identified for comment. .They are:

Alternative 1, status quo. The rule would continue to require reporting for actuation of "any ESF." The guidance would continue to infer that reporting should include the systems on a list which is similar to the list in the proposed rule.

Alternative 2, plant-specific list The rule would require that licensees develop a plant-specific, risk-informed list 7

I Alternative 3, pre-1998 practice. The rule would continue to require reporting actuation of Many ESF . The guidance would indicate that this Includes those systems identified as ESF's in each plant's final safety analysis report (FSAR).

The comments may be summarized as follows:

(1) Most commenters objected to the proposed rule, which would replace the term "any engineered safety feature (ESF), including the reactor protection system (RPS)" with a list of specific systems. The reasons cited by the commenters include the following:

(a) Providing an all-inclusive list of systems in the rules is inappropriate.

(b) Each facilitys FSAR specifies equipment that is designated as ESF equipment 4t (c) Plant-specific differences exist in the safety-related status of their systems.

(d) The risk-significance of a particular system can vary greatly between plants, due to a wide variety of design differences. An all-inclusive list would increase the burden for some plants whose equipment on the list was not ESF equipment or equipment with a suitably high risk-significance.

(e) There are a number of specific problems with the proposed list. Specific examples were provided.

(2) Most commenters recommended in favor of Alternative 3, returning to the pre-1998 practice of reporting actuation for only those systems that are designated as ESFs in each facilitys FSAR. They stated that this option best meets the goal of clarity and simplicity.

(3) One commenter recommended in favor of Alternative 1 (status quo), where the reporting guidelines contain a list of systems similar to the list proposed for the rule. It stated that the facility's internal reporting procedures already reflect the current practice. Any benefit that might be obtained by returning to the pre-1998 practice would be so slight that It would not justify the cost of changing the procedures.

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(4) Some commenters Indicated that there are problems with the status quo that need to be solved. For example, the reporting guidelines should exclude reporting of reactor water cleanup system (RWCU) Isolations that routinely occur during system restoration following maintenance outages, due to rapid pressurization following valve opening.

(5) Most commenters objected to Alternative 2 (developing a plant-specific, risk-informed list of systems). They stated that this would require a significant expenditure of resources and it is unclear as to whether or how it would meet the NRC's needs better than Alternative 3 (returning to pre-1998 practice). They also noted that there Is a separate initiative to "risk-inform" 10 CFR Part 50. This may result in development of plant-specific lists of systems based on risk significance. However, the commenters do not believe the necessary criteria have been adequately established to make that shift as part of this rulemaking~to modify 10 CFR 50. 72 and 50. 73. They recommended that later, as part of the rule change to "risk-inform" Part 50, the NRC should evaluate whether or not it is appropriate to "risk-inform" V ESF systems subject to the event reporting requirements of 10 CFR 50. 72 and 50. 73.

Response: (1) The NRC believes providing a list of systems is the best approach because it will obtain consistent reporting of events that result in actuation of highly risk-significant systems. Consistent reporting for such events is needed to support estimating eguipment reliability parameters and is important to several aspects of the NRC's general move towards more risk-informed regulation.

Commenters stated that the risk-significance of the systems varies depending on plant design. As discussed below under the headings "(e)O}" through "(e)(vii)," a number of items have been removed from the list based on specific comments. The NRC believes that these systems remaining on the list are of sufficient risk significance to warrant reporting of a system actuation. The principal reason for reporting an actuation of one of these systems is that it is 9

indicative of an unplanned plant transient that the NRC needs to evaluate to determine if action is necessary to address a safety problem. In this context, the NRC's need to evaluate the event is independent of classificatlon of the system. For example, a valid actuation of the auxiliary feedwater {AFW) system at a pressurized water reactor (PWR) means there was a transient that involved an abnormal plant parameter, such as low steam generator level, which initiated the actuation. This is the reason the NRC needs to evaluate the event, and it Is independent of how the AFW system happens to be classified at the particular plant.

The classification of systems in the FSA Rs has evolved over the years. For example, in earlier PWR designs the auxiliary feedwater system was not considered to be an ESF, and this is reflected in early FSARs. Later, although the system's function and importance did not change, it came to be considered an ESF, and this is reflected in later FSARs. Since the function and importance is the same regardless of classification, it does not make sense to exclude reporting for actuation of the auxiliary feedwater system based on its classification in the FSAR.

Furthermore, this approach is estimated to result in a net reduction in the number of events reported under this criterion. Some licensees will make additional reports involving highly risk-significant systems. However, these additional reports will be outweighed by the elimination of reports involving systems with lesse~ risk-significance.

(a) Commenters indicated that providing an all-inclusive list of systems in the rules is inappropriate. However, the NRC does not believe the list is all inclusive. It contains only systems that are highly risk-significant and omits systems of lesser risk-significance, even if the systems of lesser risk-significance are designated as ESFs. The NRC also believes the list is appropriate because it provides consistent reporting of events that result in actuation of these highly risk-significant systems and, at the same time, a net reduction in reporting burden.

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(b) Commenters stated that each facility's FSAR specifies equipment that is designated as ESF equipment However, the NRC believes that those lists are not consistent or risk-informed. For example, at several plants, emergency diesel generators (EDGs), which are highly risk-significant, are not identified as ESFs. At several pressurized water reactors (PWRs), the AFW system which Is highly risk-significant, is not Identified as an ESF. At most boiling water reactors (BWRs), the reactor core isolation cooling (RCIC) system, which Is highly risk-significant, is not identified as an ESF. On the other hand, most plants identify systems with lesser risk-significance, such as fuel building ventilation and filtration systems, as ESFs.

(c) Commenters stated that plant-specific differences exist In the safety-related status of systems. However, the NRC does not believe that this fact bears directly on the question of which system actuations should be reported. There is no need to report the actuation of all safety related systems, and there is no reason to exclude reporting for the actuation of a non-safety-related system if it is highly risk-significant simply on the basis that it has not been classified by the licensee as an ESF.

(d) Commenters stated that the risk-significance of a particular system can vary greatly among plants. They further stated that an all-inclusive list would therefore increase the burden for some plants whose equipment on the list was not ESF equipment or equipment with a

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suitably high risk-significance. The NRC agrees with the general statement that the risk-significance of a particular system can vary greatly among plants. However, the systems on the list are virtually always of high risk-significance. While it is true that, as a result of the list, some licensees will make additional reports, any additional reports will involve systems thaf are highly risk-significant Also, these additional reports will be outweighed by the elimination of reports involving systems with lesser risk-significance. Thus, the net effect is a reduction in reporting.

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(e) Commenters provided several specific examples of items they considered to be problems with the list. These examples are:

(i) In the proposed rule, the feedwater coolant injection (FWCI) system was characterized as an example of an emergency core cooling system (ECCS). Commenters stated that FWCI systems are not considered to be ECCS. The NRC believes that clarification is ~rranted. In the final rule, FWCI is not characterized as an ECCS. However, it is included as a separate item in the list.

(ii) The proposed rule would have required reporting actuations of the RCIC system.

Commenters stated that RCIC is included in the Improved Standard Technical Specifications (ISTS) because it meets criterion 4 of 10 CFR 50.36, based on its contribution to the reduction of overall plant risk. They further stated that RCIC is not credited in the plant's safety analysis.

The NRC believes that RCIC is highly risk-significant and, therefore, it remains on the Jist In the final rule.

(iii) Commenters stated that non-reportable exceptions should be allowed for systems that are considered to be ESFs, yet have lower levels of risk significance (control room ventilation systems, reactor building ventilation systems, fuel building ventilation systems, auxiliary building ventilation systems, RWCU isolations during restoration from maintenance, etc.). The NRC agrees. The final rule eliminates unnecessary reporting for systems that are considered to be ESFs, yet have lower levels of risk significance. It also eliminates reporting for RWCU lsolation-s during restoration from maintenance because they are routine and are of low risk and safety significance.

(iv) Commenters stated that the li,st inappropriately includes "associated support s*ystems" for BWR Division 3 EDGs. The NRC agrees. In the final rule the term "associated support systems" has been eliminated for BWR Division 3 EDGs, and other EDGs as well.

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(v) Commenters stated that the list inappropriately includes station blackout diesel generators {and black start gas turbines that serve a similar purpose) that are not safety related. The NRC agrees. The final rule does not require reporting 'tor station blackout diesel generators (and black start gas turbines that serve a similar purpose).

(VI) Commenters stated that although the term *anticipated transient without scram (ATWS) mitigating systems" is clear to those licensees that have dedicated systems O.e.

AMSAC), a great deal of confusion exists for those that have no dedicated system. Due to the lack of clarity, it could be interpreted that any system that might be used during an A TWS would fall into this category (i.e. feedwater systems, berating systems, control rods, etc.).

Extensive clarification would be needed to eliminate this ambiguity. The NRC agrees that clarification is warranted. In the final rule this item has been eliminated. Reporting is *not needed for actuations for a system such as AMSAC. The reports needed for*other systems are captured by other items on the list.

(vii) Commenters stated that it is unclear as to whether the service water entry applies only to emergency service water systems (i.e., those that don't operate unless there is an accident} or also to the standby service water systems that only run to remove heat from the residual heat removal (RHR) heat exchangers. The NRC agrees. In the final rule this item has been clarified to indicate that reporting is required for emergency service water (ESW) systems that do not normally run and that serve as ultimate heat sinks. In addition, this item has been deleted from the list of systems for which telephone notification is required under section 50. 72 because an ESW actuation by itself does not indicate the type of transient that the NRC needs to evaluate. However, ESW system actuations are reportable only under section 50.73 because the information is needed to support the NRC staff's equipment reliability estimates.

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(2) As stated by commenters, Alternative 3 would provide clarity and simplicity.

However, the NRC believes that adoption of the list of systems in the final rule also provides clarity and simplicity.

(3) Although one commenter recommended in favor of Alternative 1, the NRC believes that this alternative would Invite variable interpretation. The event reporting guidelines would contain a list of systems, whereas the rule would require reporting the actuation of "any ESF."

(4) Some commenters stated that, under previous requirements it was necessary to report reactor water cleanup system isolations that routinely occur during system restoration following maintenance outages, due to rapid pressurization following valve opening. The list of systems eliminates these unneeded reports because it limits the reporting of containment isolation signals to those that affect multiple systems or multiple main steam isolation valves (MSIVs).

(5) As indicated in the comments, with respect to Alternative (2), the project to "risk-inform* 10 CFR Part 50 may, in the future, lead to development of plant-specific lists of systems based on importance to risk and, as part of that project, it may be appropriate to consider whether or riot the applicability o( this reporting criterion, as well as other reporting criteria, should be based on such lists. It is expected that at that time the criteria necessary for development of the list will have been adequately established.

Comment C (Eliminate reporting for historical events): The proposed rule would have eliminated the requirement for a telephone notification, under 10 CFR 50.72, for.

(1) "Any event, found while the reactor is shutdown, that, had it been fol;lnd while the reactor was in operation, would have resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded or being in an unanalyzed condition that significantly compromises plant safety," and 14

(2) "Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown conditions; (B) remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident" if the condition no longer exists at the time of discovery.

The proposed rule would also have eliminated the requirement for a written licensee event report (LER), under 10 CFR 50.73, for:

(1) "Any operation or condition prohibited by the plant's Technical Specifications," if the condition has not existed within three years of the date of discovery, and (2) "Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor ... : (B)*remove residual heat; (C) Control the release of radioactive material; or {D) Mitigate the consequences of an accident," if the condition has not existed within three years of the date of discovery.

With regard to 10 CFR 50. 73, public comment was specifically invited on whether such historical events and conditions should be reported (rather than being excluded from reporting, as proposed). Public comment was also invited on whether the three year exclusion of such historical events and conditions should be extended to all written reports required by section 50.73(a) (rather than being limited to these two specific reporting criteria, as proposed).

Most commenters supported the revisions to 10 CFR 50. 72 that eliminate reporting of historical events. They stated that no safety significance exists for 10 CFR 50. 72 reporting of historical events.

Most commenters also supported: (1) the elimination of written LERs for historical events for the two cases proposed; (2) extending the exclusion to all written reports required under section 50.73{a); and (3) using two years as a cutoff point, rather than three years.

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They stated that two years encompasses one refueling cycle of operation. Significant effort can be expended searching back in history for historical events. Reporting historical events more than two years old provides a low safety benefit and unnecessarily increases the reporting burden. It was recognized that three years is consistent with the time period that performance indicators are tracked under the new oversight process. However, most commenters stated that no safety significance exists for 10 CFR 50. 73 reporting of historical events which occurred more than two_ years ago.

Response: The final rule eliminates the requirement to provide a .telephone notification or a written LER for a historical event for the reasons discussed above.

The cutoff date for reporting of historical events remains at 3 years, as*was indicated in the proposed rule. The 3-year cutoff is necessary because the NRC staff tracks performance I

indicators for a period of 3 years, in order to include a refueling outage as well as an extended period of operations, which provides more stable performance indicators. The additional burden of searching back for 3 years to determine if a condition existed within three years of the date of discovery, instead of only 2 years, is minimal because this type of event is rarely identified. Thus, it is consid_ered justified in order to provide better performance indicators.

Comment D (Time limits for reporting): The proposed rule would have continued to _

require reporting within one hour after occurrence for declaration of an Emergency Class, or for deviation from the plant's Technical Specifications authorized pursuant to 10 CFR 50.54{x).

Reporting of other events that are reportable by telephone under 10 CFR 50. 72 would be reportable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after occurrence, rather than within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as was previously required. Submittal of written LERs would be required within 60 days after discovery, rather than within 30 days as previously required.

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Public comment was speciftcallY, invited on the questiqn of whether additiona1*1evels should be used to better correspond to particular types of events. For example, 10 CFR 50.72 previously required reporting within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for ,events that involve low levels of radioactive releases, and events related to safety or environmental protection that involve a press release or notification of another government agency. These types of events could be maintained at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months so that information is available on a more timely basis to respond to heightened public concern about such events. In another example, events related to environmental protection are sometimes reportable to another agency, which is the lead agency for the matter, with a different time limit, such as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . These types of events could be reported to the NRC at approximately the same time as they are reported to the other agency Most comments on the proposed rule supported the proposal to use just three basic r

levels of required reporting times in 10 CFR 50.72 and 10 CFR 50.73 (1 hour, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; and 60 days), as indicated in the proposed rule, in the interest of simplicity. They indicated that additional levels of reporting are not needed. They also agreed with the revised reporting times based on importance to risk and extending the required reporting times consistent with the need for prompt NRC action. Additionally, they noted that the increased time for submittal of LERs will allow for completion of required engineering evaluations after. event disco*very, provide for more complete and accurate LERs, and result in fewer LER revisions and supplemental reports.

One comment letter, from the State of North Carolina, recommended maintaining the required reporting time at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for:

(1) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , results in concentrations in an unrestricted area that exceed 20 times the applicable concentration specified in Appendix B to Part 20, Table 2, Column 1; 17

(2) Any liquid effluent release that, when averaged over a time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds

20 times the applicable concentration specified in Appendix B to Part 20, Table 2, Column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases; (3) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment; and (4) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.

The letter indicated that the information from such events are of interest to the public and public officials. Furthermore, the State's Division of Radiation Protection (DRP) provides independent advice to State decision-makers as part of its emergency preparedness function.

Any delay in providing the information to the DRP may prevent or delay decisions on public

. health or public announcements. S~ate agencies may be able to get the information from

licensees, even under the proposed rule. However, this can be difficult to do when an incident is actually occurring unless the NRC's rules mandate the reporting within a prompt and well-defined period of time.

Similar comments were received from the State of Illinois regarding the ANPR.

Response: After consideration of the comments, and the potential need for NRC action, the final rule employs four basic levels of required reporting times in 10 CFR 50. 72 and 10 CFR 50.73 (1 hour, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and 60 days). Although this is not as simple as using just three levels, as was indicated in the proposed rule, it allows more flexibility in matching the required reporting time to the potential need for NRC action.

18

The final rule requires 4-hour reporting, If the event was not reported In 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , for an event or situation, related fo the health and safety of the public or onslte personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials. This is the same as previously required. These reports are needeq promptly because they involve events where there may be a need for the NRC to respond to heightened public concern.

The final rule also requires 4-h0l1'r reporting, if the event was not reported in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , for e unplanned transients. These are events where there may be a need for the NRC to take a reasonably prompt action, such as partially activating its response plan to monitor the course

.J:

of the event. In summary, they are:

(a) An event that resulted or,should have resulted in ECCS discharge into the reactor coolant system (RCS) as a result of a valid signal, except when it results from and is part of a pre-planned sequence during testing or operation. Previously this was a 1-hour report (b) Initiation of a shutdown required by the plant's Technical Specifications. Previously this was a 1-hour report.

(c) A reactor scram or reactor trip when the reactor is critical, except when it results from and is part of a pre-planned sequence during testing or operation. Previously, actuation of any engineered safety feature (ESF), including the reactor protection system (RPS), was a 4-hour report Three criteria are deleted from §SO.72 because they are not needed in' order to obtain prompt notification of events. They are retained in §50.73, however, because they are needed in order to obtain written LERs. In summary, they are:

J 19

(a} A natural phenomenon or other external event that poses an actual threat to plant safety, or significantly hampers site personnel in the performance of duties necessary for safe operation. Events of this type are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(b) An internal event that poses an actual threat to plant safety, or significantly hampers site personnel in the performance of duties necessary for safe operation, including fires, toxic gas releases, or radioactive releases.. Events of this type are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(c) An airborne radioactive release, or liquid effluent release, that exceeds specific e limits. Releases that are large enough to warrant prompt notification are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the declaration.

Releases that involve a public announcement or notification to another agency are reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the announcement or notification.

For the remaining events reportable under §50.72, the final rule requires 8-hour reporting, if not reported in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . These are events where there may be a need for the NRC to take an action within about a day, such as initiating a special inspection or investigation. In summary, they are:

{a} The plant including its principal safety barriers being in a seriously degraded condition, or the plant being in an unanalyzed condition that significantly degrades plant safety.

(b} A valid actuation of any system listed in paragraph {b)(3)(iv)(B), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

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(c) An event or condition that at the time of discovery could have prevented fulfillment of the safety function of structures or systems needed to shut down the reactor, remove residual heat, control the release of radioactive material, or mitigate an accident (d) Transport of a radioactively contaminated person to an offsite medical facility.

{e) A major loss of emergency assessment capability, offsite response capability, or offsite communications capability.

Comment E (Eliminate all reporting of invalid ESF actuations): The proposed rule would have eliminated the requirement for a telephone notification, under 10 CFR 50.72, for an ESF actuation if it is an invalid automatic actuation or an unintentional manual actuation. It was stated that invalid actuations are generally less significant than valid actuations because they do not involve plant conditions (e.g., low reactor coolant system pressure) that would warrant system actuation. Instead, they result from other causes (such as a dropped electrical lead during testing).

The proposed rule would not have eliminated the requirement for a written LER for such events. It was stated that there is still a need for reporting, because the reports are used in making estimates of equipment reliability parameters, which in tum are needed to support the Commission's move towards risk-informed regulation.

Most commenters indicated that 'invalid ESF actuations should not be reported under 10 CFR 50.73 unless the actuation impacts the plant such that other reporting criteria are independently met. They stated that contrary to the NRC's expectations, reporting of invalid actuations will not provide the information needed to estimate equipment reliability parameters.

This information should be collected by other less burdensome mechanisms, such as the Equipment Performance and Information Exchange (EPIX) system and Maintenance Rule reports.

21

Response: The NRC disagrees with many of the comments. Invalid actuations do provide information needed in estimating equipment reliability because they constitute unplanned demands. The response to unplanned demands may or may not differ significantly from those of planned test demands._ Thus, in making reliability estimates, the results from unplanned demands are compared against those from planned test demands to determine whether or not it is appropriate to combine them. As indicated in the Commission Paper SECY-97-101, May 7, 1997, "Proposed Rule, 10 CFR 50.76, Reporting Reliability and Availability Information for Risk-significant Systems and Equipment,* Attachment 3, this is one of the categories of information that the NRC relies upon in order to make equipment reliability 4t estimates.

As also discussed in SECY-97-101, EPIX is a voluntary program which does not provide a break out of invalid actuations and their results. The fact that ESF actuations are reported in written LERs was one of the key factors in making the determination that the NRC could work around weaknesses in the EPIX data in order to develop reliability estimates.

Reports developed under the maintenance rule, 10 CFR 50.65, are not submitted to the NRC.

Regardless, the Commission agrees'that a reduction In unnecessary burden is warranted. Accordingly, the final rule takes the following approach:

(a) The requirement to provide a telephone notification under §50. 72 for an Invalid ESF actuation is eliminated, as was indicated in the proposed rule.

(b) The requirement to report these events under §50.73 is retained. However, the licensee is given the option of providing a telephone report rather than a written LER. This is far less burdensome. In this case, the telephone notification has the same due date as the LER would have (60 days) because the information is not needed Immediately.

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Comment F (Eliminate reporting of high pressure coolant injection (HPCI) lnoperability):

As indicated in the 1983 Statements of Considerations for 10 CFR 50.72 and 50.73, failure or inoperabillty of a single train system, such as the HPCI system in BWRs, Is considered to constitute an "event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: {A) Shut down the reactor ... ; {B)

Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident" Most commenters indicated that inoperability of HPCI does not of itself constitute a

- . condition that would prevent the fulfillment of a s~fety function. Therefore, there Is no benefit in reporting of HPCI inoperability jf it has no affect on the ability to fulfill a safety function.

BWR design considers HPCI inoperability and provides supporting systems such as reactor core isolation cooling (RCIC), Core Spray, and automatic depressurization system (ADS). This is supported by the relatively long Allowed Ou~ge Time for HPCI in the Standard Technical Specifications {i.e., 14 days). If, in the event of HPCI inoperability, it can be shown that these systems are available and capable of fulfilling the safety function without HPCI, the event should not be reportable. Reporting HPCI inoperability in these cases has no meaning_ for event reporting and appears to be solely a data gathering exercise.

Additionally, the reactor oversight process uses a performance indicator for Safety System Functional Failures based on 10.CFR 50.73 reports. These indicators count failures of single train systems (such as HPCI}, assuming that the event report documents a safety system failure. Reporting HPCI inoperability when there is no impact on the overall capability to fulfill the safety function (e.g., remove residual heat) will result in an overly conservative and detrimental assessment of-this indicator.

23

Response: As indicated in the 1983 Statements of Considerations for 10 CFR 50. 72 and 50.73, the purpose of this reporting criterion is to capture failure, inoperability, etc. on the basis of a structure or system. Thus, if an event br condition could have prevented fulfillment of the safety function of a ~ystem (i.e., by that system}, it is reportable even If other system(s}

could have performed the same safety function(s}.

Also, in its assessment of plant performance, the NRC uses a performance Indicator that includes failure or inoperability of single train systems such as HP~I. Thus, elimination of the requirement to report such events would be contrary to one of the objectives of the rulemaking - to maintain consistency with the NRC's actions to improve integrated plant performance.

Comment G (Allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after discovery for telephone reporting): Section 50.72(b)(3) states" ... the licensee shall notify the NRC as soon as practical and in all cases, within eight hours of the occurrence of any of the following: ..... " The comment letter states that this should be revised to say" ... the licensee shall notify the NRC as soon as practical and in all cases, within eight hours of th'e occurrence or discovery of any of the following: .... " The addition of the term "or discovery" provides for those events that are discovered to have occurred in the past, remained undetected for sometime, and presently exist Response: The NRC disagrees. Addition of the term "or discovery," as suggested by the comment, is not necessary. As they have in the past, the reporting guidelines address those limited cases, such as discovery of an existing but previously unrecognized condition, where it may be necessary to undertake an evaluation in order to determine if an event or condition is reportable. In other cases, where telephor:ie reporting is required, the event should be reported as soon as practical and not later than the specified time limit.

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Comment H (Eliminate telephone reporting for non-critical scrams): Most commenters recommended that telephone reporting of RPS actuation (reactor scrams) be limited to those occurring from a critical condition.

Response: The NRC partially disagrees. A valid scram, even from a subcritical condition, is indicative of an event with enough significance that the NRC should screen and/or review it on the day it occurs, rather than waiting for submittal of a written LER. However, telephone reporting under section 50. 72 has been eliminated for invalid scrams from a subcritical condition.

_Comment I (Umit reporting to conditions that do prevent fulfillment of a required function): Regarding section 50.72(b)(2)(v), which indicates that licensees shall report: "Any event or condition that at the time of discovery could have prevented the fulfillment of tl)e safety function of structures or systems that are needed to: ... ," this should be revised to read as follows: "Any event or condition that at the time of discovery is preventing the ability to fulfill the safety function of structures or systems that are needed to: ... "

This change is required to reflect the correct tense of the existence of an event or condition, rather than past speculation. Because of past confusion pertaining to the interpretation of this area, it is suggested that further discussion be included in the statements of consideration explaining that "is preventing" represents actual conditions and does not imply that further failures should be speculated.

Response: The NRC does not agree. The term "could have prevented" reflects the meaning of the rule. It means that, at the time of discovery, the condition could have prevented fulfillment of the function (for example, had there been a demand for the function).

This includes but is not limited to the case where, at the time of discovery, the condition is actually preventing fulfillment of the function.

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a, ,. * .,

This Statement of Considerations and the reporting guidelines indicate that. in evaluating reportability under this criterion, it is not necessary to postulate an additional random single failure.

Comment J (Human performance data in LERs): Section 50.73(b)(2)(ix)(J) previously required that the narrative section of an LER include the following specific. information as appropriate for the particular event

"(1) Operator actions that affected the course of the event, including operator errors, procedural deficiencies, or both, that contributed to the event (2) For each personnel error, the licensee shall discuss:

(Q Whether the error was a cognitive error (e.g., failure to recognize the actual plant condition, failure to realize which systems should be functioning, failure to recognize the true nature of the event) or a procedural error, OQ Whether the error was contrary to an approved procedure, was a direct result of an error in an approved procedure, or was associated with an activity or task that was not covered by an approved procedure; (iii) Any unusual characteristics of the work location (e.g., heat, noise) that directly contributed to the error, and (iv) The type of personnel involved (i.e., contractor personnel, utility-licensed operator, utility non-licensed operator, other utility personnel)."

The proposed amendment would have char.tged section 50.73(b)(2}0i)(J) to simply state: "For each human performance related problem that contributed to the event. the licensee shall discuss the cause(s) and circumstances." It was stated that the current rule is more detailed than necessary. Details would continue to be provided!" the reporting guidelines, as indicated in section 5.2.1 of the draft of Revision 2 to NUREG-1022.

26

Most commenters recommended that, instead of adopting the wording In the proposed rule, section 50.73(b)(2)(i~(J) be revised to state: "For each root cause personnel error, the licensee shall discuss the cause(s) and circumstances." They stated that the shift from apersonnel error" and the implied aroot cause" to ahuman performance related problem" and "contribciting factors" would greatly increase the scope of investigation and burden to the licensee. They also stated that it is only appropriate to require discussion of personnel error root causes.

Response: The intent of the proposed change was to clarify the requirements, not to expand them. Accordingly, the final rule states "For each human performance related root cause, the licensee shall discuss the cause(s) and circumstances." This limits the requirement to discussion of root causes of the event It would not be appropriate, or consistent wi~ the previous requirement discussed above, to limit the requirement to discussion of personnel error root causes, as opposed to procedural deficiency root causes, for example.

)

Comment K (Do not require additional availability data in LERs): Section 50.73(b)(3) l requires that the assessment of safety consequences in an LER include the availability of systems or components that could have performed the same functions as systems or ,

components that failed during the event. Proposed section 50.73(b)(3)0~ would add a requirement that the assessment also include the availability of systems or components that "Are included in emergency or operating procedures and could have been used to recover from the event in case of an additional failure in the systems actually used for recovery."

Most commenters objected to this new provision, on the grounds that it adds significant burden without adding value. They stated that reporting should be based on existing plant conditions. Emergency operating procedures provide direction for use of many plant systems.

If additional failures must be postulated, multiple systems would be required to be included in 27

r- . .

the LER for each safety function. There exists an infinite combination of failures that could be postulated. This unbounded requirement would result In a large amount of additional information that would be of minimal use. The assessment of the safety consequences and implications of the event would become cluttered with hypothetical additional failures and possible plant responses. Some commenters stated that the proposed requirement would require licensees to speculate on actions that could have been taken, and It would add significant burden with no added value.

Response: The purpose of the proposed change was to ensure that LERs contain sufficient information to support a risk assessment of the event Usually there is enough information, or there is nearly enough information and the NRC staff can telephone the licensee to obtain any additional information needed. Section 50)3(b)(2)(6) requires that LERs include "The name and telephone number of a person within the licensee's organization who is knowledgeable about the event and can provide additional information concerning the event and the plant's characteristics. Further, Section 50.73(c) provides that the NRC may require submittal of additional information if necessary for complete understanding of an unusually complex or significant event.

However, for those events that occur when the plant is shutdown, it hi;!s been difficult to obtain enough information because it cannot be assumed that equipment that is normally operable and available during operation is available during plant shutdown. Accordingly, in the final rule there is a requirement for additional availability information. To eliminate unnecessary burden, the requirement for additional availability data is limited to shutdown events. Also, it is revised to simply require providing the availability of systems needed to shut down the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate an accident. -This will eliminate potential difficulties 28

I in deciding what combinations of failures should be postulated for the purpose of deciding which systems to address.

Comment L (The rule should stand alone): Licensees must use both the rule and NUREG-1022, Rev. 2, to determine reportability of conditions. The rule should_ be a stand-alone document written simply enough to be understood without the need for a 100+

page guidance document.

Response: The NRC does not agree that it is necessary to eliminate the detailed event reporting guidelines and/or include a similar level of detail in the rule. Generally speaking, the rule language cannot be precise enough to cover all the situations that might be governed by the rule and require clarification. Furthermore, in response to the ANPR, most commenters expressed the need for timely guidance on the final rule. Finally, the NRC has reviewed the guidelines and modified them where necessary to ensur:e they are consistent with the final rule.

Comment M (The terms "significant" and "serious" are not defined in the rule): One commenter stated that the terms Hsignificantly affects" and Hseriously degraded" are not defined anywhere in the proposed rule.

Response: The NRC does not agree that it is necessary to define these terms in the rule. The term "unanalyzed condition that significantly affects plant safetyt which was used in the proposed rule, is changed to "unanalyzed conditipn that significantly degrades plant safety" in the final rule, to make it clear that only matters with a negative effect on safety are reportable. Its meaning is defined by the same examples that have served since 1983 to define the term "unanalyzed condition that significantly compromises plant safety." These are:

(1) multiple functionally related safety grade components out of service; (2) accumulation of voids that could inhibit the ability to adequately remove heat from the reactor core, particularly 29

under natural circulation conditions; and (3) voiding In instrument lines that results In erroneous indication causing the operator to misunderstand the true condition of the plant. Also, two new examples have been added. They are: (1) discovery that a system required to meet the single failure criterion does not do so; and (2) discovery that the fire protection system does not protect at least one safe shutdown train in the event of fire in a given area. All of these examples are discussed in the Statement of Considerations for the final rule as well as the reporting guidelines.

The term "condition of the nuclear power plant, including its principal safety barriers, being seriously degraded" is defined by guidance that is very similar to the guidance which has e defined it since 1983. Specifically, the guidance* states that this criterion applies to material (e.g., metallurgical or chemical) .problems that cause abnormal degradation of or stress upon the principal safety barriers (i.e., the fuel cladding, reactor coolant system pressure boundary, or the containment) such as:

(1) Fuel cladding failures in the reactor, or in the storage pool, that exceed expe.cted values, or that are unique or widespread, or that are caused by unexpected factors.

(2) W~lding or material _defects in the primary coolant system which cannot be found acceptable under ASME Section XI, IWB-3600, "Analytical Evaluation of Flaws"*or ASME Section XI, Table IWB-3410-1, "Acceptance Standards."

(3) Serious steam generator tube degradation.

(4) Low temperature over pressure transients where the pressure-temperature relationship violates pressure-temperature limits derived from Appendix G to 10 CFR Part 50 (e.g., TS pressure-temperature curves).

30

(5) Loss of containment function or integrity, including containment leak rate tests where the total containment as-found, minimum-pathway leak rate exceeds the limiting condition for operation (LCO) in the facility's TS.

This guidance is discussed in further detail below under the heading "Principal safety barrier seriously degraded."

Comment N (False elevated sense of problems): In addition to the points discussed above under the heading "Comment£," some commenters stated that reporting of invalid actuations will convey a false elevated sense of problems to the general public, causing undue alarm for situations that actually represent little or no safety or risk significance. Therefore, the new rule should not require invalid actuations to be reported.

Response: The NRC does not'agree that it is necessary to eliminate reporting foi:l' invalid actuations in order to avoid conveying a false elevated sense of problems to the general public. As discussed in the response to Comment E, there is a need for reporting of these events because they are used in making estimates of equipment reliability parameters, which in tum are needed to support the NRC's move towards risk-informed regulation. Invalid actuations have been reportable since 1983 under the previous rules, pursuant to both ....:

sections 50.72 and 50.73. No undue public alarm about such invalid actuations has been apparent to the NRC. The commenters did not identify any specific situation or provide any anecdotal evidence that reporting such invalid actuations has caused undue public alarm.

Comment O (Eliminate reporting of missing fire barriers): One commenter stated that the proposed rule notice at Page 36299, first column, the example pertaining to missing or degraded fire barriers basically equates such conditions with degraded principal safety barriers 0.e., fuel cladding, reactor coolant pressure boundary, and containment). This is inappropriate and should be deleted.

31

I.,. ,..

Response: The NRC does not agree. The example indicates that a condition is rep9rtable, as an unanalyzed condition that significantly affects plant safety, "if fire barriers are found to be missing such that the required degree of separation for redundant safe shutdown trains is lacking." This would mean that, if a fire occurs In the given area, no safe shutdown trains would be protected to an acceptable degree. Because Probabilistic Risk Assessment (PRA) studies continue to indicate that fire is a dominant contributor to risk, the inability to guarantee one train of safe shutdown capability, as required, Is considered to be a condition

\

that significantly degrades safety.

Comment P (Applicability of the backfit rule - no basis was stated): One commenter -

stated that In the proposed rule at Page 36303,Section VI., Backfit Analysis, the NRC stated that 10 CFR 50. 1.09 does not apply without giving any basis for the claim.

Response: The discussion below, entitled Backfit Analysis, has been modified to provide the basis for the conclusion that 10 CFR 50.109 does not apply.

Comment Q (Modify *unanalyzed condition that significantly affects safetyj: Most commenters stated that in section 50.72(b)(2)0i)(B), the phrase "significantly affects plant safety" has no positive or negative conno(ation. Reword the section to read, "The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety."

Response: The NRC agrees. The phrase is revised as recommended for the reason stated.

Comment R (Recognize risk-significance factors): One commenter stated that Section 50.73(a)(1) fails to recognize any risk significance factors.

Response: The NRC does not agree. Section 50.73(a)(1) is general in nature and indicates ~at, unless otherwise specified in section 50. 73, the licensee shall report an event if

\

it occurred within the last three years regardless of the plant mode or power level, and 32

regardle~s of the significance of the structure, system, or component that initiated the event Risk factors are recognized elsewhere in section 50.73. For example, the requirement to report an event or condition that could have prevented fulfillment of the safety function of structures or systems Is limited to those structures or systems that are needed to perform specific safety functions. The list of systems for which actuation must be reported is based on risk-significance. Lack of significance is the reason for the elimination of reporting for late surveillance tests where the equipment, when tested, Is functional. It is also the basis for eliminating several other requirements, such as immediate notification under section 50. 72 for

- many invalid actuations.

Comment S (Modify "operation or condition prohibited by TSj: Section 50.73(a)(2)0)(B) should be revised to read, "Any operation or condition occurring within three years of the date of discovery which was prohibited by th.e plant's CURRENT Technical Specifications." This rewrite would direct plants that recently converted to Improved Standard Technical Specifications to apply the current requirements to the identified condition, rather than having to consider the previous req1:1irements under old Technical Specifications which are no longer applicable.

Response: The NRC agrees. The issue involves the following scenario. A licensee discovers a historical operation or condition that was prohibited by the TS in effect at the time the operation or condition occurred. However, the prohibition has subsequently been removed from the TS. The event is not considered significant because subsequently the operation or condition was found to be acceptable and the Technical Specifications have been revised to permit It. Accordingly, the final rule eliminates the requirement to report such events.

Comment T (Reporting burden would not be decreased): In addition to the points discussed above under the heading "Comment A," one commenter disagreed with the NRC's 33

assessment that the proposed rule would represent an overall decrease in burden. This disagreement was based on the following points:

(a) (Telephone notifications are Jess burdensome than written LERs): Although the proposed rule would have decreased the number of phone-in reports pursuant to 10 CFR

50. 72, the commenter believes this burden is very small when compared with the burden of processing and submitting Licensee Event Repprts (LERs) pursuant to 10 CFR 50.73.

(b) (Actuation of systems that are currently excluded systems would become reportable): In the proposed rule, systems that were excluded from reporting requirements via previous rulemaking because they represented little or no saf~ty significance have been -

reinstated (e.g., Reactor Water Cleanup System). Such action will now lead to reporting all isolations, even those with no safety significance.

(c) (Systems not classified as ESF would be treated as ESF): Systems that are not classified as Engineered Safety Features (ESF) will now be treated as ESF (e.g., Reactor Core Isolation Cooling System).

(d)-(Jnvalid actuations would be added to the reporting requirements): Invalid actuations are now included in the reporting requirements. The impact of this change is that the clarifications for what used to be reportable have been deleted. Therefore, the proposed rule would treat all isolations or movements of a component as reportable regardless of safety significance.

(e) (The requirements for human performance data would be increased): The scope of information requested for human performance events has substantially Increased, going well beyond previous direct root cause to now include associated contributing factors.

Response: The NRC believes that reporting burden will be decreased for the reasons described in the regulatory analysis. With regard to the specific bases cited for this comment:

34

(a) The NRC agrees that a telephone notification is less burdensome than a written LER. However, this does not mean that the reporting burden would be increased, or maintained, unless there is some Increase in the number of LERs required under the final rule.

This is not the case.

{b) The NRC does not agree that the proposed rule would have made actuation of previously excluded systems reportable. The previously excluded syste*ms are: (i) reactor water clean-up system; (ii) control room emergency ventilation system; Oii) reactor building

, ventilation system; {iv) fuel building ventilation system; or (v) auxiliary building ventilation system. None of these appeared on the proposed list of systems for which actuation would be reportable.

(c) The NRC believes that system actuations added by adoption of the proposed list of systems are outweighed by system actuations eliminated.

(d) The NRC does not agree that invalid actuations are being added to the reporting requirements, because they were already in the reporting requirements.

{e) See the response to Comment J.

Comment U (Incentive to disable safety systems): In addition to the points discussed above under the heading "Comment E," one commenter indicated that reporting of invalid system actuations provided an incentive to disable safety systems.

Response: The NRC does not agree that it is necessary to eliminate reporting for invalid actuations to avoid creating an incentive to disable safety systems during maintenance activities to avoid the possibility of reporting an inadvertent actuation.

As discussed in the response to Comment E, there is a need for reporting of these events because they are used in making estimates of equipment reliability parameters, which in tum are needed to support the NRC's move towards risk-informed regulation. Also, in the 35

final rule, licensees are not required to provide an immediate notification under Section 50. 72 for an Invalid system actuation. Furthermore, in the final rule licensees have the option of providing a telephone notification within 60. days, rather than submitting a wri~en LER, for an invalid system actuation. These changes provide a drastic reduction in the burden of reporting for invalid system actuations. This burden reduction mitigates against any incentive to disable safety systems during maintenance in order to avoid the possibility of reporting an invalid actuation.

Comment V (Amend 10 CFR 76.120(d)(2) to allow 60 days): One commenter noted that the NRC plans to consider the idea of expanding the 60-day deadline for written reports to 4t other regulations. The commenter recommended amending 10 CFR 76.120(d)(2) to allow 60 days for written reports required under that regulation.

Response: The NRC continues to plan to evaluate the need for rulemaking to modify 10 CFR Parts 72 and 73, including the suggestion that 60 days be allowed for written reports required under 10 CFR 72.75 and 73.71. As part of that effort, the NRC will also consider the suggestion that 60 days be allowed for written reports required under 10 CFR 76.120(d)(2).

Comment W (Enforcement levels): Some commenters indicated that the proposed characterization of Enforcement Level Ill for failure to provide a required 1-hour or 8-hour non-emergency telephone notification is too harsh in most cases. They indicated that in most cases the information provided in these non-emergency notifications has low safety significance.

Response: As discussed further below under the heading "Enforcement/ the philosophy of the Enforcement Policy changes is to base the significance of the reporting violation on its impact on the NRC's ability to provide proper oversight of licensee activities.

36

Accordingly, in some cases, Severity Level Ill is appropriate for failure to make a required telephone notification and in other cases it is not Comment X (LER fonnat and content): One commenter recommended that the NRC reconsider a "check the box approach. The commenter indicated that such an approach could be crafted to make LER data entry easier, more consistent, and less ambiguous, without making LERs more difficult for the general public to understand.

Response: The NRC does not believe it is feasible to adopt a "check the box" in the final rule because the proposed rule did not include a proposal along those lines and development of a sound system would take considerable time, delaying issuance of the final rule.

Comment Y (Coordinate with perfonnance indicator efforts): One commenter suggested careful coordinated consideration among the NRG staff responsible for this rulemaking and those responsible for performance indicator efforts to ensure that reports submitted under 10 CFR 50.73(a)(2)(v) are not being misapplied.

Response: The NRC agrees and the suggested coordination has taken place, and will continue in the future as well. As a result, it is not expected that the NRC will misapply reports submitted under 10 CFR 50. 73(a)(2)(v).

Comment Z: One commenter recommended that t.elephone notifications due within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should only be required when activation of the NRC emergency response organization is actually required.

Response: The NRC does not agree that this is a fea~ible approach because activation of the NRC's emergency response organization is not a simple function of the reporting criterion under which an event is considered to be reportable. For example, the emergency response organization is sometimes activated for events which, at the time of 37

reporting, are considered to correspond to lower levels of Emergency Classes or non-emergency reporting criteria.

Comment AA (Do not include criteria for reporling degraded steam generator tubes):

The Statement of Considerations for the proposed rule and the Draft Revision 2 to NUREG-1022 would indicate that steam generator tube degradation is considered serious, and thus reportable as a seriously degraded reactor coolant system boundary, if the tubing fails to meet specific performance criteria involving margin against burst and accident induced leakage rate.

Most commenters proposed that this guidance be deleted. They stated that the position was based on a Draft Regulatory Guide {DG-1074, Steam Generator Tube Integrity) that has not -

been approved. Discussions between the industry and the NRC are being held to define the steam generator program and Technical Specification requirements. Some of the examples provided in the proposed section are contrary to agreements that have been made between the industry and the NRC staff. Recognizing that these agreements are still evolving, the proposed revisions to the rule(s) and NUREG-1022 must agree with the final positions on steam generator issues.

Response: The details have been removed from the Statement of Considerations.

The details in the final Revision 2 to NUREG-1022 have been modified to reflect the NRC staffs current thinking. The guidance is consistent with the steam generator tube integrity performance criteria and reporting guidelines currently under discussion. This reporting is needed to permit the staff to determine if further inquiry or action might be needed before the plant is restarted.

The NRC does not agree that it is necessary to delay issuance of this reporting guidance pending staff endorsement of the NEI 97-06 initiative. The NUREG-1022 guidance merely provides reasonable examples of degraded steam generator tube conditions which the 38

"' '1 NRG needs to evaluate. If it is detennined in the future that different detailed guidance is needed, it can be issued at that time.

Ill. DISCUSSION

1. Objecti\/8s The purposes of sections 50. 72 and 50. 73 remain the same because the basic needs remain the same. The essential purpose of section 50.72 is" ... to provide the Commission with immediate reporting of ... significant events where immediate Commission action to protect the public health and safety may be required or where the Commission needs timely and accurate information to respond to heightened public concem." (48 FR 39039; August 29, 1983). Section 50.73 " ... identifies the types of reactor events and problems that are believed to be significant and useful to the NRG in its effort to identify and resolve threats to public safety. It is designed to provide the information necessary for engineering studies of operational anomalies and trends and pattems analysis of operational occurrences. The same information can be used for other analytic procedures that will aid in identifying accident precursors.M (48 FR 33851; July 26, 1983)..

The objectives of these final amendments are as follows:

(1) To better align the reporting requirements with the NRC's needs for information to carry out its safety mission. An example is extending the required initial reporting times for some events, consistent with the time at which the reports are needed for NRG action.

(2) To reduce unnecessary reporting burden, consistent with the NRC's needs. An example is eliminating the reporting of design and analysis defects and deviations with little or no risk- or safety-significance.

39

(3) To clarify the reporting requirements where needed. An example is clarifying the criteria for reporting design or analysis defects or deviations.

(4) Any changes should be consistent with NRC actions to improve integrated plant assessments. For example, reports that are needed in the assessment process should not be eliminated.

2. Section by Section Discussion of Final Amendments General requirements and reportable events {section 50. 72(a)(1) and section

50. 73(a)(1)]. The term "if it occurred within 3 years of the date of discovery is added to eliminate reporting for conditions that have not existed during the three years before discovery.

Such a historical event has less significance, and assessing reportability for earlier times can consume considerable resources. For example, assume that a procedure is found to be

.unclear and, as a result, a question is raised as to whether the plant was ever operated in a prohibited condition. If operation in the prohibited condition is likely, the answer would be reasonably apparent based on the knowledge and experience of the plant's operators and/or a review of operating records for the past three years. The effort required to review all records older than three years in order to rule out the possibility is not warranted.

A sentence is added to indicate that for an invalid actuation reported under section 50.73(a)(2)(iv) the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event in lieu of submitting a written LER. For this type of event, a telephone notification will provide the information needed and impose less burden than an LER.

40

General requirements [section 50. 72(a)(5)]. The requirement to inform the NRC of the type of report being made (I.e., Emergency Class declared, non-emergency 1-hour report, or non-emergency 8-hour report) is revised to refer to paragraph (a)(1) Instead of referring to paragraph (a)(3) to correct a typographical error.

Required initial reporting times [sections 50. 72(a)(5), (b)(1), (b)(2), and new section

50. 72(b)(3); and sections 50. 73(a)(1) and (d)J. In the final amendments, declaration of an Emergency Class continues to be reported immediately after notification of appropriate State or local agencies and not later than 1-hour after declaration. This includes declaration of an Unusual Event, the lowest Emergency Class.

Deviations from Technical Specifications authorized pursuant to 10 CFR 50.54(x) continue to be reported as soon as practical and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of occurrence.

These two criteria capture those events where there may be a need for immediate action by the NRC to protect public health and safety.

The requirement to report an event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been renumbered from section' 50.72(b)(2)(vi) to section 50.72(b)(2)(xQ. In other respects this reporting criterion is unchanged, and the event is reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , if not reported within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This provides the information at the time it may be needed to respond to heightened public concern.

The requirement to report a natural phenomenon or other external event that poses an actual threat to plant safety or significantly hampers site personnel in the performance of duties necessary for safe operation in section 50.72(b)(1)(iiQ is deleted. Events of this type are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

41

The requirement to report an internal event that poses an actual threat to plant safety, or significantly hampers site personnel In the performance of duties necessary for safe

operation, including fires, toxic gas releases, or radioactive releases in section 50.72(b){1){vi) is deleted. Events of this type are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The requirement to report an airborne radioactive release or liquid effluent release that exceeds specific limits in section 50.72 {b)(2){lv) is deleted. Releases that are large enough to warrant prompt notification are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the declaration. Releases that involve a news release or notification to other government agencies are reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the occurrence.

The remaining non-emergency events that are reportable by telephone under 10 CFR

50. 72 are reportable as soon as practical and in all cases within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months {instead of within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as was previously required). This reduces the unnecessary burden of rapid reporting, while:

(1) Capturing, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , those events where there may be a need for the NRC to take a reasonably prompt action, such as partially activating its response plan to monitor the I

course of the event (2) Capturing, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , those events where there may be a need for the NRC to take an action within about a day, such as initiating a special inspection or investigation.

See the response to Comment D, above, for further discussion.

Written LERs are due within 60 days after discovery of a reportable event or condition (instead of within 30 days as was previously required). Changing the time limit from 30 days to 60 days does not imply that licensees should take longer than they previously did to develop 42

and implement corrective actions. They should continue to do so on a time scale commensurate with the safety significance of the issue. However, for those cases where it does take longer than thirty days to complete a root cause analysis, this change will result in fewer LERs that require amendment (by submittal of an amended report).

The term "within 30 days of the discovery of a reportable event or situation" is deleted from section 50.73(d). This provision is redundant to the provisions of section 50.73(a)(1),

which requires that a licensee submit an LER within 60 days after discovery of an event described in section 50.73(a). Retaining the time limit, which is now 60 days, in section 50.73(d) would create a conflict with sections 20.2201 and 20.2203 which require licensees to submit LERs for the events described in those sections within 30 days and in accordance with section 50.73(d).

Operation or condition prohibited by technical specifications [section 50. 73(a)(2)(1)(8)].

This criterion-is modified to eliminate reporting if the Technical Specification is administrative in nature. Violations of administrative Technical Specifications have generally not been considered to warrant submittal of an LER, and since 1983 when the LER rule was issued the I

NRC's event reporting guidelines have excluded almost all cases of such reporting. This change makes the plain wording of the rule consistent with that guidance.

Also, this criterion is modified to eliminate reporting if the event consisted solely of a case of a late surveillance test where the oversight is corrected, the test is performed, and the equipment is found to be functional. This type of event has not proven to be significant because the equipment rer,nained functional.

Finally, this criterion is modified to eliminate reporting of an operation or condition that occurred in the past and was prohibited at that time if, prior to discovery of the event, the Technical Specifications were revised such that the operation or condition is no longer 43

prohibited. Such an event would have little or no significance because, by the time of discovery, the operation or condition would have been determined to be acceptable and thus permissible under current Technical Specifications.

The NRC expects licensees to include violations of the Technical Specifications in their corrective action programs, which are subject to NRC audit.

Condition of the nuclear power plant, including its principal safety barriers, being seriously degraded [former sections 50. 72(b)(1)(il) and (b)(2)(i), replaced by new section

50. 72(b)(3)(il)(A); and section 50. 73(a)(2)(i1}, renumbered to 50. 73(a)(2)(it)(A)]. Previously, 10 CFR 50.72(b)(1)(ii) and {b)(2)(i) provided the following distinction. During operation, a seriously degraded plant, including its principal safety barriers, was reportable within one hour.

An event discovered while shutdown that had it been discovered during operation would have resulted in a seriously degraded plant, iricluding its prinqpal safety barriers, was reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The new 10 CFR 50.72(b)(3)(ii)(A) eliminates the distinction because there are no longer separate 1-hour and 4-hour categories of non-emergency reports for this criterion.

There are only 8-hour non-emergency reports for this criterion.

Unanalyzed condition that significantly degrades plant safety [sections 50. 72(b)(1)(il)(A) and (b)(2)(i), replaced by new section 50. 72(b)(3)(it)(B); and section 50. 73(a)(2)(ii)(A),

renumbered to 50. 73(a)(2)(ii)(B)]. Previo,usly, 10 CFR 50.72(b)(1)(ii)(A) and (b)(2)(i) provided the following distinction. During operation, an unanalyzed condition that significantly compromised plant safety was reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . An event discovered while shut down that had it been discovered during operation would have resulted in an unanalyzed condition that significantly compromised plant safety was reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The new 10 CFR 50.72(b}(2)(ii)(B) eliminates this distinction because there are no longer separate 1-hour and 44

4-hour categories of non-emergency reports for this reporting criterion. There are only 8-hour

~ ,

non-emergency reports for this criterion.

In addition, the new 10 CFR 50. 72(b)(2)~0(8) and 50. 73(a)(2)(ii)(B) refer to a condition that significantly degrades plant safety rather than a condition that significantly compromises plant safety. This is an editorial change Intended to b(;Jtter reflect the nature of the criterion.

Condition that is outside the design basis of the plant [old sectiof) 50. 72(b)(2)(il)(B); and old section 50. 73(a)(2)(ii)(B)J. This criterion is deleted. A condition outside the design basis of the plant is still required to be reported if it is significant enough to qualify under one or more of

- the following criteria.

Plant safety significantly degraded. If a condition outside the design basis of the plant (or any other unanalyzed condition) is significant enough that, as a result, plant safety is significantly degraded, the condition is reportable under sections 5O.72(b)(2)(ii)(B) and*

50. 73(a)(2)(ii){B) ~.e., an unanalyzed condition that significantly degrades plant safety].

As was previously indicated in the 1983 Statements of Considerations for 10 CFR 50.72 and 50.73, with regard to an unanalyzed condition that significantly compromises plant safety, "The Commission recognizes that the licensee may use engineering judgment and experience to determine whether an unanalyzed condition existed. It is not intended that this paragraph apply to minor variations in individual parameters, or to problems concerning single pieces of equipment For example, at any time, one or more safety-related components may be out of service due to testing, maintenance, or a fault that has not yet been repaired. Any trivial single failure or minor error in performing surveillance tests could produce a situation in which two or more often unrelated, safety-grade components are out-of-service. Technically, this is an unanalyzed condition. However, these events should be reported only ifthey involve 45

functionally related components or if they significantly compromise plant safety," (48 FR 39042; August 29, 1983 and 48 FR 33856, July 26, 1983).

"When applying engineering judgment, and there is a doubt regarding whether to report or not, the Commission's policy is that licensees should make the report," (48 FR 39042; August 29, 1983).

"For example, small voids in systems designed to remove heat from the reactor core which have been* previously shown through analysis not to be safety significant need not be reported. However, the accumulation of voids that could inhibit the ability to adequately remove heat from the reactor core, particularly under natural circulation conditions, would constitute an unanalyzed condition and would be reportable," (48 FR 39042; August 29, 1983 and 48 FR 33856, July 26, 1983).

"In addition, voiding in instrument lines that results in an erroneous indication causing the operator to misunderstand the true condition of the plant is also an unanalyzed condition and should be reported," (48 FR 39042; August 29, 1983 and 48 FR 33856, July, 26, 1983).

Furthermore, beyond the examP,les given in 1983, examples of reportable events include discovery that a system required to meet the single failure criterion does not do so.

In anotherexample, if fire barriers are found to be missing, such that the required degree of separation for redundant safe shutdown trains is lacking, the event is reportable. On the other hand, if a fire wrap, to which the licensee has committed, is missing from a safe shutdown train but another safe shutdown train is available in a different fire area, protected such that the required separation for safe shutdown trains is st!II provided, the event is not reportable.

Structure or system not capable of performing its specified safety function. If a design or analysis defect or deviation (or any other event or condition) is significant enough that, as a 46

result, a structure or system is not capable of performing its specified safety functions, the

'condition is reportable under sections 50.72(b)(3){v) and 50.73(a)(2){v) [i.e., an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: shut down the reactor ... ; remove residual heat; control the release of radioactive material; or mitigate the consequences of an accident].

For example, in one case an annual inspection indicated that some bearings were wiped or cracked on both emergency diesel generators (EDGs). Although the EDGs were running prior to the inspection, the event was reportable because there was reasonable doubt e about the ability of the EDGs to operate for an extended period of time, as required.

Train inoperable longer than allowed. If a design or analysis defect or deviation {or any other event or condition) is significant enough that, as a result, one train of a multiple train system controlled by the plant's TS is not capable of performing its specified safety functions for a period of time longer than allowed by the TS, the condition is reportable under section

50. 73(a)(2)(i)(B). CT.e., an operation or condition prohibited by TSJ.

For example, if it is found that an exciter panel for one EOG lacks appropriate seismic restraints because of a design, analysis, or construction inadequacy and, as a result, there is reasonable doubt about the EDG's ability to perform its specified safety functions during and after a Safe Shutdown Earthquake (SSE), the event would be reportable.

Or, for example, if it is found that a loss of offsite power could cause a loss of instrument air and, as a result, there is reasonable doubt about the ability of one train of the auxiliary feedwater system to perform its specified safety functions for certain postulated steam line breaks, the event would be reportable.

Principal safety barrier seriously degraded. If a condition outside the design basis of the plant (or any other event or condition) is significant enough that, as a result, a principal 47

safety barrier is seriously degraded, It is reportable under sections 50.72(b)(3)(ii)(A) and 50.73(a)(2)(iQ(A) [i.e., any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degradedf This reporting criterion applies to material (e.g., metallurgical or chemical) problems that cause abnormal degradation of or stress upon the principal safety barriers (i.e., the fuel cladding, reactor coolant system pressure boundary, or the containment) such as:

(i) Fuel cladding failures in the reactor, or in the storage pool, that exceed expected ,

values, or that are unique or widespread, or that are caused by unexpected factors.

(ii) Welding or material defects in the primary coolant system which cannot be found -

acceptable under ASME Section XI, IWB-3600, "Analytical Evaluation of Flaws" or ASME Section XI, Table IWB-3410-1, "Acceptance Standards."

(iii) Serious steam generator tube degradation.

(iv) Low temperature over pressure transients where the pressure-temperature relationship violates pressure-temperature limits derived from Appendix G to 10 CFR Part 50 (e.g., TS1pressure-temperature curves).

(v) Loss of containment function or integrity, including containment leak rate tests where the total containment as-found, minimum-pathway leak rate exceeds the limiting I

condition for operation (LCO) in the facility's TS. 1 Common cause inoperability of independent trains or channels. If a condition outside the design basis of the plant (or any other event or condition) is significant enough that, as a result, independent trains or channels become inoperable, it would be reportable under section 1

The LCO typically employs La, which is defined in Appendix J to 10 CFR Part 50 as the maximum allowable containment leak rate at pressure Pa, the calculated peak containment internal pressure related to the design basis accident Minimum-pathway leak rate means the minimum leak rate that can be attributed to a penetration leakage path; for example, the smaller of either the inboard or outboard valve's individual leak rates.

48

50. 73(a)(2)(vii) p.e., an event where a single cause or condition caused independent trains or channels to become inoperable}. For example, In one case it was found that independent circuit breakers, required to operate after a Loss-of-Coolant Accident (LOCA), were not qualified for the expected radiation levels (and were thus considered inoperable). In another example, a wiring error caused Independent containment isolation valves to be incapable of properly closing (i.e., they would not dose tightly because they would stop closing based on limit switch operation rather than torque).

Single Cause that Could Have Prevented Fulfillment of the Safety Functions of Trains or Channels in Different Systems. Finally, a condition outside the design basis of the plant (or any other event or condition) would be reportable if it is significant enough that, as a result of a single cause, it could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident.

This new criterion is contained in sections 50.73(a)(2)(ix){A) and {B), as discussed below.

, Single Cause that Could Have Prevented Fulfillment of the Safety Functions of Trains or Channels in Different Systems. [new sections 50. 73(a)(2)(ix)(a) and (B)]. This new criterion requires reeorting any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdown condition; 49

(2) Remove residual heat; (3) Control the release of radioactive material; or

{4} Mitigate the consequences of an accident.

Events covered by this new criterion may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to this criterion if the event results from:

(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.

Subject to the two exclusions stated above, this criterion captures those events where a single cause could have prevented the fulfillment of the safety function of multiple trains or channels, but the event:

(1) Would not be captured by§§ 50.73(a)(2)(v) and 50.72(b)(3){v) [event or condition that could have prevented fulfillment of the safety function of structures and systems needed to ... ] because the affected trains or channels are in different systems; and (2) Would not be captured by§ 50.73(a)(2)(vii) [common cause inopera.bility of independent trains or channels] because the affected trains or channels are either:

(i) Not assumed to be independent in the plant's safety analysis; or

{ii) Not both considered to be inoperable.

This new criterion is closely related to§§ 50.73(a)(2)(v) and 50.72(b)(3){v) [event or condition that could have prevepted fulfillment of the safety function of structures and systems needed to: shut down the reactor and maintain it in a safe shutdown condition; remove 50

residual heat; control the release of radioactive material; or mitigate the consequences of an accident]. Specifically:

The meaning of the term "could have prevented the fulfillment of the safety function" is essentially the same for this new criterion as It is for§§ 50.73(a)(2)(v) and 50.72(b)(3)(v) D.e.,

there was a reasonable expectation of preventing the fulfillment of the safety function(s) involved]. However, in ,contrast to§§ 50.73(a)(2)(v) and 50.72(b)(3)(v), reporting under this new criterion applies to trains or channels in different systems. Thus, for this new criterion, the safety function that is affected may be different in different trains or channels.

In contrast to§§ 50.73(a)(2)(v) and 50.72(b)(3)(v), reporting under this new criterion applies only to a single cause. Also, in contrast to§§ 50.73(a)(2)(v) and 50.72(b)(3)(v), this new criterion does not apply to an event that results from a shared dependency among trains or channels that is a natural or expected consequence of the approved plant design. For example, this new criterion does not capture failure of a common electrical power supply that disables Train A of AFW and Train A of High Pressure Safety Injection (HPSI), because their shared dependency on the single power supply is a natural or expected consequence of the approved plant design.

Similar to§§ 50.73(a)(2)(v) and 50.72(b)(3)(v), this new criterion does not capture events or conditions that result from normal and expected wear or degradation. For example, consider pump bearing wear that is within the normal and expected range. In the case of two pumps in different systems, this new criterion categorically excludes normal and expected wear. In the case of two pumps in the same system, normal and expected wear should be adequately addressed by normal plant operating and maintenance practices and thus should not indicate a reasonable expectation of preventing fulfillment of the safety function of the system.

51

This criterion pertains only to written LERs required by 10 CFR 50.73. Telephone notifications are not required under this criterion.

It is esti~ated that the combination of removing the previous requirement to report a condition outside the design basis of the plant and adding this criterion will, on balance, result in fewer reports. In addition, the events reportable under this criterion are events that would likely have been considered reportable under the previous requirement to report a condition outside the design basis of the plant An example. of an event that would be reportable under this criterion is as follows.

During testing, two containment isolation valves failed to function as a result of improper air gaps in the solenoid operated valves that controlled the supply of instrument air to the containment isolation valves. The valves were powered from the same electrical division.

Thus, § 50. 73(a){2)(vii) [common cause inoperability of independent trains or channels] would not apply. The_ two valves isolated fluid process lines in two different systems. Thus

§ 50.73(a)(2)(v) [condition that could have prevented fulfillment of the safety function of a structure or system] would apply only if engineering judgment indicates there was a reasonable expectation of preventing fulfillment of the safety function for redundant valves within the same system. 2 However, this new criterion would certainly apply if a single cause (such as a design inadequacy)-induced the improper air gaps, thus preventing fulfillment of the safety function of two trains or channels in different systems.

Another example of an event reportable under this criterion is as follows. A motor operated valve in one train of a system was found with a chick 75 percent through the stem.

Although the valve stem did not fail, engineering evaluation indicated that further cracking 2

Or, alternatively, there was reasonable doubt that the safety function would have been fulfilled if the affected trains h.ad been called upon to perform them.

52

would occur which could have prevented fulfillment of its safety function. As a result, the train was not considered capable of performing its specified safety function and the valve stem was replaced with a new one.

The root cause was determined to be environmentally assisted stress corrosion cracking which resulted from installation of an inadequate material some years earlier. The same inadequate material had been installed in a similar valve in a different system at the same time. The similar valve was exposed to similar environmental conditions as the first valve.

4t The condition is reportable under this new criterion if engineering judgment indicates that there was a reasonable expectation of preventing fulfillment of the safety function of both affected trains. This depends on details such as whether the second valve stem was also significantly degraded and, if not, whether any future degradation of the second valve stem would have been discovered and corrected, as a result of routine maintenance programs, before it could become problematic.

Additional examples may be found in event reporting guidelines in NUREG-1022, I

Revision 2.

Condition not covered by the plant's operating and emergency procedures [former section 50. 72(b)(1)(iij(C); and former section 50. 73(a)(2)(il)(C)J. This criterion is deleted because it does not result in worthwhile reports aside from those that are captured by other reporting criteria such as:

(1) An unanalyzed condition that significantly degrades plant safety; (2) An event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: shut down the reactor and maintain it in a safe 53

shutdown condition; remove residual heat; control the release of radioactive material; or mitigate the consequences of an accident; (3) An event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; (4) An operation or condition prohibited by the plant's TS; (5) An event or condition that results in actuation of an ESF; (6) An event that poses an actual threat to the safety of the nuclear power plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the nuclear power plant External event that poses an actual threat or significantly hampers personnel [former section 50. 72(b)(1)(iil), deleted; and section 50. 73(a)(2)(iii)J. This criterion requires reporting a natural phenomenon or other external event that poses an actual threat to plant safety, or significantly hampers site personnel in- the performance of duties necessary for safe operation.

Section 50.72(b)(1)0iQ is deleted because it is redundant to section 50.72(a)(1)(i). That is, events of this type are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Section 50.73(a)(2){iii} is retained in order to ensure submittal of an LER. This provision is not redundant because there is no criterion in section 50.73 that generally requires an LER for declaration of an Emergency Class.

System actuation [old sections 50. 72(b)(1)(iv) and (b)(2)(il), replaced by new sections

50. 72(b)(2)(iv)(A), (b)(2)(iv)(B), and (b)(3)(iv); and section 50. 73(a)(2)(iv)]. Previously, sections 50.72(b)(1)0v) and (b)(2){ii) provided the following distinction: an event that results or should have resulted in ECCS discharge into the reactor coolant system as a result of a valid signal was reportable wjthin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; any other engineered safety feature (ESF) actuation, including reactor protection system (RPS) actuation, was reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The new 10 CFR 54

50.72(b)(2)0v}(A) requires reporting an event that results or should have resulted in ECCS discharge into the reactor coolant system as a result of a valid signal within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The new section 50. 72(b)(2)(iv)(B) requires reporting a reactor scram during critical operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The new section 50.72(b)(3)0v} requires reporting other ESF actuations within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

See the response to Comment D, above, for further discussion.

The new section 50.72(b){2)0v} eliminates telephone reporting for invalid actuations, except for actuation of the RPS when the reactor is critical. These events are not significant and thus telephone reporting is not needed. The final amendments do not eliminate the requirement for reporting of an invalid actuation under 10 CFR 50.73. There is still a need for reporting of these events because they are used in making estimates of equipment reliability parameters, which in tum are needed to support the Commission's move towards risk-informed regulation. However, for an invalid actuation reported under section 50.73(a)(2)0v), other than actuation of the RPS when the reactor is critical, section 50.73(a)(1) provides the option of making a telephone report to the NRC Operations Center within 60 days instead of submitting a written LER. The telephone report is far less burdensome. Sixty days is an appropriate time because the information is not needed immediately. (See the response to Comment E above for further discussion of this need.)

Previously, the rules generally required reporting the actuation of any ESF including the RPS. The final rule, instead, generally requires reporting for actuation of specific listed systems. These systems are:

(1) Reactor protection system (RPS) including: reactor scram or reactor trip.

(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

55

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs)

including: high-head, intermediate-head, and tow-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(4) ECCS for boiling water reactors (BWRs) including: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.

(6) PWR auxiliary or emergency feedwater system.

(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

{8} Emergency ac electrical power systems, including: emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.

(9) Emergency service water (ESW) systems that do not normally run and that serve as ultimate heat sinks. ESW system actuations are reportable under section 50.73 only.

This approach provides for consistent reporting of actuations for these highly risk-significant systems. At the same time, it eliminates reporting of actuations for systems of lesser risk-significance, such as actuation of ventilation systems that are considered to be ESFs.

Section 50. 72 excludes reporting for an actuation that resulted from and was part of a pre-planned sequence during testing or reactor operation. It further excludes reporting of an invalid actuation, except for a reactor scram or reactor trip when the reactor is critical.

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A valid actuation Is one that results from either a "valid signal" or an intentional manual Initiation. A "valid signal" is one that results from actual plant conditions or parameters satisfying the requirements for system actuation. An invalid actuation is one that does not meet the criteria for being valid.

Section 50.73 also excludes reporting for an actuation that resulted from and was part of a pre-planned sequence during testing or reactor operation. It further excludes reporting of an invalid actuation that occurred when the system was properly removed from service or an invalid actuation that occurred after the safety function had been already completed.

For those invalid actuations which are reportable under section 50. 73, a licensee may provide a telephone notification within 60 days, rather than submitting an LER. This option to provide a telephone notification rather th~n an LER does not apply, however, to a reactor scram or reactor trip that occurs while the reactor is critical.

Event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: shut down the reactor and maintain it in a safe shutdown condition; remove residual heat; control the release of radioactive material; or mitigate the consequences of an accident [former section 50. 72(b)(2)(iil), replaced by new sections 50. 72(b)(3)(v) and (b)(3)(v,); and sections 50. 73(a)(2)(v) and (a)(2)(v,)]. The phrase "event or condition that alone could have ptevented the fulfillment of the safety function of structures or systems .... " is clarified by deleting the word "alone". This clarifies the requirements by more clearly reflecting the principle that it is necessary to consider other existing plant conditions in determining the reportability of an event or condition under this criterion. For example, if one train of a two train system is incapable of performing its safety function for one reason, and the other train is incapable of performing its safety function for a different reason, the event is reportable.

57

The term "at the time of discovery" is added to section 50.72(b)(3)(v) to eliminate telephone notification for a condition that no longer exists or no longer has an effect on required safety functions. For example, it might be discovered that at some time in the past both trains of a two train system were incapable of performing their safety function, but the condition was subsequently corrected and no longer exists. In another example, while the plant is shutdown, it might be discovered that during a previous period of operation a system was incapable of performing its safety function, but the system is not currently required to be operable. These events are considered significant, and an LER is required, but there is no need for telephone notification. e A new paragraph, section 50.72(b)(3)(vi) is added to clarify section 50.72. The new paragrap~ explicitly states that telephone reporting is not required under section 50.72(b)(2)(v) for single failures if redundant equipment in the same system was operable and available to perform the required safety function. That is, although one train of a system may be incapable of performing its safety function, reporting is not required under this criterion if that system is still capable of performing the safety function. This is the same principle that was and continues to be stated explicitly in section 50.73(a)(2)(vi) with regard to written LERs.

Airborne radioactive release or liquid effluent release [former section 50. 72(b)(2)(iv),

deleted; and section 50. 73(a)(2)(viii), retained; and former section 50. 73(a)(2)(ix), deleted}.

These criteria require reporting releases of radioactive material at a very low level because, for a power reactor, such a release would indicate a breakdown in the licensee's programs to control releases - not because of the impact of such a release.

Section 50.72(b)(2)(iv) is deleted because immediate notification is not needed for releases at such a low level. Declaration of an Emergency Class, which occurs at a somewhat higher (but still low) level, captures releases that are large enough to warrant immediate 58

notification. Declaration of an Emergency Class is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months under section

50. 72(a)(1 )(i).

Section 50.73(a)(2)(vlii) Is retained in order to ensure submittal of an LER. Even if the release is very small, the NRC needs to review the event and consider whether action is needed to ensure the cause is addressed at other plants as appropriate. There is no criterion in section 50. 73 that generally requires an LER for declaration of an Emergency Class.

Section 50. 73(a)(2)(vix) is deleted because it is not correct. It indicated that reporting under section 50.73{a)(2)(viii) satisfied the requirements of section 20.2203(a)(3). However, some events captured by section 20.2203(a)(3) are not captured by se~on 50.73(a)(2)(vilQ.

Internal event that poses an actual threat or significantly hampers personnel [former section 50. 72(b)(1)(v,), deleted; and section 50. 73(a)(2)(x)]. This criterion requires reporting an internal event that poses an actual th~at to plant safety, or significantly hamp~rs site personnel in the performance of duties necessary for safe operation, including fires, toxic gas releases, or radioactive releases. Section 50.72(b)(1}(vi) is deleted because it is redundant to section 50.72{a)(1)(i). That is, events of this type are captured by declaration of an Emergency Class, which is reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Section 50.73(a)(2)(x) is retained in order to ensure submittal of an LER. This provision is not redundant because there is no criterion in section 50.73 that generally requires an LER for declaration of an Emergency Class.

Major loss of emergency assessment capability, offsite response capability, or communication capability [former section 50. 72(b)(2)(v), replaced by new section

50. 72(b)(3)(xit)}. The new section is modified by adding the word "offsite" in front of the term "communications capability" to make it clear that the requirement does not apply to internal plant communication systems.

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Contents of LERs [section 50. 73(b)(2)(il)(F)J. Paragraph (F) is revised to correct the address of the NRC Library.

Spent fuel storage cask problems [fonner sections 50. 72(b)(2)(vil) and 72.216(a)(1),

(a)(2), and (b)J. The provisions of section 50.72(b)(2){vii) are deleted because these reporting criteria are redundant to the reporting criteria contained in sections 72.216{a)(1) and (a)(2).

Repetition of the same reporting criteria in different sections of the rules added unnecessary complexity and was Inconsistent with the current practice in other areas, such as reporting of safeguards events as required by section 73.71.

Sections 72.216(a){1) and {a)(2) place upon general licensees the same reporting criteria as are placed on specific licensees under sectiohs 72.75(b)(2) and (b)(3). To avoid duplication in Part 72, sections 72.216(a)(1) and (a)(2) are deleted and section 72.216(c) is abridged to simply require that the general licensee shall make initial and written reports in accordance with sections 72. 7 4 and 72. 75. These changes eliminate a reference in section 72.216(a) to section 50.72(b)(2){vii), now deleted, which had established the time limit for initial notification by general licensees. The same time limit is placed on general licensees by including them within the scope of section o/2.75{b). Section 72.216(b) is also deleted because its requirements for a written repo~ are encompassed by section 72.75(d)(2).

Exemptions [section 50. 73(f)J. The provisions of this section are deleted because the exemption provisions in section 50.12 provide for granting of exemptions when they are warranted. Including another, section-specific exemption provision in section 50.73 adds unnecessary complexity to the rules.

3. Revisions to Event Reporting Guidelines in NUREG-1022 60

-A report NUREG-1022, Revision 2, "Event Reporting Guidelines, 10 CFR 50.72 and 50.73," is being made available concurrently with the final amendments to 10 CFR 50.72 and 50.73. The report is available for inspection in the NRC Public Document Room or it may be viewed and downloaded electronically via the interactive rulemaking web site established by the NRC for this rulemaking, as discussed above under the heading ADDRESSES. Single copies may be obtained from the contact listed above under the heading "For Further Information Contact." In the report, guidance that is considered to be new or different in a meaningful way, relative to that provided in NUREG-1022, Revision 1, is indicated by underlining the appropriate text.

4. Reactor Oversight The NRC is implementing revisions to the process for oversight of operating reactors, including inspection, assessment, and enforcement processes. In connection with this effort, the NRC has considered the kinds of event reports that would be eliminated by the final rules and concluded that the changes are consistent with the oversight process.

In connection with the proposed rule, public comment was invited on whether or not this is the case. In particular, it was requested that if any examples to the contrary are known they be identified. None were identified.

5. Enforcement The NRC intends to modify its existing enforcement policy in connection with the final amendments to sections 50.72 and 50.73. The philosophy of the changes is to base the 61

significance of the reporting violation on the impact on the NRC's ability to provide proper oversight of licensee activities. For example, a late report may impact the ability of the NRC to fulfill its obligations of fully understanding issues that are required to be reported in order to accomplish its public health and safety mission, which in many cases involves reacting to reportable issues or events. As such, the NRC intends to revise the Enforcement Policy, NUREG-160a3 as follows:

(1) Supplement LC-Examples of Severity Level Ill violations.

(a) Example 11 will be revised to read as follows-A failure to provide the required 1-hour telephone notification of an emergency action taken pursuant to 10 CFR 50.54(x).

(b) An additional example will be added that will read as follows-A failure to provide a required 1-hour, 4-hour or 8-hour non-emergency telephone notification pursuant to 10 CFR

50. 72, that substantially impacts agency response.

(c) An additional example will be added that will read as follows-A late 4-hour or 8-hour notification that substantially impacts agency response.

(2) Supplement I. D-~amples of Severity Level IV violations.

(a) Example 4, will be revised to read as follows-A failure to provide a required 60-day written LER pursuant to 10 CFR 50.73.

These changes in the Enforcement Policy will be consistent with the overall objective of the rule change of better aligning the reporting requirements with the NRC's reporting needs.

The Enforcement Policy changes will correlate the Severity Level of the infractions with the relative importance of the information needed by the NRC.'

3 The examples refer to those published in NUREG-1600, "General Statement of Policy and Procedure for NRC Enforcement Actions," dated May 1, 2000.

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Section IV.A.3 of the Enforcement Policy provides that the Severity Level of an untimely report may be reduced depending on the Individual circumstances. In deciding whether the Severity Level should be reduced for an untimely 1-hour, 4-hour, or 8-hour non-emergency report, the impact that the failure to report had on any agency response will be considered. For example, If a delayed 8-hour reportable event impacted the timing of a followup inspection that was deemed necessary, then the Severity Level will not normally be reduced. 1 Similarly, a late notification that delayed the NRC's ability to perform an engineering analysis of a condition to determine if additional regulatory action was necessary will generally not be considered for disposition at a reduced Severity Level.

6. Electronic Reporting The NRC is currently in the process of implementing an electronic document management and reporting program, known as the Agency Wide Document Access and Management System (ADAMS) that will provide for electronic submittal of many types of reports, including LERs. Accordingly, no separate rulemaking effort to provide for electronic submittal of LERs is necessary.

7. State Input Many States (Agreement States and Non-Agreement States) have agreements with power reactors to inform the States of plant issues. State reporting requirements are frequently triggered by NRC reporting requirements. Accordingly, the NRC sought State comment on issues related to the proposed amendments via letters to State Liaison Officers 63

as well as by a specific request in the Federal Register notice on the proposed rule.

Comments on the proposed rule were received from one State agency, as discussed above under the heading "Comment D."

8. Plain Language The President's Memorandum dated June 1, 1998, entitled, "Plain Language in Government Writing," directed that the Federal Government's writing be in plain language.

The NRC requested comments on the proposed rule specifically with respect to the clarity and effectiveness of the language used. A number of suggestions aimed at improving the clarity and effectiveness of the language used were received and incorporated into the final rule.

IV. ENVIRONMENTAL IMPACT: CATEGORICAL EXCLUSION The NRC has determined that this proposed regulation is the type of action described in categorical exclusion 10 CFR 51.22(c)(3)(iii). Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this proposed regulation.

V. BACKFIT ANALYSIS The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to information collection and reporting requirements such as those contained in the final rule because they do not impose backfits as defined in 10 CFR 50.109(a)(1). Therefore, a backfit analysis has not been prepared. However, as discussed below, the NRC has prepared a 64

regulatory analysis for the proposed rule, which examines the costs and benefits of the proposed requirements in this rule. The Commission regards the regulatory analysis as a disciplined process for assessing information collection and reporting requirements to determine that the burden imposed is justified in light of the potential safety significance of the information to be collected.

VI. REGULATORY ANALYSIS The NRC prepared a draft regulatory analysis for the proposed rule to examine the costs and benefits of the alternatives considered by the NRC, and public comments on this analysis were requested in connection with the proposed rule. As discussed above under the heading "Comment T," some commenters disagreed with the proposition that the rule would reduce reporting burden. These comments were addressed by incorporating changes into the final rule, such that the assumptions in the draft regulatory analysis are sustained, and no changes have been made to the regulatory analysis. The regulatory analysis is available for inspection in the NRC Public Document Room or it may be viewed and downloaded electronically via the interactive rulemaking web site established by NRC for this rulemaking, as discussed above under the heading ADDRESSES. Single copies may be obtained from the contact listed above under the heading "For Further Information Contact."

VII. PAPERWORK REDUCTION ACT STATEMENT This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This rule has been reviewed and 65

approved by the Office of Management and Budget, approval numbers 3150-0011 and 3150-0104.

The annual public reporting burden for the currently existing reporting requirements in 10 CFR 50.72 and 50.73 is estimated to average about 700 hours0.0081 days 0.194 hours 0.00116 weeks 2.6635e-4 months per nuclear power reactor, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collection. It is estimated that the proposed amendments would impose a one-time implementation burden of about 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months per reactor. The recurring annual information collection burden is estimated to be reduced by 132 hours0.00153 days 0.0367 hours 2.18254e-4 weeks 5.0226e-5 months per reacto"t. 4t Send comments on any aspect of this information collection, including suggestions for reducing this burden, to the Records Management Branch (T-6E6), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 or by Internet electronic mail to BJS1@NRC.GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011 AND 3150-0104); Office of Management and Budget, Washington, DC 20503.

Public Protection Notification If a means used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, an information collection.

66

J J "'

VIII. REGULATORY FLEXIBILITY ACT CERTIFICATION In accordance with the Regulatory Flexibility Act (5 U.S.C. 605(b)), the Commission certifies that this rule does not have a significant economic impact on a substantial number of small entities. This proposed rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entitie.s" set forth in the Regulatory Flexibility Act or the size standards established by the NRC (10 CFR 2.810).

IX. SMALL BUSINESS REGULATORY ENFORCEMENT FAIRNESS ACT In accordance with the Small Business Regulatory Enforcement Fairness Act of 1996, the NRC has determined that this action is not a major rule and has verified this determination with the Office of Information and Regulatory Affairs of 0MB.

X. NATIONAL TECHNOLOGY TRANSFER AND ADVANCEMENT ACT The National Technology Transfer and Advancement Act of 1995, Pub. L 104-113, requires that Federal agencies use technical standards developed or adopted by voluntary consensus standards bodies unless the use of such a standard is inconsistent with applicable law or otherwise impractical. There are no consensus standards regarding the reporting of

,..../

safety information by nuclear power plant licensees to regulatory authorities that would apply to the requirements imposed by this rule. Thus, the provisions pf the Act do not apply to this rule.

XI. FINAL AMENDMENTS List of Subject T arms for Parts 50 and 72 67

t 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire prevention, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

10 CFR Part 72 Criminal penalties, Manpower training programs, Nuclear materials, Occupational safety and health, Reporting and recordkeeping requirements, Security measures, Spent fuel.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C.

553, the NRC is adopting the following amendments to 10 CFR Part 50 and 10 CFR Part 72.

PART SO-DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

1. The authority citation for Part 50 continues to read as follows:

Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stal 444, as amended (42 U.S.C. 2132, 2133,2134,2135,2201,2232,2233,2236,2239,2282);secs.201,asamended,202,206, 88 Stal 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C.

5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 955 as amended (42 U.S.C.

2131, 2235), sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections so.1'3, 68

50.54(O.D.), and 50.103 also issued under sec. 108, 68 Stat 939, as amended (42 U.S.C.

2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 l).S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 184, 68 Stat 954, as amended (42 U.S.C. 2234). Appendix Falso issued under sec. 187, 68 Stat. 955 (42 u.s.c. 2237).

2. Section 50.72 is amended by revising paragraphs (a) and (b) to read as follows:

§ 50. 72 Immediate notification requirements for operating nuclear power reactors.

(a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under

§ 50.21 (b) or§ 50.22 of this part shall notify the NRC Operations Center via the Emergency Notification System of:

(i) The declaration of any of the Emergency Classes specified in the licensee's approved Emergency Plan; 2 or (ii) Those non-emergency events specified in paragraph (b) of this section that occurred within three years of the date of discovery.

(2) If the Emergency Notification System is inoperative, the licensee shall make the required notifications via commercial telephone service, other dedicated telephone system, or 1

Other requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained elsewhere in this chapter, in particular §§ 20.1906, 20.2202, 50.36, 72.216, and 73.71.

2 These Emergency Classes are addressed in Appendix E of this part.

69

any other method which will ensure that a report is made as soon as practical to the NRC Operations Center. 3 (3) The licensee shall notify the NRC immediately after notification of the appropriate State or local agencies and not later than one hour after the time the licensee declares one of the Emergency Classes.

(4) The licensee shall activate the Emergency Response Data System (ERDS) 4 as soon as possible but not later than one hour after declaring an Emergency Class of alert, site area emergency, or general emergency. The EROS may also be activated by the licensee during emergency drills or exercises if the licensee's computer system has the capability to -

transmit the exercise data.

(5) When making a report under paragraph (a)(1) of this section, the licensee shall identify:

(i) The Emergency Class declared; or (ii) Paragraph (b)(1), "One-hour reports," paragraph (b)(2), "Four-hour reports," or paragraph (b)(3), "Eight-hour reports," as the paragraph of this section requiring notification of I

the non-emergency event (b) Non-emergency events- (1) One-hour reports. If not reported as a declaration of an Emergency Class under paragraph (a) of this section, the licensee shall notify the NRC as soon as practical and in all cases within one hour of the occurrence of any deviation from the plant's Technical Specifications authorized pursuant to § 50.54(x) of this part 3

Commercial telephone number of the NRC Operations Center is (301) 816-5100.

4 Requirements for EROS are addressed in Appendix E,Section VI.

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(2) Four-hour reports. If not reported under paragraphs (a) or {b)(1) of this section, the licensee shall notify the NRC as soon as practical and In all cases, within four hours of the occurrence of any of the following:

(i) The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.

{iQ -(iii) [Reserved]

(iv)(A) Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the

- actuation results from and is part of a pre-planned sequence during testing or reactor operation.

(B) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

{v) - (x) [Reserved]

(xi) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.

{3) Eight-hour reports. If not reported under paragraphs (a), (b)(1) or (b)(2) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following:

(i) [Reserved]

(ii) Any event or condition that results in:

71

r (A) The condition of the nuclear power plant, including Its principal safety barriers, being seriously degraded; or

{B) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

{iii) [Reserved]

{iv)(A) Any event or condition that results in valid actuation of any of the systems listed in paragraph (b){3){iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

{B) The systems to which the requirements of paragraph (b)(3)(iv)(A) of this section apply are:

(1) Reactor protection system (RPS) including: reactor scram and reactor trip. 5 (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). *

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(4) ECCS for boiling water reactors (BWRs} including: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; isolation condenser system;.and feedwater coolant injection system.

(6) PWR auxiliary or emergency feedwater system.

5 Actuation of the RPS when the reactor is critical is reportable under paragraph (b)(2){iv)(B) of this section.

72

(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

(8) Emergency ac electrical power systems, including: emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.

\

(v) Any event or condition that at the time of discovery-could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition;

{B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident (vi) Events covered in paragraph (b)(3)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (b)(3)(v) of this section if redundant equipment in the same system was operable and available to perform the required safety function.

(vii) - (xi) [Reserved]

(xii) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.

(xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability {e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system) .

73

r

3. Section 50.73 is amended by revising sections {a), {b){2)0i)(F), (b){2)0i)(J), (b)(3), (d),

and (e) and by removing and reserving paragraph (f) to read as follows:

§ 50. 73 Licensee event report system.

(a) Reportable events. (1) The holder of an operating license for a nuclear power plant (licensee) shall submit a Licensee Event Report (LER} for any event of the type described in this paragraph within 60 days after the discovery of the event In the case of an Invalid actuation reported under§ 50.73{a}(2){iv), other than actuation of the reactor protection system {RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. Unless otherwise specified in this section, the licensee shall report an event if it occurred within three years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event.

(2) The licensee shall report:

(i)(A) The completion of any nuclear plant shutdown required by the plant's Technical Specifications.

(B) Any operation or condition which was prohibited by the plant's Technical Specifications except when:

(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions: or 74

(3) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event (C) Any deviation from the plant's Technical Specifications authorized pursuant to

§ 50.54(x) of this part OQ Any event or condition that resulted in:

(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (B) The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.

(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

(iv)(A) Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section, except when:

(1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or (2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed.

{B) The systems to which the requirements of paragraph (a)(2)(iv)(A) of this section apply are:

(1) Reactor protection system (RPS) including: reactor scram or reactor trip.

(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

75

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs)

Including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(4) ECCS for boiling water reactors (BWRs) Including: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.

(6) PWR auxiliary or emergency feedwater system.

(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

(BJ Emergency ac electrical power systems, including: emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.

(9) Emergency service water systems that do not normally run and that serve as ultimate heat sinks.

(v) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident (vO Events covered in paragraph (a)(2)(v) of this section may include one or more

./

procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, 76

,._1, ".

,..,i1 ,.4 .11 <1'l -

construction, and/or procedural Inadequacies. However, Individual component failures need not be reported pursuant to paragraph (a)(2){v) of this section If redundant equipment in the same system was operable and available to perform the required safety function.

(vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or

{D) Mitigate the consequences of an accident (viii)(A) Any airborne radioactive release that. when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 I

times the applicable concentration limits specified in appendix B to part 20, table 2, column 1.

(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds 20 times the applicable concentrations specified in appendix 8 to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.

(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Sh,ut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident 77

(B) Events covered in paragraph (ix)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to paragraph (ix)(A) of this section if the event results from:

(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.

(x) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

(b) * * *

(2) * *.

(F) The Energy Industry Identification System component function identifier and system name of each component or system referred to in the LER.

(1) The Energy Industry Identification System is defined in: IEEE Std 803-1983 (May 16, 1983) Recommended Practice for Unique Identification in Power Plants and Related Facilities-Principles and Definitions.

(2) IEEE Std 803-1983 has been approved for incorporation by reference by the Director of the Federal Register.

(3) A notice of any changes made to the material incorporated by reference will be published in the Federal Register. Copies may be obtained from the Institute of Electrical and Electronics Engineers, 345 East 47th Street, New York, NY 10017. IEEE Std 803-1983 is available for inspection at the NRC's Technical Library, which is located in the Two White Flint 78

North Building, 11545 Rockville Pike, Rockville, Maryland; and at the Office of the Federal Register, 1100 L Street, NW, Washington, DC.

(J) For each human performance related root cause, the licensee shall discuss the cause{s) and circumstances.

(3) An assessment of the safety consequences and implications of the event. This assessment must include:

(i} The availability of systems or components that could have performed the same function as the components and systems that failed during the event, and (ii) For events that occurred when the reactor was shutdown, the availability of systems or components that are needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.

(d) Submission of reports. Licensee Event Reports must be prepared on Form NRC 366 and submitted to the U.S. Nuclear Regulatory Commission, as specified in§ 50.4.

{e) Report legibility. The reports and copies that licensees are required to submit to the Commission under the provisions of this section must be of sufficient quality to permit legible reproduction and micrographic processing.

(f) {Reserved]

79

PART 72-LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE

4. The authority citation for part 72 continues to read as follows:

Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 189, 68 Stat.

929, 930, 932, 933, 934, 935, 954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.2071,2073,2077,2092,2093,2095,2099,2111,2201,2232,2233,2234,2236, 2237, 2238, 2282); sec. 274, Pub. L.86-373, 73 Stat. 688, as amended (42 U.S.C. 5841, 5842, 5846); Pub. L. 95-:601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-486, sec. 7902, 106 Stat. 3123 (42 U.S.C. 5851); sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C.

4332); secs. 131, 132, 133, 135, 137, 141, Pub. L.97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, Pub. L. 100-203, 101 Stal 1330-235 (42 U.S.C. 10151, 10152, 10153, 10155, 10157, 10161, 10168).

Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat. 955 (42 U.S.C. 2239); sec. 134, Pub. L.97-425, 96 Stat. 2230 (42 U.S.C. 10154)'. Section 72.96(d) also issued under sec. 145(g), Pub. L. 100-203, 101 Stat.

1330-235 (42 U.S.C.10165(g)). SubpartJ also issued under secs. 2(2), 2(15), 2(19), 117(a),

141(h), Pub. L.97-425, 96 Stat. 2202, 2203, 2204, 2222, 2224, (42 U.S.C. 10101, 10137(a),

10161(h)). Subparts Kand Lare also issued under sec. 133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 (42 U.S.C. 10198).

80

5. Section 72.216 is revised to read as follows:

§ 72.216 Reports.

(a) [Reserved]

(b) [Reserved]

(c) The general licensee shall make initial and written reports in accordance with

§§ 72.74 and 72.75.

Dated at Rockville, Maryland, this lt>thday of Oc...tobvi..., , 2000.

For the Nuclear Regulatory Commission.

LL.I/,,= -l.;V Annette L. Vietti-Cook, Secretary of the Commission.

/

81

NORTH CAROLINA DEPARTMENT OF ENVIRON~,_.1fA18f3 NATURAL RESOURCES l:>MSION OF RADIATION PROTECTION October 12, 1999

99 DC, 2i P 4 :12 Mr. Dennis P. Allison -**

US Nuclear Regulatory Commission 0 *-

Washington, DC 20555-0001 R',.,

JAMES B. HUNT JR;' ADJl GoVERNOR DOCKET NUMBER

Dear Mr. Allison:

PROPOSED RULE p 50 .J- 1~

to F~ 3ha.'IJ This letter is in regard to the proposed changes to 10 CFR 50.72 and 10 CFR 50.73 presented in the July 6, 1999 Federal Register. These proposed changes would increase the reporting times for certain incidents. There are several events whose required reporting times would increase from four hours to eight hours under the proposed rules. These are:

1. An airborne radioactive release that results in concentrations over 20 times allowable levels in an unrestricted area.

2. Liquid effluent in excess of 20 times allowable concentrations released to an unrestricted area.

3. Radioactively contaminated person transported to an off-site medical facility for treatment.

4. News release or other government agency notification related to the health and safety of the public or on-site personnel, or protection of the environment.

For each of these events, the North Carolina Division of Radiation Protection believes that the required reporting time should be maintained at four hours.

The basis for taldng the above position is that information on any such events will be of interest to the public and public officials. Furthermore, the Division of Radiation Protection provides independent technical advice to State decision-makers as part of its emergency preparedness function. As such, any delay in providing information to the Division of Radiation Protection may prevent or delay decisions on public health and delay announcements to the public on risks due to the incident. State agencies may be able to get such information from licensees, even under the proposed rule.

This can be difficult to do when an incident is actually occurring unless NRC rules mandate that the licensee report events in timely fashion and within a prompt and well-defined period of time.

Thank you for allowing the Division of Radiation Protection to comment on these proposed rules. Please do not hesitate to call me if you have questions.

Sincerely,

(if1!l"o/ ~edged by cam

. t'T ? 6 1999 ue _.,. *--*

Richard M. Fd?cHP 382!5 BARRETT DR. RALEIGH, NORTH CAROLINA 27609-722I PHONE 919-!571-4141 FAX 919-!571-4148 AN EQUAL OPPORTUNITY / Al'FIRMATIVE ACTION EMPLOYER * !50% RECYCLED/I 0% POST-CONSUMER PAPER

I U.S. NU<liARAEGUlATORY COMMISSfON RUI.SMICNll&ADIIDCAIDS STAFF Off1CE CF1ttlll!DETARY CFTtECDlSION D>a nMal:s POIINlltDla 10/ 21 /q, /(ujc(_ ~ ~ ~

Cq)laeAa /

Ackfl

10/20/99 I. Carol Gallagher ,

Z. Rulemakings and Adjudications Staff Please enter the attached letter from Richard M Fry of the State of NC into the comments on the proposed rule to modify the event reporting requirements in 10 CFR 50.72 and 50.73. I received it today via US Mail.

Dennis Allison

W. Glenn Warren, Chairman BW_R OWNERS' GRQUpDOCh::_fED Tel: (205) 992-5940 Fax: (205) 992-0391 wgwarren@southernco.com c/o Southern Nuclear* 40 Inverness Center Parkway* PO Box 1295

Birmingham, AL 35242

99 OCT -4 ~ S :06 Project No. 691

\...,

BWROG-99072 September 17 , 1999 DOCKET NUMBER Secretary PROPOSED AU 50.J--1 :i.

U.S. Nuclear Regulatory Commission Washington, DC [ t,,'f ~R 3<,,.;:>91)

Attention: Rulemaking and Adjudication Staff

Subject:

Federal Register Volume 64, Number 128, dated July 6, 1999, Proposed Rule for Reporting Requirements for Nuclear Power Reactors The purpose of this letter is to transmit Boiling Water Reactor Owners' Group (BWROG) comments on the proposed rule for Reporting Requirements for Nuclear Power Plants.

The Boiling Water Reactor Owners' Group (BWROG) appreciates the NRC effort to resolve comments on this proposed rule. In general the proposed rule is an improvement over the current regulation, however there are some areas where further clarification of detail is needed. Clarification will minimize future regulatory issues with interpretation and ensure regulatory burden is commensurate with the safety significance of reported events .

The BWROG Licensee Event Reporting (LER) Committee has reviewed the proposed rule and the Draft revision of NU REG 1022. The committee has found the proposed rule to be more useable and less burdensome than the previous rule. The committee endorses the concept of revising both the rule and the NUREG concurrently. This will facilitate a more comprehensive understanding of the changes and their impact.

The BWROG seeks a clear and unambiguous guidance document that will reduce the regulatory interpretation issues associated with reporting events. If needed, the BWROG supports delaying the issuance of the final rule to accommodate an additional workshop.

As discussed with Dennis Allison of the staff at the public meeting on August 3, 1999, the BWROG LER Committee is ready to support an additional meeting or workshop in order to better define examples for inclusion in NU REG 1022.

Specifically, in response to the Federal Register Notice, the LER committee provides comments enclosed as Attachment 1. In addition to these comments, the committee has identified specific comments on individual sections and paragraphs. These are included as Attachment 2 to this letter.

JJ.S. UCLEAR REGULATORY COMMISSI01i ULEMAKINGS&ADJUO i\FF OFFICE OF THE OF THE COMMISSIOO Document Sfatistics Postmark Date _ _ _ __ __

Cop1 Received _ _ _ __ _ _

Add'I Coples Reproduced _ _ _ __

Special Distribution,_ _ _ __

BWROG-99072 September 17, 1999 Page2 If you have any questions, please contact Lonnie Daughtery, Entergy, (601--437-2334) or the undersigned.

Very truly yours, W. Glenn Warren, Chairman BWR Owners' Group cc: RM Pulsifer, NRC BWROG Licensee Event Reporting Committee

BWROG Primary Representatives BWROG Executive Oversight Committee JM Kenny, BWROG Vice Chairman LF Daughtery, Entergy TG Hurst, GE KR Fletcher, GE GVine, EPRI A Marion, NEI 2

Attachment 1 This Attachment provides BWROG feedback in several areas as requested in the Federal Register Notice.

First was the request to discuss the proposed reporting times and the need for additional levels. In response to this request the committee recommends that only three levels be utilized. This will minimize operator burden and will facilitate implementation upon issuance of the final rule. Additionally, we think these levels better define the significance of the initial report and will aid the NRC in responding to events in the future.

Secondly public comment was sought concerning the proposal to require reporting of actuations of a prescribed list of systems. Of the options listed in the notice, the committee recommends selecting option (3) return to the pre-1998 situation. This option with the exclusions currently allowed would be the easiest and most straightforward option for implementation.

Public comment was also sought concerning reporting of historical problems. The committee recommends that historical look backs be limited to two years or one operating cycle whichever is shortest. This will enhance the importance of reported events to other utilities while minimizing the burden to the utility making the report. The committee supports adding this limitation to all criteria which requires a review of historical data.

Finally the notice sought public comment on the reporting of component problems. The committee has reviewed this criteria and strongly recommends that it be deleted from the final rule. This criteria reinserts ambiguity and interpretation problems, lowers the significance level of the reporting criteria, adds a tremendous amount of burden back on

the licensees and is redundant in many ways to other reporting mechanisms and criteria .

BWR Owners' Group ATTACHMENT2 Specific comments on 10 CFR 50. 72 SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS 50.72(a)(4) The licensee shall activate the Emergency Response 50.72(a)(4) The licensee shall activate the Emergency Response Data System (ERDS)5 as soon as possible but not later than one Data System (ERDS)5 as soon as possible but not later than one hour after declaring an emergency class of alert, site area hour after declaring an emergency class of alert, site area emergency, or general emergency. The ERDS may also be emergency, or general emergency. Tue ~RDS may also be activated by the licensee during emergency drills or exercises if aetivated b~, the lioensee during emergeney drills or e~Eereises if the licensee's computer system has the capability to transmit the the lioensee' s oomputer system has the oapabili-ty to tnmsmit the exercise data. eKereise data This last sentence should be deleted. It is superfluous information that adds no value to the discussion of reportability.

50. 72(b )(2)(ii)(B) The nuclear power plant being In an unanalyzed 50. 72(b )(2)(ii)(B) The nuclear power plant being lin an condition that significantly affects plant safety. unanalyzed condition that significantly a:ffeots degrades plant safety.

The phrase significantly affects plant safety" has no positive or negative connotation. It is therefore recommended that the term degrades be substituted in place of affects. In addition, a lower case editorial change should be made.

50.72(b)(2)(v) Licensees shall report: "Any event or condition 50. 72(b )(2)(v) Licensees shall report: "Any event or condition that at the time of discovery could have prevented the fulfillment that at the time of discovery is oould have preventmgtea the of the safety function of structures or systems that are needed to: ... ability to fulfill the safety function of structures or systems that are needed to: ...

Editorial change to reflect the correct tense of the existence of an event or condition, rather than past speculation.

AITACHMENT2

BWR Owner's Group Specific comments on 10 CFR 50 73 SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS 50.73(a)(2)(i)(B) Any operation or condition occurring within 50.73(a)(2)(i)(B) Any operation or condition occurring within three years of the date of discovery which was prohibited by the two years of the date of discovery which was prohibited by the plant's Technical Specifications, except when: plant's current Technical Specifications, except when:

(i) The technical specification is administrative in nature; or (i) The technical specification is administrative in nature; or (ii) The event consists solely of a case of a late surveillance test (ii) The event consists solely of a case of a late surveillance test where the oversight is corrected, the test is performed, and the where the oversight is corrected, the test is performed, and the equipment is found to be capable of performing its specified safety equipment is found to be capable of performing its specified functions. safety functioM; or (iii) A previously existing condition does not comply with a newly implemented more restrictive specification.

As previously stated, we believe that two years is the appropriate time period for reporting historical events.

The addition of current" allows for plants that have recently converted to Improved Standard Technical Specifications to apply the current requirements to the identified condition rather than considering Technical Specifications which are no longer applicable.

"Safety function" should be singular to accommodate equipment with only one safety function.

The third provision eliminates the need for a utility to do an unnecessary historical search for conditions that complied with a previous Technical Specification, but do not comply with a newly implemented, more restrictive Technical Specification.

2

BWR Owner's Group ATIACHMENT2 50.73(a)(2)(ii)(B) The nuclear power plant being in (A) an 50.73(a)(2)(ii)(B) The nuclear power plant being in Wan unanalyzed condition that significantly affected plant safety unanalyzed condition that significantly affeeted degraded plant safety.

The phrase "significantly affected plant safety" has no positive or negative connotation. It is therefore recommended that the term degraded be substituted in place of affected. In addition, (A) should be deleted as a minor editorial chanf!e.

50.73(a)(2)(ii)(C) A component being in a degraded or non- S0,'73(a)(l)(ii)(C) l*L 0OfflflOB.Bnt aeiB.g iB. a degraded or aoa conforming condition such that the ability of the component to eoafofl'B:ffig eoadition seee. that the ability of the eOIRflOBBB.t to perform its specified safety function is significantly degraded and perfofffl its speeified safety funetion is signilieaatly degraded the condition could reasonably be expected to affect other similar end the eonditioa eould reasonably ea BKpBeted to a.ffeet other components in the plant. similar eomponents in the plant.

stated, we believe that this new criterion should be deleted.

50.73(a)(2)(v) Licensees shall report: "Any event or condition 50.73(a)(2)(v) Licensees shall report: "Any event or condition occurring within three years of the date of discovery that could occurring within two years of the date of discovery that eeu-14 have prevented the fulfillment of the safety function of structures would have prevented the fulfillment of the safety function of or systems that are needed to: ... structures or systems that are needed to: ...

As previously stated, we believe that two years is the appropriate time period for reporting historical events.

Editorial change to reflect the existence of an event or condition, rather than past speculation.

3

ATIACHMENT 2

BWR Owner's Group 50.73(a)(2)(ix)(J) For each human performance related problem 50.73(a)(2)(ix)(J) For each personnel error human performaBee that contributed to the event, the licensee shall discuss related problem that eoatributed to the ev:ent, the licensee shall discuss The shift from personnel error" and the implied "root cause" to "human performance related problem" and "contributing factors" greatly increases the scope of investigation and burden to the licensee. It is only appropriate to require discussion of personnel error root causes.

50.73(a)(3)(ii) are included in emergency or operating procedures S0,'73(a)(3)(ii) are iaeluded iH emergeaey or opemtiag and could have been used to recover from the event in case of an proeedures and eould hlwe beea used to reeo:i,*er from the e,,,ent additional failure in the systems actually used for recovery. iH ease of an edditioaal failure iH the s~*stems eemally used for reeovery.

We recommend that this new criterion be deleted Emergency operating procedures provide direction for use of many plant systems. If an additional failure must be postulated for every event, multiple systems would be required to be included in the LERfor each safety function. For example, if the reactor scrammed due to personnel error andfeedwater was used to recover from the event, an additional failure of loss of o.ffsite power would require alternate injection methods that could include service water and fire water even though they are systems of last resort.

This rule change would result in a large amount ofadditional information that would be of minimal use. The assessment of the safety consequences and implications of the event would become cluttered with hypothetical additional failures and possible plant responses.

4

ATTACHMENT 2

BWR Owner's Group Specific comments on NUREG 1022, Revision 2 (draft)

SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS HPCS- High Pressure Core Spray LPCI- Low Pressure Core Injection Abbreviations and LPCS- Low Pressure _Core Spray should be added to the list of abbreviations NP RDS and SALP should be deleted since they are no longer used and are not referenced in the NUREG.

Section 2.5, Time Limits for Reporting For example, if a technician sees a problem, but a delay occurs For example, if a technician sees a problem, but a delay occurs before an engineer or supervisor has a chance to review the before an engineer or supervisor has a chance to review the situation, the discovery date (which starts the 6 ~ y clock) is situation, the discovery date (which starts the 30-day clock) is the the date that the technician sees a problem.

date that the technician sees a problem.

Change is needed to align written guidance with the proposed change to required LER submission times.

Section 3.2.1, Plant shutdown required by TS If the shutdown is completed, licensees are required to submit an If the shutdown is completed, licensees are required to submit an LER within 6~ days.

LER within 30 days.

Change is needed to align written guidance with the proposed change to required LER submission times.

5

ATTACHMENT 2

BWR Owner's Group Section 3.2.2, Technical Specification Prohibited It is recommended that this sentence be deleted since this Operation or Condition, TS Surveillance statement appears to contain no basis. The deficiency should be Requirements Discussion assumed to have occurred at the time of discovery (as is the Otherwise, the deficiency should be assumed to have occurred current practice) unless firm evidence exists to the contrary. At halfway between the last successful test or use and the untimely the very least, it should be assumed that the deficiency occurred test that revealed the deficiency. at the time that the late surveillance was due.

Section 3.2.2, Technical Specification Prohibited STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent establishes Operation or Condition, Entry into STS 3.0.3 requirements for actions when: (1) an LCO is not met and the Discussion associated ACTIONS are not met, (2) an associated ACTION is STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent, establishes not provided, or (3) as directed by the associated ACTIONS requirements for actions when an LCO is not met and no action themselves. Entry into STS 3.0.3 (ISTS LCO 3.0.3) for either of statement is provided. Entry into STS 3.0.3 is considered to the first two above reasons are generally reportable under this indicate that a condition existed longer than allowed by TS. Thus, criterion. However, when the plant TS specify the entry into entry into STS 3.0.3 (ISTS LCO 3.0.3) for any reason or 3.0.3 as the required ACTION and that action and its completion justification is reportable. time are met, the event is not reportable under this criterion.

Also, momentary (less than approximately 5 minutes) entries into TS 3.0.3 regardless of the reason, are not reportable under this criterion. Any TS 3.0.3 entry involving actual plant shutdown should be reviewed for reporting under 50.72(b)(l)(i)(A) and 50.73(a)(2)(i)(A) for a plant shutdown required by TS.

The proposed wording is suggested to replace the existing outdated guidance. The proposed wording reflects the current operating philosophy associated with implementation of the Standard Technical Specifications.

6

ATTACHMENT2

BWR Owner's Group Section 3.2.2, Technical Specification Prohibited Sections 50.55a(g) and 50.55a(f) require the implementation of Operation or Condition, Missed Tests Required by ISI and IST programs in accordance with the applicable edition ASM::E Section XI Discussion of the ASME Code for those pumps and valves whose function Sections 50.55a(g) and 50.65a(f) require the is required for safety. STS Section 4.0.5 (or an equivalent) implementation of ISI and IST programs in accordance covers these testing requirements. (Geaerally, there is ao with the applicable edition of the ASME Code for those eompamble nffg seetion.) A missed ...

pumps and valves whose function is required for safety.

This statement should be deleted since ISTS Section 5, STS Section 4.0.5 (or an equivalent) covers these testing "Programs and Manuals," has a section/or these requirements requirements. (Generally, there is no comparable ISTS in "Jnservice Testing Program".

section.) A missed ...

Section 3.2.2, Technical Specification Prohibited Also, if a fire protection deficiency results in the inability to Operation or Condition, Fire Protection Systems preserve at least one safe shutdown train in the event of a fire, it When Required by TS Discussion should be reported evaluated for reporting as an unanalyzed Also, if a fue protection deficiency results in the inability to condition that significantly affects plant safety, as discussed in preserve a safe shutdown train in the event of a fue, it should be Section 3.2.4 of this report.

reported as an unanalyzed condition that significantly affects plant safety, as discussed in Section 3.2.4 of this report. The guidance should be modified to indicate the need to evaluate for reporting, rather than indicate that reporting is definitely required since other trains or systems may be available to preserve safe shutdown.

7

A ITACHMENT 2 BWR Owner's Group Section 3.2.2, Example (3) Entering STS 3.0.3 (3) Entering STS 3.0.3 due to lack of specific TS actions (3) Entering STS 3.0.3 With essential water chillers (A) and (B) out of service, the only With essential water chillers (A) and (B) out of service, the only remaining operable chiller (A/B) tripped. This condition caused remaining operable chiller (A/B) tripped. This condition caused the plant to enter STS 3.0.3 (equivalent to ISTS LCO 3.0.3) for 1 the plant to enter STS 3.0.3 (equivalent to ISTS LCO 3.0.3) for 1 hour, until chiller (A) was restored to service and the hour, until chiller (A) was restored to service and the temperature temperature was restored to within TS limits.

was restored to within TS limits.

An LER is required for this event because, STS 3.0.3 1Nas An LER is required for this event because STS 3.0.3 was entered entered in this case, there were no actions provided in the plant TS for that condition and STS 3.0.3 was entered for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The proposed wording reflects the current operating philosophy associated with implementation of the Standard Technical Specifications. An additional example is also included to clarify this position:

Entry into STS 3.0.3 when the plant TS specify 3.0.3 entry During a surveillance test on the A train of a two-train Standby Gas Treatment (SBGT) system, a condition was discovered on the B train that rendered it inoperable. The test was halted and steps taken to return the A train to a standby readiness condition.

During the restoration, switch manipulations momentarily rendered the A train inoperable. With both trains inoperable, the plant TS specify immediate entry into LCO 3.0.3. The entry into LCO 3.0.3 was logged and then exited within 1 minute once switch manipulation on the A train was completed.

This event is not reportable under this criterion because all the actions specified by the plant TS were completed within the 8

ATIACHMENT2

BWR Owner's Group required completion times. There was no operation or condition prohibited by TS. Also, momentary entries into STS 3.0.3 are not reportable.

Section 3.2.2, Example (5) Seismic Restraints Assume that during an NRG e*,,aluation i.lt is found that an Assume that during an NRC evaluation it is found that an exciter exciter panel for one EDG had lacked appropriate seismic panel for one EDG had lacked appropriate seismic restraints since restraints since the plant was constructed, because of a design, the plant was constructed, because of a design, analysis or analysis or construction inadequacy. Also assume that, u Qpon construction inadequacy. Also assume that, upon evaluation, there evaluation, the EDG is determined to be inoperable there is is reasonable doubt about the EDG' s ability to perform its reasonable doubt about the EDG's ability to perforre its specified safety functions during and after an SSE. specified safety functions dur.ng a-nd after a-n ggE.

The recommended wording deletes superfluous information and equates the ability to perform a specified safety function to operability.

Section 3.2.2, Example (6) Vulnerability to Loss of A.ssume that dDuring a design review it is found that a loss of Offsite Power offsite power could cause a loss of instrument air and, as a Assume that during a design review it is found that a loss of offsite result, auxiliary feedwater (AFW) flow control valves could fail power could cause a loss of instrument air and, as a result, open. Then for low steam generator pressure, such as could auxiliary feedwater (AFW) flow control valves could fail open. occur for certain main steam line breaks, high AFW flow rates Then for low steam generator pressure, such as could occur for could result in tripping the motor driven AFW pumps on thermal certain main steam line breaks, high AFW flow rates could result overload and were therefore declared inoperable. The single in tripping the motor driven AFW pumps on thermal overload. turbine driven AFW pump would not be affected.

The single turbine driven AFW pump would not be affected.

The recommended wording deletes superfluous information clearly indicates that the equipment is determined to be inoperable (for a time greater than allowed by Technical Specifications).

9

ATTACHMENT 2 BWROwner's Group Section 3.2.4, Degraded Condition, Example (1) .Efollowing the end of cycle shutdown, ...

following the end of cycle shutdown, ... The event is reportable because the cladding failures exceed The event is reportable because the cladding failures exceed expected values fil1Q;- are unique or widespread, and are eaHsed expected values, are unique or widespread, and are caused by by 1:HleKpeeted faetors.

unexpected factors.

Minor editorial change (capitalization). Whether the factors are unexpected or not, are not relevant to reporting.

Section 3.2.4, Degraded Condition, Example (2) The event is reportable beeause Eee degradation eannet ae The event is reportable because the degradation cannot be eeBflidered aeeeptable as isof the generic implications.

considered acceptable as-is.

Many conditions identified by the licensee are not acceptable as-is. It is therefore more appropriate to indicate that the event is reportable due to the generic implications.

Section 3.2.5, External Threat to Plant Safety, It is proposed that these examples be deleted The licensee was Examples (2) and (3) required to make a report due to entry into the Emergency Plan (2) Hurricane making reporting under this category unnecessary.

A licensee in southern Florida declared an Unusual Event after a hurricane warning was issued by the ...

(3) Fire With the unit at 100-percent power, the control room was notified that a forest fire was burning ....

ATIACHMENT2 BWR Owner's Group Section 3.2. 7, Event or Condition That Could ... regardless of whether the system was needed at the time of discovery.

Prevent Fulfillment of a Safety Function, Addition is needed to be consistent with the narrative m the next paragraph Discussion

... regardless of whether the system was needed at the time.

Section 3.2. 7, Event or Condition That Could This portion ofthe discussion should be deleted. Operability determinations are performed by the licensee using Generic letters 9 /-/ 8 and 91-18 Rev. I.

Prevent Fulfillment of a Safety Function, It is inappropriate to include operability determinaJton guidance in this Discussion reportability guideline.

The staff believes that the conditions necessary to consider the redundant train operable and available, for this purpose, should include the following:

in cases where the redundant train should operate automatically, it is capable of timely and correct automatic operation, or in cases where the redundant train should be operated manually, the operators would detect the need for its operation and initiate such operation, using established procedures for which they are trained, within the needed time frame, without the need for troubleshooting and repair, and;

the redundant train is capable of performing its safety function for the duration required, and; there is not a reasonable expectation of preventing fulfillment of the safety function by the redundant train 11

BWR Owner's Group ATTACHMENT2 Section 3.2.7, Event or Condition That Could Prevent No, a RCIC failure is not reportable. RCIC is included in ISTS Fulfillment of a Safety Function, Example (2) because it meets criterion 4 of 10CFRS0.36 based on its If the plant's safety analysis considered RCIC as a system needed contribution to the reduction Qf overall plant risk. RCIC is not to mitigate a rod ejection accident (e.g., it is included in the credited in the plant's safety analysis. If the plaefs safe~,

Technical Specifications) then its failure is reportable under this aeal;csis eeesiaefeel RGIG as a s;cetem eeeeleel te mitigate a fee criterion; otherwise, it is not reportable under this section of the ejeetiee aeeiaem: Ee.g., it is ieehiaeel ie the +eebH::ieal rule. Speeit:ieatiees) theft its fail'Hfe is Fepeftaele 'Heaef this efitefiee; ethern*ise, rt is eat Feperta-ble 'Heaef this seetien ef the rule.

This clarification will correct problems associated with RCIC being included in TS, but not credited in a plants safety analysis.

Section 3.2.7, Event or Condition That Could Prevent The independent failure (e.g., excessive set point drift) of a Fulfillment of a Safety Function, Example (13) single pressure switch is not reportable unless it alene 0ffilla The independent failure (e.g., excessive set point drift) of a single would have caused a system to fail to fulfill its safety function, pressure switch is not reportable unless it alone could have caused or is indicative of a generic problem that could have resulted in a system to fail to fulfill its safety function, or is indicative of a the failure of more than one switch and thereby cause one or generic problem that could have resulted in the failure of more more systems to fail to fulfill their safety function.

than one switch and thereby cause one or more systems to fail to fulfill their safety function. Wording change is required to coincide with the proposed wording change of the rule (i.e. deleting "alone" and replacing "COU Id" Wl*th "WOU Id") .

Section 3.2.7, Event or Condition That Could Prevent . . . event or condition that alene--could prevent fulfillment of a Fulfillment of a Safety Function, Example (17) safety function). (This Feperting efitefiee is d.iseussea ie Seetien

... event or condition that aleH:e--could prevent fulfillment of a 3 .3 .3 ef this Fepert.)

safety function). (This reporting criterion is discussed in Section 3.3.3 of this report.) This line which refers to section 3.3.3 should be delete~ This section no longer exists in NUREG 1022.

12

BWR Owner's Group ATIACHMENT 2 Section 3.2.7, Event or Condition That Could Prevent Examples (17), (18), (19) and (20) should be renumbered as .

Fulfillment of a Safety Function, Examples (17), (18), (19) and (18), (19), (20) and (21) due to the duplication of example (17).

(20)

Section 3.2.11, Contaminated Person Requiring A contract worker experienced a back injury lifting a tool while Transport Offsite, Example (1) working in a contaminated area the reaetor eonta-mmeB-t and was A contract worker experienced a back injury lifting a tool while considered potentially contaminated because his back could not working in the reactor containment and was considered potentially be surveyed.

contaminated because his back could not be surveyed.

The example should be changed such that the employee was working in a contaminated area. In the given example it is possible that the employee was working in a radiologically controlled area that was not contaminated It could therefore be assumed that the worker was not contaminated, eliminating the need to make a report.

5.2.1(1), Narrative Description or Text (NRC Form 366A, Item For equipment that was inoperable at the start of the event,

17) provide an estimate of the time the equipment became For equipment that was inoperable at the start of the event, provide inoperable and the last time the equipmeB-t 1,i;:as kBown to be an estimate of the time the equipment became inoperable and the operable. Indicate the basis for this conclusion (e.g., a test was last time the equipment was known to be operable. Indicate the successfully run or the equipment was operating). For basis for this conclusion (e.g., a test was successfully run or the equipment that failed, provide the failure time and the last time equipment was operating). For equipment that failed, provide the the equipment was known to be operable. Also provide the basis failure time and the last time the equipment was known to be for the last time known operable.

operable. Also provide the basis for the last time known operable.

Deletion of the last time equipment was known to be operable is necessary to conform with !SI'S philosophy of equipme_nt being operable since the last performance of the surveillance test (unless firm evidence exists to the contrary.

13

/Jiff OOCKET~D Northern States Power Company USNRC 414 Nicollet Mall Minneapolis, MN 55401

. Sff 29 p 3 :4OTelephone (612) 330-5500 99 September 20, 1999 r

Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 5"0 J- '1 :2..

( 6'l/FR3&, ~t/ I)

ATTN: Rulemaking and Adjudication Staff

SUBJECT:

Proposed Rule for Reporting Requirements for Nuclear Power Reactors -- 64 Federal Register 36293 -- July 6, 1999 Northern States Power Company (NSP) has reviewed the Proposed Rule for Reporting Requirements. NSP supports the industry comments submitted by NEI in their September 1999 submittal on the subject matter.

NSP has no additional comments on the subject at this time. There is one section specific to NUREG 1022, Revision 2 (draft) that is worth additional attention.

3.2.4, Degraded Condition, (A) Nuclear power plant, including its principal safety barriers, being seriously degraded, introduces a new section (3):

NSP supports NEI' s proposal to delete this section since this position is based on a Draft

Regulatory Guide {DG-1074, Steam Generator Tube Integrity) that has not been approved. Discussions between the industry and the NRC are being held to define the steam generator program and Technical Specification requirements. Recognizing that these agreements are still evolving, the proposed revisions to the rule(s) and NUREG 1022 must agree with the final positions on steam generator issues.

We respectively request that our comments be considered in future Commission action on this matter.

Sincerely, Lt/1,/t/l!!f~

Mark V. McKeown NSP Nuclear Generation c: Roger 0. Anderson Richard Pearson Ralph Beedle (NEI)

ctm edged by card._..._ __

PostmtHk Date q /;). o / q 1 Copies Rec *ved I eprdur,.. L/-

- - ~

-$,~~~~,PD£.~ /llr I

Charles H. Cruse Calvert Cliffs Nuclear Power Plant Vice President 1650 Calvert Cliffs Parkway Nuclear Energy Lusby, Maryland 20657 nuo~l* .. I :... J 410 495-4455 A Member of the Constellation Energy Group September 22, 1999 V "KET UMBER U. S. Nuclear Regulatory Commission PROPOSED RU Washington, DC 20555 ATTENTION: Secretary, Rulemakings and Adjudications Staff

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Proposed Rulemaking for Reporting Requirements for Nuclear Power Reactors; 64 Federal Register 36293, July 6, 1999 Baltimore Gas and Electric Company submits comments on the subject rulemaking for 10 CFR 50.72 and 50.73, Reporting Requirements for Nuclear Power Plants, and associated guidance, NUREG-1022, Revision 2, Event Reporting Guidelines, in Attachment A.

In general, we support the proposed changes to these rules noticed in 64FR36293 - July 6, 1999 and NUREG-1022, Revision 2. We also applaud the Nuclear Regulatory Commission staff efforts to work with interested parties, including the nuclear power industry and the Nuclear Energy Institute, to develop a consensus concerning the subject rulemaking and changes to NUREG-1022.

We endorse the comments made by Nuclear Energy Institute concerning the rulemaking and associated guidance, and add the comments in Attachment (1) concerning the proposed rulemaking. We also have added comments concerning draft NUREG-1022, Revision 2 in Attachment (2).

Should you have questions regarding this matter, we will be pleased to discuss them with you.

~1;;"'~--h C~arle:~. Crus:

Vice President-Nuclear Energy 1?

CHC/CDS/bjd Attachments: (1) Comments on Proposed Rulemaking for Reporting Requirements for Nuclear Reactors; 64 FR 36293 (2) Comments on NUREG-1022, Revision 2 S£ 3 0 1999

~cknowledged by card ._...........--""'

1.R REGULATORY COMMISSIO,~

~.

,,,stm*n< Dot* q/,?. t/ /ci :J Cf ip1es FtOO:3/4ivea -- - I -:-r --

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Rulemakings and Adjudications Staff September 22, 1999 Page2 cc: Document Control Desk, NRC H. J. Miller, NRC R. S. Fleishman, Esquire Resident Inspector, NRC J. E. Silberg, Esquire R. I. McLean, DNR S. S. Bajwa, NRC J. H. Walter, PSC A. W. Dromerick, NRC

ATTACHMENT (1)

COMMENTS ON PROPOSED RULEMAKING FOR REPORTING REQUIREMENTS FOR NUCLEAR REACTORS; 64 FR 36293 Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant September 22, 1999

ATTACHMENT (1)

COMMENTS ON PROPOSED RULEMAKING FOR REPORTING REQUIREMENTS FOR NUCLEAR REACTORS; 64 FR 36293

1. Licensee Event Report Format and Content We believe that utilization of the "check-the-box" format for Licensee Event Report (LER) reporting should be reconsidered. The "check-the-box" approach would reduce the burden on licensees for LER report preparation while still supplying the Nuclear Regulatory Commission (NRC) with the information they need to conduct their mission. A "check-the-box" format could be crafted that would make LER data entry easier, more consistent, and less ambiguous. We believe that a "check-the-box" format can be developed that will not make LERs more difficult for the general public to understand, but actually increase public understanding by providing more concise and consistent information across the industry concerning reportable events.

2. Use of 10 CFR 50.73(a)(2)(v) as a Direct Input to NRC Performance Indicator Program Through our participation in the development of the new NRC Performance Indicator Program, we have become aware that each event reported under 10 CFR 50.73(a)(2)(v), "Any event .. . that alone could have prevented the fulfillment of the safety function of structures or systems ... " will be counted as a Safety System Functional Failure. It appears that the reportability criteria and the Safety System Functional Failure Performance Indicator are not entirely analogous. The Performance Indicator Program is being proposed to track actual Safety System Functional Failures while 10 CFR 50.73(a)(2)(v) requires reporting of events that alone could have caused a failure of a safety system. Many of the events reported under 10 CFR 50.73(a)(2)(v) do not result in actual safety system failures but potential safety system failures. We understand the need of NRC for information regarding potential safety system failures. However, we suggest careful and coordinated consideration among the NRC staff responsible for this rule and the Performance Indicator Program to ensure that use of 10 CFR 50.73(a)(2)(v) reports as indicators of safety system functional failures is reviewed to ensure that the reports are not being misapplied.

3. Enforcement Criteria Discussion in Proposed Rulemaking We believe that the proposed categorization of failures to provide required 1-hour or 8-hour non-

emergency telephone notifications pursuant to 10 CFR 50.72 as Severity Level III violations is too harsh in most cases. In most cases, the information provided in these non-emergency notifications has low safety significance. We suggest that Section IV.7 Enforcement, paragraph (l)(b) be modified to include only missed notifications that have safety significance; or relocated to an example of a Level IV violation; or deleted.

ATTACHMENT (2)

COMMENTS ON NUREG-1022, REVISION 2 Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant September 22, 1999

ATTACHMENT (2)

COMMENTS ON NUREG-1022, REVISION 2 Section 2.5, Time Limits for Reporting The fact that reporting times in 10 CFR 50.72(b)(l) and (b)(2) are keyed to the "occurrence of the event or condition" and LER submittals are keyed to "discovery of the event or condition" is confusing. Often, "discovery" of conditions reportable under 10 CFR 50.72 occurs at a time that is greater than the allowed time period for making a telephone notification. For example, it is rare for reportable events involving engineering analysis or equipment fabrication issues to be discovered at the same time that they actually occur. Such delayed discoveries are common and would result in frequent violations of the 1-hour and 8-hour notification requirements by licensees. We feel that requiring the 1-hour and 8-hour clock to start "from the occurrence" of an event is often an unrealistic or impossible requirement to meet. We suggest the wording of the rule be changed to include "of the discovery" in place of "of the occurrence." [see paragraphs 10 CFR 50.72(b)(l) and (b)(2)]

Section 3.2.4, Degraded Condition We have always viewed 10 CFR 50.72(b)(l)(ii) and 50.73(a)(2)(ii) as very serious events with high thresholds. Significant degradation of principal safety barriers and unanalyzed conditions that significantly affect plant safety should include events where the plant's safety analyses would not be met in the event of a design basis accident and cases where unanalyzed conditions significantly affect plant safety.

One of the examples of unanalyzed events does not appear to significantly affect plant safety. The example is a missing fire barrier, such that the required degree of separation for redundant safe shutdown trains is lacking (pg 33 of draft NUREG-1022, Revision 2). While we agree this is an unanalyzed condition, we question whether this type of event "significantly" affects plant safety. Under the example cited above, most cases of inoperable Appendix R fire penetration seals would be considered reportable events. Such events do not strike us as meeting the threshold of "significantly" affect plant safety, because they are normally mitigated by the availability of suppression and/or detection systems as well as other fire prevention strategies such as control of combustible loading, etc. We suggest the example be deleted.

Section 3.2.8, Common Cause Failures of Independent Trains or Channels Examples (1), (2), and (4) of this section would, in our judgement, also be reportable under 50.72(b)(2)(iii) and 50.73(a)(2)(v), "Any event or condition that ... could have prevented fulfillment of the safety function ... " We suggest that more concise examples be used that are reportable only under 10 CFR 50.73(a)(2)(vii). Use of such examples will serve to better distinguish the differences between these reporting criteria and make this rule easier to understand by licensees.

Section 4.2.4, ENS Event Notification Worksheet (Form 361)

The ENS Event Notification Worksheet appears to have two errors. On page 1 of the sheet under the section titled, "4-Hr Non-Emergency 10 CFR 50.72(b)(2)" are two line items, "(iv)(A) Air Release> 2x App B" and "(iv)(B) Liq. Release> 2x App B." These line items should read "(iv)(A) Air Release > 20x App B" and "(iv)(B) Liq. Release> 20x App B." In addition, the Event Notification Worksheet must be revised to reflect the changes to 10 CFR 50. 72 and 73 and Section 4.2.4 of NUREG-1022, Revision 2 must be revised to reference the new version of the Event Notification Worksheet.

OOCKF. ro

J ALLIANT ENERGY ..

us' '

IES Utilities Inc.

99 SEP 21 P2 :c4 Duane Arnold Energy Center 3277 DAEC Road September 20, 1999 Palo, IA 52324-9785 NG-99-1296 0 Office: 319.851.7611 11 Fax: 319.851.7986 Secretary, U. S. Nuclear Regulatory CommissicfriPL www.alliant-energy.com Washington, DC 20555-0001 ATTN: Rulemakings and Adjudications Staff DOCKET NUMBER

Subject:

Duane Arnold Energy Center (DAEC) PROPOSED RULE Docket No: 50-331 Op. License No: DPR-49 DAEC Specific Comments on 10 CFR 50.72 and 10 CFR 50.73 Rulemaking and Draft Revision 2 to NUREG 1022.

File: A-100

Dear Sir:

Federal Register (FR) Vol. 64, No 128 dated July 6, 1999, published draft rules and guidance associated with the event reporting requirements of 10 CFR 50.72 and 10 CFR 50.73. The DAEC participated in and fully endorses the comments being forwarded to your staff through the Nuclear Energy Institute. Enclosed are the DAEC specific comments on the proposed rule changes and associated guidance discussed in the FR.

We are providing a mark-up to draft Revision 2 to NUREG 1022 with a description and basis for each change for your review and consideration.

If you have any questions regarding this matter, please contact my office at 319-851-7801.

Sincerely,

~~~4 Kenneth E. Peveler, Manager, Regulatory Performance Attachments:

(1) Descriptions and Bases for Proposed Changes (2) DAEC Mark-up to NUREG-1022, Revision 2 draft cc: NRC Document Control Desk D. Allison, NRC-NRR DOCU

~eknowfedged by card ...~.~J.Q ..l!!L.,x,

U.S. UCLEAR REGULATORYCONIMISlSION RULEMAKINGS &AOJUDICATDtB Sl'AFF OFACE OF THE SECRETARY OFTHE DocllnentSlfitlstts Postmark Date q ~0 opies Received __________

Add'I .

~ ~ ~ ~ ~~ t:..l::.~'-

Attachment 1 to NG-99-1296 Page 1 of3 Descriptions and Bases for Proposed Changes

1. Editorial.

2. The DAEC endorses NEI comments regarding the deletion of this new reporting criterion regarding component degradation and generic implications. If however, deletion does not occur, we are submitting a suggested re-wording of the rule and NUREG markup of the new examples that we feel captures a similar intent, but at a more clearly defined higher threshold than what the draft rule requires. This particular reporting criterion, by the significance implied in the first two paragraphs (A and B), carries (and always has carried) a context of an event with significance higher than most of the remaining criteria. We believe our proposed wording better aligns the significance of this new criterion with that implied by the existing two paragraphs. Also, as written, the draft rule and guidance are redundant to the guidance already provided by 10 CFR 50.72(b)(2)(iii), 10 CFR 50.73(a)(2)(v) and 10 CFR 50.73(a)(2)(vii) and their associated portions ofNUREG 1022. If the staff does not believe this to be the case, perhaps added guidance to those criteria would be better than creating a new criterion.

Also included by means of our markup, is an attempt to allow operability determinations required by GL 91-18 (and the associated inspection manual guidance for operability determinations) to be the determining factor for event reporting. Event reporting should be based on the conclusions made via the operability determination process. In this draft, as well as other areas of this guidance, many terms such as "significantly degraded safety function" are used which have no definitions other than in this event reporting guidance. The event reporting thresholds should be defined such that when functions are actually lost or a Structure, System, or Component (SSC) is declared inoperable, then applicable event reports are made. Anything less results in event reports with little to no actual safety significance, which does not

3.

contribute to the NRC "cornerstone" of developing public trust This added exception is to eliminate reporting of invalid actuations ofESF components when the safety function was not required, such as during a refueling outage. Many events that otherwise are reported during such plant conditions have absolutely no significance. An example, which is also provided later in this submittal, is actuation of a containment isolation signal that occurs while the primary containment is open for personnel access during an outage.

4. Of the three choices discussed in the FR notice, the DAEC prefers to return to the pre-1998 guidance associated with the ESF actuation criterion. Comments #5 and #6 are provided in the event that the list of risk significant systems is used in the final rule.

5. As written, the proposed new rule deleted the previous exclusion for RWCU system isolations; a subject of previous rulemaking. It is our desire to keep this exclusion.

6. If this list were to stay, we recommend deletion of the service water systems. There are many "operational" features of these systems that result in automatic starts that are not safety related but would become reportable under this rule (e.g. pump start on high suction strainer differential pressure during normal system operation or testing).

Attachment 1 to NG-99-1296 Page 2 of 3

7. Change wording to a present tense context that is applicable to 50.72 event reporting.

8. Change to would" versus "could" to reflect a conclusion or an absolute condition, especially for historical events. The threshold should be based on what "would" have happened not "could," which is always the subject of interpretation. Consider the same change to the sister requirement in 50.72.

9. Editorial change. The phrase is redundant to the term operable.

10. Adding the three-year limit to be consistent with proposed change to 50.73(a)(2)(v).

11. Editorial, to be consistent with change to 60 days.

12. Editorial, to be consistent with change to 60 days.

13. Change to make the statement become actual NRC guidance. As written, a statement of what another licensee did or how they reported is not guidance. Though it is usually implied as the correct report, it is requested that the staff endorse the decision.

14. See comment #13 above.

15. Editorial.

16. Change is to answer the question without adding caveats that take away from clarity.

17. Recommend deletion. This guidance adds to the reporting burden when compared to previous guidance regarding "firm evidence." There is no documented basis for the "halfway" or the associated increase in reporting that it implies.

18. The current guidance that all 3.0.3 entries are reportable, is not supported by the rule itself. If the TS are followed when the TS require entry into 3.0.3 as an action statement, how can there have been an operation or condition prohibited by TS? This insert is based on guidance from Standard TS.

19. Change to reflect the need to evaluate under a separate criterion. As written, the guidance should be relocated. Also, the conclusion stated does not appear to allow credit for designs with multiple trains of safe shutdown. Again, if that function is not lost, what are we reporting? TS define limits for how long a plant can operate with one train inoperable. How can you be in compliance with your TS, but be considered in "an unanalyzed condition that significantly affects plant safety?"

20. Adding two examples to build upon comment #18.

21. Recommend deletion. This guidance adds to the reporting burden when compared to previous guidance regarding "firm evidence." There is no documented basis for this guidance to make such a conclusion.

22. Change to implement the determination of operability and inoperability as the threshold for reporting.

23. Change to implement the determination of operability and inoperability as the threshold for reporting.

24. Change to make the statement become actual NRC guidance. As written, a statement of what another licensee did or how they reported is not guidance. Though it is usually implied as the correct report, it is requested that the staff endorse the decision.

25. See comment #2.

26. Delete reference to "unique." Unique is open for interpretation and does not add clarity to the guidance.

27. Changes to example to coincide with comment #2.

Attachment 1 to NG-99-1296 Page 3 of3

28. Delete regardless of other changes. This example is more related to the quality and adequacy of that licensee's operability determination rather than reporting guidance.

29. Changes to example to coincide with comment #2.

30. Editorial.

31. Delete the word unique per comment #26, and editorial to reflect tense.

32. The answer or position given adds confusion, especially the term "as-is," which is a term applied to non-conformance dispositions.

33. Change to make the statement become actual NRC guidance. As written, a statement of what another licensee did or how they reported is not guidance. Though it is usually implied as the correct report, it is requested that the staff endorse the decision.

34. See comment #3.

35. See comment #5.

36. See comment #6.

37. Change to keep the context of risk significance, if the list were to remain in the rule .

38. Change to keep the context of risk significance, if the list were to remain in the rule.

39. Change to improve the definition of"invalid" by adding examples.

40. Delete portion of definition that adds no value.

41. Editorial and additional guidance added to build upon comment #3.

42. Change to alter the conclusion stated which contradicts the conclusion given in example (2) on page 44. Note that the procedural step referenced in this example (2) does not state "definitely" expected to occur (and rod blocks are not). The conclusion given in example (7), as written, adds a higher threshold to the guidance that is not supported by the rule. In order to maintain the implied threshold, the rule would need to be changed to something like, " ... except when: (1) The actuation resulted from and occurred precisely as stated in a pre-planned sequence during testing or reactor operation. "

43. See comment #8.

44. Delete. This guidance appears to be more applicable to GL 91-18 and associated guidance for operability determinations and, as such, should be deleted or relocated to those documents.

45. For the content of this guidance to stand as-is, such that loss of or unplanned RCIC LCOs are reportable as a single train safety system, the Rule would need to be changed. Currently the Rule does not support the more stringent requirement implied by the guidance because RCIC typically does not fulfill one of the listed safety functions (A through D). According to STS, NUREG-1433, Rev.I dated 04/07/95, Bases section 3.5.3, "The RCIC system is not an Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC system operation."

46. Editorial.

47. Change to be consistent with the staffs proposed change which deletes the reference to "alone could have."

48. Editorial.

49. See comment #10.

Attachment 2 to NG-99-1296, DAEC Mark-up to NUREG-1022, Revision 2 Draft

(B) AnY operation or condition occurmg within three years of the date of cDscoyery which was prohibited by the pwrt's Technical Specifications~

except~

Q1 The te<:hnical specification is adrnJofstlat;ve in natuta: or

@ n,e event consists so1e1v ot a case ot a 1ate fY{V8illance test whD the QY91'IA'lt ii corrected, the test is pe,:formed, and the aaujpment is found to be capabJe of pertqnuna its specified safety functions.

[see {bX1), llbove, for the 50.72 criterion (C) Any deviation from the plant's Technical regarding 50.54(x}J Specfficatlons authorized pursuant to §50.54(x) of this part (il1 Any event or condition dttring opel'a-tion that (ii} Any event or condition that resulted in:

results in: W The condition of the nuclear power plant,

.{!l The condition of the nuclear power plant, including its principal safety barriers. being seriously including its pnncipal safety b a r r i eg r s,~-

serio degraded;_or that rastttled in degradec:t_or ,eStllts irl #1, fill The nudear power plant being in (A) an

!ID The nuclear power plant * . an unanalyzed condition that signfflcan1fy compromisod unanalyzed condition that significantly compremtaes ~

affects plant safety. jf i~sut .1-(B} In a condition hlt i:s 01:1tside the design basis of the plant: or basis of the ple:i ,t; or f6~ In a eondition not cowered~ tf<1e plant's (0) In a condition not CO' ered by the pfarltls operating and emergency procedures.* operating m,d eme1geney procedmes."

[relocated trom fbX1J(liJ, 1 houri (ii) Any natural phenomenon or other external (ii) Any natural phenomenon or other external condition that poses an actual threat to the safety of condition that posed an actual ttveat to the safety of the nuc1ear power plant or signiflcantty hampers site the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for personne! in the performance of duties necessary for the safe operation of the plant. the safe operation of the nuclear power plant.

[relocated tram fbX1 )(1il}, 1 houri

~ Any event or condtion that results in a- frv}(A} Any event or concition that resulted in 1!l"'

intentiona! manual actuation or valid automatic manual or automatic actuation of any actuation of any ..

engineered safety fealule fESA, inch2dinR the reactor eiigineered 38fety fetllure (CSA, iltcitldlnr, the reactor p1otectiou s,stem (RPS} of1hesystems listed in protection s,stem (RPS' of 1he systems listed in paraq:aph (b){2)flV}(B) of this seetlon except~ paragraph (a}{2)(iy)(B) of this section except when:

W the actuation resutts from and is part of a fA:rfJl The actuation resutted from and was part pre-planned sequence during testing or reactor of a pre-planned sequence during testing or reactor operation;-. operation;_gr

fBJ The acttJstien is i1he:fid mitt ~ The actuation was invalid amt.

(1) Occurs ,mile the S)Steffl is properly ffrDJ. Occurred while the system was property removed from :sec , iee, removed from servtce;_gr (fl) Ocet:!rs after the :safeey fttl ietion has been -(e}-/81 Occurred after the safety function had already eempiete~ er been already cornpteted...er ~ or (jii)

Draft NUREG~1022, Rev. 2 6

Insert 1, page 6, NUREG 1022, Rev.2 draft (C) Multiple similar components being in a degraded or non-conforming condition such that the ability to perform a specified safety function was actually lost .

Insert 2, page 6, NUREG 1022, Rev .2 draft (iii) Occurred during a plant operating mode when the safety function was not required."

1<aJ tn,ohos onl) the following :,peeifie ESFs 1([JJ ln,ohed o,q the foffe-oing :,peeifie ESFs er their equi,afeut systems. ffl Reaeter wate, elem, ttp or their eciui,alent syeteffl!t fiJ Reeeter uate1 elem, up system: fiiJ Oonttol roo,n emergency ,entilstien e)stem, fH) Oentrol roerr, erneigenq ,entilatien 9'f9leffl': (iiiJ Awtei1 building ,entiletion system, tr;) eystem, {iii; Reeeter bttikiug ,eoliation :syetern. (w; Ftlel bttilclin; wefttilatio:n a,etem, or 1',j Ati:,cilie~ Ft1el building .entilat1011 system, or M A:imlil!tl'y bt:lildi:ftg ,enttletiou system. buildi~ ,entiletion.

{8) The systems to whjch the reaukements of paragraph la)(2)fry)(Al of 1his1ection ar,pty are; ft I Reactor motectfon awtem {reactor scram.

reactor trio),

(2) Emergency core coofina systems lECCSl for pressurized water reactors (pWRs) incluc:ing: high-head, lnten'nediate-h atd 10!:btod iniectjon systems and the low DJ'9HUffl injection function of residual (decay) heat remoyal systems, f3J ECCS for bojlng water reactors lBWRsl includjna; blatHJressur9 and kJw-oresaure core spray systems: higtH)ressure coolant iniedion swtem; feedwatercoo'8nt iniection mtem; low Dl'19S8Ul9 pectfon furdion of the residual heat removal swtem:

and autpmatic dapreplN'7atfoo system,

£41 BWB ieqation conclfflp system and reactor core l§ofaUor) cooling mtem.

RayxlHaryfeedwater system. l5JPWRauxiliaryfeedwatersystem.

smms fncludlng: £BJ cgutai,nnent mtems inc!udinq; containment and .

the M Anv event or cotdtion fulfillment l/e8(S of1:he date of disc0vefY that safety that an, prewnted the fulfillment " the safety needed to: structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a (A) Shut down the reactor and maintain it in a safe shutdown condition. safe shutdown condition; (B) Remove residual heat, (B) Remove reskiJal heat; (C} Control the release of radioactive material, (C} Control the release of ratloac1ive material; or or (D) Mitigate the consequences of an acc:ident (D) Mitigate the consequences of an accident.

7 Draft NUREG-102:Z Rev. 2

(vii} Any event where a single cause or cond1tion caused at least one independent train or channel to become inoperable In multiple systems or two independent trms or channets to become inoperabJe in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition (B) Remove residual heat; (C) Control the release of radioactive material; or (0) Mitigate the consequences of an accident.

~ (A) Any airborne racioactive release 1hat, (viii)(A) Any airborne radioactive release 1hat, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , results in when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , resulted In concentrations in an unntStiicted area that exceed 20 airborne radionuctide concet 1trations in an unrestricted times the applicable concentration specified in area that exceeded 20 times the applicable appendix B to part 20, table 2, column 1. concentration limits specified in appendix B to part 20, table 2, column 1.

(B) Any Uquid effluent re&ease that, when {B) Any liquid effluent release that. when averaged over a time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds 20 times the averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds applicable concentration specified in appendix B to pan 20 times the appficable concentrations specified in 20, tabMt 2, cokann 2, at the point of entry into 1he appendix B to part 20, table 2, column 2, at the pomt of receiving waters (Le., un.estJictad area) for aD entry into the receiving waters fut.. unrestricted area) radionudides except tritium and dissolved noble gases. for aft radionuclides mccept tritium and dissolved noble gases.

Immediate "elifieaticns made ttndef this pmagnq,h

satist, ti'le reeitti1ements ef §28.:2282 of this ehepter.)

(ix) Reports stibnlitted tc ti'le Gemmiasicn in accordance Nilh pmagiltJ'h (a-)f.2,triij of lhi9 oeeMll1 a:lse meet !he efflttel'lt rek:ae repotting r e ~ t s et §20::2209fa,(8) of this ehapter.

~ Any event that poses an actual threat to (ix) Any event that posed an actual threat to the the safety of the nuclear power plant or significantly safety of the nuclear power p!ant or sigrrifican1ly hampers site persomel in the perfonnance of duties hampered site personnel In the pedormance of duties necessary for the safe operation of the nuclear power necessary for the safe operation of the nuclear power plant inclU<ing fires, toxic gas releases, or radioactive plant indudlng fires, toxic gas releases, or radioactive releases releases.

£relocated from fbX1 XvfJ. 1 hourJ ftt:00 Any event requiring the transport of a radioacliYety contaminated person to an offsite medicaJ facility for treatment.

Draft NUREG-J 022, Rev. 2 8

(!Jll>

situation, the discovery date (which starts the@day clock) is the date that the technician sees a problem.

In some cases, such as discovery of an existing but previously unrecognized condition, it may be necessary to undertake an evaluation in order to determine if an event or condition is reportable.

If so, the guidance provided in Generic Letter 91 ~ 18, "Information to Licensees Regarding two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability,* which applies primarily to operabirlty determinations, is appropriate for reportability determinations as well. This guidance indicates that the evaluation should proceed on a time scale commensurate with the safety significance of the issue and, whenever reasonable expectation that the equipment in question is operable no longer exists, or significant doubts begin to arise, appropriate actions, inctuding reporting, should be taken. In such cases. if a telephone notification of the concfltion is required under §50.72, it should be made as soon as practical and in all cases within s hours after the reasonable expectation of operabiflty no longer exists.

2.6 Events Discussed with the NRC Staff

On occasion, some licensee personnel have erroneously believed that if a reportable event or condition had been discussed with the resident inspector or other NRC staff, there was no need to report under 10 CFR 50.72 and 50.73 because the NRC was aware of the situa~on:* sbme licensee personnel have also expressed a similar misunderstanding for cases in which the NRC staff identified a reportable event or condition to the licensee via inspection or assessment activities. Such conditions do not satisfy §§50. 72 and 50.73. Sections 50. 72 and 50. 73 specifically require a. telephone notification via the ENS and/or submittal of a written LER for an event or condition that meets the criteria stated in those rules.

2.7 Voluntary Reporting Information that does not meet the reporting criteria of 10 CFR 50.72 and 50.73 may be reportable under other requirements such as 10 CFR 50.9, 20.2202, 20.2203, 50.36, 72.74, 72.216, 73.71, and Part 21. In particular, 10 CFR 50.9 (b) states *Each applicant or licensee shaJI notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security.* This applies to information which is not already required by other reporting or updating requirements. Notification must be made to the Administrator of the appropriate Regional Office within two working days of identifying the information. Reporting pursuant to

§50.9 is required, not voluntary.<11 Voluntary reporting, as discussed in the following paragraphs, pertains to information of lesser significance than described in §50.9(b).

Licensees are permitted and encouraged to report any event or condition that does not meet the criteria for required reporting, if the licensee believes that the event or condition might be of safety significance or of generic interest or concern. Reporting ~irements aside, assurance

<11 As inaicated in the Statement of Considerations for §50.9, ~A licensee cannot evade the rule by never 'finding' infonnation to be significant. The fact that a licensee considers information to be significant can be established, for example, by the actions taken by the licensee to evaluate that information.* 59 FR 49362, December 31, 1987.

Draft NUREG*1022, Rev. 2 14

\II-\

3.2 Specific Reporting Criteria 3.2.1 Plant Shutdown Required by Technical Specifications

§50.72(b)(+~(i){,tq §50.73(a}(2)(i)(A)

Licensees shall report: "The initiation of any Licensees shaJI report '"The completion of nuclear plant shutdown required by the any nuclear plant shutdown required by the plant's Technical Specifications.* plant's T echnicaJ Specifications.*

If not reported as an emergency under §50.72(a), licensees are required to report the initiation of a plant shutdown required by TS to the NRC via the ENS as soon as practical and in aJI cases within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the initiation of a plant shutdown required by TS to t h e f via the ENS. tf the shutdown is completed, licensees are required to submit an LEA within ys.

Discussion W !fe,.

The §50. 72 reporting requirement is intended to capture those events for which TS require the initiation of reactor shutdown to provide the NRC with early warning of safety significant conditions serious enough to warrant that the plant be shut down.

For §50.72 reporting purposes, the phrase *initiation of any nuclear plant shutdown* includes action to start reducing reactor power, i.e., adding negative reactivity to achieve a nuclear plant shutdown required by TS. This includes initiation of any shutdown due to expected inabintv to restore equipment prior to exceeding the LCO action time. As a practical matter. in order to meet the time limits for reporting under §50.72, the reporting decision should sometimes be based on such expectations. {See Example 4.)

The *initiation of any nuclear plant shutdown* does not include mode changes required by TS if

initiated after the plant is already in a shutdown condition.

A reduction in power for some other purpose, not constituting initiation of a shutdown required by TS, is not reportable under this criterion.

I For §50.73 reporting purposes, the phrase acompletion of any nuclear plant shutdown* is defined as the point in time during a TS required shutdown when the plant enters the first shutdown condition required by a limiting condition for operations (LCO) [e.g., hot standby (Mode 3) for PWRs] with the standard technical specifications (STS). For example, if at 0200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months a plant enters an LCO action statement that states, "restore the inoperable channel to operable status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,* the plant must be shut down (i.e., at least in hot standby) by 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months . An LER is required if the inoperable channel is not returned to operable status by 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months and the plant enters hot standby.

An LEA is not required if a failure was or could have been corrected before a plant has completed shutdown (as discussed above) and no other criteria in §50.73 apply.

19 Draft NUREG-1022, Rev. 2

Examples (1) Initiation of a TS-Required Plant Shutdown While operating at 100-percent power, one of the battery chargers, which feeds a 125 Vdc vital bus, failed during a surveillance test. The battery charger was declared inoperable, placing the plant in a 2-hour LCO to return the battery charger to an operable status or commence a TS-required plant shutdown. Licensee personnel started reducing reactor power to achieve a nuclear plant shutdown required by a TS when they were unable to complete repairs to the inoperable battery charger in the 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allowed. The cause of the battery charger failure was subsequently identified and repaired. Upon completion of surveillance testing, the battery charger was returned to service and the TS required plant shutdown was stopped at 96-percent * ~J f&

r u,rea c,::..,>

The licensee made an ENS n

because of the initiation of a T8-required plant shutdown. An LER was nat(!~~!!9;t.Jnder this criterion since the failed battery charger was corrected before the plant completed shutdown.

(2) Initiation and Completion of a TS-Required Plant Shutdown During startup of a PWR plant with reactor power in the intermediate range, two of the four reactor coolant pumps (RCPs) tripped when the station power transformer supplying power de-energized. With less than four RCPs operating, the plant entered a 1-hour LCO to be in hot standby. Control rods were manually inserted to place the plant in a shutdown concfition.

The licensee made an ENS notifi!tion because of the initiation of a T8-required plant shunjown. An LER ~~ days because of the completion cl the TS-required plant shutdown. r:e~-ir~ /;},. _

(3) Failure that was or could have been corrected before a plant h1n completed shut down !m§ required.

Question:

What about the situation where you have seven days to fix a component or be shut down, but the planfmust be shut down to fix the component? Assume the plant shuts down, the component is fixed, and the plant returns to power prior to the end of the ,

seven day period. Is that situation reportable?

Answer.

No. If the shutdown was not required by the Technical Specifications, it need not be reported. However. other criteria in 50.73 may apply and may require that the event be reported *

Question:

Suppose that there are seven days to fix a problem and it is likely the problem can be fixed during this time period. However, the plant management elects to shut down and Draft NUREG-_1022, Rev. 2 20

fix this problem and other problems.

Answer:

fil Initiation of plant shutdown in anticipation of LCO reglrired shutdown.

The plant lost one of tvvo sources of offsite power due to overheating In the main transformer, *The TS allow 72"hours to restore the source or initiate a shutdown and be HOT STANDBY in the next 6:hours and COLD SHUTDOWN in the followina 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The Hcensee estimated that:the transfonner problem couJd not be corrected within the LCO action time. Therefore the decision was made to start a shutdown soon after the transfonner

problem was discovered, The shutdown was uneventful and was completed, with the plant in HOT STANDBY, prior to the expiration of the LCO action time. After the plant reached HOT STANDBY, further evaluation indicated that the transformer problem could not be corrected prior to ttm requirement to place the plant in COLO SHUTDOWN. Based on this time estimate. it was decided to place the unit in COLD SHUTDOWN.

The event is reportable under §50,Z2Cbl'2lffi as the initiation of plant shutdown reauired by TS because, at the time the shutdown was initiated, and the time the report was due. ft was not expected that the eauioment WQUld be restored to ooerable status within the reauired time, This is based on the fact that the reoortina reauirement is intended to capture those events for wmch TS require the initiation of a reactor shutdown.

The event is reportable under §50.73lal@ffiCAl because the otant shutdown was completed when 1he otant reached HOT STANDBY CMode 3). Had the transformer been repaired and the shutdown process tenninated before the plant reached Mode 3, the event woufd not be reportable under §50.73{a)(2)ffi(A).

21 Draft NUREG-1022, Rev. 2

of time longer than allott'ed by TS. The event i:s repormble e't'en thous-:,h the sttnieillance is sttbseciuentl'f' satisfactorily performed. it indicates that equipment {i.e., one train ot a multiple train system) was not caoab!e of oertormina its specified safetyfunctions land thus was inoperable) for a period of time tonaer than alJowed bv TS. If the untimely test indicates that the

* * * * *

8 \eta.-

11 I?

Some plants have TS which allow a delay of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in declaring an LCO or TS requirements not met if it is found that a surveillance was not performed within its specified frequency or interval. However, as disaJssed above, failure to perform a surveillance within its frequency or interval is still reportable if it indicates that eauioment O.e,. one train of a multiple train svsteml was not caoable of oertonnina tts soecified safetviunctions and. 1hus. was inoperable for a period of time longer than allowed by TS. The additional delay in dedaring the LCO not met does not change the fact that the condition existed longer than allowed by TS. The delay merely specifies appropriate remedial action

As specified in the rule, the event is not reportable if it consists solely of a case of a late surveillance test where the oversiQht is corrected, the test is performed, and the equipment is found to be capable of performing its specified safety functions.

\

t+}Oesian Featum and Analysis Defects and Deyiations Section 50.36(e){4) indicates that design feattires to be inc!uded in TS are those features of the facility such 8.5 materials of eonstrtletion or geometric amtngements which, if attered or modified, wotdd have a signifiamt effect on safety and ate not CO\'ered b~ items (1) through {0} abc't1e.

Reportabilit)' requirements related to desi~n features are included in other sections of 10 GFA

50. 72 and 50. rS. A desian or analysis defect or deyiatipn is reportable under this giterion if. as a i resufl eauioment O,e., one train of a multiole train svstero> was not capable of oerformina its i

specified safety functions land. thus, was inoperable) tor a period ot time lonaer than allowed by TS, Since design and analysjs conditions are Iona-lasting, the essential guesffon is whether the equipment involved was capable of performing its specified safety functions.

{SrTS Administrative Requirements. lnctuding Radiological Controls. Required by Section 6 of the STS, or Equivalent t

i f

I Section 6 of the STS (Section 5 of ISTS), or its equivalent, has a number of administrative I

and the condition was rectified immediately after d_~isco~*-te::.:.r,::.:i._*- - - - - - - - - - ' - - - -

is only intendsd

. n <* .

Cl..n.lt<:Jnrs TS* . It is ~~ln1'a e * *,or TS SR 3.0.3. *---------------------

be shutdowfi. These matters are discussed further in GL 87-09, GL 91-18 Draft NUREG-_1022, Rev. 2 24

requirements such as organizational structure, the required number of personnel on shift, the maximum hours of work permitted during a specific interval of time, and the requirement to have; maintain, and implement certain specified procedures. Violation of an s:dministrati'Q'e TS ~

administrative jn nature in and of itself does not neeessarir, eonetitt:1te a reportable condition C-operstion or condition prohibited by the pfant's TS*). is not reportable. This reJ'ortfng requirement des:b with matters affecting plant operation more St:Jbstartttafly and more diredt)*

than matters that are mainly administrath'e.ifl- F=ailure to ffloet adfflinisu=a.twe TS FOquiFOments is reportable only if it results in .Jiolation:s of equipment operahilit) requirements or had a similar detriment8:I effect on a licensee's ability to SB:fely operate the plant For example, operation with less than the required number of people on shift Aotlld constitttte operation prohibited by the TS. I lon1et1er, a dlango in the plant's organizational structure that has not yet been approved as a Technical Specification change would not be reportable.

An administrative procedure violation or failure to implement a procedure, such as failure to lock a high radiation area door, that does not have a direct impact on the safe operation of the plant, is generally not reportable under this criterion .

Radiological conditions and events that are reportable are defined in 10 ,CF~Wj~

20.2203. Redundant reporting is not required.

,r;;,

/ ~~~~:..z;..;==

c.,f-t J: 3 I (l.,r-ffl-Missed Tests Required by ASME Section XI 1ST and ISi and by STS 4.0.5, or Equivalent

Sections 50.55a(g) and 50.55a(f) require the implementation of ISi and 1ST programs in accordance with the applicable ecfrtion of the ASME Code for those pumps and valves whose function is required for safety. STS Section 4.0.5 (or an equivalent) covers these testing requirements. (Generally, there is no comparable ISTS section.) A.missed or fate IST/ISVASME test is reportable when the test intel"'tal pit.IS any allowable extension plus the LOO action time has been exceeded it indicates that egyioment {i.e.. one train of a multiple train system) was not capable of oerforming its specified safety functions and, thus. was inoperable for a period of time longer than allowed by TS.

--tef The proposes FUlo ._.<<>ufd ha¥e fE!EIUired FOpor=ting vlhen *a TS aotien staiefflent ia Aot mee The 'H'ording of the final rule requires reporting *Any operation or condition prohibited by the plant's Technicm Specifications.* The Ste:temeflt of Considerations for the final rule indicates that this change wms made to accommodate plants that did not ha'te requirements specifiea:tly defined as action statements (46 m 90855, July .ea, 1sea~.

25 Draft NUREG-1022, Rev. 2

Insert 3, page 25, NUREG 1022, Rev .2 draft Entry into STS 3.0.3 STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent, establishes requirements for actions when:

(1) an LCO is not met and the associated ACTIONS are not met, (2) an associated ACTION is not provided, or (3) as directed by the associated ACTIONS themselves.

Entry into STS 3.0.3 (ISTS LCO 3.0.3) for either of the first two above reasons are generally reportable under this criterion. However, when the plant TS specify the entry into 3.0.3 as the required ACTION and that action and its completion time are met, the event is not reportable under this criterion. Also, momentary (less then 5 minutes) entries into TS 3.0.3 regardless of which reason, are not reportable under this criterion. Any TS 3.0.3 entry involving actual plant shutdown should be reviewed for reporting under 50.72(b)(l)(i)(A) and 50.73(a)(2)(i)(A) for a plant shutdown required by TS.

~ire Protection Systems When Required by TS When operability requirements for fire protection systems are specified in TS they are within the Examples (1) LCO Exceeded In conducting a timely 30-day surveillance test a licensee found a standby component with a 7-day LCO allowed outage time and associated 8-hour shutdown action statement to be Inoperable. ch:1ring a 30-eiay s1:1rteiUance test (This is equivalent to a 7-day restoration completion time and an 8-hour action completion time in ISTS.) Subsequent review indicated that the component was assembled Improperly during maintenance conducted 30 days previously and the post-maintenance test was not adequate to identify the error. Thus, there was firm evidence that the standby component had been inoperable for the entire 30 days.

An LER was required because the 7-day LCO allowed outage time and the shutdown action statement time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months were exceeded. Had the inoperability been identified and corrected within the 7-day LCO allowed outage time plus the 8-hour shutdown action statement. the event would not be reportable.

(2) Missed Surveillance Tests A ficensee, with the plant in Mode 5 following a 10-month refueling outage, detennined that certain monthly TS surveillance tests, which were required to be perfonned regardless of plant mode, had not been perfonned as required during the outage. The STS 4.02 (equivalent to ISTS SR 3.02) extension was also exceeded. The surveillance tests were immediately performed.

An LEA is reqtJired beem:tSe the time inteM!I, inel1:1ding extensions pem,itted by TS, exceeded the TS stn~eillance intertal pll:IS the LOO ftetion smtement times (eei1:1i\.aient to ISTS eomi,letion times~. No LER wouk:tbe required if the test-showed the equipment was sir] operable. On lhe QJl]Qr hand, jf the tg_st shQw,,d U,e equjim,_ent was not c a n a b l e ~

performing its specified safety functions and, thus, wasjnoperabJe, the eventwould ;, J ~

reportable. ; ,r:1-(1.,, ~

3) Ente . g ST .0.3 -

r this event because STS 3.0.3 was entered.

Draft NUREG-1022, Rev. 2 26

Insert 4, page 26, NUREG 1022, Rev.2 draft (3) Entering STS 3.0.3 due to lack of specific TS actions With essential water chillers (A) and (B) out of service, the only remaining operable chiller (A/B) tripped. This condition caused the plant to enter STS 3.0.3 (equivalent to ISTS LCO 3.0.3) for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , until chiller (A) was restored to service and the temperature was restored to within TS limits.

An LER is required for this event because, in this case, there were no actions provided in the plant TS for that condition and STS 3.0.3 was entered for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(4) Entry into STS 3.0.3 when the plant TS specify 3.0.3 entry During a surveillance test on the A train of a two-train Standby Gas Treatment (SBGT) system, a condition was discovered on the B train that rendered it inoperable .

The test was halted and steps taken to return the A train to a standby readiness condition. During the restoration, switch manipulations momentarily rendered the A train inoperable. With both trains inoperable, the plant TS specify immediate entry into LCO 3.0.3. The entry into LCO 3.0.3 was logged and then exited within 1 minute once switch manipulation on the A train was completed.

This event is not reportable under this criterion because all the actions specified by the plant TS were completed within the required completion times. There was no operation or condition prohibited by TS. Also, momentary entries into STS 3.0.3 are not reportable.

~-4) Missed Tests Required by ASME Section XI 1ST and ISi, and by STS 4.0.5, or Equiwrlent Deamples of potentially reporumle conditions are failures to perform required activities 'frithin specified times for those components govemed by TS. Such activities include stroke testing valves, testing V'8Hes in the position required for the performance of their safety function, verifying motor operated vah,e stroke times for both topen ~ closed~ directions, using the proper test pressures to property classify a:nd test aetroe ro*alves snd to inereese test frequency subsequent to obtaining test results that were belo'it' certain threshold values. A

missed test is reportable when the test intel"'0'81 plus any e:Uowsble extension plus the LCO action time is e,cceeded.

(S) J4-rMultiple Test Failures An example of multiple test failures involves the sequential testing of safety valves.

Sometimes multiple valves are found to lift with set points outs4de of TS limits.

If the discrepancies are large enough that multiple valves are inoperable the event may also be reportable under §50.73{a)(2)(vii) *Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or tvvo independent trains or channels to become inoperable in a single system ....*

( I,J ,@ Seismic Restraints Assu . . . . .

An LEA woutd be required because the EOG was inoperable for a period of time longer than {}

allowed by TS. £E:'::i

( 1 ) 1fil Vulnerabffity to Loss of Offsite Power Assume that during a design review it is found that a loss of offsite oower: could cause a loss of instrument air and. as a result. auxiliary feedwater (AFW} flow control valves could fail open. Then for low steam generator pressure, such as could occur for certain main steam

* *

the mo10r driven AfW pumps on would not be ected.

uired because of AFW was i erable for a W~f~ -t1it,tfvt~ dQdc..ud 27 Draft NUREG-1022, Rev. 2 i Vl .,-~ i,\t.

3.2.3 TechnicaJ Specification Deviation per §50.54(x)

§SO.72(b)(1 )ffl{tij §50.73(a)(2)(i)(C)

Licensees shall report "Any deviation from Licensees shall report *Any deviation from the plant's TechnicaJ Specifications the plant's Technical Specifications authorized pursuant to §50.54(x) of this authorized pursuant to §50.54(x) of this part. part."

If not reported as an emergency under §50.72(a), licensees are required to report any such deviation to the NRC via the ENS as soon as practical and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Licensees are required to submit an LEA within a&gQ days.

Discussion

10 CFR 50.54(x) generally permits licensees to take reasonable action in an emergency even though the action departs from the license conditions or plant technical specifications if (1) the action is immediately needed to protect the public health and safety, inciuding plant personnel, and (2) no action consistent with the license conditions and technical specifications Is immediately apparent that can provide adequate or equivalent protection. Deviations authorized pursuant to 10CFR 50.54(x) are reportable under this criterion.

Example With the plant at 1QO-percent power, the upper containment airlock inner door was opened to allow a technician to exit from the containment while the upper airlock outer door was inoperable, resulting in the loss of containment integrity. The upper airlock door was inoperable pending retests following seaJ replacement The technician was inside containment when the lower airlock failed, requiring the technician to exit through the upper door.

The licensee decided to exercise the option allowed for under 1o CFA 50.54(x) and open the upper containment airlock inner door. In this instance, immediate action was considered necessary to protect the safety of the technician. The upper airlock was not scheduled to be returned to operability for another 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months and the time to repair the lower air1ock door was unknown.

29 Draft NUREG-1022, Rev. 2

3.2.4 Degraded Condition

§50.72(b)(+ID(II) §50.73(a)(2)(il)

Licensees shall report: *Any event or Licensees shall report: *Any event or condition during operation that result§ in: condition that resulted in:

!Al The condition of the nuclear power !Al The condition of the nuclear power plant, including its principal safety barriers, plant, including its principal safety barriers, being seriously degraded; or results in being seriously degraded; or tne.t resulted in

!ID The nuclear power plant bein~ In .an The nuclear power plant being in ~

an unanaJyzed condition that significantly an unanalyzed concfltion that significantiy compromises affects plant safety.* comDromised affected plant safety_; or

(} In a condition that is outside the design b ~ of the plant; or (6~ In e. condition not CCYoered by the (6) In a condition not CO'f ered by the pltmt's operating and emergency plant's operating and emergency procedures.* procedures.'

If not reported as an emergency under §50. 72(a), licensees are required to report a seriously degraded principal safety barrier or an unanaJyzed condition that significantly affects plant safety to the NRC via the ENS as soon as practical and in all cases within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Licensees are required to submit an LER within 60 days*

In the case of a component with a sianfficantly degraded ability to oerfonn a safety function, where the condition couJd affectfother similar components in the plant licensees are required to submit an LEA within 60 days.

Discussion fAJ Nuclear power plant including its principal safety barriers. being seriously degraded:

This criterion applies to material (e.g., metallurgicaJ or chemicaQ problems that cause abnonnaJ degradation of or stress upon the principal safety barriers (i.e., the fuel cladding, reactor coolant system pressure boundary, or the containment) such as:

(1) Fuel claddin~ failures in the reactor, or in the storage pool, that exceed expected values, or that are~~'-.i:j_f~d, or that are caused by unexpected factors.

!2) Welding or materiaJ defects in the primary coolant system which cannot be foand Draft NUREG-1_022, Rev. 2 30

Insert 5, page 30, NUREG 1022, Rev.2 draft .

(C) Multiple similar components being in a degraded or non-conforming condition such that the ability to perform a specified safety function was actually lost.

apply to minor variations in individual parameters, or to problems concerning single pieces of equipment For example, at any time, one or more safety-related components may be out of service due to testing, maintenance, or a fault that has not yet been repaired. Any trivial single failure or minor error in perfonning surveillance tests could produce a situation in which tvvo or more often unrelated, safety-grade components are out-of-service. Technically, this is an unanalyzed condition. However, these events should be reported only if they involve functionally related components or if they significantly compromise plant safety.*{9]

"When applying engineering judgment, and there is a doubt regarcfmg whether to report or not, the Commission's poticy is that licensees should make the report. *<111J

For example, small voids in systems designed to remove heat from the reactor core which have been previously shown through anaJysis not to be safety significant need not be reported. However, the accumulation of voids that could inhibit the ability to adequately remove heat from the reactor core, particularly under natural circuJation concfrtions, would constitute an unanalyzed condition and would be reportable. ll{nJ

*1n addition, voiding in instrument lines that results in an erroneous indication causing the operator to misunderstand the true condition of the plant is also an unanalyzed condition and should be reported. "l12J Furthermore, beyond the examples given in 1983, examples of reportable events would include discovery that a system required to meet the single failure criterion does not do so.

In another example, if fire barriers are found to be missing, such that the required degree of separation for redundant safe shutdown trains is lacking, the event would be reportable. On the other hand, If a fire wrap, to which the licensee has committed, is missing from a safe shutdown train but another safe shutdown train is available in a different fire area, protected such that the required separation for safe shutdown trains is stiU provided, the event would not be reportable.

~

(CJ Significantly degraded component(s):

f9J 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

<10* 48 FR 39042, August 29, 1983.

c111 48 FR 39042. August 29, 1983 and 48 FR 33856, July 26, 1983.

1

< 21 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

Draft NUREG-1022, Rev. 2 32

implications. reporting would not be required under this crtterion.

Jn a further example, whjle processing calculations it was detennined that four motor operated yaJves within the reactor building were located below the accident flood !evel and were not aualified for 1hat condition. Pending reolacement with auafified equipment. the licensee determined that three of the valves had sufficiently short opening time that their afetv function would be comcleted before thev were submerged. The fourth wtve was

¥.~u:*me=w~s

~ ~r==saC = =~:~&

An example of an event that would not be reportable is as follows. The motor on a motor-ocerated valve burned out after reoeated cycling for testing. This event woutd not b;!

a~~<lv~T~

rePQrtable because jt is a single component fafture, and while there might be similar MOYs in I , L M'rnor switch adiustments on MOVs wouJd not be reoorted where they do not skJnificanttv dect the abifrty of the MOV to carry out its design-basis function and the cause of the adjustments is not a generic concern, At one p!ant the switch on the radio transmitter for the auxiliary bwldina crane was used to handle a spent fuel cask while two protective features had been defeated by wiring errors. A new radio control transmitter had be@n proct.[ed and placed in service, Beca1 ss, the new 33 Draft NUREG-1022. Rev. 2

controller was Wired differently than the old one. the drum oyer-soeed protection and spent fuel pool root slot limit switch were inadvertently defeated. Whjle the crane was found to be outside its desian basis, this condition would not be reportable because switch wiring

deficiency could not reasonably be expected to affect any other components at the plant Examples (1) Significant Degradation Failure§ of Reactor Fuel Rod Cladding Identified During Testing of Fuel Assemblies Radio-chemistry data for a particular PWR Indicated that a number of fuel rods had failed during the first few months of operation. Projections ranged from 6 to 12 failed rods. The end of ~ e reactor cootant system iodine-131 activity averaged 0.025 micro curies per m Riter. *ng the end of cycle shutdown, iodine-131 spiked to 11.45 micro curies per milliliter. e cause was due to a significant number of failed fuel rods. Inspections revealed that 136 of the total 157 fuel assemblies contained failed fuel (approximately 300 fuel rods had through-wall penetrations), far exceeding the anticipated number of failures. The

defects were generally pinhole sized. The fuel cladding failures were caused by long-term fretting from debris that became lodged between the lower fuel assembly nozzfe and the first spacer grid, resulting in penetration of the stainless-steel fuel cladding. The source of the debris was apparently a machining byproduct from the thermal shield support system repairs during the previous refueUng outage.

An ENS notification is reqt.tired beeatlse a principal safet.\,' barrier {the ftJel cladcfin~~ was /\

fotffld serio_t:ISfrt* de~raded: A:n LEA is l'eelt1ited:1be~.

cladding failures exceed expected vallJltp ! ' ~ ~ a n d ~ ~ by reQOrtable ~becaJ='= ffi unexpected factors. . ~

(2) Reactor Coolant System Pressure Boundary Degradation due to Corrosion of a Control Rod Drive Mechanism Flange WhHe the plant was in hot shutdown, a total of six control rod drive mechanism (CROM) reac;tor vessel nozzte flanges were identified as leaking. Subsequently one of the flanges was found eroded and pitted. While removing the nut ring from beneath the flange, it was discovered that approximately 50 percent of one of the nut ring halves had corroded away and that two of the four bolt holes in the corroded nut ring half were degraded to the point where there was no bolt/thread engagement An inspection of the flanges and spiral wound gaskets, which were removed from between the flanges, revealed that the cause of the leaks was the gradual deterioration of the gaskets from age. A replacement CROM was installed and the gaskets on all six CRDMs were replaced with new design graphite-type gaskets.

Draft NUREG*1022. Rev. 2 34

(3) Sig"ifieant Degradation of Reactor Fuel Rod Cladding Identified During Fuel Sipping Operations With the plant in cold shutdown, fuel sipping operations appeared to Indicate a significant portion of cycle 2 fuel, type alYP,

had failed, i.e., four confirmed and twelve potentiaJ fuel leakers. The potentiaJ fuel leakers had only been sipped once prior to making the ENS notification. The licensee contacted the fuel vendor for assistance on-site in evaluating this problem.

An ENS notification was made because the fuel cladding degradation was thought to be widespread. However, additionaJ sipping operations and a subsequent evaluation by the licensee's reactor engineering department with vendor assistance concluded that no additional fuel failures had occurred, i.e., the abnormaJ readings associated with the potential Office, the licensee retracted this aw ~de w~$ -ti.e r

fuel leakers was attributed to fission products trapped in the crud layer. Based on the results of the evaluation the licensee concluded that the fuel cladding was not seriously degraded and that the ewnt was not reportabte~1enflll, ailllr d ~ Ille ~nal ;

..,~;,,.fe, d~;s,0 8

35 Draft NUREG-1022. Rev. 2

3.2.6 System Actuation

§50.72(b)(2~ §50.73(a)(2)fw){A}

Licensees shall report *Any event or Licensees shall report *Any event or condition that results in a-intentional manual condition that resulted In a-manual or actuation or valid automatic actuation of any automatic actuation of any engineered safet)* feattJre (ESF), includin~ engineered safety feature tESF), ineludlnr.,

the reactor protection sr,stem (RPS1/4 of the the reactor prcteetion S'fStem fRPS} ~

systems listed in paragraph Cb){?l(iv}{B} of systems Hsted tn paragraph Ca){2lf1V}CB} of this section except w h ~ the actuation this section except when:

results from and is part of a pre-planned tA)-f11 The actuation resulted from and sequence during testing or reactor was part of a pre-planned sequence during operatio~. testing or reactor operation; 2r fO) The aettlation is invalid and: tBrfZl The actuation was invalid and; (1) Ocet:1rs while the system is properly flj-fll Occurred while the system was remO"ted frcm service: properly removed from service; QC (2) Oca:1rs after the safety function he:s ffet(uJ Occurred after the safety function been mready eompletecl, or had been already completed. er (3) lrr,ok.es on!) the following specific (3} Involves on!) the folloW'ring specific ESFs or their equhta:lent systems. @ ESFs or their equi'v'alent systems. fij Reactor water clean up system, fii) Control Reactor water clean ttp system, (ii) Oontrol rcom emergency ventilation system; (iii} room emergency ro1entilation system; fiiiJ Reactor building ventilation system, (fvj Fuel Reactor building oentil8tion system. (w) Fuel building 'tentilation system, or~) Auxilisry

~

bailding , r e n t i ~ l i a r f buildi11g rtentilation system.

§50.72(b){2)(1V){Bl b*Htfi1:.F;,,pr;~

§50.73(a)(2)(iv)(B}

'"Toe systems to which the requirements of "The systems to which the requirements of paragraph (b}(2)frvl{A} of this section apply paragraph {a)(g)[ly)fAl of this section apply are: are; a, Reactor orotection system <reactor 11} Reactor protection system {reactor scram. reactor triDt sgam. reactor trio).

f2J Emeraencv core cooling svstems t2J Emergency core cooling svstems CPWBs> indudina: hiah-hea,d. intennecHate- <PWRs) including: high-head. intennediate-head and kJw-head iniection svstems and head, and low-head iniection swtems and the tow pressure lniection function of the tow pressure injection function of residual Cdecavl heat remoyaJ svstems. residual ldecavl heat removal systems.

(3) ECCS for boiling water reactors {3J ECCS for b01lina water reactors

{BWRsl including; high-pressure and low- CBWRsl inciudina: high-pressure and low:

pressure core spray systems: high-pressure oressure core sorav systems; high-pressure coolant iniection system: feedwater coolant coolant injection system; feedwater coolant (continued on next oaae l (continued on next oaae) 39 Draft NUREG-1022, Rev. 2

Insert 6, page 39, NUREG 1022, Rev.2 draft (iii) Occurred during a plant operating mode when the safety function was not required .

(continued from previous page) (continued from previous page) injection system: tow pressure iniection injection svstem: row pressure inlectton function of the residual heat remoyaJ system; function of the residual heat removaJ system; and automatic deoressunzation svstem. and automatic denresaw:imtipn swtem.

141 BWR isolation condenser system and f41 BWR isolation condenser system and reactor core isolation cooling svstem. reactor core isolation cootina svstem.

/SJ PWR awaliary feedwater system. f5J PWR awalary feedwater svstem.

(6) Containment systems including: fBJ Containment swtems including:

containment and reactor vesseJ ISQlatioo containment and reactor vessel isolatioJJ svstems <aenera.J coutairn,mnt isolation swtems laeneral containment isolation mnats affecting . numerous valves and. maJn signals affectina numerous ya1yes and main

I C!l*l"H'tl:ll,lc!,

containment may and fan cooler systems. containment spray and fan cooler systems.

mEmergency ac etectrical P0ml'.

mEmergency ac electricaJ oower systems. inc!udina: . emergency diesel systems. .iocludina; emeraencv diesel generators fEDGs} and their associated aenerators <EDGs) and their assodated support systems; hydroelectric facilities used SUDQOrt systems; lwdJpelectric facilities used in lieu of EDGs at the Oconee Station: satetv in lieu of EDGs at the Oconee Station; safety related aas turbine generators: BWR related aas turbjne generators: BWB dedicated DMsiQn a EDGs aod ttum dedicated OMsion 3 EDGs and their associated suooort systems; and station associated supgort systems: and station blackout diesel generators <and black-start blackout diesel generators land black-s1art aas turbines that serve a similar ouroose) aas turbines that serve a similar purpose) which are started from the control room and which are started from the control room and included in the pJanfs operating ang included in the plant's operating and emergency procedures. emeraencv procedures.

fBJ Anticipated transient withgyt scram fBJ Anticipated transient without scram ATW . ATW . .

If not reported under §50.72{a) or (b)(1), licensees are required to report arr, engineered safety featttre actuation of a system listed in the rule. indudinR the reactor protection system, to the NRC via the ENS as soon as practical and in all cases within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the event Licensees are required to submit an LEA within 60 days.

Discussion The Statements of Sonsidenmons indicate that this These paragraphs require events to be reported ** -=--.......~==..:.==~ tes either manually or

premise that these systems

, therefore: (1) they should ed frequently or unnecessari n is interested both in events where a system was needed to 40

mitigate the consequences of an event (whether or not the equipment performed property) and events where a system actuated unnecessarily.

In discussing the reporting of aet.1::18:tions which ere part of prei,lanned procedures, the Statements of Considerations alse state th~Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event (e.g., at the discretion of the licensee as part of a preplanned procedure).

This indicates e:n :II:~_intent is to require reporting actuation of systems that mitigate the consequences of significant events. Usually, the staff would not consider this to include single component actuations because single components of complex systems, by themselves, usually do not mitigate the consequences of significant events. However, in some cases a component would be sufficient to mitigate the event (i.e., perform the safety function) and its actuation

~

would, therefore, be reportable. This position is consistent with!!,jlljl.~lD.gND~~IO!. reporting

~ u i~~ e~ the e,_~that these systems are

,~of_Jn;lQRificatrt~~ .,,-1/

(1 .-;;;+,

m s, "' ( ,_I.MA}_,

£

,,a

.:JI>

Single trains do mitigate the consequences of reportable.

events. and, thus.

In this regard, the staff considers actuation of a diesel-generator to be actuation of a train-not actuation of a single component - because a diesel generator mitigates the event (performs the safety function for plants at which diesel generators me classified as ESF systems). (See Example 3 below.)

The staff also considers intentional manual actions, in which one or more ESF-cornponents are actuated in response to actual plant conditions resulting from equipment failure or human error, to be reportable because such actions would usually mitigate the consequences of a significant event This position is consistent with the statement that the Commission is interested in events where a system was needed to mitigate the consequences of the event. For example, starting a safety injection pump in response to a rapidly decreasing pressurizer level or starting HPCI in response to a loss of feedwater would be reportable. However. shifting alignment of makeup pumps or closing a containment isolation valve for normal operationaJ purposes would not be reportable.

The Statement of Considerations tUso indica.-tes that Actttstion ef mtdtiehmlnel ESF e:ctusticn systems is defined a:s aetue:tion of enough ch8nnels to complete the minimtJm e:C'tt:le:tion logic.

Therefore, single chm,nel aetue:tions. whether C8:ttsed by failures or otherwise, 8:1'e not reportable if they do not complete the minimum scb::t8tion logie. Note, ho*MWer. that if only a single logic ehe:nnel actttates *ut,en, in fact; the system should ha'te aetuated in r~pcnse to plant pel"flffleters, this would be reportable. The e~ent would be reportable 1:1nder these criteria (ESF actuation) as v.1ell as under 10 om 50.72tl))(2)6ii) and 10 CFR 50.73(a)(2)(*rj (e*.rem or conditi~n tUone). This position is consistent vtith the statement that the Commission is interested in e~ents where an ESF 'ft'8S needed to mitigate the consequence5, -.*i*l'tetl"ler or not the equipment perfem,ed preperiy.

With regard to preplanned actuations, the Smtements of Consideration indica.-te that operation of a system as part of a planned test or operational evolution need not be reported. Preplanned 41 Draft NUREG-1022, Rev. 2

actuations are those which are expected to actually occur due to preplanned activities covered by procedures. Such actuations are those for which a procedural step or other appropriate documentation indicates the specific ESF-actuation that-is actually expected to occur. Control room personnel are aware of the specific signal generation before its occurrence or indication in the control room. However, if during the test or evolution, the system actuates in a way that is not part of the planned evolution, that actuation should be reported. For example, if the nonnal reactor shutdown procedure requires that the control rods be inserted by a manual reactor scram, the reactor scram need not be reported. However, if unanticipated conditions develop during the shutdown that cause an automatic reactor scram, such a reactor scram should be reported. The fact that the safety analysis assumes that a system will actuate automatically during an event does not eliminate the need to report that actuation. Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event (e.g., at the discretion of the licensee as part of a planned evolution).

Note that if an operator were to manually scram the reactor in anticipation of receiving an automatic reactor scram, this would be reportable just as the automatic scram would be reportable

Valid ESF actuations are those actuations that manual initiation, unless it is part of a preplann initiated in response to actual plant conartio

~*-l'l--ft-~-~~l't-9""""1tfftj*o"lfn,..is. * .3

Thus, uwalid e not intentional In general. invalid actuations are not reportable by telephone under §50.72{b)(2)(w). In addition.

invalid actuations that occur when the system is already property removed from service are not reportable if all requirements of plant procedures for removing equipment from service have been met This includes required clearance documentation, equipment and ~ I board J tagging, and properly positioned vatves and power supply breakers@ ~~mvalid actuations that occur after the safety function has already been complet are not reportab An example actuation after the control rods have already been inserted into the

£I Ll core. ~~,+ JJ;,-1 - -,

Rn8:lly, inoalid aet:us:tions for several specific systems or their e~uimlent s:re not reportable.

'fhese systems are the reactor water clean up s:ystem in boiling water reactors tlYNRs), the control room emergene, ventilation ~stem, the reactor building oentilstion system (RBVS), the fuef building oentilation system, and the a:wdliary building ventilation system. Thus, rei:,orting of inv81id acttJatioi,s for these specific systems due to signals the:t originated from non ESF circuitry Draft NUREG-J022. Rev. 2 42

Insert 7, page 42, NUREG 1022, Rev.2 draft In addition, invalid actuations that occur when the plant is in an operating mode where the safety function is not required, are not reportable. An example of this is an invalid actuation of containment isolation logic during a refuel outage when the containment is open for personnel access .

ESF, in response to plant conditions, to mitigate the consequences of an event. *A:s indicated in the Statements of GonS1deratioi is for the l'tdes As cflSCUSSed previously jn this section.

actuations that need not be reported are those that are initiated for reasons other than to mitigate the consequences of an event (e.g., at the discretion of the licensee as part of a planned procedure or evolution).

(7) ESF-Actuation During Maintenance Activity At a BWR, a maintenance activity was under way invoMng placement of a jumper to avoid ESF unintended actuations. The maintenance staff recognized that there was a high potentiaJ for a loss of contact with the jumper and consequent ESF actuation. This potential was explicitly stated in the maintenance work request and on a risk evaluation sheet The operating staff was briefed on the potential SF-actuations prior to start of work. During the event, a loss of continuity did occur and the ESF actuations occurred, involving isolation, standby gas treatment start, closing of some valves in the primary containment isolation system (recirculation pump seal mini-purge valve, nitrogen supply to drywall valve, and containment atmospheric monitoring valve) ocettrrecl.

The staff has concluded that the e,ent 't\'Ould not be reportable if the ~1ent were de9Cribed in appropriste documentation as cleffnitel

ex;,eeted to occur. I IO'Ne'fer since the er+ent ,.,

not listed e definitely expected to * -

Emergency Core Gooling Systems (EGOS) for PreSSt:Jrized Water Reactors (PWRs):

reaaer coolant system aeeumtJlaters

boron ir,jeetior, system

nigh , intermediate , ar,d low head i"iection systems, ineh:1ding systems for charging tJSing centrift:19al chstging pt:1mps, safety injection systems, and residt:181 (decay) heat remo'f'al systems EGOS for Boiling Water Reactors (IYh'As):

high and low, pressttre core spray systems high preesure coolant injection system, feedvte:ter coolant injeciion system, residttal heat remo,'81 sy9tem (low r,resst:1re injeeiion portion)

iselatior, condenser system, reactor core isolation cooling system

a:utomatic depressurimtior, system Draft NUREG:1022, Rev. 2 46

32. 7 Event or Condition That Could Prevent FulfiDment of a Safety Function

§50.72(b)(2)(ffi-!) §50.73(a)(2)(v)

Licensees shall report: *Any event or Licensees shall report: *Any event r con *on that alone at the time of discovery conditlon=ow~l-::.l=~~.,a.:;=:.=..=

ve prevented the fulfillment of the date of discovery that atonet~!Jd:jt,ave ety function of structures or systems that prevented the fulfillment of the safety are needed to: function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it (A) Shut down the reactor and maintain it in a safe shutdown condition; in a safe shutdown condition; (8) Remove residual heat; (8) Remove residual heat; (C) Control the release of radioactive (C) Control the retease of radioactive material; or material; or (D) M"rtigate the consequences of an (D) Mitigate the consequences of an accident. accident*

§50.72(b}(2)(vi) §50.73(a)(2)(vi)

Eyents coyered in paragraph lb}(2llv} of *events covered in paragraph (a)(2)(v) of this this section may include one or more section may Include one or more personnel procedural errors. egyjpment failures, and/or errors, equipment failures, and/or discovery discovery of design. analysis. fabrication, of design, analysis, fabrication, construction, construction, and/or procedural and/or procedural inadequacies. However, inadeauacies, However, individual individual component failures need not be ou=,:~:raaa:,,nw:=:

eauioment in the same system was operable reported pursuant to this paragraph if redundant equipment in the same system was operable and availabfe to perform the and available to perform the required safety required safety function.*

function.

[The Statements of Consideration for 10

*uording stmilar to those If not reported under §50.72(a) or (b)(1), ricensees shall notify the NRC via the ENS as soon as practical and In all cases within ~hours of discovery of the event or condtion and submit an LEA

_within §Q_days.

Discussion The level of judgment for reporting an event or condition under this criterion is a reasonable expectation of preventing fulfillment of a safety function. In the discussions which follow, many of which are taken from the Gtatemertt of 0onStder8tions or from previous NUREG guidance, several different expressions such as "would have,* *could have,* *atone could have,* and Draft NUREG-1022, Rev. 2 48

component lewl.

Single Failure These reporting criteria 8:1 e not meant to require reporting of a single, independent (i.e.,

random) componel'lt failure that makes only one functionally redundant train inoperative unless it is indieafu1e of a generic J'roblem (i.e., has common mode failure implications).

As indicated in Paragraph 50.73(a)(2)(vi) * .* .individuaJ component failures need not be reported pursuant to this paragraph if redundant equipment In the same system was operable and available to perform the required safety function."

The staff considers application of this principle to include cases where one train of a lVIO train system is:

failed, or;

otherwise incapable of performing its function be.,cause of factors such as operator error or design, analysis, fabrication, construction and/or procedural inadequacies, or,

in the case of a train which should be running, otherwise not performing its function because of factors such as operator error or design, analysis, fabrication, construction and/or procedural inadequacies, or;

otherwise subject to a reasonable expectation of being prevented from fulfilling Its safety function

A single failure that defeats the safety function of a redundant system is reportable even if the l171 For example, conditions that would indicate a need for operation of the redundant train are regularly monitored and instrumentation used to monitor these conditions is capable and available.

l1SJ For example, this means that the exclusion from reporting single component failures under this criterion (i.e., Paragraphs 50.72(b)(2)(vi), and 50.73(a)(2)(vr1) should not apply when there is a reasonable expectation of failure of the redundant train as a result of the same cause.-

Apr,lieation of this r,rincq.,le is illusti atect in settera.l pane of this seetion, including. (1) the immediately proceeding quotations from 40 m ~ m,d 40 m 93858: {2} the immediately follo'lt'ing disctl'SSion of commor, cause failures, and; (3} the discussions in Examples 12 and 10.

As indicated in the first paragraph of this seetion, the event should be reported under this enterion if there is a reasonable e:,q,eetation of pre\ienting fulfillment of the safety fune'tion.

51 Draft NUREG-1022, Rev. 2

(2) Failure of a Single--Train Non-Safety System Question:

analysis, are failures and unavailability of. this syste Answer:

. 5tf I JJz in le? M If RCIC is not a *safety system* in that no credit for its operation is taken in the safety l/5 R asa n

., it included the T erio ; otherwise, it is not (3) Failure of a Single--Train Environmental System Question:

There are a number of environmental systems in a ptant dealing with such things as low level waste (e.g., gaseous radwaste tanks). Many of these systems are not required to meet the single failure aiterion so a single failure results in the loss of function of the system. Are all of these systems covered within the scope of the LER rule?

Answer:

If such systems are required by Technical Specifications to be operational and the system is needed to fulfill one of the safety functions identified in this section of the rule then system level failures are reportable. If the system is not covered by Technical Specifications and is not required to meet the single failure criterion, then failures of the system are not reportable under this aiterion .

Loss of Two Trains (4) Loss of Onsite Emergency Power by Multiple Equipment lnoperability and Unavailability During refueling, one emergency diesel generator (EOG) in a two train system was out of service for maintenance. The second EOG was dec1a.red inoperable when it failed its survemance test An ENS notification is required and an LER is required. As addressed in the Disa 1SSion section above, loss of either the onsite power system or the offsite power system is reportabte under this criterion.

(5) Procedure Error Prevents Reactor Shutdown Function The unit was in mode 5 (95°F and 0 psig; before initial criticality) and a post-modification test was in progress on the train A reactor protection system (RPS), when the operator observed that both train A and B source range detectors were disabled. During post-

-modification testing on train A RPS, instrumentation personnel placed the train B input error Draft NUREG-1022, Rev. 2 54

Insert 8, page 54, NUREG 1022, Rev.2 draft No. Unplanned inoperability ofRCIC is not reportable under this criterion as the loss of a single train safety system. RCIC is not an ECCS system. The typical BWR RCIC system does not fulfill either of the specified safety functions (A_D) listed in this reporting criterion.

Loss of One Trajn (9) Oversized Breaker Wiring Lugs Situation:

During testing of 480 volt safety-related breakers, one breaker would not trip electrically.

Investigation revealed that one wire of the pigtail on the trip coil, although still in its lug, was so loose that there was no electrical connection. The loose connection was due to the fact that the pigtail lug was too large (No. 14-16 AWG), whereas the pigtail wire was No. 20 AWG. A No. 18-22 lug is the acceptable industry standard for a No. 20 AWG wire.

Since the trip coils were supplied pre-wired, all safety-related breakers utilizing the trip coil were inspected. All other breakers inspected had No. 14-16 AWG lugs. No lugs were found with loose electrical connections. Nevertheless, all No. 14-16 AWG lugs were replaced with acceptable industry Standard No. 18-22 AWG lugs.

Comment:

The event is reportable because the incompatible pigtails and lugs could have caused one or more safety systems to fail to perform their Intended function [50.72<b)l2)Cv) and 50.73(Q)(2)(v)].

(10) Contaminated Hydraulic Fluid Degrades MSIV Operation Situation:

During a routine shutdown, the operator noted that the #11 MSIV ctosing time appeared to be excessive. A subsequent test revealed the #11 MSIV shut within the required time, however, the #12 MSIV ctosing time exceeded the maximum at 7.4 sec. Contamination~v/r;.(

the hydraulic fluid in the valve actuation system had caused the system's check valves to stick and delay the transmission of hydraulic pressure to the actuator. Three more filte

@urchased providing supplemental filtering for each MSIV. Aner filters will be used in pump suction filters to remove the fine contaminants. The #12 MSIV was repaired and /\

returned to service. Since the valves were not required for operation at the time of ~

discovery, the safety of the public was not affected.

Comments:

The event is reportable under 50,73(a)(2)(v) because a single~ condition could have prevented fulfillment of a safety function ~-73(a)(2H'v)1. The fact th8t the condition mis discovered v1hen the 'vafrtes .. ere not required for oe,eration does not affect the reportability of the condition. The event is not reportable under 50.72<b}(2)ly) because, at the time *of discovery. the plant was shutdown and the MSIV's were not required to be operable, Draft NUREG-_1022. Rev. 2 56

(13) Generic Set-point Drift

Situation:

With the plant in steady state operation at 2170 MWt and while performing a Main Steam Line Pressure Instrument Functional Test and Calibration, a switch was found to actuate at 853 psig. The Tech Specs limit is 825 +15 psig. The redundant switches were operable. The cause of the occurrence was set point drift. The switch was recalibrated and tested successfully per HNP-2-5279, Barksdale Pressure Switch Calibration, and returned to service.

This is a repetitive event as reported in one previous LER. A generic review revealed that these type switches are used on other safety systems and that this type switch is subject to drift. An investigation will continue as to why these switches drift, and if necessary, they will be replaced.

Comments:

The event is not reportable due to the drift of a single pressure switch.

The event is reportable if it is indicative of a generic and/or repetitive problem with this type of switch which is used in several safety systems [50.73(a)(2)(vi} or (vii)].

Question:

Are set point drift problems with a particular switch to be reported if they are experienced more than once?

Answer. J._\,.+e. £.

The independent f a ; ~ £ ~ e set point drift) of a Single pressure switch is n0I

(14) reportable unless ~ ~flused a system to fail to fulfill its safety function, or is indicative of a genenc problem that could have resulted in the failure of more than one switch and thereby cause one or more systems to fail to fulfill their safety function.

Maintenance Affecting Only One Train Question:

Suppose the wrong lubricant was installed in one pump, but the pump in the other train was correctly lubricated. Is this reportable?

Answer:

Engineering judgement is required to decide if the lubricant could have been used on the other pump, and, therefore, the system function would have been lost If the procedure called for testing of the first pump before maintenance was performed on the second pump and testing ciearly identified the error, then the error would not be reportable. However, if the procedure called for the wrong lubricant and eventually both pumps would have been Draft NUREG-1022, Rev. 2 58

IN 85-27 described multiple failures of a reactor protection system during control rod insertion testing of a reactor at power. One of the control rods stuck. Subsequent testing identified 3 additional rods that would not Insert (scram) into the core and 11 control rods that had an initial hesitation before insertion. The licensee considered each failure as a single random failure; thus each was determined not to be reportable. Subsequent assessments indicated that the instrument air system, which was to be oil-free, was contaminated with oil that was causing the saam solenoid valves to fail. While the failure of a single rod to insert may not cause a reasonable doubt that-about the ability Qf other rods v.rottld fail to insert, the affected, lhU9 afleeting !he !llllety lanetion of the rods. I::) f failure of more than one rod does cause a reasonable doub~~ other rods cottld be

~

As indicated in IN 85-27, muttiple failures of redundant components of a ety system are sufficient reason to expect that the failure mechanism, even though not known, could prevent the fulfillment of the safety function.

(17) Potential Loss of High Pressure Coolant Injection

During normaJ refueling leak testing of the upstream containment Isolation check vatve on the High Pressure Coolant Injection (HPCI) steam exhaust, the case of the non-containment isolation check valve was found lodged in downstream piping. This might have prevented HPCI from functioning if the disc had blocked the line. The event was caused by fatigue failure of a disc pin.

Following evaluation of the condition, the event was determined to be reportable because the HPCI couk:f have been prevented from pertorming its safety function if the disc had blocked the line. In addition, the event is reportable if the fatigue failure is inaicative of a common-mode failure.

(1 a, Defeethe Gomponent Deli'vered btrt not lnstailed Question!

I lorn shot:lld a plant report a defecti-o'e component that was eJelM!red, but not installed9 A single defedi*te component Yt'Ottld not generally be r9"ort8:ble t&SSt1ming that the problem has no generic implications). A generic problem or a number of defecth1e components wot:Jld probably constittite a eondltion that eould ha't-e pnJJ-ented fulfillment of a safety function, and, if so, nrould be reportable. Engineering judgment is reqttired to determine if the defeas could ha'te escaped detection prior to installation and operation. As a minimttm, any generic problem may be reported as a *t10IUrrtarf LEA. In addition, stieh a condition may be reportable ttf'1der 10 Offi Part 21.

(18) Operator Inaction or Wrong Action Question: In some systems used to control the release of radioactivity, a detector controls certain equipment In other systems, a monitor is present and the operator is required to initiate action under certain conditions. The operator is not "wired* in. Are failures of the Draft NUREG-1022, Rev. 2 60

10 CFR 50.72 §50.73(a)(2)(vii)

(No corresponding Licensees aJI report *Any event where a single cause or Part 50.72 condltio caused at least one independent train or channel to requirement.] become operable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident*

Licensees are required to report a common-cause failure as an LEA within 98-60 days.

Discussion This criterion requires those events to be reported where a single cause or cooortion caused independent trains or channels to become inoperable. Common-causes may include such factors as high ambient temperatures, heat up from energization, Inadequate preventive maintenance, oil contamination of air systems, incorrect lubrication, use of non-qualified components or manufacturing or design flaws. The event Is reportable if the independent trains or channels were inoperable at the same time, regardless of whether or not they were discovered at the same time. (Example (2) below illustrates a case where the second failure was discovered 3 days later than the first)

An event or failure that results in or involves the failure of independent portions of more than one train or channel in the same or different systems is reportable. For example, if a cause or

condition caused components in Train *A* and *s* of a single system to become inoperable, even if additional trains (e.g., Train *c*) were still available, the event must be reported. In addition, if the cause or condition ca.used components in Train *A* of one system and in Train

s* of another system {i.e., train that is assumed in the safety analysis to be independent) to

become inoperable, the event must be reported. However, if a cause or condition caused components in Train *A* of one system and Train *A* of another system (i.e., trains that are not assumed in the safety analysis to be independent), the event need not be reported unless it meets one or more of the other reporting .criteria.

Trains or channels for reportability purposes are defined as those redundant, independent trains or channels designed to provide protection against single failures. Many engineered safety systems containing active components are designed with at least a two-train system. Each independent train in a two--train system can normally satisfy all the safety system requirements to safely shut down the plant or satisfy those criteria that have to be met following an accident This criterion does not include those cases where one train of a system or a component was rem_oved from service as part of a planned evolution, in accordance with an approved procedure, Draft NUREG-1022, Rev. 2 62

TXU Electric C. Lance Terry Comanche Peak Senior Vice President & Principal Nuclear Officer Steam Electric Station P.O. Box 1002 *99 Sff 27 P 2 :C 4 Glen Rose, TX 76043 Tel: 254 897 8920 Fax: 254 897 6652 lterryl@txu.com Log# TXX-99213 File# 10185 Ref. # 10CFR50. 72 10CFR50.73 DOCKET NUMBER PROPOSED RU so .J- 1.:2

(~ 'IFR3/,~'llj September 20, 1999 Secretary of the Commission U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn: Rulemakings and Adjudications Staff

SUBJECT:

COMMENTS ON PROPOSED RULE 10 CFR PARTS 50 AND 72, "REPORTING REQUIREMENTS FOR NUCLEAR POWER REACTORS," AND DRAFT NUREG-1022, REVISION 2, "EVENT REPORTING GUIDELINES - 10 CFR 50.72 AND 50.73" REF: Federal Register Volume 64, No. 128, Pages 36291-36307, dated July 6, 1999 TXU Electric appreciates the effort that the Nuclear Regulatory Commission (NRC) has put into revising 10 CFR 50.72 and 10 CFR 50.73 and allowing for industry involvement. Specifically, the table top exercises, held on November 13, 1998, as discussed in the Reference page 36291 were of great benefit and allowed for an open exchange of information. TXU Electric supports the NRC's position to reduce or eliminate the reporting burden associated with events of little or no safety significance. The proposed rule has made progress toward this goal. However, there are specific portions of the rule and associated NUREG-1022, Revision 2, that detract from this goal and, if implemented, may increase the reporting burden of the licensee.

This proposed rule was reviewed collectively by several utilities (AmerenUE, Wolf Creek Nuclear Operating Company, Pacific Gas and Electric Company, STP Nuclear Operating Company and TXU Electric) and our areas of concern include:

TXX-99213 Page 2 of 4 Significantly Degraded Components Most significant is the recent addition of a new reporting criterion for Significantly Degraded Component(s). TXU Electric finds the new criterion as written to be unclear in focus and subject to widely varying interpretation.

The attempt to capture items (components) which are seriously degraded, but not necessarily enough to render a system inoperable, is far below the current reporting threshold and represents a significant increase in licensee burden.

Further, even if an operability statement is added to the criterion, as suggested at the tabletop, held on August 3, 1999, TXU Electric believes that a component degradation significant enough to render a system inoperable would be captured by other reporting criteria, making this new criterion redundant and unnecessary.

In addition, the new criterion as discussed by the staff at the table top exercise on August 3, 1999, appeared to have been added primarily as a data collection mechanism for motor operated valves. This is contrary to the stated objectives of the proposed rule which focuses on events of risk and safety significance.

If component data is needed, there are other resources available to the NRC to obtain the information such as maintenance rule reports and the Equipment Performance Information Exchange (EPIX).

TXU Electric recommends that the new criterion not be added to the rule. As discussed at the table top, held on August 3, 1999, significant component degradations that are risk and/or safety significant will be reported under other criteria such as loss of a function, common-mode failure, Part 21 , or as a Technical Specification violation.

ESF Actuations The addition of a specific list of systems which must be reported as Engineered Safety Features will increase reporting by plants whose licensing basis does not include those specific systems. TXU Electric recommend a return to the pre-1998 practice of relying on each facility's Final Safety Analysis Report (FSAR) and shift to a risk-informed approach when such criteria are fully developed. As part of the effort to risk inform Part 50, this section should be changed to be more risk informed.

. TXU T:XX-99213 Page 3 of 4 Invalid ESF actuations are still included in the proposed rule change reporting requirements for the written LER. TXU Electric recommends that the existing clarifications for not reporting certain invalid actuations be retained in the guidance. Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event. Specifically, actuations when the system is already properly removed from service, part of a planned evolution, that occur after the safety function has already been completed, or single component actuations of complex systems, which do not by themselves mitigate the consequences of significant events, are not reportable.

LER Information The scope of information requested for human performance events has increased by shifting from "personnel error" and the implied "root cause" to "human performance related problem" and "contributing factors." It is more appropriate to require discussion of personnel error root causes.

TXU Electric recommends deletion of the new criterion, 10 CPR

50. 73(b )(3)(ii), which requires a discussion of emergency or operating procedures that could have been used to recover from an event be included in a LER. The proposed rule change would result in a large amount of additional information that would be of minimal use. The safety consequences discussion would be cluttered with hypothetical failures and speculated plant responses.

Historical Limitations TXU Electric supports the new historical reporting limitation for operations prohibited by the plant's Technical Specification and conditions that could have prevented fulfillment of a safety function, but believes this limitation should be applied to the whole rule.

TXX-99213 Page 4 of 4 In addition to the general comments above, TXU Electric endorses the comments submitted to the NRC by the Nuclear Energy Institute (NEI).

Sincerely, 6(1~~

C. L. Terry By:~

Roger~alker

&:411Jilk Regulatory Affairs Manager TAH/grj cc: Mr. E. W. Merschoff, Region IV Mr. J. I. Tapia, Region IV Mr. D. H. Jaffe, NRR Resident Inspectors, CPSES Mr. Bob Post, NEI

Pacific Gas and Electric Company Lawrence F. Womack Diablo Canyon Power Plant Vice President P.O. Box 56 Nuclear Technical Services Avila Beach, CA 93424

99 SU' 27 AJO :S 1 805.545.6000 September 20, 1999 OFF., ,

f 1!.

PG&E Letter DCL-99-123 AOJI_ L Secretary DO" ,r- UMBER U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 PROPOSED RU So" 1:2-Attn. : Rulemakings and Adjudications Staff l (,l/F~3, ~e,i)

Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Comments on proposed rule 10 CFR Parts 50 and 72, "Reporting Requirements for Nuclear Power Reactors," and Draft NUREG-1022, Revision 2, "Event Reporting Guidelines - 10 CFR 50. 72 and 50. 73"

Reference:

Federal Register Volume 64, No. 128, Pages 36291-36307, dated July 6, 1999

Dear Secretary:

PG&E appreciates the effort that the NRC has put into revising 10 CFR 50.72 and 10 CFR 50. 73 and allowing for industry involvement. Specifically, the table top exercises, held on November 13, 1998, as discussed at Reference page 36291 were of great benefit and allowed for an open exchange of information. PG&E supports the NRC's position to reduce or eliminate the reporting burden associated with events of little or no safety significance. The proposed rule has made progress toward this goal. However, there are specific portions of the rule and associated NUREG-1 022, Revision 2, that detract from this goal and if implemented may increase the reporting burden of the licensee. This proposed rule was reviewed collectively by several utilities (AmerenUE, Wolf Creek Nuclear Operating Company, Pacific Gas and Electric Company, STP Nuclear Operating Company and TXU Electric). Our areas of concern include:

Rulemaking and Adjudication Staff PG&E Letter DCL-99-123 September 20, 1999 Page2 Significantly Degraded Components Most significant is the recent addition of a new reporting criterion for Significantly Degraded Component(s). PG&E finds the new criterion as written to be unclear in focus and subject to widely varying interpretation.

The attempt to capture items (components) which are seriously degraded, but not necessarily enough to render a system inoperable, is far below the current reporting threshold and represents a significant increase in licensee burden. Further, even if an operability statement is added to the criterion, as suggested at the tabletop, held on August 3, 1999, PG&E believes that a component degradation significant enough to render a system inoperable

- would be captured by other reporting criteria, making this new criterion redundant and unnecessary.

In addition, the new criterion as discussed by the staff at the table top exercise on August 3, 1999, appeared to have been added primarily as a data collection mechanism for motor operated valves. This is contrary to the stated objectives of the proposed rule which focuses on events of risk and safety significance. If component data is needed, there are other resources available to the NRC to obtain the information such as maintenance rule reports and the Equipment Performance Information Exchange (EPIX).

PG&E recommends that the new criterion not be added to the rule. As discussed at the table top, held on August 3, 1999, significant component degradations that are risk and/or safety significant will be reported under other criteria such as loss of a function, common-mode failure, Part 21 , or as a Technical Specification violation.

ESF Actuations The addition of a specific list of systems which must be reported as Engineered Safety Features will increase reporting by plants whose licensing basis does not include those specific systems. PG&E recommends a return to the pre-1998 practice of relying on each facility's Final Safety Analysis Report (FSAR) and shift to a risk-inform ed approach when such criteria are fully developed. As part of the effort to risk inform Part 50, this section should be changed to be more risk informed.

Invalid ESF actuations are still included in the proposed rule change reporting requirements for the written LER. PG&E recommends that the existing clarifications for not reporting certain invalid actuations be retained in the

Rulemaking and Adjudication Staff PG&E Letter DCL-99-123 September 20, 1999 Page3 guidance. Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event. Specifically, actuations when the system is already properly removed from service, that are part of a planned evolution, that occur after the safety function has already been completed, or single component actuations of complex systems, which do not by themselves mitigate the consequences of significant events, are not reportable.

LER Information The scope of information requested for human performance events has increased by shifting from "personnel error and the implied "root cause" to "human performance related problem" and "contributing factors. " It is more appropriate to require discussion of personnel error root causes.

PG&E recommends deletion of the new criterion, 10 CFR 50.73 (b)(3)(ii),

which requires a discussion of emergency or operating procedures that could have been used to recover from an event be included in a LER. The proposed rule change would result in a large amount of additional information that would be of minimal use. The safety consequences discussion would be cluttered with hypothetical failures and speculated plant responses.

Historical Limitations PG&E supports the new historical reporting limitation for operations prohibited by the plant's Technical Specification and conditions that could have prevented fulfillment of a safety function, but believes this limitation should be applied to the whole rule.

In addition to the general comments above, PG&E endorses the comments submitted to the NRC by the Nuclear Energy Institute (NEI).

Sincerely, Lawrence F. Womack

Rulemaking and Adjudication Staff PG&E Letter DCL-99-123 September 20, 1999 Page4 cc: Steven D. Bloom Ellis W. Merschoff David L. Proulx Diablo Distribution Robert E. Post (NEI)

RLR

r~ ' ' '\,,_

~ Northeast Rope Ferry Rd. (Route 156), Waterford, CT 06385

-if Nuclear Energy fJO i;,. L Millstone Nuclear Power Station Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385-0128 (860) 447-1791

99 ~C'"' 24 P2 :33 Fax (860) 444-4277 The Northeast Utilities System o.

r, SEP 2 3 1999 Al 817889 DOCKET NUMBER Re: 10CFR50.72 PROPOSED RULE 10CFR50.73 Secretary, U.S. Nuclear Regulatory Commission Attention: Rulemaking and Adjudications Staff Washington, DC 20555-0001 Millstone Nuclear Power Station Comments on the Proposed Rule for Reporting Requirements for Nuclear Power Reactors These comments are submitted on behalf of Northeast Nuclear Energy Company's Millstone Station in response to the Federal Register notice concerning proposed rulemaking on Reporting Requirements for Nuclear Power Reactors (64 Federal Register 36291 of July 6, 1999). Northeast Nuclear Energy Company (NNECO) endorses the comments provided by the Nuclear Energy Institute (NEI) on behalf of the nuclear industry. NNECO also supports the efforts made by the NRC Staff with this proposed rulemaking in reducing the reporting burden on licensees and the NRC with regard to reports that have little or no safety significance.

NNECO wants to emphasize the concern raised by NEI regarding the last minute addition of the requirement to report significantly degraded components in Section 50.73(a)(2)(ii)(C). This reporting criterion does not meet the stated objectives of the rule change and should be deleted. The lack of specificity in the reporting threshold associated with this aspect of the rule change renders this criterion ambiguous, and thus, subject to widely varying interpretations. As a result, the reporting burden of non-safety-significant events or conditions will very likely increase. Further, based on the discussions surrounding this criterion at the workshop of August 3, 1999, it is NNECO's opinion that the information sought by the staff can be obtained through other reporting criteria and may be available currently through other measures such as the Equipment Performance Information Exchange or the NRC Inspection Program.

EP 2 9 1999

. . . . . . . .INIIIIIIIIII I ~

053422-5 REV. 12-95

LEARREGU J GS& UD OFTflES OFTHECC Ml

U.S. Nuclear Regulatory Commission B17889\Page 2 NNECO is also commenting on the change from the current requirement to report a plant being in an "unanalyzed condition that significantly compromises plant safety" to the proposed wording in Section 50.72(b)(2)(ii)(B) to report being in an "unanalyzed condition that significantly affects plant safety". If implemented this change could broaden the scope of scenarios subject to reporting. Although this change is being characterized as editorial , the wording of the proposed revision invites a broader application of what could be considered reportable, and thus, has the potential of not meeting the stated objectives.

NNECO agrees with extending the non-emergency prompt notifications to eight hours and the LER reporting to 60 days. This will help to eliminate unnecessary reports and retractions as well as unnecessary LER supplements. NNECO also agrees with eliminating the "outside design basis" reports. This will result in a significant burden reduction to both the NRC and Licensees .

If you have any questions regarding these Ms. Mari Jaworsky at (860) 447-1791 , extension 5379.

Very truly yours, comments, please contact NORTHEAST NUCLEAR ENERGY COMPANY R. P. Nacci Vice President - Nuclear Oversight and Regulatory Affairs

cc: H. J. Miller, Region I Administrator L. L. Wheeler, NRC Project Manager, Millstone Unit No. 1 D. P. Beaulieu, Senior Resident Inspector, Millstone Unit No. 2 R. 8. Eaton, NRC Senior Project Manager, Millstone Unit No. 2 A C. Cerna, Senior Resident Inspector, Millstone Unit No. 3 J. A Nakoski, NRC Project Manager, Millstone Unit No. 3 P. C. Cataldo, NRC Resident Inspector

Union Electric PO Box 620 Callaway Plant DOCKET NUMBER Fulton, MO 65251 PROPOSED RULE DOC- -J

'r

99 s:" 23 P: *os r

I..,

,\l _ September 22, 1999 Secretary U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn: Rulemakings and Adjudications Staff

~~

meren UE ULNRC-04117 Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1, UNION ELECTRIC CO.

FACILITY OPERA TING LICENSE NPF-30

Reference:

Federal Register Volume 64, No . 128, Pages 36291-36307, dated July 6, 1999

Subject:

Comments On Proposed Rule IO CFR Parts 50 And 72, "Reporting Requirements For Nuclear Power Reactors," And Draft NUREG-1022, Revision 2, "Event Reporting Guidelines -

10 CFR 50.72 And 50.73 "

Union Electric (UE) appreciates the effort that the Nuclear Regulatory Commission (NRC) has put into revising 10 CFR 50.72 and 10 CFR 50.73 and allowing for industry involvement. Specifically, the table top exercises, held on November 13, 1998, as discussed in the Reference page 3 6291 were of great benefit and allowed for an open exchange of information. UE supports the NRC ' s position to reduce or eliminate the reporting burden associated with events of little or no safety significance. The proposed rule has made progress toward this goal.

EP l 1999

~cfmo edged by card I I IIIFSIIII 91 . . . . aw a subsidiary of Ameren Corporation

LIS RU

ULNRC-041 17 Callaway Plant Page2 However, there are specific portions of the rule and associated NUREG-1022, Revision 2, that detract from this goal and, if implemented, may increase the reporting burden of the licensee. Our areas of concern include:

Significantly Degraded Components Most important is the recent addition of a new reporting criterion for Significantly Degraded Component(s). UE finds the new criterion, as written, to be unclear in focus and subject to widely varying interpretation. The attempt to capture items (components) which are seriously degraded, but not necessarily enough to render a system inoperable, is far below the current reporting threshold and represents an increase in licensee burden. Further, even if an operability statement is added to the criterion, as suggested at the August 3, 1999 tabletop, UE feels that a component degradation significant enough to render a system inoperable would be captured by other reporting criteria, making this new criterion redundant and unnecessary.

In addition, the new criterion as discussed by the staff at the table top exercise on August 3, 1999, appeared to have been added primarily as a data collection mechanism for motor operated valves. This is contrary to the stated objectives of the proposed rule which focuses on events of risk and safety significance. If component data is needed, there are other resources available to the NRC to obtain the information, such as maintenance rule reports and the Equipment Performance Information Exchange (EPIX).

UE recommends that the new criterion not be added to the rule. As discussed at the table top, held on August 3, 1999, significant component degradations that are risk and/or safety significant will be reported under other criteria such as loss of a function, common-mode failure, Part 21 , or as a Technical Specification violation.

ESF Actuations The addition of a specific list of systems which must be reported as Engineered Safety Features will increase reporting by plants whose licensing basis does not include those specific systems. UE recommends a return to the pre-1998 practice of relying on each facility's Final Safety Analysis Report (FSAR) and shift to a risk-informed approach when such criteria are fully developed. As part of the effort to risk inform Part 50, this section should be changed to be more risk informed.

Invalid ESF actuations are still included in the proposed rule change reporting requirements for the written LER. UE recommends that the existing clarifications for not reporting certain invalid actuations be retained in the

ULNRC-04117 Callaway Plant Page 3 guidance. Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event. Specifically, actuations when the system is already properly removed from service, part of a planned evolution, that occur after the safety function has already been completed, or single component actuations of complex systems, which do not by themselves mitigate the consequences of significant events, should not be reportable.

LER Information The scope of information requested for human performance events has increased by shifting from "personnel error" and the implied "root cause" to "human performance related problem" and "contributing factors." It is more appropriate to require discussion of personnel error root causes .

UE recommends deletion of the new criterion, 10 CFR 50.73(a)(3)(ii), which requires a discussion of emergency or operating procedures, that could have been used to recover from an event, be included in a LER. The proposed rule change would result in a large amount of additional information that would be of minimal use. The safety consequences discussion would be cluttered with hypothetical failures and speculated plant responses.

Historical Limitations UE supports the new historical reporting limitation for operations prohibited by the plant's Technical Specification and conditions that could have prevented fulfillment of a safety function. UE recommends this limitation be

applied to the active rule.

In addition to the general comments above, UE endorses the comments submitted to the NRC by the Nuclear Energy Institute (NEI).

Sincerely, (A)~,1-,t(J#

Warren A. Witt Assistant Plant Manager WAW/MAR/mib

ULNRC-04117 Callaway Plant Page4 cc: U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop Pl-137 Washington, DC 20555-0001 Mr. Ellis W . Merschoff Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N . Donohew (2 copies)

Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop OWFN-4D3 Washington, DC 20555-2738 Manager, Electric Department

Missouri Public Service Commission PO Box 360 Jefferson City, MO 651 02 Superintendent, Licensing Wolf Creek Nuclear Operating Corporation POBox411 Burlington, KS 66839

D "KE U BER PROPOSED RU A CMS Energy Company Palisades Nuclear Plant Tel: 616 764 2276

99 S[~ 23 A1: :_6 27780 Blue Star Memorial Highway Fax: 616 764 3265 Covert, Ml 49043 llathan L. Haskell Director, Licensing Cr Fi ADJ September 17, 1999 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN : Rulemakings and Adjudications Staff

SUBJECT:

Comments on Proposed Rule for Reporting Requirements for Nuclear Power Reactors -- 64 Federal Register 36293 -- July 6, 1999 Consumers Energy Company is pleased to provide the following comments in response to the Proposed Rule for Reporting Requirements for Nuclear Power Reactors, published on July 6, 1999 in 64 Federal Register 36293.

In a letter dated on or about September 17, 1999, the Nuclear Energy Institute will be submitting comments on the proposed rulemaking on behalf of the nuclear industry .

Consumers Energy Company has participated in the development of those comments, and hereby endorses the NEI letter in its entirety.

We are particularly concerned about the new requirement proposed to be added as 50.73(a)(2)(ii)(C) to report a component being in a degraded condition. This is truly a new requirement that has not been adequately evaluated for need and burden. We disagree with NRC staff statements made during a public meeting on August 3, 1999, that the proposed component reporting criterion already exists as a subset of the current criterion for reporting conditions "outside the design basis of the plant". The new criterion to report SSC degradation that has potential generic implications represents a reporting threshold far below that previously required. Degradation, however severe, that does not impair a component's (and by extension its train's and system's) ability to meet its functional requirements (ie, component remains operable) ,

is not reportable under 10 CFR 50.72 or 50.73, unless that degradation results in a condition prohibited by Technical Specifications. Furthermore, such degradation that does not impair a component's ability to complete its required functions can not have EP 2 8 \99

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any risk significance. Including the reporting of non-risk-significant matters in 10 CFR

50. 72/73 is inconsistent with the NRC's stated objectives for the rule change.

Component failure data is already being provided by licensees through existing systems such as the Equipment Performance Information Exchange (EPIX). If NRC needs component data, utilization of EPIX or other sources would be more efficient and would provide more uniform and comprehensive data than reports filed under 10 CFR 50.72/73.

It is also recommended that the NRC reconsider the reporting of invalid ESF actuations, and revise the reporting criterion so that only valid actuations are required to be reported . Invalid ESF actuations are, by their nature, not risk significant. Utilization of ESF actuation data as some measure of demands made on ESF components is flawed because the reported data fails to include the majority of demands, those resulting from surveillance, post maintenance, and post modification testing . Since utilization of invalid actuation data to derive ESF reliability still does not provide a credible measurement, the rule should not require it to be reported.

Nathan L. Haskell Director, Licensing

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JnnstJrook Technical O ..,ter 5000 Dominion Boulevard Glen Allen, Virginia 23060 IIIRGINIA POWER September 21 , 1999 GL99-041 DOCKET NUMBER

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Secretary U.S. Nuclear Regulatory Commission Washington , DC 20555 ATTN: Rulemaking and Adjudication Staff REPORTING REQUIREMENTS FOR NUCLEAR POWER REACTORS; 10 CFR PARTS 50 AND 72 Virginia Power appreciates the opportunity to comment on the proposed rule to amend the event repo rting requirements for nuclear power reactors. Notice of the proposed rule was published in the Federal Register (Vol. 64, No. 128) on July 6, 1999.

We endorse the industry comments subm itted by the Nuclear Energy Institute, NEI.

If you need further information, please contact Gwen Newman, at Gwen_Newman@vapower.com or (804) 273-4255.

Respectfully, EP 28 1r

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ilt/r-R3b~C,1) September 20, 1999 JPN-99-029 IPN-99-103 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attention: Rulemaking and Adjudications Staff

SUBJECT:

Indian Point 3 Nuclear Power Plant Docket No. 50-286 James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Comments on Reporting Requirements For Nuclear Power Plants

REFERENCES:

1. Federal Register, July 6, 1999, 64 Fed. Reg. 36293

2. NEI to USNRC dated September 17, 1999 regarding proposed rule for reporting requirements for nuclear power reactors - 64 Federal Register 36293 - July 6, 1999 .

Dear Sir:

The Authority is submitting the comments summarized below on the proposed event-reporting requirements for nuclear power plants, Reference 1 .

Overall, these changes represent an improvement for both the NRC and licensees. With these changes, the NRC has made a significant step towards reducing the reporting burden by reducing or eliminating the necessity for reports with little or no safety significance.

The rule' s extended reporting times will allow Authority personnel to better manage their resources, while potentially reducing the number of amended reports. In short, these changes go a long way towards accomplishing the NRC's objective while meeting their information needs for prompt action and fulfilling its safety mission.

However, the Authority has a few remaining concerns regarding the rule as currently proposed. Of particular concern is a recently added requirement pertaining to degraded or non-conforming components.

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Necessity of Reportin9 Degraded Components The proposed rule would require the reporting of components .,being in a degraded or non-conforming condition ... " 10 CFR 50.73(a)(2)(ii)(C). The Authority agrees with NEI (Reference 2). This late addition to the rule does not meet the overall objectives of the rulemaking package and should be deleted. This provision does not ( 1) better align reporting needs, (2) reduce the burden of reporting items of little safety significance, or (3) improve clarity in reporting.

This provision lacks clarity and is, in at least part, redundant to other reporting requirements. Lack of clarity was the reason reporting equipment .,outside the design basis" was eliminated from the current rule. This provision is equally, if not more unclear and represents the opportunity for even greater inconsistencies in reporting. In addition, it is difficult to hypothesize a scenario where a risk-significant, degraded or nonconforming component would not be reported under some other provision of either this rule, or some other reporting requirement -- such as 10 CFR 21 .

Part 21 already covers component failures with generic implications that .,could reasonably be expected to affect other similar components ... "

Reporting Invalid ESF Actuations As currently proposed, the rule would eliminate the necessity for telephone notification of an invalid ESF actuation, but would retain requirements for a written report. The FR notice (Reference 1) states that written reports are necessary to make .,estimates of equipment reliability parameters, which in turn are needed fo support the Commission's move towards risk-informed regulation." The requirement for written reports should also be eliminated.

Other mechanisms for collecting equipment reliability data are already available and would better serve the intended purpose.

The Authority supports the concept of risk-informed regulation and recognizes the necessity for reliable supporting data. The research and preparation of these is a resource intensive process. However, data collection could be collected and submitted through existing programs without imposing the burden of regulatory reporting requirements on licensees. Industry programs, such as INPO's EPIX, already exist to provide information on nuclear power plant equipment reliability. If invalid ESF actuations are not adequately reported under those programs, consideration should be given to revising them so that they can be. Industry initiatives to enhance current data collect channels are already under discussion. These initiatives could be used as a vehicle for implementing improved ESF system reliability reporting.

Standardized Definition of ESF The proposed rule contains a detailed list of ESF systems for reporting and there are benefits to standardizing reporting requirements. However, all plants are not the same.

As recent work in the area of PRA has shown us, seemingly subtle differences in environment or operating procedures can have significant effects on the risk contributions of similar equipment and systems. For this reason, the Authority supports NEl's position (Reference 2) that a detailed list of standardized ESF equipment not be included in the rule. -

2

We recognize that the adoption of plant-unique equipment lists may run counter to the benefits of a common set of ESF systems.

As suggested by NEI, reporting requirements should continue to use the long-standing practice of relying on each plant's UFSAR. Ultimately, any reliance on the ESF category could be eliminated and replaced with reporting criteria based on a risk-informed ranking of plant-specific structures, systems and components.

In general, the Authority supports NEl's comments (Reference 2) on this rulemaking.

This letter does not contain any new commitments. If you have any questions regarding this matter, please contact the Director - Nuclear Licensing, Ms. C. D. Faison.

Senior Vice President and Chief Nuclear Officer cc: Next page 3

cc: Regional Administrator U.S. Nuclear Regulatory Commission 4 75 Allendale Road King of Prussia, PA 19406 Office of the Resident Inspector U.S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Office of the Resident Inspector U.S. Nuclear Regulatory Commission Indian Point 3 Nuclear Power Plant P. 0. Box 337 Buchanan, NY 10511

Mr. George F. Wunder, Project Manager Project Directorate I Division of Licensing, Project Management U. S. Nuclear Regulatory Commission Mail Stop 8C4 Washington, DC 20555 Mr. Guy Vissing, Project Manager Project Directorate I Division of Licensing, Project Management U. S. Nuclear Regulatory Commission Mail Stop 8C2 Washington, DC 20555 4

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Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20055-0001 ATTN: Rulemakings and Adjudications Staff Gentlemen:

Subject:

Comments on Proposed Rule for 10 CFR 50.72 and 50.73, Reporting Requirements, and Draft NUREG-1022, Revision 2 Cooper Nuclear Station, NRC Docket 50-298, DPR-46

Reference:

1. 64 Federal Register 36291, dated July 6, 1999, Proposed Rule, "Reporting Requirements for Nuclear Power Reactors" The Nebraska Public Power District (District) hereby submits comments on the proposed rule in Reference 1, including the draft report NUREG-1022 Revision 2, for Nuclear Regulatory Commission (NRC) consideration. The District agrees with the stated objectives for the proposed rule, specifically better aligning the reporting requirements with the NRC's current reporting needs, reducing reporting burden, and clarifying reporting requirements where warranted. The District also appreciates the NRC' s efforts in revising the guidelines, NUREG-1022, concurrent with the proposed rule change, as these guidelines are an essential component to interpreting and implementing the rule(s).

The Nuclear Energy Institute (NEI) sponsored an industry task force providing comments to the NRC regarding the proposed rule and draft report NUREG-1022 Revision 2 in a public workshop held August 3, 1999 at NRC headquarters, and in a subsequent letter from NEI. The District endorses the NEI task force comments; therefore, they will not be ~ted in this correspondence. However, the District respectfully submits additional comments to emphasize and support the industry position.

Reporting of Component Level Failures The District fully support's NEI's comments in strong opposition to the reporting of component level failures (proposed rule 10 CFR 50.73 (a)(2)(iiXC)). We agree that this was a last-minute Cooper Nuclear Station P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 8.25-3811 I Fax: (402) 8.25-5211 http//www nppd com

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NLS990098 September 20, 1999 Page 2 of 4 addition in the process and does not meet the NRC's stated objectives for this proposed rule:making. An informal query of the Cooper Nucl~ Station (CNS) condition reporting system was conducted to ascertain the impact this change could have. The query revealed that, from January 1, 1998 to September 2, 1999, there were at least 93 occurrences where particular components were determined INOPERABLE due to a degraded or nonconforming condition but -

the system remained OPERABLE. We agree that these 93 occurrences would potentially result in additional Licensee Event Reports (LERs), or at the very least require evaluation for reporting, should this new criterion be added. We believe that the reporting of these types of events does not afford additional insight on risk impacts and therefore does not meet the threshold of items requiring reporting to the NRC.

The Federal Register notice states that these types of events would be reportable ~use the "capability of several similar components to perform their specified safety functions could be sigpificantly degraded. It should be noted that "could be" has ~ distinct meaning; based on the reasons for reporting events (i.e., significance to public health and safety) po'stulated events or conditions are not appropriate to be reported unless there is firm evidence to demonstrate that the capability of several similar components to ,perform their specified safety functions would be significantly degraded. Reporting based on speculation of how a discovered condition could affect similar components, without firm evidence, does not serv~ the purpose of providing valuable information with which the NRC may evaluate and analyze precursors to' potential severe core damage. It is also in opposition to the philosophy underlying Improved Technical Specifications and the provisions of Generic Letter 91-18, Revision 1, which provides specific guidance on treatment of degraded or nonconforming conditions in the arena of operability &nd corrective action. Granted that in many cases thei:e exists a link between operability and reportability, the balance of the reporting criteria sufficiently covers when a report would be required: common cause, loss of safety function, a condition of the* plant (including principal safety barriers) being significantly degraded or the plant being in an unanalyzed condition that significantly degrades plant safety.

For -example, if a failure occurs on a component critical to operability of an emergency diesel generator (EDG), it is prudent to *evaluate if there is a common cause which would cause a failure in the redundant EDG. Note that this should be limited to those events which would cause failures in redundant or similar equipment, not those failures that only "could." Where would one draw the line on what "could" p.appen? If a diesel fuel oil pump fails, it is reasonable to assume that the other fuel oil pump could at some point fail for the same reason. However this needs to be carefully evaluated to understand the failure modes and the relative risk presented to the redundant fuel oil pump for these failure modes to occur. In this example, it could simply be that.the diesel fuel pump which failed had reached the end of,its useful life, but this does not always mean that the same holds: true for the redundant pump. If, however, it is discovered that the underlying cause of the failure could have caused a failure in the redundant train, and there is reasonable doubt that the redundant train would complete its safety function if called upon, the event would be reportable as a condition which would have resulted in the loss of a safety function.

NLS990098 September 20, 1999 Page 3 of 4 Engineered Safeguards Features (ESP) System Actuations We agree that the all-inclusive list is inappropriate since risk-significance of a particular system can vary greatly between plants. The NRC states in the Federal Register notice that the listing of specific systems would "result in consistent reporting of events that result in actuation of these highly risk-significant systems." We are unaware of how the NRC arrived at the conclusion that the listed systems are in fact "highly risk-significant" at all plants. A prime, example is the feedwater coolant injection system listed for Boiling Water Reactors-we agree with NEI's comments on this in that it is not an Emergency Core Cooling System.

Rather than listing the specific systems, a more appropriate approach may be to state the functions. However, this would still require that all the necessary criteria would have to be adequately established as part of risk-informing 10 CPR Part 50. Until Part 50 is risk-informed, we agree that the Option where reporting would be required for the actuation of "any ESP" as is defined in each facilities FSAR is the appropriate option at this time.

Reporting of Historical Events The current NEI letter concludes this section by stating: "Reporting historical events more than two years old provides a low safety benefit and unnecessarily increases the reporting burden.

This exclusion of such historical events and conditions should be extended to all written reports required by section 50.73(a)." The question becomes, what is the value of reporting historical events even within two years? Is it not more of regulatory significance than of safety significance? As Part 50 becomes more risk informed, the approach would be that cases such as this be dealt with in the Corrective Action program at each facility; there are other ways to communicate events to the industry and the NRC which would not require the same expenditure of resources (both licensee and NRC) as submitting an LER For example, LERs on Technical Specification violations should only apply to real-time plant issues, and an LER would be submitted after the event is over and the facts are understood. The LER content requirements could be slightly modified to ensure that licensees research their corrective action programs for historical events and provide that in the "previous similar events" category. The District recognizes that this is an "out-of-the-box" suggestion which may not be appropriate to consider at this juncture, but perhaps should be considered in future changes to risk-inform Part 50.

Additional, specific comments are contained in the Enclosure to this letter. The District appreciates the NRC's efforts in this area and the opportunity to respond to the proposed rule and accompanying reporting guidelines.

Should you have any questions concerning this matter, please contact me at (402) 825-5551, or Linda Dewhirst at (402) 825-5009.

NLS990098 September 20, 1999 Page4 of4 Sincerely,

~~

Larry Newman Acting Manager, Nuclear Licensing and Safety

/lrd Enclosure cc: NPG Distribution

Enclosure to NLS990098 September 20, 1999 Page 1 of7 PROPOSED NUREG-1022 REV. 2 TEXT COOPER NUCLEAR STATION COMMENTS ABBREVIATIONS p. ix-xi Editorial Comment: Formatting on a few of the abbreviations to align the margins:

Generic letter, incident investigation team, information notice, inservice inspection, and technical specifications Section 2.8 Retraction or Cancellation of Event Reports Example 3 in Section ~ 3 .2.4 illustrates a case where there Example 3 in Section 3.3.1 illustrates a case where there were were sound reasons for a retraction. The laat e¥ent Hfl:der sound reasons for a retraction. The last event under Example 1 in e*0:E1::1ple l mgeetioa ~ .~ .~ HlastFates e: ease 1i'ffi@fe the FeasoB:S Section 3.3.2 illustrates a case where the reasons for retraction fof fetraetioa v.refe aot adeEtMffie.

were not adequate.

Referenced sections need to be corrected to coincide with the appropriate sections in the Revision 2 of the NUREG. In the latter example, the draft NUREG has deleted the discussion on the inappropriate retraction (see last event under Example 1 in Section 3.2. 6; as such, either the strikeout that was done in that section needs to be restored OR the above second sentence needs to be deleted). NOTE: if the discussion on the inappropriate retraction is left in the discussion for the last event in Section 3.2.6, then the above needs to reflect the appropriate section, i.e.,

" .... Example 1 in Section 3.2.6 ... "

Section 3.2.2 Technical Specification Prohibited Operation or For a discrepancy discovered during a timely surveillance test Condition, TS Surveillance Requirements (i.e., a surveillance test that was performed within the TS ~uired For a discrepancy discovered during a timely surveillance test, it frequency), it should be assumed that the discrepancy should be assumed that the discrepancy occurred ..... occurred .....

Editorial addition for clarity. Although later in the NUREG the definition of "timely" is implied, it seems prudent to specifically indicate what is meant by a "timely" surveillance test.

Section 3.2.2 Technical Specification Prohibited Operation or However, the existence of similar discrepancies in multiple valves Condition, Example (4), Multiple Test Failures is could be an indication that the discrepancies arose over a period

Enclosure to NLS990098 September 20, 1999 Page 2 of7 PROPOSED NUREG-1022 REV. 2 TEXT COOPER NUCLEAR STATION COMMENTS However, the existence of similar discrepancies in multiple valves of time. and the failure modes should be evaluated to make this is an indication that the discrepancies arose over a period of time. determination.

Changed "is" to "could be " and added a statement that an evaluation should determine this conclusion. While in most cases this statement may be true, it should still be evaluated to reach this conclusion.

Section 3.2.4 Degraded Condition If not reported as an emergency under 50.72(a), licensees are If not reported as an emergency under 50.72(a), licensees are required to report a seriously degraded principal safety barrier or required to report a seriously degraded principal safety barrier or an unanalyzed condition that significantly a.ff@eta degrades plant an unanalyzed condition that significantly affects plant safety .... safety ....

Consistent with other areas of the comment letter, the focus should not simply be on "affects" plant safety but "degrades plant safety. As stated elsewhere in the NEI comments: The phrase "significantly affects plant safety" has no positive or negative connotation. It is therefore recommended that the term degrades be substituted in place of "affects."

Section 3 .2.4 External Threat to Plant Safety. Discussion Usually, with the passage of time, it is apparent that an actual Usually, with the passage of time, it is apparent that an actual threat did not occur, and thus no LER in sabmitted required (see threat did not occur, and thus no LER is submitted (see Example Example 1). In some cases, with the passage of time, it is judged 1). In some cases, with the passage of time, it is judged that an that an actual threat did occur and, thus, an LER is sab1mtted actual threat did occur and, thus, an LER is submitted (see required (see Example 2).

Example2).

Editorial comment to be consistent with other parts ofthe NEI comment letter in terms of "required" vs. "submitted" Section 3.2.6~ System Actuation.. Discussion This discussion in this section appears to be reversing previous interpretations of this part of the rule which even though Part 50 is not yet risk-informed, were reasonable approaches from a

Enclosure to NLS990098 September 20, 1999 Page 3 of7 PROPOSED NUREG-1022 REV. 2 TEXT COOPER NUCLEAR STATION COMMENTS risk/safety significance perspective. For example, the section no longer recognizes that "system actuations" need not be reported for those signals not related to safety, such as Reactor water clean up system, Control Room emergency ventilation system, reactor building ventilation system, fuel building ventilation system, or auxiliary building ventilation system, which were also invalid signals or occur after the safety function has already been completed We believe that this is contrary to the intent of the rule and will result in significantly more reports, with little safety or risk significance.

The discussion on Actuation ofmultichannel ESF actuations systems is proposed as being deleted, however it appears that this should be le.ft in the NUREG as it provides valuable guidance in dealing with actuation logic issues, especially where maintenance was involved It is also unclear why discussions on "valid" actuations remain, yet many of the statements of consideration or other text regarding "invalid" actuations are being deleted, especially when later in the NURF,G it is stated that invalid actuations are not required to be reported under 50. 72 however it is implied that an LER under 50. 73 would be required Further, in the second to the last paragraph before the Examples, the NURF,G states "when invalid actuations excluded by the conditions described above occur ... "yet the majority of"conditions" are being deleted Again, there seems to be little safety benefit in reporting certain

Enclosure to NLS990098 September 20, 1999 Page 4 of7 PROPOSED NUREG-1022 REV. 2 TEXT COOPER NUCLEAR STATION COMMENTS invalid actuations at all, provided they meet certain criteria, as has been past practice.

Section 3.2.6 System Actuation. ExamQle (1), RPS Actuation This example is inconsistent with the changes made in the With the BWR defueled, an invalid signal actuated the RPS. Discussion of the NUREG, as the text supporting this example is There was no component operation because the control rod drive proposed as being deleted It is recommended that the supporting system had been properly removed from service. This event is not text on invalid actuations and exclusions from reporting be reportable because (1) the RPS signal was invalid and (2) the restored (see above discussion).

system had been properly removed from service.

Section 3.2.6 System Actuation. ExamQle (1), RPS Actuation This event is reportable as a manual RPS actuation. Even though This event is reportable as a manual RPS actuation. Even though the reactor scram was in response to an existing written the reactor scram was in response to an existing written procedure, this event does not involve a preplanned sequence procedure, this event does not involve a preplanned sequence because the loss of recirculation pumps and the resultant off-because the loss of recirculation pumps and the resultant off- normal procedure entry were event driven, not pre-planned. AB normal procedure entry were event driven, not pre-planned. AB LER is reqHH"ed. In this case, the licensee initially retracted the l:,ER: is reEtQired. Ia teis eaee, ilie lieeaaee ~aY~ retraeted ilie ENS notification believing that the event was not reportable.

@+~ B:etifieatieB: eelie1,cmg iliat the e1-*0B:t 1r..aa B:et repertaele. After staff review and further discussion, it was agreed that the

A....-'ter sta.ff i:e1rie 1rt aa:d further dise:yasieB:; it :i,1,1ae agreed that ilie event is reportable for the reasons discussed above.

0 1.. eB:t is repertable fer t;ee reaeeaa dise:yased a-be1,ce.

Recommend leaving the discussion regarding an inappropriate retraction in this section PROVIDED a reference is made to it in Section 2. 8 above.

Section 3.2.6 System Actuation, Example (6), manual Editorial: Do not delete the "of' in the title.

Actuation ef "ESF Component in Response to Actual Plant Condition In this example the emphasis seems to be a component level The staff considers the manual start of the charging pump (which actuation, which is clearly not the intent as generally single also serves as an ECCS pump, but with a different valve lineup) components do not mitigate an event. As such to ensure this in response to dropping pressurizer level to be an intentional example is consistent it should be clearly explained if this single manual actuation ef a::B: :gsp in response to equipment failure or component is similar to the EDG example.

human error and reportable because it constitutes deliberate

Enclosure to NLS990098 September 20, 1999 Page 5 of7 PROPOSED NUREG-1022 REV. 2 TEXT COOPER NUCLEAR STATION COMMENTS manual actuation of a single component af BB ESF, in response to plant conditions.

Section 3.2. 7, Event or Condition That Could Prevent If the event or condition 0ffiH0: would prevent fo1:611ment of the Fulfillment of a Safety Function, Discussion safety function at the time of discovery, it would be reportable If the event or condition could 12revent fu1:611ment of the safm under 50.72(b)(2)(v) (ENS notification). If it eeffia-would have function at the time of discovery, it would be ~ortable under prevented fu):611ment of the safety function at any time within 50.72(b)(2)(v) (ENS notification). If it could have ImlVented ~ two years of the date of discovery, it would be reportable fu):fi))ment of the safm function at any time within three years of under 50.73(b)92)(v) (written LER).

the date of discovea. it would be feJ)Ortable under 50.73<b)22)(v)

(written LER). These criteria cover an event or condition where structures, components, or trains of a safety system 0ffiH0: would have failed These criteria cover an event or condition where structures, to perform their intended functions ......

components, or trains of a safety system could have failed to perform their intended functions ...... If so, this would constitute an event that "ee1/4HQ would have prevented" the fo)fi))ment of a safety function, and, accordingly, Last sentence ofthe paragraph following the paragraph on GDC must be reported.

17: If so, this would constitute an -event that "could have prevented" the fu1:611ment of a safety function, and, accordingly, A design or analysis defect or deviation is reportable under this must be reported. criterion if it 001/4HQ would prevent fn):fi))ment of the safety function of structures or systems defined in the rules.

A desig!! or analysis defect or deviation is reQQrtable under this criterion if it could 12revent fu1fi11ment of the safm function of Should replace "could" with "would" to reflect the existence of structures or ~stems defined in the rules. an event or condition, rather than past speculation. Also should reflect two years irutead ofthree, as this is the NEUindustry proposal.

Section 3.2.7, Event or Condition That Could Prevent So would reactor core isolation cooling NOT be required to be Fulfillment of a Safety Function, Discussion reported as a loss ofsafety function if the system is not credited in The definition of the systems included in the scope of these an analysis and would not be required to perform one ofthe four criteria is provided in the rules themselves. It includes ~stems .functioru (A) through (D) as specified in the rule?

Enclosure to NLS990098 September 20, 1999 Page 6 of7 PROPOSED NUREG-1022 REV. 2 TEXT COOPER NUCLEAR STATION COMMENTS

~uired by the TS to be OQera.ble to Qerform one of the four functiQns (A) throulili (D) s~cified in the rule.

Section 3.2.7, Event or Condition That Could Prevent It is recommended that the appropriate reporting criteria be Fulfillment of a Safety Function included; i.e., 50. 72 vs. 50. 73 to clearly illustrate which one All Examples applies (or if both apply) for each of the examples.

Section 3.2.7, Event or Condition That Could Prevent This event is reportable because disabling the somce range Fulfillment of a Safety Function, Example (5) Procedure detectors ~ would have prevented fulfillment of the safety Error Prevents Reactor Shutdown Function function to shut down the reactor.

This event is reportable because disabling the source range detectors could have prevented fnl:fi))ment of the safety function Change "could" to "would" for reasons stated previously.

to shut down the reactor.

Section 3.2.7, Event or Condition That Could Prevent Thus, in this case, because of the use of the wrong lub.ricant, the Fulfillment of a Safety Function, Example (8), Maintenance system "eould have" or "would haveZ! failed and is therefore Affecting Two Trains reportable under 10 CFR 50.73(a)(2)(v).

Thus, in this case, because of the use of the wrong lubricant, the system "could have" or would have" failed. Again, change "could" to "would" Also should add that this is

reportable; this appears to be an example of a 50. 73 report so it should be stated as such Section 3.2. 7, Event or Condition That Could Prevent Following evaluation of the condition, the event was determined J?nlfillment of a Safety Function, Example (17- to be reportable because the HPCI 8ffiHQ would have been DUPLICATE), Potential Loss of High Pressure Coolant prevented from performing its safety function ....

Injection Following evaluation of the condition, the event was determined Again, change "could" to "would" to be reportable because the HPCI could have been prevented from performing its safety function ....

Enclosure to NLS990098 September 20, 1999 Page 7 of7 PROPOSED NUREG-1022 REV. 2 TEXT COOPER NUCLEAR STATION COMMENTS Section 3.2.10~ Internal Threat to Plant Safm In plant releases must be reported if they require evacuation of In plant releases must be reported if they require evacuation of rooms or buildings and, as a result, the ability of the operators to rooms or buildings and, as a result, the ability of the operators to perform necessary duties for safe operation of the plant is perform necessary duties is significantly hampered. significantly hampered.

Editorial addition "for safe operation ofthe plant" consistent with other parts ofthis section.

Section 4.2.1, Timeliness See Section~ 2.5 for further discussion of reporting See Section 2.11 for further discussion of reporting timeliness. timeliness.

Section 2.11 no longer exists; 2. 5 is the appropriate reference.

Section 5.1.1, Submission of LERs See Section~ 2.5 for further discussion of discovery c4lte.

See Section 2.11 for further discussion of discovery date.

Section 2.11 no longer exists; 2. 5 is the appropriate reference.

Section 4.2.3, ENS Notification Retraction This section refers to Section 2.10 for further discussion, however no such section exists. It appears to have been replaced by Section 2.8. However, se*ction 2.8 refers the reader to sections 4.2.3 and 5.1.2. This is a circular reference!!

Section 5.2.1, Narrative Description or Text (NRC Form See Section ~ 2.5 for further discussion of discovery date.

366A, Item 17), 50.73(b)(2)(iI)(C)

See Section 2.11 for further discussion of discovery date. Section 2.11 no longer exists; 2. 5 is the appropriate reference.

CP&L Carolina Power & Light Company PO Box 1551 411 Fayetteville Street Mall PE&RAS-99-077 Raleigh NC 27602 September 20, 1999 Secretary U.S. Nuclear Regulatory Commission OCKET UMBE POSED RULE 50 J- 1 :i.

7fo'-IFR 3, :i.9/)

Washington, DC 20555-0001 Attn. : Rulemakings and Adjudications Staff

Subject:

Proposed Rule for Reporting Requirements for Nuclear Power Reactors, (64 FR 36291 - July 6, 1999)

Ladies and Gentlemen:

These are Carolina Power & Light Company (CP&L) comments related to the Proposed Rule for Reporting Requirements for Nuclear Power Reactors, noticed in 64 FR 36291 on July 6, 1999.

CP&L encourages the NRC to consider those comments made by the uclear Energy Institute (NEI) on behalf of the nuclear power industry.

Specifically, CP&L concurs with the po ition that the new requirement to report degraded components added by section 50.73(a)(2)(ii)( C ), does not meet the stated objective of the rule change and should be deleted. This attempt to capture components that are degraded without rendering the associated system inoperable, is subject to various levels of interpretation. This would increase the burden to licensees without a commensurate improvement in safety.

CP&L also supports the position that the proposed rule should not contain a detailed list of engineered safety features (ESF) systems for reporting. Each facility FSAR specifies equipment that is designated as ESF, and there are plant specific differences in the safety-related status of each facility's systems. Therefore, an all-inclusive list of systems in the regulation is not appropriate. It is our position that we should continue to rely on each facility's FSAR.

Please contact me at (919) 546-4579 if you have questions.

,471y,

~f~~

Regulatory Affairs Sf P 2 8 199

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SEP 2 2 1999 RUI.EMAKINGS AN s:F Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402-2801 September 17 , 1999 Secretary of the Commission U.S. Nuclear Regulatory Commission ATTN: Rulemaking and Adjudication Staff Washington, D.C. 20555-0001

Gentlemen:

NUCLEAR REGULATORY COMMI SSION (NRC) - OPPORTUNITY FOR PUBLIC COMMENTS ON PROPOSED RULEMAKING, "REPORTING REQUIREMENTS FOR NUCLEAR POWER REACTORS" On July 6, 1999, NRC published a proposed rulemaking related to reporting requirements for nuclear power reactors for public comment (64 FR 36291). Along with the change to the rule, NRC made available a draft revision of the accompanying reporting guidelines, i.e., NUREG 1022, Revision 2, "Event Reporting Guidelines, and 10 CFR 50.72 and 50.73." TVA has reviewed the proposed rule and the draft NUREG revision and has several comments and

recommendations for your consideration.

TVA agrees with the stated objectives for the proposed rule; namely, better aligning the reporting requirements with the NRC's current reporting needs, reducing the reporting burden, and clarifying the reporting requirements where needed.

Additionally, TVA appreciates NRC's efforts to provide the revised guidelines concurrently with the rule change. This is particularly important for event reporting because of the essential part that the guidelines play in interpreting and implementing the rule(s) . TVA is participating in an industry task force sponsored by the Nuclear Energy Institute (NEI). The task force provided comments regarding the proposed rule and the draft NUREG 1022 Revision 2 to NRC staff in a public workshop at NRC headquarters on August 3, 1999, and in a subsequent letter from NEI.

SEP 2 8 1999 ed by ca.rel ...............,*** ~ ~ A At Printed on recycled -

UCLEAR REGULATORY t,t, nv1.~ir,KINGS & ADJUDICATIO FFICE OF THE SECRET OF THE COM~IS

Secretary of the Commission Page 2 September 17, 1999 TVA endorses the task force comments; therefore, they will not be repeated in this letter. However, some additional comments on the same issues are included to emphasize our support and provide TVA's perspective.

Specifically, TVA would like to emphasize and support NEI's comments in strong opposition to the proposed component-level reporting criterion (i.e., 10 CFR 50.73 (a) (2) (ii) (C)). This newly-proposed criterion was a last-minute addition to the process and does not meet any of NRC's stated objectives for the reporting rule changes.

Conversely, it is evident that the criterion represents another attempt to reimpose requirements for a data collection effort that does not relate to event reporting and that has been previously denied by NRC senior management.

There are several elements of the proposed rule and the draft NUREG 1022, Revision 2, that were not addressed by the industry task force upon which TVA would like to comment. Additionally, we have included comments on Section III, "Analysis of Comments," issued in the Federal Register notice along with the proposed rule. All of our comments are provided in the enc l osure to this letter.

We appreciate NRC's efforts in this area and the opportunity

to respond to the proposed r ule.

Sincerely,

'!~)f?:::~l

Mana~;

Nuclear Licensing Enclosure cc (Enclosure):

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Secretary of the Commission Page 3 September 17, 1999 FCM:BJG Enclosure cc (Enclosure):

T. E. Abney, PAB lG-BFN J.E. Blackburn, OSA lE-BLN P. L. Pace, ADM lL-WBN Pedro Salas, OPS 4C-SQN EDMS, WR 4Q-C

Enclosure Federal Register Volume 64, No. 128 - Proposed Rule - RIN 3150-AF98 (36291)

Section III *Analysis of Comment~*

Comment 7 The industry's comment recommended eliminating all reporting of invalid actuations of Engineered Safety Feature actuations. NRC's response to the industry's comment accepted the recommendation with regard to 10 CFR 50.72 ENS notifications of invalid actuations, however, declined the recommendation to eliminate the requirement for a written report of an invalid actuation under 10 CFR 50.73. NRC's stated reason for continuing the requirement for the written report of invalid actuations states:

~There is still a need for reporting of invalid actuations because they are needed to make estimates of equipment reliability parameters, which in turn are needed to support the Commission's move towards risk-informed regulation. This is discussed further in a May 7, 1997 Commission paper, SECY-97-101, Proposed Rule, 10 CFR 50.76, Reporting Reliability ahd Availability Information for Risk-significant Systems and Equipment, Attachment 3.w TVAN Comment- The data needed for equipment rellabllity estimates Is already collected via INPO's EPIX system and made available to the NRC In lieu of the proposed data rule.

As such, there Is not need to make these reports via 10 CFR 50.73. Also, note that the purpose of the LER described here Is not Included In the response to Comment 33 (Section Ill, uAnalysls of Comments"). Therefore, we recommend that the response to this industry comment be reconsidered and the requirement for reporting Invalid and unintentional actuations be ellmlnated as recommended In the earller comments.

Comment 22 The comment suggested the idea of changing the wording to allow waiting until the beginning of the next business day for non-emergency 50.72 phone calls. This was based on the observation that NRC was unlikely to take action on the events during non-business hours. NRC's response indicates that the reports are needed in case additional information is required so that they might ~initiate a special inspection or an investigation, if warranted, within a day.w TVAN Comment-This Is Inconsistent with the Revised Reactor Oversight Process In which special Inspections or Investigations will not be Initiated based on the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report.

Instead, the basellne inspection program has an event follow-up review that will use the significance determination process to assess the event and determine the significance of the event, Follow-up Inspections will be Initiated in accordance with the action matrix consistent with the significance (color of finding) of the event.

1

Comment 28 This comment suggested eliminating two of the requirements for innnediate followup notification during the course of an event, section 50.72(c) (2) (i), the results of ensuing evaluations or assessments of plant conditions, and section S0.72(c) (2) (ii), the effectiveness of response or protective measures taken. The comment indicated that the requirements continue to apply after the event and that they require reporting even if, for example, the result of a further analysis does not change the initial report. NRC's response stated that follow-up reports are needed while the event is ongoing. For example, if an analysis is completed during an ongoing event, and it confirms an earlier estimate of how long it will take to uncover the reactor core if electric power is not restored, that information may very well be useful for the purpose of evaluating the need for protective measures (evacuation) .

TVAN Comment-The assessment of Comment 28 ls Inappropriate. The 10 CFR 50.72 process Is not used to detennine or communicate protective measures. The responsibility for these activities, the criteria for decisions, and the associated communication responslbllltles are described In the utlllty's radlologlcal emergency plan. The NRC assessment should be withdrawn and reconsidered approprlately.

Section IV, *oiscuaaion*

Subsection 2 *section by Section Discussion of Proposed Amandmante*

Paragraph 3 states, ~Non-emergency events that are reportable by telephone under 10 CFR 50.72 would be reportable as soon as practical and in all cases within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (instead of within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as is currently required).

TVAN Comment- TVAN Is concerned that th'e words 11 as soon as practlcal" might result In Inappropriate emphasis or second-guessing on exactly when the report was made.

Paragraph 3 also states, ~This [reporting non-emergency events in 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s] would reduce the burden of rapid reporting, while still capturing those events where there may be a need for the NRC to contact the plant to find out more about the event and/or initiate a special inspection or investigation within about a day.

TVAN Comment-As stated In TVAN's comments on Comment 22 (see above) this appears to be inconsistent with the criteria for Initiating special Investigations or Inspections as described In the Revised Reactor Oversight Process.

Paragraph 6 reflects a concern with the potential impact on the Performance Indicator program by extending the LER submittal time to 60 days. The concern states that *Adding 30 days to the delivery of data supplying these programs would result in the reduction in the currency and value of these indicators to senior managers. With respect to the Accident Sequence Precursor program, the additional 30 days will add a commensurate amount of time to each individual event assessment since Licensee Event Reports (LERs) are the main source of data for these 2

analyses. The delivery date for the annual Accident Sequence Precursor report would also slip accordingly. The NRC staff would have to make more extensive use of Immediate Notifications (10 CFR 50.72) and event follow-up to compensate in part for the Licensee Event Report (LER) reporting extension.*

TVAN Comment- Data reporting for the performance Indicator program In the Revised Reactor Oversight Process Is independent of LER reporting time frames. The performance Indicator data wm be reported within 15 days from the end of the quarter and wlll Include events that have occurred (or found) In that quarter since the Indicators use the date occurrence for reporting. Allowing 60 days for submitting an LER provides reasonable time to perform a corrective action evaluation commensurate with the safety significance.

This time frame Is reasonable based on experience. The 60 day Orne frame will not affect reporting of performance indicator data or any NRC decisions based on asaessments of the Indicator data.

Paragraph 7 invites public connnent regarding whether additional levels should be introduced to better correspond to particular types of events.

Examples are given which would maintain one type of report at the current 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months requirement and another at a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months limit.

TVAN Comment-With the stated exception of our recommendation to further relax non-emergency reporting until "the next business day," TVAN supports the currently proposed three levels of reporting (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and 60 days) as opposed to the suggestion In the comment to reintroduce a 4-hour notification requirement. As listed In the comment In response to Paragraph 6 above, the change from 30 to 60 days does not Impact reporting of performance Indicators as Implied. The 1-hour and 8-hour requirements are,consistent with the significance of the events whlle the additional complexity associated with other levels of reporting do not appear to provide any measurable addltlonal level of,beneflt for NRC response.

Paragraph 13 would word the requirement in the new Sections S0.72(b) (2) (ii) (B) and S0.73(a) (2) (ii} (B) to report ~a condition that affects plant safety- rather than the current wording of sections SO. 72 (b) (1) (ii) (A) and 50. 73 (a) (2) (ii) (A) which require reporting of a condition that *compromises plant safety.* NRC comments indicate ~hat the revision is fteditorial* and is intended to ~better reflect the nature of the criterion.*

TVAN Comment- TVAN does not agree that the comment can be interpreted as merely editorial. Rather, It can be construed to be a change In the intent of the criterion which could slgnlflcantly lower the threshold of reporting under this criterion. The proposed wording would result in the need for reporting any affect no matter how sllght (even theoretically In the positive direction). It appears that the current wording is better suited In that it allows the licensee some room for Judgment as to whether the effect Is sufflclently deleterious as to cause plant safety to be compromised.

Paragraph 30 includes a discussion of the new criterion in Section 50.73(a) (2) (ii) {C). That is ~Licensee shail report *Any event or condition that resulted in:_ (C) A component being in a degraded or non-conforming condition such that the ability of the component to perform its specified safety function is significantly degraded and the condition could reasonably be expected to affect other similar components in the plant.* The paragraph states, *In particular, the 3

proposed amendments will help ensure that significant safety problems that could reasonably be expected to be applicable to similar components at the specific plant or at other plants will be identified and addressed although the specific licensee might determine that the system or structure remained operable, or that technical specification requirements were met.w TVAN Comment-TVAN strongly supports the comments made by the Nuclear Energy Institute regarding this newly-proposed criterion. TVAN has concluded that there Is no basis for establishing a new reporting criterion where, "the condition could reasonably be expected to affect other slmllar components In the plant." It Is our concluslon that components that are found to be In a degraded and or non-conforming condition such that the ability of the component's ability to perform Its specified safety function Is slgnlflcantly degraded are substantively addressed as part of our 10 CFR 50 Appendix B Corrective Action Program. This program requires that, as part of determining the cause of the condition, the extent of that condition be Identified (i.e., other components which could be affected by that same condition) in order to preclude recurrence of the degraded or non-conforming condition .

Addltlonally, and more specifically, the statement In this paragraph, uthat could reasonably be expected to be appllcable to slmllar components at the specific plant or at other plants" -

should be removed. This statement could be interpreted to imply that the decision to report would include an assessment as to whether the Identified condition applies to other plants. While our 10 CFR 50 Appendix B Corrective Action Program Includes an assessment of potential generic applicability to other TVAN nuclear sites, it is not reasonable for TVAN personnel to accurately assess whether a condition might apply to nuclear plants outside of TVAN.

Furthermore, TVAN Is concerned with the use of the words "significant or "slgnlflcantly" In the description of this criterion. Experience has shown that these words are hlghly Interpretable and subjective. TVAN Is concerned that use of an unquantifiable term such as "significant" without a specific definition would lead to unproductive second-guessing.

Paragraphs 32 through 34 contain examples of conditions that NRC provided to illustrate the criterion for a significantly degraded component(s).

TVAN Comment- These examples show the high level of "lnterpretablllty" or subjectivity In the criterion that TVA feels makes this criterion problematic.

In the case of the normally-open valves In the safety injection system, It Is debatable that, given that the valves would be able to reopen (/.e., their safety function) that they should be considered to be significantly degraded.

In the example of the grease-contaminated jumper wires, despite their exposure to grease and the contention that the electrical components were not qualified for exposure to grease, there was no mention In the example that any of the valves had failed to operate or that testing had show any loss of continuity as a result of the contamination. Without those conditions, it would be debatable as to whether this condition resulted in a significant degradation in the components' ablllty to perform their safety function.

And In the example of the unquallfled valves discovered to be below flood level, the assessment showed that all four valves would Indeed perform their safety function. In that 4

case, If the valves would be able to perfonn their safety function, It Is debatable whether they could properly be assessed as significantly degraded.

Paragraph 40 regards a new section, 50.72 (b) 2 (iv) which would eliminate telephone reporting for invalid automatic actuation or unintentional manual actuation of Engineered Safety Features. The new rule would still require a written report of these actuations under 10 CFR 50.73. NRC states ~There is still a need for reporting of these events because they are used in making estimates of equipment reliability parameters, which in turn are needed to support the Commission's move towards risk-informed regulation. (See SECY-97-~0l, May 7, 1997, Proposed Rule, 10 CFR 50.76, Reporting Reliability and Availability Information for Risk-significant Systems and Equipment, ) ."

TVAN Comment-The data needed for equipment rellablllty estimates Is already collected via INPO's EPIX system and made available to the NRC In lieu of the proposed data rule.

As such, there Is not need to make these reports via 10 CFR 50.73. Also, note that the purpose of the LER described here Is not included In the response to Comment 33 (Section Ill, "Analysis of Comments"). Therefore, we recommend that the requirement for reporting Invalid and unintentional actuations be eliminated as recommended In the earller

comments.

Paragraph 41 through 46 propose the elimination of the term ~ESF" and proposes to add a list of systems (similar to the list currently provided in the reporting guidelines, i.e., NUREG 1022 Revision 1) to, the rule. Paragraph 45 sets forth three alternative strategies ,to the proposed rule. The first alternative is to ~Maintain the status quo."

The rule would continue to require reporting for actuation of any ESF. The guidance would continue to indicate that reporting should include as a minimum the system on the list. The second alternative would require use of a plant-specific, risk-informed list. Under this alternative, the list of systems would be risk-informed, and plant-specific. Licensees would develop the list based on existing PRA analyses, judgment, and specific plant design. No list would be provided in the rule. Alternative 3 would return to the pre-1998 situation (i.e.,*

before publication of the reporting guidance in NUREG-1022, Revision 1).

Under this alternative, the rule would continue to require reporting for actuation of any ESF. The guidance would indicate that reporting should include those systems identified as ESF's for each particular plant (e.g., in the FSAR). Paragraph 45 specifically invites comments on these alternatives.

TVAN Comment-TVAN supports alternative 3. As stated In Paragraph 46, this alternative was In use from 1984 through 1997 without any major problems. This alternative provides the hlghly-valued advantages of simplicity and clarity. Licensees are famlllar with their Safety Analysis Reports, the systems llsted as ESF systems, and what the safety analyses credit the systems for under specmc conditions. The perfonnance Indicators In the Revised Reactor Oversight Process and the avallable equipment performance Information in EPIX provide sufficient Information for NRC to detennine if actuations of safety systems are being caused by generic equipment performance problems. ft would however be important to include examples of non-reportable exceptions In the implementing guidance for systems that are considered to be ESFs, yet have lower levels of risk significance (control room ventilatlon systems, reactor building ventilatlon systems, fuel building 5

ventilation systems, auxlllary building ventllatlon systems, RWCU isolatlons during restoration from maintenance, etc.) as were implemented In Revision 1 of NU REG 1022.

Subsection 6, *Reporting of Component Problems*

TVAN Comment- Equipment rellability data Is collected via INPO's EPIX system and made avallable to the NRC In lleu of the proposed data rule. Also, NRC has a considerable number of hours under the basellne inspection program to conduct reviews of utility actions taken for equipment problems that affect the maintenance rule scope of equipment Any findings can be assessed using the significance detennlnatlon process to assess the significance of the problem. Follow-up Inspections will be initiated In accordance with the action matrix consistent with the significance (color of finding) of the event. As such, there Is not a clear purpose to require these reports via 10 CFR 50.73.

Subsection 7, *ED.forcement*

The discussion in this section describes NRC's intention to modify the

Enforcement Policy in connection with the proposed amendments to sections 50.72 and 50.73. The described revisions give as examples of Severity Level III and IV violations, failures or delays in reporting.

TVAN Comment- In the Revised Reactor Oversight Process special Inspections or Investigations will not be Initiated based on the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report Instead, the baseline inspection program has an event follow-up review that wlll use the significance detennlnatlon process to assess the event and detennlne the significance of the event, Follow-up Inspections wlll be Initiated In accordance with the action matrix consistent with the significance (color of finding) of the event. As such, the vlolatlon level for untimely reports should be pegged at Level N and only escalated under extremely unusual cases where the element of willfulness and a clear safety Impact has been demonstrated.

Additionally, this subsection states ~Repetitive failures to make LER reports indicative of a licensee's inability to recognize reportable conditions, such that it is not likely that the NRC will be made aware of operational, design and configuration issues deemed reportable pursuant to 10 CFR 50.73, will be considered for categorization at Severity Level III.w TVAN Comment- The use of a repetitive criterion for escalation of nonsafety-slgnlficant matters Is Inconsistent with the Commission's latest pronouncement precluding the aggregation of non-safety significant vlolatlons Into a safety significant problem, as well as the approach embodied In the Revised Reactor Oversight Process. That process does not use a repetitive criteria when considering whether to Issue a Level N violation for significance detenninatlon process Items. Slmllarly, It should not be used for escalation for non-significance determination process Items. Also, the rationale for grouping a series .

of non-significant LER-related issues for consideration of escalated enforcement Is strongly llnked to NRC's practice of pursuing escalated enforcement of non-safety-slgnlflcant items on the basis of "regulatory concern." This practice was also recently rejected by the Commission for future enforcement consideration. The escalation criteria for repetitiveness should be deleted.

6

Detailed Comments on Draft NUREG 1022 Revision 2 Section 2.s, Tim* Limit* fo~ Reporting (page 13)

This section states that 10 CFR 50.72 reporting times have some flexibility because a licensee heeds to ensure that reporting does_not interfere with plant operation. However, the guidelines go on to state that the reporting time flexibility does not mean that a licensee should automatically wait until close to the time limit expiration ,before:

reporting. The section contains an example of reporting of~ small radioactive release and concludes that the ENS notification should have 1 been made at about the same time as the report to the State.

TVAN Comment - One of the stated Objectives for extending the reporting time was to allow more complete investigation of the event In order to avoid unproductive supplemental notifications. NRC has already determined that 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was a sufflclently

tlmely period to meet NRC's event reporting needs. The example provided In the proposed amendment could have the effect of ellmlnatlng this benefit by opening the llcensee-to second-guessing and thereby forcing the llcensee to make the notification when delaying It, yet staying within the 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months would provide a more accurate, complete report.

Section 3.2.2 Technical Specification Prohibited Operation or,Condition Reporting Unaucce*aful Untimely Surveillance Teats (page 24)

The revision to this section introduces new guidance for determining the time of,inoperability of equipment when the results of an untimely surveillance test show that the equipment woul~ have been unable to perform its safety function. If the surveillance test had been timely, the time of inoperabili ty is the time of discovery (i.e.-, when the - test failed). Otherwise, the guidance states that the deficiency should be assumed to have occurred halfway between the last successful* test or use and the untimely test that revealed the deficiency.

TVAN Comment - It Is recommended that this new guidance be retracted since this .

statement appears to contain no basis. The deficiency should be assumed to have ,

occurred at the time of discovery (as Is the current practice) unless firm evidence exists to the contrary.

Bntry into STS 3.0.3 (page 25)

STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent, establishes requirements for actions when an LCO is not met and no action statement is provided.

Entry into STS 3.0.3 is considered to indicate that a condition existed longer than allowed by TS. Thus, entry into STS 3.0.3 (ISTS LCO 3~0.3) for any reason or justification is reportable.

TVAN Comment-TVAN does not agree with the prlnclple that entry Into STS 3.0.3 automatically constitutes a Technlcal Specification Prohibited Operation or Condition. STS 3.0.3 provides Technlcal Specification guidance for actions to be taken when the conditions In the appllcable LCO and its associated actions do not cover the existing condition. Some LCOs cover the range of posslble conditions and prescribe a specmc set 7

of actions for each. In many of these cases, the actions and times are consistent with those In STS 3.0.3. In other cases, adoption of STS, has resulted In replacing specific action statements such as "be In hot shutdown In 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months " with "Enter LCO 3.0.3."

If entry Into STS 3.0.3 results In Initiation of a shutdown, it will be reported under 10 CFR 50.72(b)(2)(I). If the tech-spec required shutdown Is completed, a Licensee Event Report would be required by 10 CFR 50.73{a)(2)(I)(A). Requiring an LER for any entry into STS 3.0.3 appears to Introduce an unnecessary report.

Mi*@td Tests Required by ASME Section ll IST and ISI and by STS 4.0.5, or Equivalent [page 25]

The section contains a sentence that states ~A missed or late IST/ISI/ASME test is reportable when it indicates that equipment (i.e.,

one train of a multiple train system) was not capable of pe~forming its specified safety functions and, thus, was inoperable for a period of time longer than allowed by TS.w TVAN Comment- Ttils sentence should be reworded to state, "A missed or*Iate ISTnSI/ASME test Is reportable when, once perfonned, It Indicates that equipment (I.e., one train of a multiple train system) was not capable of performing Its specified.safety functions and, thus, was Inoperable for a period of time longer than allowed by TS." This wording makes the guidance consistent with the guidance provided earlier In the section for missed surveillance tests. ;-

Rxample (3) (page 26)

Example 3 is provided as an example of an event which is reportable due to the consequent entry into STS 3.0.3.

TVAN Comment-This example should be removed consistent with the removal of the wording as Indicated In the comment above.

3.2.4 Degraded Condition (page 30)

This section states ~Licensees shall report: "Any event or condition that resultfil! in:

ffi1 The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded;_

(3) Steam generator tube degradation in the following circumstances:

(a) The severity of degradation corresponds to failure to maintain structural safety factors. The structural safety factors implicit in the licensing basis are those described in Regulatory Guide 1.121.

These safety factors include a margin of 3.0 against gross failure or burst under normal plant operating conditions, including startup, operation in the power range, hot standby, and cool-down, and all anticipated transients that are included in the plant design specification."

8

TVAN Comment -The use of normal full power operating pressure differential has historically been used as the basis for compllance with the three DP criteria. Technical Specification plugglng llmlts are typically based on full power differential pressure.

Design reports for minimum wall calculations for tubing and sleeves and Regulatory Gulde 1.121 compllance calculations were based on normal full power differential pressure. As such, adding the words "including startup, operation In the power range, hot standby, and cool-down, and all anticipated transients that are Included In the plant design specification" appears to constitute a regulatory position that is either new or different from a previously appllcable staff position, which represents a backflt. The phrase should be removed.

Section 50, 73 (a) (2) (ii) (C) [page 30]

A component being in a degraded or non-conforming condition such that the ability of the component to perform its specified safety function is significantly degraded and the condition could reasonably be expected to affect other similar components in the plant .

TVAN Comment - Thia newly-Introduced criterion adds a great deal of subjectivity to reporting, requiring speculation regarding generic appllcablllty. Comments on this criterion have been previously addressed In the section of this submlttal deallng with the proposed rule as presented in the Federal Register notice. Each of the examples Included with this criterion are addressed In the following paragraphs of this submlttal. Overall, our comment Is that, rather than pointing out a need for a new criterion, the examples tend to Indicate that the NRC staff disagreed with the licensee's anessment of operablllty In the cases described in the example. If the decision had been made that the components were degraded *to the point of system lnoperablllty, the conditions would have been reportable under existing reportability criteria. -;

Paragraph 1, page 33 - This section provides an example regarding ,

several normally open valves in the low pressure safety injection system that were routinely closed by a particular plant in order to support quarterly surveillance testing of the system. In reviewing the design basis and associated calculations, it was determined that the capability of the valves to open in the event of a large break LOCA combined with degraded grid voltage during a surve'illance test was significantly.

degraded. The licensee concluded that the valves would still be able to reopen under the postulated conditions and considered them operable.

However, that conclusion could not be supported using the conservative standards established by Generic Letter 89-10. Pending determination of final corrective action, administrative procedures- were implemented to preclude closing the valves. The event would be reportable because the capability of a component to perform its specified safety functions was significantly degraded and the same condition could reasonably be*

expected to apply to other similar components.

TVAN Comment- It Is debatable that, given that the valves would be able to reopen (/.e.,

their safety function) that they should properly be considered to be significantly degraded.

Paragraph 2, page 33 - In another example, during a routine periodic inspection, jumper wires in the valve operators for three valves were found contaminated with grease which was leaking from the limit switch gear box. The cause was overfilling of the grease box, as a result of following a generic ~intenance procedure. The leakage resulted in 9

contamination and degradation of the electrical components which were not qualified for exposure to grease. This could result in valve malfunction{s). The conditions were corrected and the maintenance procedures were changed. The event would be reportable because the capability of several similar components to perform their specified safety functions could be significantly degraded.

TVAN Comment - In this example, despite their exposure to grease and the contention that the electrical components were not qualtfled for exposure to grease, there was no mention In the example that any of the valves had failed to operate or that testing had shown any loss of continuity as a result of the contamination. Without those conditions, It would be debatable as to whether this condition resulted in a significant degradation in the components' ablllty to perform their safety function.

Paragraph 3, page 33 - In a further example, while processing calculations it was determined that four motor operated valves within the reactor building were located below the accident flood level and were not qualified for that condition. Pending replacement with qualified equipment, the licensee determined that three of the valves had sufficiently short opening time that their safety function would be completed before they were submerged. The fourth valve was normally open and could remain open. After flooding, valve position indication could be lost, but valve position could be established indirectly using process parameter indications. The event would be reportable because the capability of several similar components to perform their specified safety functions could be significantly degraded.

TVAN Comment - In this example, the assessment showed that all four valves would perform their safety function. If the valves would be able to perform their safety function, it is debatable whether they could properly be assessed as significantly degraded. *It could be reasonably determined by conducting an Engineering time-dependent study (similar to those conducted to assess the Environmental Qualificatlon of components) that there was no challenge to the safety function of the valves by this condition, therefore there was no significant degradation and no reason to report.

Paragraph 3, page 33 - An example of an event that would not be reportable is as follows. The motor on a motor-operated valve burned out after repeated cycling for testing. This event would not be .

reportable because it is a single component failure, and while there might be similar MOVs in the plant, there is not a reasonable basis to think that other MOVs would be affected by this same condition. On the other hand, if several MOVs had been repeatedly cycled and then after some extended period of time one of the MOVs was found inoperable or significantly degraded because of that cycling, then the condition would be reportable.

TVAN Comment - This example further Illustrates the subjectivity of the criterion. The lnltlal condition Is determined to be not reportable because of the single fallure nature of the condition. However, If other MOVs falled to operate or were found In a degraded condition because of repeated cycllng then the condition becomes reportable. Neither the example nor the guidance accompanying the criterion offer a definition of what comprises an "extended pertod of time." Similarly, no guidance is offered as to the number of fallures which would lead to a consistent Interpretation of the repeatablllty of the condition.

Page 34, Example 1 - Si~ifiecra.E De~rade:eien Failures of Reactor Fuel Rod Cladding Identified During Testing of Fuel Assemblies Radio-chemistry data for a particular PWR indicated that a number of fuel rods had failed during the first few months of operation.

Projections ranged from 6 to 12 failed rods. The end of cycle reactor coolant system iodine-131 activity averaged 0.025 micro curies per milliliter. following the end of cycle shutdown, iodine-131 spiked to 11.45 micro curies per milliliter. The cause was due to a significant number of failed fuel rods. Inspections revealed that 136 of the total 157 fuel assemblies contained failed fuel (approximately 300 fuel rods had through-wall penetrations), far exceeding the anticipated number of failures. The defects were generally pinhole sized. The fuel cladding failures were caused by long-term fretting from debris that became lodged between the lower fuel assembly nozzle and the first*spacer grid,,

resulting in penetration of the stainless-steel fuel cladding. The source of the debris was apparently a machining byproduct from the thermal shield support system repairs during the previous refueling outage.

The event is reportable because the cladding failures exceed expected values, are unique or widespread, and are caused by unexpected factors.

TVAN Comment - The revision to this example seems contrary to one of the stated purposes for the overall change to the reporting rule (i.e., reporting significant events or conditions). The revision changes the example from "Significant Degradation" to "Fallures." The Justification for reporting In this example adds unnecessary subjectivity to the example, because the three listed factors (expected values exceeded, the fallure be of a unique or widespread nature, and be caused by unexpected factors) In the revised Justification could all be met, and the fuel aaaembly barrier properly assessed as not significantly degraded.

Page 34, Bxample :2 - Reactor Coolant System Pressure Boundary Degradation due to Corrosion of a Control Rod Drive Mechanism Flange While the plant was in hot shutdown, a total of six control rod drive mechanism (CRDM) reactor vessel nozzle flanges were identified as leaking. Subsequently one of the flanges was found eroded and pitted.

While removing the nut ring from beneath the flange, it was discovered that approximately 50 percent of one of the nut ring halves had corroded away and that two of the four bolt holes in the corroded nut ring half were degraded to the point where there was no bolt/thread engagement.

An inspection of the flanges and spiral wound gaskets, which were removed from between the flanges, revealed that the cause of the leaks was the gradual deterioration of the gaskets from age. A replacement CRDM was installed and the gaskets on all six CRDMs were replaced with new design graphite-type gaskets.

The event is reportable because the degradation cannot be considered acceptable as-is.

TVAN Comment - The revision of the Justification of this example seems contrary to one of 11

the stated purposes for the overall change to the reporting rule (I.e., reporting significant events or conditions) by lowering the threshold for reporting. The fact that a condition Is determined to unacceptable as Is, does not in and of Itself Indicate that the condition represents a significant degradation of the barrier.

Page 35, Example 3 - Degradation of Reactor Fuel Rod Cladding Identified During Fuel Sipping Operations This section contains an example of a plant which while in cold shutdown, was conducting fuel sipping operations the results of which appeared to indicate a significant portion of cycle 2 fuel, _type "LYP,"

had failed, i.e., four confirmed and twelve potential fuel leakers. The potential fuel leakers had only been sipped once prior to making the ENS notification. The licensee contacted the fuel vendor for assistance on-site in evaluating this problem.

An ENS notification was made because the fuel cladding degradation was thought to be widespread. However, additional sipping operations and a subsequent evaluation by the licensee's reactor engineering department with vendor assistance concluded that no additional fuel failures had occurred, i.e., the abnormal readings associated with the potential fuel leakers was attributed to fission products trapped in the crud layer.

Based on the results of the evaluation the licensee concluded that the fuel cladding was not seriously degraded and that the event was not reportable. Consequently, after discussion with the Regional Office,,

the licensee retracted this event.

TVAN Comment-The Introduction of the phrases "appeared to Indicate" and 11 thought to be widespread," Introduce additional subjectivity to the example which could lead to unproductive second-gu888lng. The previous wording to the example implied that reporting was accomplished when there was a reasonable level of confirmation of the condition. The new wording implies that the reporting should take place at an earlier stage when the cause and extent of the problem are speculative.

Section 3.2.6 System Actuation 150.72 (b) (2) (iv) (B) and 850.73 (a) (2) (iv) (B)

The systems to which the requirements of paragraph (a) (2) (iv) (A) of this section apply are; (1) Reactor protection system (reactor scram, reactor trip).

(2) Emergency core cooling systems (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(3) ECCS for boiling water reactors (BWRs) including: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; feedwater coolant injection system; low pressure injection function of the residual heat removal system; and automatic depressurization system.

(4) BWR isolation condenser system and reactor core isolation cooling system.

(5) PWR auxiliary feedwater system.

(6) Containment systems including: containment and reactor vessel 12

isolation systems (general containment isolation signals affecting numerous valves and main steam isolation valve [MSIV] closure signals in BWRs) and containment heat removal and depressurization systems, including containment spray and fan cooler systems.

(7) Emergency ac electrical power systems, including: emergency diesel generators (EDGs} and their associated support systems; hydroelectric facilities used in lieu of EDGs at the Oconee Station; safety related gas turbine generators; BWR dedicated Division 3 EDGs and their associated support systems; and station blackout diesel generators (and black-start gas turbines that serve a similar purpose) which are started from the control room and included in the plant's operating and emergency procedures.

(8) Anticipated transient without scram (ATWS), mitigating systems.

(9) Service water (standby emergency service water systems that do not normally run).

TVAN Comment - TVAN Is opposed to the incluslon of a specific list of systems as a substitution for "actuation of any engineered safety feature (ESF) *.* ". The list will cause reporting of actuation of systems that are not listed In UFSARs for TVAN plants as Engineered Safety Features, which are not credited for design basis event mitigation, and are of llmlted safety significance. The concept of codifying a specific 11st of systems In the CFR and In the reporting guidelines (NUREG) is contrary to the stated objectives of the proposed rule change In that addltlonal reports would be flied for systems which have little

'bearing on ,safety.

The Federal Register Notice accompanying the proposed rule changes to 10 CFR 50.72 and 50.73 and NUREG 1022 Invited comments regarding three potential alternatives for

accomplish Ing NRC's objectfve for obtaining the needed information regarding actuation '

of safety systems. Aa stated In JVAN's comments on the corresponding section of the rule change, TVAN supports th&'alternatlve of returning to the pre-1998 version of the-rule

, which relies on the utilities' Updated Final Safety Analysls report to determine which .

systems are considered to be ESF systems (and therefore reportable under this criterion) for that specific plant. This ensures that only actuations for systems which have a recognized contribution to safety for that plant are reported.

If the alternative chosen results In Implementation of the 11st (or one similar) to that In the draft NUREG, TVAN recommends that Item (4) "BWR Isolation condenser system and reactor core Isolation coollng system" be removed from the list. TVAN's Bolllng Water Reactor (BWR) facility does not have a system which meets a description of "Isolation condenser system" and TVAN is concerned that If left as worded, the terminology could be used to other systems fulfllllng slmllar functions. Furthermore, the reactor core isolation cooling system (RCIC) Is not considered a decay heat removal system In the plant's UFSAR, I.e., no credit is taken for that function in accident analysis.

Addltlonally, Item (3) should be amended to remove "feedwater coolant Injection system" as this system Is not an Emergency Core Coollng System.

If Implemented, TVAN Is also concerned with the descriptive phrasing for other systems such as in item (6) "containment heat removal and depressurlzation systems" and "containment spray and fan cooler systems." These phrases lead to potential subjectivity.

and "lnterpretability" which could result in unproductive discussions regarding the need for reporting systems (such as feedwater Injection) which are not credited as safety systems in plant safety analyses, and have llttle or no Impact on safety.

13

TVAN's Comment on maintaining 10CFR50.73 reportability of invalid ESF actuations Is provided In the comments on the proposed rule and accompanying Federal Register Notice.

Paragraph 5, page 42 and Paragraph 1, page 43 TVAN Comment-These paragraphs previously provided exemptions for certain BWR systems for Invalid actuations caused by signals originated by non-ESF circuitry. The draft NU REG 1022 Revision 2 eliminates this paragraph, presumably because these systems are not Included in the proposed list (I.e., the exemption Is no longer needed). If another alternative Is chosen (other than the list), this exemption should be considered for reinstatement.

Section 3.2.7, Bvent or Condition That Could Prevent J!'ulfillment of a Safety Function [page 49]

This section provides guidance for reportability of a case when offsite power is unavailable. The conclusion of the guidance is that, if either offsite power or onsite emergency power is unavailable to the plant, it is reportable regardless of whether the other system is available ..

TVAN Comment -This guidance should be deleted. The Interpretation that loss of offstte power Is* a condition that, of itself, could prevent fulfillment of a safety function is an overly-conservative and narrow view of a safety function. The safety functions defined In the rule, (I.e., Shut down the reactor and maintain It In a safe shutdown condition, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident) can be fulfilled In the event of a loss of offsite power. Loss of offslte power Is governed by applicable Technical Specifications, and actual losses are reported by ESF actuation reporting (loss of offsite power results in reactor trip and actuation of Emergency Diesels).

This section also provides the following guidance regarding reportability of inoperability of single train systems, specifically High Pressure Core Injection (HPCI).

There are a limited number of single-train systems that perform safety functions (e.g., the High Pressure Coolant Injection System in BWRs). For such systems, loss of the single train would prevent the fulfillment of the safety function of that system and, therefore, is reportable even though the plant technical specifications may allow such a condition to exist for a limited time."

TVAN Comment - Requiring reporting of HPCI lnoperability for all cases Indicates a narrow view of the term safety function as defined in this section. Previously, safety function has been defined as the ability to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.

lnoperabllity of HPCI does not of itself constitute a condition that would prevent the fulfillment of a safety function. BWR design considers HPCI inoperablllty and provides alternate systems such as RCIC, Core Spray, and ADS. This Is supported by the relatively long Allowed Outage Time (Aon for HPCI In Standard Technical Specifications (I.e., 14 days). If, In the event of HPCI lnoperablllty, It can be shown that these systems are 14

available and capable of fulfilling the safety function (e.g., remove residual heat) without HPCI, the event should not be reportable. Reporting HPCI lnoperabillty in this case has no meaning for event reporting and appears to be solely a data gathering exercise. This purpose would be more-appropriately accomplished by other means (i.e., EPIX).

Additionally, the Revised Reactor Oversight Process as described In NEI 99-02 (Draft Rev.

B) prescribes a performance Indicator for Safety System Functional Failures based on 10 CFR 50.73 reports. These indicators count failures of single train systems (such as HPCI),

assuming that the event report documents a safety system failure. Reporting HPCI lnoperabillty when there is no Impact on the overall capability to fulfill the safety function

,(e.g., remove residual heat) will result In an overly conservative and detrimental assessment of this Indicator*

15

DOCKET NUMBER P OPOSED RU A. Edward Scherer Doc ~,_ 1 M~qager of I\ - Ntle'lear Regulatory Affairs

'JI I r

An EDISON INTERNATIONAL'" Company September 20, 1999 *99 SEF' 21 p 1 :57 The Secretary of the Commission, '- ,

U.S. Nuclear Regulatory Commission, An Washington, DC 20055-0001 Attention: Rulemaking and Adjudication Staff

Subject:

Proposed Rule - Reporting Requirements for Nuclear Power Reactors, 10 CFR Parts 50. 72 and 10 CFR Part 50. 73 (64 FR 36291 , dated July 6, 1999)

Gentlemen:

In the subject Federal Register Notice, the NRC solicited comments on proposed rulemaking regard ing changes to reporting requ irements for nuclear power reactors. This letter provides Southern California Edison's (SCE) comments on the subject proposed rule (enclosed). In general , SCE supports the proposed changes which will reduce or eliminate the reporting burden associated with events of little or no safety significance.

SCE also believes the NRC should fully apply the reporting burden reduction goal to short term NRC notifications - immediate notification (phone reports due within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) should be required only when activation of the NRC emergency response organization is actually required. Finally, SCE's enclosed comments supplement and support the comments submitted by the Nuclear Energy Institute.

If you have any questions, please feel free to contact me or Mr. Clay E. W illiams at (949) 368-6707.

Jlly,

Enclosure:

As Stated cc: D. P. Allison , Offi ce for Analysis and Evaluation of Operational Data J. A Sloan, NRC Senior Resident Inspector, San Onofre Units 2 and 3 L. Raghavan , NRC Project Manager, San Onofre Units 2 and 3 P. 0 . Box 128 San Clemente, CA 92674-0128 949-368-7501 Fax 949-368-7575

(99_ F

-/. - -

~~~~~ ~ :RIZ5s

ENCLOSURE Specific Comments on proposed Reporting Requirements for Nuclear Power Reactors As detailed below, this enclosure provides Southern California Edison's (SCE's) comments on the NRC's proposed rulemaking for Reporting Requirements for Nuclear Power Plants. Also provided, as requested in the Federal Register, are SCE's specific comments on NUREG-1022, Rev. 2, "Event Reporting Guidelines 10 CFR 50.72 and 50.73." These comments are consistent with the industry comments provided by the Nuclear Energy Institute. Where appropriate, comments have been keyed to the relevant section of the text in the FRN and are indicated by brackets ( ):

GENERAL;

1. Reporting of component level failures SCE strongly disagrees with the new requirements which would be added by 50.73(a)(2)(ii)(C) to report a component being in a degraded condition. This last minute addition does not meet any of the NRC's stated objectives to align reporting to needs, reduce the burden of reporting items of little safety significance, and clarity in reporting. Rather, it appears to be a data collection effort that does not relate to event reporting. This component level information, if proven necessary, should be addressed through other existing reporting systems.

The proposed change would add a new requirement to report if a component is in a degraded or non-conforming condition such that {a) the ability of the component to perform its specified safety function is significantly degraded; and

{b) the condition could reasonably be expected to apply to other similar components in the plant.

This newly p*roposed criterion was not included in the advance notice of proposed rulemaking (ANPR) published in July 1998. The criterion also was not included for discussion in the public meeting on August 21, 1998, the stakeholders meeting on September 1, 1998, nor the public meeting held on November 13, 1998. This attempt to capture items (components) which are degraded, but yet do not render a system inoperable, is a proposed addition that lacks clarity and is subject to widely varying levels of interpretation.

Some propose that this information is already being submitted under the current "outside the design bases of the planr (0D8) criterion, but this is not clear. As stated in the ANPR, one of the reasons for deleting the ODB criterion is the confusion and controversy over the meaning of the requirement Even with the deletion of the ODS criterion, conditions will continue to be reported if they result in a loss, or partial loss, of capability to perform a safety function. The degree of uncertainty and lack of clarity which wood exist in this newly proposed requirement, would greatly increase the number of unnecessary LERs thu~

frustrating the NRC's stated intent.

1

ENCLOSURE In addition, the newly proposed criterion is contrary to the stated objectives and diverges from the intent of the proposed rule change:

The rule change is intended to obtain reporting that is more consistent with the risk significance of the systems involved, yet the proposed section is focused on the component level as opposed to the stated objective of risk significance of the systems.

The rule change is intended to reduce reporting burden, yet many increases in burden can be identified:

This would significantly impact the engineering staff. Each component failure would need to be evaluated for "significance" of the "degradation" even though the system remained operable. Also, since this rule will likely require more LERs to be written the engineering personnel would be involved distracted from risk significant systems to conduct these evaluations. Finally, a greater number of evaluations and LERs wo_uld generate a greater amount of senior management distractions.

Due to the lack of clarity in the criterion, widely varying interpretations would require LERs to be written for non-safety significant events.

These include writing LERs for maintenance preventable functional failures (MPFFs), minor differences in equipment qualification (EQ) life span, and trivial differences in seismic qualification that does not render the system inoperable.

j

The rule change is intended to clarify the reporting requirements.

Unfortunately, as written the proposed criteria is not clear and is subject to varying interpretation.

While the stated objective of this criterion is to "allow" continued reporting, there are no current requirements to report component failures or degradations that do not impact operability. Previously established guidance in Generic Letter 91-18 and 91-18 Rev. 1 provide for determining the operability of structures, systems or components (SSCs). By using these guidelines, SSCs may be determined to be operable, but degraded. Because these SSCs categorizations do not affectsystem or train operability, no reporting in accordance with 10 CFR 50.72 or 50.73 has been previously required, except for conditions identified as being prohibited by Technical Specifications. The new criterion to report degraded SSCs that have generic implications represents a reporting threshold far below that previously required.

It is additionally important to note that the newly proposed criterion is somewhat redundant with existing reporting guidance. The discussion in NUREG 1022, pertaining to 10 CFR 50.73(a){2)(vii) states: "In addition if the cause or condition caused components in Train A of one system and in Train 8 of another system 2

ENCLOSURE (i.e., train that is assumed in the safety analysis to be independent) to become inoperable, the event must be reported." Significant component defects are also covered under 10 CFR Part 21 and additional discussion on componen, failures is also provided under conditions prohibited by Technical Specifications.

If component data is needed, it is currently available through other sources.

Efforts should be made to obtain data from the Equipment Performance Information Exchange (EPIX) and Maintenance Rule reports. Component failure data is also available to the industry via the nuclear network.

2. ESF system actuations There is a separate initiative to risk inform 10 CFR part 50. As part of that initiative, reporting of ESF actuations should be based on risk significance. This is Option 2 in the proposal. SCE does not believe that all the necessary criteria have been adequately established to make the shift as part of this rulemaking.

The optimum approach is to return to the longstanding practice that uses the licensees FSAR to define the ESF systems. As part of the future rule change to risk inform Part 50, this sections should be changed to include a risk informed list.

In the proposed amendments the term "any engineered safety feature (ESF),

including the reactor protection system (RPS)," which currently defines the systems for which actuation must be reported .in section 50.72(b)(2)0v) and section 50.73(a)(2)0v), would be replaced by a specific list of systems. In addition to this proposed list of systems, three principal alternatives to the proposed rule have been identified for comment. These include (1) maintaining the status quo, (2) requiring use of a plant-specific, risk-informed list, or (3) returning to the pre-1998 situation before publication of the event reporting guidelines in NUREG-1022, Revision 1.

As stated earlier, Option 3 best meets the near term goal of clarity and simplicity.

This returns to the pre-1998 situation whereby reporting would be required for the actuation of "any ESF" as is defined in each facilities FSAR. It would, however, be important to include examples of non-reportable exceptions in the implementing guidance for systems that are considered to be ESFs, yet have lower levels of risk significance (control room ventilation systems, reactor building ventilation systems, fuel building ventilation systems, auxiliary building ventilation systems, RWCU isolations during restoration from maintenance, etc.).

3. Reporting of historical events No safety significance exists for 10 CFR 50.72 reporting of historical events. SCE therefore believes that 10 CFR 50.72 should clearly indicate that this is for current conditions and not require reporting of historical events.

3

ENCLOSURE The proposed amendments would add provisions to sections 50.73(a)(2)(i)(B) and 50.73(a)(2)(v) to eliminate reporting of a condition or event that did not occur within three years of the date of discovery. SCE believes that two years is the appropriate time period for 10CFRS0. 73 reporting of historical events.

Although three years is consistent with the time period that performance indicators are tracked under the new oversight process, SCE continues to believe that no safety significance exists for 10 CFR 50. 73 reporting of historical events which occurred more than two years ago. Two years encompasses one refueling cycle of operation. Significant effort can be expended searching back in history for historical events. Reporting historical events more than two years old provides a small safety benefit and unnecessarily increases the reporting burden. This exclusion of such historical events and conditions should be extended to all written reports required by section 50.73(a) .

4. Required initial reporting times In the interest of simplicity, the proposed amendments should maintain just three basic levels of required reporting times in 10 CFR 50.72 and 10 CFR 50.73 (1 hour, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (or the end of the next business day), and 60 days).

SCE basically agrees with the revised reporting times based on importance to risk and extending the required reporting times consistent with the need for prompt NRC action. Nevertheless, SCE recommends that to improve accuracy and clarity without adding unnecessary delay, the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report should be required at the close of the next business day.

Additionally, the increased time for submittal of LERs will allow for completion of required engineering evaluations after event discovery, will provide for more complete and accurate LERs and result in fewer LER revisions and supplemental reports.

5. Invalid ESF Actuations In response to recommendations to stop reporting invalid ESF actuations, the proposal stated; "The comments are partially accepted. The proposed amendments would eliminate the requirement for telephone notification of an invalid actuation under 10 CFR 50.72. Invalid actuations are generally less significant than valid actuations because they do not involve plant conditions (e.g., low reactor coolant system pressure) conditions that would warrant system actuation. Instead, they result from other causes such as a dropped electrical lead during testing). However, the proposed amendments would not eliminate the requirement for a written report of an invalid actuation under 10 CFR 50.73."

SCE sees little safety benefit in reporting invalid ESF actuations and believes that the above response supports the contention that they are not significant. Invalid ESF actuations should not be reported.

4

ENCLOSURE Reporting only valid ESF actuations that address the response of the plant to actual challenges would accomplish the intended change as proposed in our response to the ANPR. Contrary to the NRC's expectations, reporting of invalid actuations will not provide the information needed to estimate equipment ~liability parameters. This information should be collected via other less burdensome mechanisms, such as Equipment Performance Information Exchange (EPIX) and Maintenance Rule reports.

SPECIFIC COMMENTS The following table provides specific comments and recommendations tied to the proposed rule or NUREG 1022 sections .

5

ENCLO Specific comments on 1 0 CFR 50. 72 SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS

50. 72{a)(1 )(i) The declaration of any of the Emergency 50.72{a)(1)(1) The declaration of any of the Emergency Classes Classes specified in the licensee's approved Emergency Plan;2 specified in the licensee's approved Emergency Plan;2 2

Those Emergency Classes are addressed in Appendix E of this 2 part. These EmeFgenoy Glasses are addressed in Append0e E of this part.

This section of the role is clear as written and does not require the footnote for clarity.

50. 72{a){4) The licensee shall activate the Emergency Response 50. 72{a)(4) The licensee shall activate the Emergency Data System (ERDS) 15 as soon as poss~le but not later than one Response Data System (ERDS) 15 as soon as possible but not hour after declaring an emergency clas of alert, site area later than one hour after declaring an emergency class of alert, emergency, or general emergency. Tho EROS may also be site area emergency, or general emergency. =t=he ERBS rnay activated by the licensee during emergency drills or exercises if also be !teti'v'ated by the licensee during emergency drills or tho licensee's computer system has the capability to transmit the exercises if the lieensee's computer system has the eep!tbiliey*

exercise data. to transmit the exercise data.

This last sentence should be deleted. It is superfluous information that adds no value to the discussion of reportability.

6

50.72{b)(2)(ii)(B) The nuclear power plant being In an unanalyzed condition that significantly affects plant safety.

ENCLOS*

50.72(b){2)(ii)(8} The nuclear power plant being tin an unanalyzed condition that significantly affem degrades plant safety.

The phrase "significantly affects plant safety" has no positive or

. negative connotation. It is therefore recommended that the term degrades be substituted in place of affects. In addition, a lower case editorial chanae should be made.

50. 72(b}(2)(v) Licensees shall report: "Any event or condition 50. 72(b l(2}(v} Licensees shall report "Any event or condition that at the time of discovery could have prevented the fulfillment that at the time of discovery is could ha*,e preveningted the of the safety function of structures or systems that are needed fulfillment of the safety function of structures or systems that to: ... are needed to: ...

Editorial change to reflect the correct tense of the existence of an event or condition, rather than past speculation.

7

ENCLOS Specific comments on 1 0 CFR 50. 73 SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS

50. 73(a){2){i)(B) Any operation or condition occurring within three 50. 73(a){2){i)CB) Any operation or condition occurring within years of the date of discovery which was prohibited by the plant's three years of the date of discovery which was prohibited by Technical Specifications, except when: the plant's current Technical Specifications, except when:

(i) The technical specification is administrative.in nature; or (i) The technical specification is administrative in nature; or (ii) The event consists solely of a case of a late surveillance test (ii) The event consists solely of a case of a late surveillance where the oversight is corrected, the test is performed, and the test where the oversight is corrected, the test is perfom,ed, and equipment is found to be capable of performing its specified the equipment is found to be capable of performing its safety functions. specified safety functio~.

The addition of "currenf' allows for plants that have recently converted to Improved Standard Technical Specfflcations to apply the current requirements to the identified condition rather than considering Technical Specifications which are no longer applicable.

"Safety function" should be singular to accommodate equipment with only one safety function.

50. 73(a){2){ii){B) The nuclear power plant being in (A) an 50.73(a){2)(11)(8) The nuclear power plant being in (At an unanalyzed condition that significantly affected plant safety unanalyzed condition that significantly affeeted degraded plant safety.

The phrase "significantly affected plant safety" has no positive or negative connotation. It is therefore recommended that the term degraded be substituted in place of affected. In addition, (A) should be deleted as a minor editorial change.

8

ENCLO 50.73{a)(2)(ii)(C} A component being in a degraded or non- SO. 73fa)f2)fii)(Gl A eemponent being in a degraded or non conforming condition such that the ability of the component to eonformir,g eonditiori st1eh that the ability of the eompor,er,t to perform its specified safety function is significantly degraded and perform it! !peeified safety ft1netion is signifieantly degraded the condition could reasonably be expected to affect other similar end the eondition eot1ld rea!onably be e'Xpeeted to affeet other components in the plant. similar eompor,ents iri the plant.

SCE believes this new criterion should be deleted.

50.73(a)(2)(v) Licensees shall report: "Any event or condition 50. 73(a)(2)(v) Licensees shall report: "Any event or condition occurring within three years of the date of discovery that could occurring within three years of the date of discovery that eottfti have prevented the fulfillment of the safety function of structures would have prevented the fulfillment of the safety function of or systems that are needed to: ... structures or systems that are needed to: ...

Editorial change to reflect the existence of an event or condition, rather than past speculation.

50. 73{a){2)(ix){J) For each human performance related problem 50.73(a){2){1x){J) For each personnel error ht1mar, that contributed to the event, the licensee shall discuss ... performar,ee related problem that eor,tributed to the event, the licensee shall discuss ...

The shift from "personnel error" and the implied "root cause" to "human performance related problem" and "contributing factors" greatly increases the scope of investigation and burden to the licensee without a commensurate safety benefit.

It is only appropriate to require discussion of personnel error root causes.

9

50. 73{a}(3}(ii) are included in emergency or operating procedures and could have been used to recover from the event in case of an additional failure in the systems actually used for ENCLO

50. 73(a}(3)(11} are included in emergency or operating procedure! and could have been u!ed to reeo*ver from the e*vent in C8!e of an additional f!lilure in the !ystems actually recovery. used for recovery.

SCE recommends that this new criterion be deleted.

Emergency operating procedures provide direction for use of many plant systems. If an additional failure must be postulated for every event, multiple systems would be required to be included in the LER for each safety function. For example, if the reactor scrammed due to personnel error and 'feedwater was used to recover from the event, an additional failure of loss of offsite power would require alternate injection methods that could include service water and fire water even though they are systems of last resort.

This rule change would result in a large amount of additional information that would be of minimal use. The assessment of the safety consequences and implications of the event would become cluttered with hypothetical additional failures and possible plant responses.

10

ENCLOSURE

Specific comments on NUREG 1022. Revision 2 (draft)

SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS NPRDS and SALP should be deleted since the~ are no long_er Abbreviations used and are not referenced in the NUREG.

Section 2.5 1 Time Limits for Reporting For example, if a technician sees a problem, but a delay For example, if a technician sees a problem, but a delay occurs occurs before an engineer or supervisor has a chance to before an engineer or supervisor has a chance to review the review the situation, the discovery date (which starts the 6oa&-

situation, the discovery date (which starts the 30-day clock} is the day clock) is the date that the technician sees a problem.

date that the technician sees a problem.

Change is needed to align written guidance with the proposed change to required LER submission times.

Section 2.5 1 Volunta!'.Y Reporting Instructions for voluntary ENS notifications and LERs are Instructions for voluntary ENS notifications and LERs are discussed in Sections 4.2.2 and 5.1.5 of this report.

discussed in Sections 4.2.2 and 5.1.5 this report.

Minor editorial change.

Section 3.2.1 1 Plant shutdown reguired bl£ TS If the shutdown is completed, licensees are required to submit If the shutdown is completed, licensees are required to submit an an LER within 6080 days.

LER within 30 days.

Change is needed to align written guidance with the proposed

- change to required LER submission times.

Section 3.2.1 1 Plant shutdown reguired bll TS 1 Example {1} An LER was not stibmitted reauired under this criterion since An LER was not submitted under this criterion since the failed the failed battery_ charoer was corrected before the giant battery charger was corrected before the plant completed comgleted shutdown.

shutdown. '

The proposed change illustrates clarity as to whether an LER was actually required to be submitted, not merely whether or not one was submitted in this particular case.

Section 3.2.1 i Plant shutdown reguired b~ TS. ExamQle {2} An LER was required stibmitted because of the completion of An LER was submitted because of the completion of the TS- the TS-required plant shutdown.

required plant shutdown.

The proposed change illustrates clarity as to whether an LER was actually required to be submitted, not merely whether or not one was submitted in this particular case.

11

It an LER reguired?

Section 3.2.1 1 Plant shutdown reguired b~ TS 1 Examgle {3)

ENCLOS I~ an LER required?

Answer:

Some jt1dgment is reqt1ired. An LER is not required. if-the Some judgment is required. An LER is not required if the sitt1etior, eot1ld he¥e beer, eorrected before the plant was situation could have been corrected before the plant was required to be sht1t down, er,d r,o other enterie ir, 50.73 epply.

required to be shut down, and no other criteria in 50. 73 apply. The shut down is reportable, however, if the situation could not The shut down is reportable, however, if the situation could not have been corrected before the plant was required to be shut have been corrected before the plant was required to be shut down, or if other eriterie of 50. 73 epply.

down, or if other criteria of 50. 73 apply.

Minor editorial change (It was replaced with Is). Deleted superfluous information alreadv stated in the question.

Section 3.2.1 1 Plant shutdown reguired b~ TS 1 Examgle {4) The plant lost one of two sources of offsite power due to The plant lost one of two sources of offsite power due to overheating in the main transformer. The TS allow 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to overheating in the main transformer. The TS allow 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the source or initiate a shutdown and be in HOT restore the source or initiate a shutdown and be HOT STANDBY ...

STANDBY. .. -

Minor editorial change.

Section 3.2.2, Technical Specification Prohibited Operation Note that there would not be a three-year limitation in this case or Condition, Safety Limit and LSSS Discussion because, in addition to the reguirements of §50.73(al(2l(il(B),

Note that there would not be a three year limitation in this case s12ecific re12orting rnguirements are stated in §50.36(c)(1}-snd because, in addition to the requirements of §50. 73(a)(2)(i)(B), the TS.

specific reporting requirements are stated in §50.36(c)(1) and the TS. Reference to TS should be deleted. The NRC approved removal of this reporting requirement from ISTS in TSTF-5.

Section 3.2.2 1 Technical S~cification Prohibited Ogeration It is recommended that this sentence be deleted since this or Condition 1 TS Surveillance Reguirements Discussion statement appears to contain no basis. The deficiency should Otherwise, the deficiency should be assumed to have* occurred be assumed to have occurred at the time of discovery (as is halfway between the last successful test or use and the untimely the current practice) unless firm evidence exists to the test that revealed the deficiency. contrary. At the very least, it sho,uld be assumed that the deficiency occurred at the time that the late surveillance was due.

12

ENCLOSURE Section 3.2.2, Technical Specification Prohibited Operation STS 3.0.3 {ISTS LCO 3.0.3), or its equivalent establishes or Condition, Entry into STS 3.0.3 Discussion requirements for actions when: (1) an LCO is not met and the STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent, establishes associated ACTIONS are not met, (2) an associated ACTION requirements for actions when an LCO is not met and no action is not provided, or (3) as directed by the associated ACTIONS statement is provided. Entry into STS 3.0.3 is considered to themselves. Entry into STS 3.0.3 (ISTS LCO 3.0.3) for either indicate that a condition existed longer than allowed by TS. of the first two above reasons are generally reportable under Thus, entry into STS 3.0.3 (ISTS LCO 3.0.3) for any reason or this criterion. However, when the plant TS specify the entry justification is reportable. into 3.0.3 as the required ACTION and that action and its completion time are met, the event is not reportable under this criterion. Also, momentary (less than approximately 5 minutes) entries into TS 3.0.3 regardless of the reason, are not reportable under this criterion. Any TS 3.0.3 entry involving actual plant shutdown should be reviewed' for reporting under 50.72(b)(1)(Q(A) and 50.73(a)(2)(i)(A) for a plant shutdown required by TS.

The proposed wording is suggested to replace the existing outdated guidance. The proposed wording reflects the current operating philosophy associated with impl~mentation of the standard Technical Specifications.

Section 3.2.2, Technical Specification Prohibited Operation Sections 50.55a(g) and 50.55a(f) require the implementation of or Condition, Missed Tests Required by ASME Section XJ ISi and 1ST programs in accordance with the applicable edition Discussion of the ASME Code for those pumps and valves whose function Sections 50.55a(g) and 50.55a(f) require the implementation of is required for safety. STS Section 4.0.5 (or an equivalent)

ISi and 1ST programs in accordance with the applicable edition of covers these testing requirements. (Generally, there is no the ASME Code for those pumps and valves whose function is comparable ISTS section.) A missed ...

required for safety. STS Section 4.0.5 (or an equivalent) covers these testing requirements. (Generally, there is no comparable This statement should be deleted since ISTS Section 5, ISTS section.) A missed ... "Programs and Manuals," has a section for these requirements in "lnservice Testina Proaram".

13

ENCLO Section 3.2.2, Technical Specification Prohibited Operation Also, if a fire protection deficiency results in the inability to or Condition, Fire Protection Systems When Required by TS preserve at least one safe shutdown train in the event of a fire, Discussion it should be reported evaluated for reporting as an unanalyzed Also, if a fire protection deficiency results in the inability to condition that significantly affects plant safety, as discussed in preserve a safe shutdown train in the event of a fire, it should be Section 3.2.4 of this report.

reported as an unanalyzed condition that significantly affects plant safety, as discussed in Section 3.2.4 of this report. The guidance should be modified to indicate the need to evaluate for reporting, rather than indicate that reporting is definitely required since other trains or systems may be available to preserve sa'fe shutdown.

Section 3.2.2, Example (3) Entering STS 3.0.3 (3) Entering STS 3.0.3 due to lack of specific TS actions

.(fil Entering STS 3.0.3 With essential water chillers (A) and (B) out of service, the only With essential water chillers (A) and (B) out of service, the only remaining operable chiller (A/8) tripped. This condition caused remaining operable chiller (A/8) tripped. This condition caused the plant to enter STS 3.0.3 (equivalent to ISTS LCO 3.0.3) for the plant to enter STS 3.0.3 (equivalent to ISTS LCO 3.0.3) for 1 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , until chiller (A) was restored to service and the hour, until chiller (A) was restored to service and the temperature temperature was restored to within TS limits.

was restored to within TS limits.

An LER is required for this event because, STS 3.0.3 was An LER is required for this event because STS 3.0.3 was entered in this case, there were no actions provided in the entered plant TS for that condition and STS 3.0.3 was entered for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The proposed wording reflects the current operating philosophy associated with implementation of the Standard Technical Specifications. An additional example is also included to clarify this position:

Entry into STS 3.0.3 when the plant TS specify 3.0.3 entry During a surveillance test on the A train of a two-train Standby Gas Treatment (SBGT) system, a condition was discovered on the B train that rendered it inoperable. The test was halted and steps taken to return the A train to a standby readiness condition. During the restoration, switch manipulations momentarily rendered the A train inoperable. Wrth both trains inoperable, the plant TS specify immediate entry into LCO 3.0.3. The entry into LCO 3.0.3 was logged and then exited 14

ENCLOS within 1 minute once switch manipulation on the A train was completed.

This event is not reportable under this criterion because all the actions specified by the plant TS were completed within the required completion times. There was no operation or condition prohibited by TS. Also, momentary entries into STS 3.0.3 are not reportable.

Section 3.2.2. Example (5) Seismic Restraints Assume that during afl NRG evaluation ilt is found that an Assume that during an NRC evaluation it is found that an exciter exciter panel for one EDG had lacked appropriate seismic panel for one EDG had lacked appropriate seismic restraints restraints since the plant was constructed, because of a since the plant was constructed, because of a design, analysis or design, analysis or construction inadeq*uacy. Also assume that, construction inadequacy. Also a~sume that, upon evaluation, tt Upon evaluation, the EDG is determined to be inoperable there is reasonable doubt about the EDG's ability to perform its there is reasonable doubt about the EDG's ability to perform its specified safety functions during and after an SSE. specified safety functions during !Ind !lfter an SSE.

The recommended wording deletes superfluous information and equates the ability to perform a specified safety function to operabilitv.

Section 3.2.2. Example (6) Vulnerability to Loss of Offsite Assume th!lt dDuring a design review it is found that a loss of Power offsite power could cause a loss of instrument air and, as a Assume that during a design review it is found that a loss of result, auxiliary feedwater (AFW) flow control valves could fail offsite power could cause a loss of instrument air and, as a open. Then for low steam generator pressure, such as could result, auxiliary feedwater (AFW) flow control valves could fail occur for certain main steam line breaks, high AFW flow rates open. Then for low steam generator pressure, such as could could result in tripping the motor driven AFW pumps on occur for certain main steam line breaks, high AFW flow rates thermal overload and were therefore declared inoperable. The could result in tripping the motor driven AFW pumP.s on thermal single turbine driven AFW pump would not be affected.

overload. The single turbine driven AFW pump would not be affected. The recommended wording deletes superfluous information clearly indicates that the equipment is determined to be inoperable (for a time greater than allowed by Technical Specifications).

15

Section 3.2.3, TS Deviation per §50.54{xl33.2.3 Technical Specification Deviation per 50.54(xl Example ENCLOS When the action was completed the control room operators notified the NRC Operations Center, in accordance with the

When the action was completed the control room operators reporting requirements of 10 CFR 50. 72, that they had notified the NRG Operations Center, in accordance with the exercised 10 CFR 50.54(x). Subsequently, an LER was reporting requirements of 10 CFR 50.72, that they had exercised required to be submitted in accordance with 10 CFR 10 CFR 50.54(x). Subsequently, an LER was submitted in 50. 73( a){2)(i) ...

accordance with 10 CFR 50.73(a}(2){i) ...

Section 3.2.4, Degraded Condition In the case of a component with a significantly degraded ability In the case of a component with a significantly degraded ability to to perform a safety function, where the condition could perform a safety function, where the condition could affect/other affectfother similar components in the plant, licensees are similar components in the plant, licensees are required to submit required to submit an LER within 60 days.

an LER within 60 days.

Minor editorial cha e extraneous "I .

Section 3.2.4, Degraded Condition. CB) Unanalyzed condition Discussion for unanalyzed condition that significantly affects that slgnificantly affects plant safety Discussion plant safety should include a memorandum from William T.

Russell to Thomas T. Martin dated April 1, 1991, that addresses the reportability of ASME Code Class Ill defects under this criterion. This memorandum was in response to a request for clarification from Public Service Electric and Gas on re ortin uirements ardin service water i i leaks.

Section 3.2.4. Degraded Condition, Example (1) Following the end of cycle shutdown, ...

Following the end of cycle shutdown, ... The event is reportable because the cladding failures exceed The event is reportable because the cladding failures exceed expected values and, are unique or widespread, and are expected values, are unique or widespread, and are caused by caused by unexpee-ted me-tors.

unexpected factors.

33.2.3 Technical Specification Deviation per 50.54(x}

Minor editorial change (capitalization). Whether the factors are une ected or not are not relevant to re ortt .

Section 3.2.4, Degraded Condition, Example {2) The event is reportable beeeuse the degradation Cl!lnnot be The event is reportable because the degradation cannot be considered aeeeptable e:!-isof the generic implications.

considered acceptable as-is.:.

Many conditions identified by the licensee are not acceptable as-is. It is therefore more appropriate to indicate that the event is re ortable due to the eneric im lications.

16

ENCLOSU Section 3.2.4, Degraded Condition, Example (3) Consequently, after discussion with the Regional Office, the Consequently, after discussion with the Regional Office, the licensee appropriately retracted this event licensee retracted this event.

Illustrates clarity as to the acceptability of the licensee's actions.

Section *3.2.5, External Threat to Plant Safety, Examples (2) It Is proposed that these examples be deleted. The licensee and (3) was required to make a report due to entry into the Emergency (2) Hurricane Plan making reporting under this category unnecessary.

A licensee in southern Florida declared an Unusual Event after a hurricane warning was issued by the ...

(3) Fire Wrth the unit at 100-percent power, the control room was notified that a forest fire was burning ....

Section 3.2.7, Event or Condition That Could Prevent . .. regardless of whether the system was needed at the time of Fulfillment of a Safety Function, Discussion discovery.

. . . regardless of whether the system was needed at the time.

Addition is needed to be consistent with the narrative in the next paragraph.

Section 3.2.7, Event or Condition That Could Prevent This portion of the discussion should be deleted. Operability Fulfillment of a Safety Function, Discussion determinations are performed by the licensee using Generic The staff believes that the conditions necessary to consider the letters 91-18 and 91-18 Rev. 1. It is inappropriate to include redundant train operable and available, for this purpose, should operability determination guidance in this reportability include the following: guideline.

in cases where the redundant train should operate automatically, it is capable of timely and correct automatic operation, or in cases where the redundant train should be operated manually, the operators would detect the need for its operation and initiate such operation, using1 established procedures for which they are trained, within the needed time frame, without the need for troubleshooting and repair, and; the redundant train is capable of performing its safety function for the duration required, and; there is not a reasonable expectation of preventing fulfillment of the safety function by the redundant train 17

Section 3.2.7 1 Event or Condition That Could Prevent Fulfillment of a Safetv Functlon 1 ExamRle {2}

If the plant's safety analysis considered RCIC as a system ENCLOSU No, a RCIC failure is not reportable. RCIC is included in ISTS because it meets criterion 4 of 10CFR50.36 based on its contribution to the reduction of overall plant risk. RCIC is not needed to mitigate a rod ejection accident (e.g., it is included in credited in the plant's safety analysis. If the pletrit1s safety the Technical Specifications) then its failure is reportable under srietlysis considered RSIS ets e system Meeded to mitigate a this criterion; otherwise, it is not reportable under this section of 1'6d eje~or, eeeider,t te*.g., it is ir,eltided ir, the feet,r,ieal the rule. Speeifieatior,s} then i~ fsilure is reportable ur,eJer this eriterior,:

otheNtise, it is not reportable ur,der this section of the rule.

This clarification will correct problems associated with RCIC being included in TS, but not credited in a plants safety analysis.

Section 3.2.7 1 Event or Condition That Could Prevent The independent failure (e.g., excessive set point drift) of a Fulfillment of a Safetv Function 1 Exam~le (13} single pressure switch is not reportable unless it 8lene eottld The independent failure (e.g., excessive set point drift) of a single would have caused a system to fail to fulfill its safety function, pressure switch is not reportable unless it alone could have or is Indicative of a generic problem that could have resulted in caused a system to fail to fulfill its safety function, or is indicative the failure of more than one switch and thereby cause one or of a generic problem that could have resulted in the failure of more systems to fail to fulfill their safety function.

more than one switch and thereby cause one or more systems to fail to fulfill their safety function. Wording change is required to coincide with the proposed wording change of the rule (i.e. deleting "alone" and replacing "could" with "woukij.

Section 3.2. 7 1 Event or Condition That Could Prevent ... event or condition that etlone could prevent fulfillment of a Fulfillment of a Safetv Function 1 ExamQle (17} safety function). (Fhis reporting eriterion is discussed in

... event or condition that alor,e could prevent fulfillment of a Seetior, 3.3.3 of this report.}

safety function). (This reporting criterion is discussed in Section 3.3.3 of this report) This line which refers to section 3.3.3 should be deleted. This section no longer exists in NUREG 1022.

Section 3.2.7 1 Event or Condition That Could Prevent Examples (17), (18), (19) and (20) should be renumbered as Fulfillment of a Safetv Function 1 ExamQles {17} 1 (18} 1 {19} and (18), (19), (20) and (21) due to the duplication of example (17).

{20) 18

Offsite, Examgle (1}

Section 3.2.11, Contaminated Person Reguiring Transgort A contract worker experienced a back injury lifting a tool while ENCLOSURE A contract worker experienced a back injury lifting a tool while working in a contaminated area the reactor eontf!linment and was considered potentially contaminated because his back working in the reactor containment and was considered could not be surveyed.

potentially contaminated because his back could not be surveyed. The example should be changed such that the employee was working in a contaminated area. In the given example it is possible that the emplo-yee was working in a radiologically controlled area that was not contaminated. It ,could therefore be assumed that the worker was not contaminated, eliminating the need to make a report.

4.2.4 ENS Event Notification Worksheet (NRC Fonn 361} The ENS Event Notification Worksheet (NRC Form 361) is-8n The ENS Event Notification Worksheet (NRC Form 361) is an atteehmer,t to lnformetior, Netiee 89 89, deteel Beeember 26, attachment to Information Notice 89-89, dated December 26, 1989, subject E*,er,t Notification V'lorksheets. The *,,ork:sheet 1989, subject Event Notification Worksheets. The worksheet provides the usual order of questions and discussion for easier provides the usual order of questions and discussion for easier communication and its use ... 34.2.4 ENS Event Notification communication and its use ... 34.2.4 ENS Event Notification Worksheet (NRC Form 361)

Worksheet (NRC Form 361)

Unless IN 89-89 is revised to incorporate a new NRG form 361, all reference to IN 89-89 should be deleted because the NRG Form 361 will be revised to reflect the changes to 10GFR50. 72 in this rulemaking.

5.1.4 Volunta~ LERs35.1.4 Volunta~ LERs See Section r.9 2. 7 for additional discussion of voluntary See Section 2.9 for additional discussion of voluntary LERs. LERs.

Minor editorial change due to renumbering of sections in the NUREG.

19

ENCLOSU.

5.2.1(1). Narrative Description or Text (NRC Form 366A, Item For equipment that was inoperable at the start of the event, 17)35.2.1 Narrative Description or Text {NRC Form 366A, provide an estimate of the time the equipment became Item 17) inoperable and the last time the equipment was known to be For equipment that was inoperable at the start of the event, operable. Indicate the basis for this conclusion {e.g., a test provide an estimate of the time the equipment became was successfully run or the equipment was operating). For inoperable and the last time the equipment was known to be equipment that failed, provide the failure time and the last time operable. Indicate the basis for this conclusion {e.g., a test was the equipment was known to be operable. Also provide the successfully run or the equipment was operating). For basis for the last time known operable.

equipment that failed, provide the failure time and the last time the equipment was known to be operable. Also provide the basis Deletion of the last time equipment was known to be operable for the last time known operable. is necessary to confonn with ISTS philosophy of equipment being operable since the last perfonnance of the surveillance test (unless finn evidence exists to the contrary.

20

DOC,":T 0 5"() .:f 7,:2..

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September 20, 1999 A.._l GDP 99-0163 Secretary, U.S. uclear Regulatory Commission Washington, D.C. 20555-0001 ATTN: Rulemakings and Adjudications Staff

Paducah Gaseous Diffusion Plant (PGDP)

Portsmouth Gaseous Diffusion Plant (PORTS)

Docket Nos. 70-7001 & 70-7002 USEC Comments on NRC's Proposed Rule, "Reporting Requirements for Nuclear Power Reactors," (64 FR 36291)

Dear Madam:

The United States Enrichment Corporation (USEC) is pleased to submit the following comments on the NRC's Proposed Rule, "Reporting Requirements for Nuclear Power Reactors."

USEC agrees with the Commission and the industry that the change to the reporting requirements of 10 CFR 50.73 allowing 60 days to submit a follow-up written event report is appropriate. However, USEC believes that the Commission' s decision to currently limit the applicability of this change to Part 50 licensees is too narrowly focused. USEC understands that the NRC is considering the feasibility of expanding the applicability of this change to other categories of licensees in the future. Written reports required by 10 CFR 76.120 are similar in nature to Licensee Event Reports required by Part 50.73, in that they also require a root cause analysis and corrective action plan development. USEC, therefore, requests that 10 CFR 76.120(d)(2) be similarly amended to allow the submittal of written event reports within 60 days of discovery in order to allow time to complete root cause analyses and corrective action plans.

USEC agrees with the Commission that this longer time allowance would reduce the number of event reports that require revision to update information resulting in resource savings with little or no reduction in safety. In keeping with the spirit of the Commission' s proposed change, USEC would continue to make every effort to complete and submit reports on a time scale commensurate with the significance of the issue.

19 vcara ...............-..... . ,

6903 Rockledge Drive, Bethesda, MD 208 17- 1818 Telephone 301-564-3200 Fax 301-564-3201 http://www.usec.com Offices in Livermore, CA Paducah, KY Porcsmouch, OH Washington, DC

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Secretary, U. S. Nuclear Regulatory Commission September 20, 1999 GDP 99-0163, Page 2 Thank you for the opportunity to provide our comments. We would be pleased to discuss these comments with you. Please contact Ms. Lisamarie Jarriel at USEC HQ (301) 564-3247, Mr. Mike Boren at PGDP (270) 441-6777, or Mr. Scott Scholl at PORTS (740) 897-2373.

Sincerely, tu..1:-i Q.. .;__Q -R.t Steven A. Toelle Nuclear Regulatory Assurance and Policy Manager cc: Robert Pierson, NRC HQ Patrick Hiland, NRC Region III Office Kenneth O'Brien, NRC Resident Inspector - PGDP David Hartland, NRC Resident Inspector- PORTS

Lewis Sumner Southern Nuclear Vice President Operating Company, Inc.

Hatch Project Support 40 Inverness Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 Fax 205.992.0341 SOUTHERN << \

COMPANY September 20, 1999 Energy to Serve Your World "'

Ms. Annette L. Vietti-Cook, Secretary U. S. Nuclear Regulatory Commission Attn: Rulemakings and Adjudications Staff Washington, D. C. 20555 5~J7r2_

( /pt/~R3'1~91 Comments on Proposed Rule "Reporting Requirements for Nuclear Power Reactors" (64 Federal Register 36291 dated July 6, 1999)

Dear Ms. Vietti-Cook:

Southern Nuclear Operating Company (Southern Nuclear), the licensed operator for the Joseph M. Farley Nuclear Plant, the Edwin I. Hatch Nuclear Plant and the Vogtle Electric Generating Plant, has reviewed the supplemental proposed rule, "Reporting Requirements for Nuclear Power Reactors," published in the Federal Register on July 6, 1999. In accordance with this request, Southern Nuclear is in total agreement with the comments that are to be provided to the NRC by the Nuclear Energy Institute.

In addition, Southern Nuclear would like to reiterate that the last minute addition of the requirement to report significantly degraded components, section 50.73(a)(2)(ii)(C),

does not meet the stated objectives of the rule and should be deleted. Also, we do not support the proposed revisions which specify the ESF systems for which reporting is required.

Respectfully submitted, H. L. Sumner HLS/JDB cc: Southern Nuclear Operating Company Mr. D. N. Morey, Vice President - Farley Mr. J. B. Beasley, Ir. Vice President - Vogtle REES File: G.02.11

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.. 2 0 1999 Entergy Nuclear Generation Company Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 A:=AND/- . September 17 , 1999 89l1HtiO~\/ Ur. 2.99.098 Secretary Docket No. 50-293 U.S. Nuclear Regulatory Commission License No. DPR-35 Washington, DC 20555

Subject:

Comments on Proposed Reporting Rule (64 FR 36293, July 6, 1999)

Dear Sir:

The following Pilgrim Nuclear Power Station comments are submitted in response to Federal Register notices regarding the proposed rulemaking on certain reporting requirements for nuclear power reactors (64 FR 36293).

The proposed changes to 10 CFR Parts 50.72 and 50.73 are considered , in general, to meet the objectives to better align the reporting requirements with NRC needs, reduce the reporting burden where there is little or no safety significance, and to improve the clarity of the reporting requirements.

Pilgrim Station regulatory staff experienced in the reporting requirements and reporting participated in workshops and table top exercises held by the Nuclear Energy Institute (NEI) and NRC . The NRC workshops and exercises proved to be insightful and useful as part of the rulemaking process and significantly improved the context of and resolution of the proposed reporting requirements.

Pilgrim Station endorses the comments submitted by NEI on behalf of the industry.

There is, however, one specific area of concern and one other comment regarding the proposed requ irements for wh ich Pilgrim Station offers the followi ng comments.

The proposed (new) reporting requirement, 10 CFR 50.73(a)(2)(i i)(C), involves a "significantly degraded" component. This proposed, new reporting requirement seems to be contrary to the objectives of the proposed rule because it imposes an unnecessary additional reporting burden. This proposed requirement appears to stem from a NRC staff perceived need for data collection. The example(s) provided in the proposed rule and discussed at the NRC workshop on August 3, 1999, indicated NRC staff concerns with evaluations performed in accordance with Generic Letter 91-18,

"... Resolution of Degraded or Nonconforming Conditions." The concern was offered by the NRC mechanical engineering section with little discussion and no specificity relative to the veracity and merit of the concerns. Presumably, and if such concerns 299098

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U.S. NUCLEAR REG OFFICE OF

U.S. Nuclear Regulatory Commission Ur. 2.99.098 Page 2 were merited, the concern would properly be the subject of the NRC enforcement process. The perceived need for data collection should not be a reason for a new reporting requirement. The addition of the proposed (new) reporting requirement, in lieu of the enforcement process, appears to be mis-directed.

The proposed (new) reporting requirement could result in the requirement to report a problem (condition) with a component for which an evaluation concludes a component, although degraded, is operable (capable of providing its required function) . The distinction between "degraded" and "significantly degraded" is vague and could be subject to varying interpretations. Thus, the proposed (new) reporting requirement, in itself, could or would likely result in reports that are not now required . One of the reasons the NRC staff proposed the deletion of the requirement to report "conditions outside the design basis" was because of the confusion and controversy over the meaning of the reporting requirement. The confusion resulted in the consequent burden by licensees to report and for the NRC to review and process the reports.

Reporting is already required for problems that adversely affect the operability of one or more channels or trains of a system(s). To proscribe the reporting at the component level, when reporting is already required for problems (conditions) that affect the operability of one or more channels or trains of a system(s), is considered to be an unnecessary expansion of the reporting burden, and contrary to at least one of the reasons for the proposed rule.

Finally, the reporting of component failures is part of the existing reporting process (EPIX). No problems were identified with the EPIX reporting process.

The following comment involves the listing of engineered safety feature (ESF) systems.

Of the options contained in the proposed rule, Pilgrim Station supports Option 1 (maintaining the existing reporting requirements) rather than Option 2 (risk-informed listing) or Option 3 (pre-1998 listing) for the following reasons. The Pilgrim Updated Final Safety Analysis Report (UFSAR) contains a listing of engineered safeguards (i.e.,

ESF systems). The existing reporting requirements and applicable UFSAR systems have been incorporated into the Pilgrim Station reporting process. The previous actions taken and experience gained resulted in a significant reduction in the number of reports for an unplanned ESF actuation. Meanwhile, the NRC and industry pilot plants are currently implementing a pilot program for the new regulatory oversight process to be implemented in April 2000. The process utilizes risk-significance. The Option 2 listing would require consequent changes to the Pilgrim Station reporting process. The risk-significant listing could change in the future because of changes to regulatory requirements other than those for reporting. The Option 3 listing would require some changes to the Pilgrim Station reporting process but the benefit of the changes is considered to be minimal to that of the existing reporting process. Therefore, Option 1 is supported by Pilgrim Station.

299098

U.S. Nuclear Regulatory Commission Ur. 2.99.098 Page2 Pilgrim Station considers the opportunities provided by the NRC in the workshops to have been very useful for this proposed rule-making.

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299098

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-.i Florida Power CORPORATION Co*W ""'"'"""

DOCKET NUMBER PROPOSED RULE , ff 5 o d-1 :z,.

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- - - Operating License No. DPR-72 September 14, 1 999 AQ._1c 3F0999-05 The Secretary of the Commission Attention: Rulemakings and Adjudication Staff U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

Comments on Proposed Rule 10 CFR Parts 50 and 72, "Reporting Requirements for Nuclear Power Reactors," and Draft NUREG-1022, Rev. 2, "Event Reporting Guidelines - 10 CFR 50.72 and 50.73"

Reference:

Federal Register Volume 64, Number 128, Pages 36291 through 36307, dated July 6, 1 999

Dear Secretary:

This letter provides Florida Power Corporation ' s (FPC's) comments in response to the Federal Register notice concerning proposed rulemaking on Reporting Requirements for Nuclear Power Reactors (64 Federal Register 36291 , dated July 6, 1999). FPC appreciates the opportunity to comment on the proposed rule. FPC commends the NRC's initiative to revise the current reporting requ irements of 10CFR50. 72/ 50. 73 and the guidance contained in NUREG-1022. FPC also supports the NRC's effort to: reduce or eliminate the reporting burden associated with events of little or no safety significance; revise reporting requirements based on importance to risk; and, extend the required reporting times.

In general, FPC believes that the proposed change meets the stated objectives to align reporting w ith needs, reduce burden where there is no safety significance and provide clarity to reporting. Extensive use of workshops and tabletop exercises during the rulemaking process have provided a valuable testing ground for proposed revisions.

However, FPC has two areas of concern:

1. The last minute addition of a requ irement to report significantly degraded components, 10CFR50. 73(a)(2)(ii)(C), does not meet the stated objectives of the rule change and should be deleted. There are no current requirements to report component failures or degradations that do not impact system train/channel operability. This new criterion represents a reporting threshold f ar below t hat currently required. The requirement to report component degradations t hat are not safety significant (the component is operable and does not render a system train/channel inoperable) is a data collection effort that runs counter to the intent of the event reporting rule.

CRYSTAL RIVER ENERGY COMPLEX: 15760 W. Power Line Street

Crystal River, Florida 34428-6708 * (352) 795-6486 A Florida Progress Company eknnwfP.d ed bv eafff ...SEP ____tt Qft"-_,

U.S. NUa.EAR REGULATORY COM S I .~

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U.S. Nuclear Regulatory Commission 3F0999-05 Page 2 of 6 The attempt to capture components which are seriously degraded, but not enough so as to render a system train/channel inoperable, is subject to widely varying levels of interpretation. Such a great degree of uncertainty and lack of clarity exists in this newly proposed requirement, that th e potential exists to increase the number of required Licensee Event Reports (LERs) far beyond those that were submitted due to being "outside the design bases." One of the reasons that the "Outside the Design Basis" criterion is proposed to be deleted is the confusion and controversy over the meaning of the requirement.

2. Second, the proposed rule contains a detailed li st of Engineered Safety Feature (ESF) systems for reporting. To support the time frame of this rulemaking initiative, FPC recommends each facility define ESF systems based on their current Final Safety Analysis Report (FSAR). This is Option 3 in the rulemaking package. Ultimately, as part of the initiative to risk-inform 10 CFR Part 50, a plant-specific, a risk-informed list should be used. Th is would include only those systems that were significant to safety. Additional effort and discussion in that area is needed. In the short-term, the NRC should continue with the longstanding practice of relying on each facility's FSAR, shifting to a risk-informed approach when criteria are fully developed.

In addition to the above and attached comments, FPC has reviewed and endorses the comments contained in a proposed draft Nuclear Energy Institute (NEil Event Reporting Task Force (ERTF) response to the Federal Register notice (64 Federal Register 36291, dated July 6, 1999). No comments are being made against the proposed changes to 10CFR72.

This letter contains no new commitments. FPC appreciates the opportunity to comment on these important NRC proposed rules and would be w illing to meet with you or your staff to discuss these comments further.

Sincerely,

-f.;~~~~

Director, Nuclear Regulatory Affairs SLB/dwh Attachment xc: NRC Document Control Desk

U.S . Nuclear Regulatory Commission 3F0999-05 Page 3 of 6 ATTACHMENT COMMENTS ON PROPOSED CHANGES These comments are in response to the Federal Register notice concerning proposed rulemaking on Reporting Requirements for Nuclear Power Reactors {64 Federal Register 36291, dated July 6, 1999).

Reporting of Component Level Failures FPC supports the NEI position that strongly objects to the new requirements added by 10CFR50. 73{a){2){ii){C) to report a component being in a degraded condition. This last minute addition does not meet any of the NRC's stated objectives to align reporting to needs, reduce the burden of reporting items of little safety sign ificance, and clarity in reporting. It is clearly a data collection effort that does not relate to event reporting. The need for this component level information should be addressed through other, existing ,

material reporting systems.

The proposed change adds a new requirement to report a component degradation or non-conforming condition, if: (a) the ability of the component to perform its specified safety function is significantly degraded; and {b) the condition could reasonably be expected to apply to other similar components in the plant. This newly proposed criterion was neither included in the advance notice of proposed rulemaking {AN PR) published in July 1998 nor included for discussion in the public meeting on November 13, 1998. This attempt to capture components which are degraded, but do not render a system inoperable, is a proposed addition that lacks clarity and is subject to varying levels of interpretation.

Some propose that this information is already being submitted under the current "outside the design bases" {008) criterion, but this is not so. As stated in the ANPR, one of the reasons for deleting the 008 criterion is the confusion and controversy over the meaning of the requirement. Even with the deletion of the 008 criterion, conditions will continue to be reported if they result in a loss or partial loss of capability to perform a safety function .

Such a great degree of uncertainty and lack of clarity exists in this newly proposed requirement, that the potential exists to greatly increase the number of required LERs .

In addition, the newly proposed criterion is contrary to the stated objectives and diverges from the intent of the proposed rule change:

The rule change is intended to obtain reporting that is more consistent with the risk significance of the systems involved, yet the proposed section is focused on the component level as opposed to the stated objective of risk significance of the systems.

The rule change is intended to reduce reporting burden , yet many increases in burden can be identified:

U.S. Nuclear Regulatory Commission 3F0999-05 Page 4 of 6 This would impact the Control Room Operators. Since the Nuclear Shift Manager or Shift Technical Advisor typically reviews all corrective action documents for reportability, every component failure would have to be scrutinized in much more detail.

This would impact the Engineering Staff. Each component failure would need to be evaluated for "significance" of the "degradation" even though the system remained operable. Also, since this rule will likely require more LERs to be written, the engineering personnel would be involved in a greater number of root cause evaluations. In addition, an increased number of root cause evaluations and LERs would require increased senior management review.

Due to the lack of clarity in the criterion, varying interpretations would require LERs to be written for non-safety significant events. These include writing LERs for maintenance preventable functional failures (MPFFs), minor differences in equipment qualification (EQ) life span, and trivial differences in seismic qualification that do not render the system inoperable.

The rule change is intended to clarify the reporting requirements, yet as written the proposed criteria is not clear and is subject to varying interpretation. Use of the terms "significantly degraded," "could reasonable be expected," and "similar components" contribute to this lack of clarity.

The stated "objective" of the criterion is to "allow" continued reporting of design basis or other discrepancies on a (component level), yet current criterion do not require this. There are no current requ irements to report component failures or degradations that do not impact operability. Previously established guidance in Generic Letter 91 - 18 and Generic Letter 91 - 18, Revision 1, provide for determining the operability of structures, systems or components (SSCs). By using these guidelines, SSCs may be determined to be operable,

but not fully qualified (degraded). Because these SSCs categorizations do not affect system or train operability, no reporting in accordance with 10CFR50. 72/50. 73 has been previously required, except for conditions identified as being prohibited by Technical Specifications . The new criterion to report degraded SSCs that have generic implications represents a reporting threshold far below that previously required.

It is important to note that the newly proposed criterion is somewhat redundant w ith existing reporting guidance. The discussion in NU REG 1022, Revision 1, pertaining to 10CFR50. 73(a)(2)(vii), states: "In addition, if the cause or condition caused components in Train A of one system and in Train B of another system (i.e., train that is assumed in the safety analysis to be independent) to become inoperable, the event must be reported. "

Significant component defects are also covered under 1 OCFR21 and additional discussion on component failures is also provided under conditions prohibited by Technical Specifications.

U.S. Nuclear Regulatory Commission

. 3F0999-05 Page 5 of 6 If component data is needed, it is currently available through other sources. Efforts should be made to obtain data from the Equipment Performance Information Exchange (EPIX) and Maintenance Rule reports. Component failure data is also available to the industry via the Nuclear Network. The licensee's use of industry information is routinely inspected by the NRC. In addition, the Resident Inspector Program closely monitors operational status and communicates it to Regional and NRR personnel. This information is reviewed for generic implications.

Engineered Safety Feature (ESF) System Actuations There is a separate initiative to risk inform 10CFR50. As part of that initiative, reporting of ESF actuations should be based on risk significance. This is Option 2 in the proposed rule.

FPC supports the NEI position that the necessary criteria have not been adequately established to make the shift as part of this rulemaking. The optimum approach is to return to the longstanding practice that uses the licensees FSAR to define the ESF systems. As part of the rule change to risk inform 10CFR50, thi s section should be changed to include a risk informed list.

In the proposed rule, the term "any engineered safety feature (ESF), including the reactor protection system (RPS)," which currently defines the systems for which actuation must be reported in 10CFR50. 72(b)(2)(iv) and 10CFR50. 73(a)(2)(iv), would be replaced by a specific list of systems. In addition to this proposed list of systems, three principal alternatives to the proposed rule have been identified for comment. These include: (Option

1) maintaining the status quo; (Option 2) requirin g use of a plant-specific, risk-informed list; or, (Option 3) returning to the pre-1998 situation before publication of the event reporting guidelines in NUREG-1022, Revision 1.

FPC supports the NEI position that Option 3 best meets the near term goal of clarity and simplicity. This returns to the pre-1998 situation whereby reporting would be required for the actuation of "any ESF" as is defined in each facility's FSAR. It would however be important to include examples of non-reportable exceptions in the implementing guidance for systems that are considered to be ESFs, yet have lower levels of risk significance (control room ventilation systems, reactor building ventilation systems, fuel building ventilation systems, auxiliary bu ilding ventilation systems, etc.).

Reporting of Historical Events FPC supports the NEI position that two years is the appropriate time period for reporti ng historical events. Although t hree years is consistent with the time period that performance indicators are tracked under the new oversight process, FPC asserts that no safety significance exists for 10CFR50. 73 reporting of historical events which occurred more than two years ago. Two years encompasses one refueling cycle of operation. Significant effort can be expended searching back in history for historical events. Reporting historical events more than two years old provides a low safety benefit and unnecessarily increases the reporting burden.

U.S. Nuclear Regulatory Commission 3F0999-05 Page 6 of 6 The proposed amendments would add provisions to sections 10CFR50. 73(a)(2)(i)(B) and 10CFR50. 73(a)(2)(v) to eliminate reporting of a condition or event that did not occur within three years of the date of discovery. This exclusion of such historical events and conditions should be extended to all written reports required by Section 50. 73(a).

No safety significance exists for 10CFR50. 72 reporting of historical events. FPC therefore believes that 10CFR50. 72 should clearly indicate that this is for current conditions and not require reporting of historical events.

Required Initial Reporting Times In the interest of simplicity, the proposed amendments should maintain only three basic levels of required reporting times in 10CFR50. 72 ( 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ), and 10CFR50. 73 (60 days) .

FPC agrees with the revised reporting times based on importance to risk and extending the required reporting times consistent with the need for prompt NRC action. Additionally, the increased time for submittal of LERs will allow for completion of required engineering evaluations after event discovery, provide for more complete and accurate LERs, and result in fewer LER supplemental reports. FPC does not believe that additional levels of reporting are required.

Comment 7 and Response (FR Page 36293)

In response to recommendations to stop reporting invalid ESF actuations, the proposal stated: "The comments are partially accepted. The proposed amendments would eliminate the requirement for telephone notification of an invalid actuation under 10CFR50. 72.

Invalid actuations are generally less significant than valid actuations because they do not involve plant conditions (e.g., low Reactor Coolant System pressure) conditions that would warrant system actuation. Instead, they result from other causes such as a dropped electrical lead during testing. However, the proposed amendments would not eliminate the requirement for a written report of an invalid actuation under 10CFR50. 73."

FPC supports the NEI position that no significance is associated with reporting invalid ESF actuations and believes that the above response supports the contention that they are not significant. Invalid ESF actuations should not be reported.

Reporting only valid ESF actuations that address the response of the plant to actual challenges would accomplish the intended change. Contrary to the NRC's expectations, reporting of invalid actuations will not provide the information needed to estimate equipment reliability parameters. This information should be collected through less burdensome mechanisms, such as Equipment Performance Information Exchange (EPIX) and Maintenance Rule reports.

(1)

North North Atlantic Energy Service Corporation P. O. Box 300 Atlantic Seabrook, NH 03874 (603) 474-9521

-99 sEri 20 mo :2s The Northeast Utilities System o,-, September 17, 1999 fI AD.

NYN-99088 DOCKET NUMBER PROPOSED RU Ms. Annette Vietti-Cook, Secretary Attention: Rulemaking and Adjudications Staff United States Nuclear Regulatory Commission Washington, DC 20555-0001

Seabrook Station Comments on Proposed Rule Federal Register Volume 64, No. 128, Pages 36291 - 36307, Dated July 6, 1999 The following comments are provided by North Atlantic Energy Service Corporation (North Atlantic) in response to a proposed rule published in the Federal Register on July 6, 1999 amending the reporting requirements for nuclear power reactors as contained in 10CFR50.72 and 50.73. North Atlantic appreciates the opportunity to comment on the proposed rule. We support the NRC Staffs efforts to reduce or eliminate the reporting burden associated with events of little or no safety significance. North Atlantic supports and endorses the comments provided by the Nuclear Energy Institute (NEI) on behalf of the nuclear industry.

However, North Atlantic is concerned that the addition of the new degraded component reporting requirements of 10CFR50.73(a)(2)(ii)(C) has the potential to require a significant increase in the number of Licensee Event Reports to be submitted each year such that the net result of the rule change is a significant increase in the reporting burden with the focus being on issues that are not safety significant. During the August 3, 1999, public workshop, NRC Staff personnel stated that the new criteria is a replacement for the outside the design basis criteria that was being deleted.

This cannot be the case since the existence of a degraded condition, during which the component is operable and capable of fulfilling its specified safety functions, cannot constitute a condition outside the design basis of the plant. Information on degraded equipment is provided via the EPIX system and is tracked by the Maintenance Rule requirements and is readily available to the NRC resident inspectors. Therefore, North Atlantic recommends that the new degraded component reporting criteria of IOCFR50.73(a)(2)(ii)(C) be deleted from the proposed rule.

In addition, the terms significant and significantly are used in the proposed rule but they are not defined in the rule nor in the revised reporting guidance of NUREG-1022. Comment No. 9 in the Federal Register Notice states that significant and significantly would be defined by examples. However, the examples provided in UREG-1022 do not define the terms and

- ~ ') f99 edged by card *-* ' II ll I ** ,,n

Ms. Annette Vietti-Cook, Secretary NYN-99088 I Page 2 contribute to the potential for interpretation of the rule by inspectors with the result being an increase in the reporting burden. North Atlantic recommends that the terms significant and significantly be clearly defined or that the reporting criteria be revised to base reporting upon a defined term such as operable.

Should you have any questions regarding our comments please contact Mr. James M. Peschel, Regulatory Compliance Manager, at (603) 773-7194.

Very truly yours, NORTH ATLANTIC ENERGY SERVICE CORP.

r ed C. Feigen aum Executive V =ident and Chief Nuclear Officer cc: United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 H.J. Miller, NRC Region I Administrator J. T. Harrison, NRC Project Manager, Project Directorate 1-2 R. K. Lorson, NRC Senior Resident Inspector

jjjljljj Omaha Public Power District DOCKET NUMBER PROPOSED RU D ('<Fl[D U 'R 444 South 16th Str991 Mall Omaha, Nebraska 68 102-2247 '9<J SEP 20 A1G :25 September 17, 1999 LIC-99-0081 Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTN: Rulemakings and Adjudications Staff

References:

1. Docket No. 50-285

Subject:

2. Federal Register Volume 64 dated July 6, 1999 (64 FR 128)

Comments on Proposed Revision to 10CFRS0.72, 10CFRS0.73 and NUREG 1022 Revision 1 The Omaha Public Power District (OPPD) has reviewed the Federal Register notice concerning proposed rulemaking on "Reporting Requirements for Nuclear Power Reactors" (64 Federal Register 36291 of July 6, 1999).

OPPD believes that the proposed change can meet the stated objectives to align reporting with needs, reduce burden where there is no safety significance and provide clarity to reporting.

OPPD fully endorses the comments to this rule that are being sent by the Nuclear Energy Institute (NEI) .

There are two areas of the proposed rule with which OPPD has significant concerns.

The first concern is the requirement to report significantly degraded components, section 50.73(a)(2)(ii)(C). This section does not meet the stated objectives of the rule change and should be deleted. The proposed requirement to report items that are not safety significant runs counter to the intent of the event reporting rule. OPPD believes that collecting data of components, as a part of this rule, is an unwarranted backfit. The requirement to capture items (components) which are seriously degraded, but not enough so as to render a system inoperable, is subject to widely varying levels of interpretation. Such a degree of uncertainty and lack of clarity exists in this newly proposed requirement, that the potential exists to greatly increase the number of required LERs beyond those that were submitted due to being "outside the design bases of the plant." This would lead to various thresholds of reporting being adopted by licensees and the Regions. This practice is not consistent with the proposed Risk Based Inspection Program.

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.- . LIC-99-0081 Page2 The second concern is that the proposed rule contains a detailed list of ESF systems for reporting. To support the time frame ofthis rulemaking initiative, OPPD recommends each facility define ESF systems based on their current USAR. This is option three in the rulemaking package. Ultimately, as part of the initiative to risk-inform 10 CFR Part 50, a plant-specific, risk-informed list should be used when criteria are fully developed. This would include only those systems that are significant to safety.

Please contact me if you have any questions.

Sincerely, 1.,n'L 1~11-11 V'} ~ K. Gambhir Division Manager Engineering & Operations Support SKG/epm c: E. W. Merschoff, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manager W. C. Walker, NRC Senior Resident Inspector Document Control Desk Winston & Strawn

/ .t Niagara G Moh~ 1

'S John H. Mueller So , 1 :z. September 1~ 1999 Phone: 315.349.7907 Senior Vice President and (&'fFRst,:}'!ij NMPlL 146-~ S[f' 2u A1L :L5 Fax: 315.349.1321 e-mail: muellerj@nimo.com Chief Nuclear Officer Secretary o.

U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 AO' ATTN: Rulemakings and Adjudications Staff RE: Nine Mile Point Unit 1 Nine Mile Point Unit 2 Docket No. 50-220 Docket No. 50-410 DPR-63 NPF-69

Subject:

Comments on the Proposed Rulemaking on 10 CFR 50.72 & 50.73 Gentlemen:

The purpose of this letter is to provide Niagara Mohawk Power Corporation's (NMPC's) comments on the proposed rulemaking in 64 Federal Register 36291, July 6, 1999 to revise 10 CFR 50.72 and 50.73. NMPC generally agrees with the comments being provided by the Nuclear Energy Institute (NEI). In addition, NMPC is very concerned that the proposed rulemaking represents a step backward from the originally proposed rulemaking approximately one year ago which resolved the concerns associated with the reporting of "outside the design basis of the plant," as exemplified by the NMPC pressure relief panel violation issued on June 18, 1996, associated with NRC Inspection Report 50-220/96-05.

As stated in Mr. L. J. Callan's letter to Mr. J. H. Mueller on March 26, 1998, regarding the pressure relief panel violation, "In the near future, we will publish an advanced notice of proposed rulemaking and conduct a public workshop to obtain public input regarding plans to modify the event reporting requirements .... 11 The proposed rule that followed that letter represented a significant step forward for the industry. The currently proposed rule leaves the objectives stated in Mr. Callan' s letter incomplete.

The fundamental regulatory issue associated with the reportability of changes in design values for the pressure relief panels was the definition of those changes which would cause a licensee to be "outside the design basis of the plant. 11 The NRC Staff, by virtue of its interpretation (which we believed was not supported by the plain language of the reporting rule that had existed for some period of time) had added the requirement to report any change in a design value, regardless of its effect on the ultimate safety design basis envelope. The advance notice of proposed rulemaking on July 23, 1998, appropriately resolved this issue. However, the proposed rule issued on July 6, 1999, expanded the scope of reporting requirements beyond issues which would be relevant to the safe operation of the facility or its ability to respond to transients or accidents. The NRC Staff thereby proposed expanding the scope of what was required to be reported to include issues that had no effect on the safe operation of the facility or its ability to respond to transients or accidents. The Staff's July 6, 1999, proposal would multiply the number of reports made, thereby diluting the Staff's ability to respond to the significant safety issues that are properly reportable.

Nine Mile Point Nuclear Station P.O. Box 63, Lycoming, New York 13093-0063

I Page2 The newly proposed Section 50. 73(a)(2)(ii)(C) would require reporting if a component is in a degraded or non-conforming condition, such that the ability of the component to perform its specified safety function is significantly degraded and the condition could reasonably be expected to affect to other similar components in the plant. However, the term "significantly degraded" is open to interpretation and inconsistent application. For example, one possible interpretation is that a deviation from any design value contained in the Final Safety Analysis Report could be viewed as a significant degradation subject to reporting. Moreover, the proposed requirement that the condition could reasonably be expected to affect other similar components also leaves much room for interpretation, confusion and inconsistency. As an example, in the case of the pressure relief panels, would this proposed reporting requirement consider the pressure relief panels as a whole or would it require consideration of each of the components that make it up. In this instance, would a licensee have to examine the possibility that bolts may be inappropriately sized on any other component of the plant?

The issue of degraded components is clearly a matter that is being satisfactorily handled within the purview of Generic Letter 91-18. By virtue of that letter and its implementation, which is well understood by licensees and the Staff, the NRC would be notified and involved as necessary when degradation which reached a level which required the participation or approval of the NRC for continued operation is discovered.

Therefore, NMPC recommends that the proposed Section 50. 73(a)(2)(ii)(C) not be promulgated, and in this respect, the reporting rule be returned to the form that resolved the pressure relief panel reporting issue in the advance notice of proposed rulemaking. This would set the reporting threshold at an appropriate safety significance threshold, help prevent a lack of clarity or ambiguous interpretations, and eliminate dilution of resources of both licensees and the NRC on issues that are below a safety significant threshold.

In addition, a portion of the new rule creates a conflict with an existing regulation. The maintenance rule, 10 CPR 50.65, gives incentives to licensees to maximize the availability of safety systems. There is an opposing incentive in the proposed 10 CPR 50. 73(a)(2)(iv) which would require reporting invalid actuations of safety systems. This requirement would have the likely effect of encouraging licensees to disable safety systems during minor troubleshooting activities in order to avoid the risk of an invalid actuation.

Thank you for the opportunity to comment on this proposed rulemaking.

Sincerely,

~8/i ueller Senior Vice President and Chief Nuclear Officer JHM/JFR/kap xc: Mr. H.J. Miller, NRC Regional Administrator, Region I Mr. S. S. Bajwa, Section Chief PD-I, Section 1, NRR Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. D.S. Hood, Senior Project Manager, NRR Document Control Desk Records Management

DOC f F['

NUCLEAR ENERGY I NS TI TUTE

'99 scr 20 Al u :~~Zr~~- Davis OPERATIONS DEPARTMENT, NUCLEAR GENERATION September 17, 1999 Secretary U .S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Rulemaking and Adjudication Staff

SUBJECT:

Proposed Rule for Reporting Requirements for Nuclear Power Reactors -- 64 Federal Register 36291 -- July 6, 1999 PROJECT NUMBER: 689 These comments are submitted on behalf of the nuclear power industry by the Nuclear Energy Institute (NEI) 1 in response to the Federal Register notice(s) concerning proposed rulemaking on Reporting Requirements for Nuclear Power Reactors (64 Federal Register 36291 of July 6, 1999).

In general, we believe that the proposed changes to 10 CFR 50. 72 and 50. 73 meet the stated objectives to better align reporting requirements with needs, reduce burden where there is no safety significance, and provide greater clarity to reporting requirements. Extensive use of workshops and tabletop exercises during the rulemaking process has provided a valuable testing ground for the proposed rev1s1ons.

We have two areas of significant concern.

First, the last minute addition of a requirement to report degraded components, section 50. 73(a)(2)(ii)(C), does not meet the stated objectives of the rule change and should be deleted. This requirement to report items that are not safety significant is a data collection exercise that runs counter to the intent of the event reporting rule. If this recent addition had been subjected to the same review and discussion 1 NEI is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including regulatory aspects of generic operational and technical issues. NEI's Members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect/engineering firms , fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy issue.

1776 I STREET, NW SUITE 400 WASHINGTON, DC 20006-3708 PHONE 202 739 8000 FAX 202.785 40 19

U:AR REGULATO

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Secretary September 17, 1999 Page 2 process as the rest of the rule, the added burden and lack of clarity would have been clearly evident to the staff. Additionally, we believe that collecting data on "degraded" components, as a part of this rule, is an unwarranted backfit. The needed component performance information can be extracted from other existing reports and databases.

The lack of clarity in this new requirement is particularly alarming. This attempt to capture components that are degraded, but not necessarily enough so as to render a system inoperable, is subject to widely varying levels of interpretation.

Additionally, the ambiguity of this provision would likely increase the number of required LERs of little safety significance far beyond those that were submitted due to being "outside the design basis of the plant," or any of the other criterion. This would further add unnecessary burden to licensees.

Second, the proposed rule contains a detailed list of engineered safety feature (ESF) systems for reporting. We do not support the proposed revisions which specify the systems for which reporting is required. An all-inclusive list of systems in a regulation is inappropriate. In the interest of maintaining clarity and simplicity, the best approach would be to select Option 3 and return to the pre-1998 situation whereby reporting would be required for the actuation of "any ESF" as is defined in each facility's FSAR. Ultimately, as part of the initiative to risk-informed 10 CFR Part 50, the possibility of using a plant-specific, risk-informed list should be explored. This would include only those systems that were significant to safety.

Additional effort and discussion are required before a risk-informed approach could be considered. In the short-term, we should continue with the longstanding practice of relying on each facility's FSAR.

Detailed comments and specific proposals are enclosed. We would be willing to meet with the NRC staff to further discuss these comments.

We appreciate the opportunity to comment on this proposed rule. Should you have further questions, please contact Bob Post of the NEI staff at 202-739-8115.

Sincerely, James W. Davis Enclosure

Enclosure Specific Comments on proposed Reporting Requirements for Nuclear Power Reactors These specific mdustry comments are provided on the proposed rulemaking for Reporting Requirements for Nuclear Power Reactors, published in Federal Register Notice 64 FR 36291 of July 6, 1999. Where appropriate, comments have been keyed to the relevant section of the text in the FRN:

GENERAL:

1. Reporting of component level failures We strongly object to the new requirements added by 50. 73(a)(2)(ii)(C) to report a component being in a degraded condition. The proposed change would add a new reporting requirement if a component is in a degraded or non-conforming condition such that: (a) the ability of the component to perform its specified safety function is significantly degraded; and (b) the condition could reasonably be expected to apply to other similar components in the plant. This last minute addition does not meet any of the NRC's stated objectives to better align reporting requirements to needs, reduce the burden of reporting items of little safety significance, or enhance clarity in the reporting requirements. It appears to be mainly a data collection effort that does not relate to event reporting. The need for this component level information should be addressed through other, existing reporting systems.

This newly proposed criterion was not included in the advance notice of proposed rulemaking (ANPR) published in July 1998. The criterion also was not included for discussion in the public meeting on August 21, 1998, the stakeholder's meeting on September 1, 1998, nor the public meeting held on November 13, 1998. This attempt to capture components that are degraded, but do not necessarily render a system inoperable, is a proposed addition that lacks clarity and is subject to widely varying levels of interpretation.

During public discussions, some have proposed that this information is already being submitted under the current "outside the design basis of the plant" (ODB) criterion. Our review found that this is not so and that there are a number of cases, that were presented as examples, that would not be reported under the current requirements. As stated in the ANPR, one of the reasons for deleting the ODB criterion is the confusion and controversy over the meaning of the requirement. Even with the deletion of the ODB criterion, conditions will continue to be reported if they result in a loss or partial loss of capability to perform a safety function. The rule change is intended to clarify the reporting requirements, yet the proposed criteria is not clear and is subject to widely varying interpretation.

This newly proposed requirement is so ambiguous, that its implementation would greatly increase the number of required LERs.

In addition, the newly proposed criterion is contrary to the intent of the proposed rule:

The rule change is intended to obtain reporting that is more consistent with the risk significance of the systems involved, yet the proposed section is focused on the component level as opposed to the stated objective of risk significance at the system level.

Although the rule change is intended to reduce licensee's reporting burden, implementation of this provision would have the opposite effect. A number of increases in burden can be identified:

The provision would impact the Control Room Operators. Since the Shift Supervisor or Shift Technical Advisor typically reviews corrective action documents for reportability, these component failures would have to be evaluated for reportability.

The provision would impact the Engineering Staff. Each component failure would need to be evaluated for "significance" of the "degradation" with respect to reportability requirements even though the system remained operable. Also, since this rule will likely require more LERs to be written, the engineering personnel would be involved in a greater number of root cause evaluations. In addition, a great number of root cause evaluations and LERs would be generated, requiring senior management review.

Due to the lack of clarity in the criterion, widely varying interpretations would require LERs to be written for non-safety significant events. These could potentially include writing LERs for all maintenance preventable functional failures (MPFFs), minor differences in equipment qualification (EQ) life span, and trivial differences in seismic qualification that do not render the system inoperable.

The stated "objective" of the criterion is to "allow" continued reporting of design basis or other "discrepancies" on a (component level), yet the current criterion does not require this.

There are no current requirements to report component failures or degradations that do not impact operability. Previously established guidance in Generic Letter 91-18 and 91-18 Rev. 1 provide for determining the operability of structures, systems or components (SSCs). By using these guidelines, licensees can determine whether SSCs are operable, but degraded. Because these SSCs categorizations do not affect system or train operability, no reporting in accordance with 10 CFR 50. 72 or 50. 73 has been previously required, except for certain conditions identified as being prohibited by Technical Specifications. The Page 2

new criterion to report degraded SSCs that have generic implications represents a reporting threshold far below that previously required.

Additionally, it is important to note that the newly proposed criterion is somewhat redundant with existing reporting guidance. For example, the discussion in NUREG 1022, pertaining to 10 CFR 50. 73(a)(2)(vii) states: In addition, if the cause or condition caused components in Train A of one system and in Train B of another system (i.e., train that is assumed in the safety analysis to be independent) to become inoperable, the event must be reported."

Significant component defects are also covered under 10 CFR Part 21 and additional discussion on component failures is also provided under conditions prohibited by Technical Specifications. -

If component data is needed, this information is already being provided by licensees through existing systems such as the Equipment Performance Information Exchange (EPIX) and Maintenance Rule reports. Component failure data is also available to the industry via the nuclear network. Reactor licensees' use of industry information is routinely inspected by the NRC. In addition, the Resident Inspector Program closely monitors operational status and communicates it to Regional and NRR personnel. This information is reviewed for generic implications. Most importantly, events or conditions of significance will still continue to be reported under the remaining criteria without the need to create any new criterion to take the place of the deleted "outside the design basis of the plant" criterion.

The examples that are provided in the NUREG, although possibly reportable under other existing criterion, do not provide credible support to justify the need to report "degraded components". Specifically, the examples tend to imply, more directly, that the staff disagreed with the licensees' assessments of operability in each of these cases. Had the assessments in the examples indicated inoperability, reporting would have been made under other existing criterion.

Following the controversy surrounding the NMPC blowout panel violation, the NRC staff acknowledged weaknesses in reporting requirements. The subsequent proposal to delete the "outside the design basis of the plant" reporting criterion was a significant step towards improving the ambiguity associated with these issues. Collecting data on "degraded" components, as a part of this rule, is an unwarranted backfit even more confusing than "outside the design basis of the plant." It represents a "step in the wrong direction" with respect to correcting these identified weaknesses in the reporting requirements.

2. ESF system actuations In the proposed rule(s), the term "any engineered safety feature (ESF), including the reactor protection system (RPS)," which currently defines the systems for which actuation must be reported in section 50. 72(b)(2)(iv) and section Page 3

50. 73(a)(2)(iv), would be replaced by a specific list of systems. In addition to this proposed list of systems, three principal alternatives to the proposed rule have been identified for comment. These include (1) maintaining the status quo, (2) requiring use of a plant-specific, risk-informed list, or (3) returning to the pre-1998 situation before publication of the event reporting guidelines in NUREG-1022, Revision 1.

Option 3 best meets the goal of clarity and simplicity. This approach is to return to the longstanding practice with reporting required for the actuation of "any ESF" as is defined in each facility's FSAR. It would however be important to continue to include examples of non-reportable exceptions in the rule and in the implementing guidance as appropriate for systems that are considered to be ESFs, yet have lower levels of risk significance (control room ventilation systems, reactor building ventilation systems, fuel building ventilation systems, auxiliary building ventilation systems, RWCU isolations during restoration from maintenance, etc.) .

However, we recommend retaining the current guidance in NUREG 1022 section 3.3.2 related to not reporting certain types of component level actuations.

Option 1, maintaining the status quo, has problems that would have to be solved.

For example, the Reactor Water Cleanup Isolations routinely occur during system restoration following maintenance outages, due to rapid pressurization following valve opening. Although they are not definitely expected to occur, a strong possibility exists that the isolation will occur. Specific provisions (such as those included in NUREG 1022, revision 1, second draft, which was published in 1994) would have to be included to provide exceptions to reporting these isolations. To do otherwise would result in the continued reporting of RWCU isolations that are not risk significant .

There is a separate initiative to "risk-inform" 10 CFR part 50. Option 2 involves reporting of ESF actuations based on risk significance. However, we do not believe that all the necessary criteria have been adequately established to make the shift as part of this rulemaking. Additionally, developing a plant-specific, risk-informed list represents a significant expenditure of resources and it is unclear as to whether or how Option 2 meets the NRC's needs better than Option 3. As part of the rule change to "risk-inform" Part 50, it should be evaluated if it would be appropriate to "risk-inform" ESF systems subject to the event reporting rule(s).

We do not support the proposed revisions which specify the systems for which reporting is required. We believe that an all-inclusive list of systems in a regulation is inappropriate. Each facility's FSAR specifies equipment that is designated as ESF equipment. Plant specific differences exist in the safety-related status of their systems. Additionally, the risk-significance of a particular system can vary greatly between plants, due to a wide variety of design differences. An all-inclusive list would therefore increase the burden for some plants whose equipment on the list was not ESF equipment or equipment with a suitably high Page 4

risk-significance. There are a number of problems with the list. Here are a few examples. There may be more:

Feedwater coolant injection systems are not ECCS systems at any BWR.

Re*actor Core Isolation Cooling System (RCIC) is included in ISTS because it meets criterion 4 of 10 CFR 50.36, based on its contribution to the reduction of overall plant risk. However, RCIC is not credited in the plant's safety analysis.

Non-reportable exceptions should be allowed for systems that are considered to be ESFs, yet have lower levels of risk significance (control room ventilation systems, reactor building ventilation systems, fuel building ventilation systems, auxiliary building ventilation systems, RWCU isolations during restoration from maintenance, etc.).

The list inappropriately includes "associated support systems" for BWR Division 3 EDGs.

The list inappropriately includes station blackout diesel generators (and black start gas turbines that serve a similar purpose) that are not safety related.

Although the term Anticipated Transient Without Scram (ATWS) mitigating systems is clear to those licensees that have dedicated systems (i.e. AMSAC),

a great deal of confusion exists for those that have no dedicated system. Due to the lack of clarity, it could be interpreted that any system that might be used during an ATWS would fall into this category (i.e. feedwater systems, boration systems, control rods, etc.). Extensive clarification would be needed to eliminate this ambiguity.

It is unclear as to whether this applies only to Emergency Service Water systems (ie. those that don't operate unless there is an accident i.e.

"emergency service water function") or also to the standby service water systems that only run to remove heat from the RHR heat exchangers.

3. Reporting of historical events No safety significance exists for 10 CFR 50.72 reporting of historical events. We therefore support the revisions to 10 CFR 50. 72 that eliminates reporting of historical events.

The proposed rule would add provisions to sections 50. 73(a)(2)(i)(B) and

50. 73(a)(2)(v) to eliminate reporting of a condition or event that did not occur within three years of the date of discovery. We believe that two, rather than three, years 18 the appropriate time period for reporting historical events.

Two years encompasses one refueling cycle of operation. Significant effort can be expended searching back in history for historical events. Reporting historical events more than two years old provides a low safety benefit and unnecessarily increases the reporting burden. As the staff has pointed out, three years is consistent with the time period that performance indicators are tracked under the Page 5

new oversight process. We continue to believe that no safety significance exists for 10 CFR 50.73 reporting of historical events which occurred more than two years ago. This exclusion of such historical events and conditions should be extended to all written reports required by section 50. 73(a).

4. Required initial reporting times In the interest of simplicity, the proposed rule should maintain just three basic levels of required reporting times in 10 CFR 50. 72 and 10 CFR 50. 73 (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and 60 days).

We agree with the revised reporting times based on importance to risk and extending the required reporting times consistent with the need for prompt NRC action. Additionally, the increased time for submittal of LERs will allow for completion of required engineering evaluations after event discovery, provide for more complete and accurate LERs, and result in fewer LER revisions and supplemental reports. We do not believe that additional levels of reporting are required.

5. Invalid ESF Actuations In response to recommendations to stop reporting invalid ESF actuations, the NRC states: "The comments are partially accepted. The proposed amendments would eliminate the requirement for telephone notification of an invalid actuation under 10 CFR 50. 72. Invalid actuations are generally less significant than valid actuations because they do not involve plant conditions (e.g., low reactor coolant system pressure) conditions that would warrant system actuation. Instead, they result from other causes such as a dropped electrical lead during testing).

However, the proposed amendments would not eliminate the requirement for a written report of an invalid actuation under 10 CFR 50. 73."

We see no reason to submit written reports on invalid ESF actuations and believe that the above response supports the contention that they are not significant.

Invalid ESF actuations should not be reported under 10 CFR 50. 73 unless the actuation impacts the plant such that other reporting criteria are independently met.

Reporting only valid ESF actuations that address the response of the plant to actual challenges would accomplish the intended change, as proposed in NEI's response to the ANPR. Contrary to the NRC's expectations, reporting of invalid actuations will not provide the information needed to estimate equipment reliability para.meters. This information should be collected via other less burdensome mechanisms, such as EPIX and Maintenance Rule reports.

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6. HPCI Inoperability Inoperability of HPCI does not of itself constitute a condition that would prevent the :fulfillment of a safety function. Therefore, we see no benefit in reporting of HPCI inoperability if it has no affect on the ability to fulfill a safety function.

Guidance provided in the NUREG needs to be modified to clarify this issue. BWR design considers HPCI inoperability and provides supporting systems such as RCIC, Core Spray, and ADS. This is supported by the relatively long Allowed Outage Time (AOT) for HPCI in Standard Technical Specifications (i.e., 14 days).

If, in the event of HPCI inoperability, it can be shown that these systems are available and capable of fulfilling the safety function without HPCI, the event should not be reportable. Reporting HPCI inoperability in these cases has no meaning for event reporting and appears to be solely a data gathering exercise.

Additionally, the Revised Reactor Oversight Process as described in NEI 99-02 (Draft Rev. B) prescribes a performance indicator for Safety System Functional Failures based on 10 CFR 50. 73 reports. These indicators count failures of single train systems (such as HPcn, assuming that the event report documents a safety system failure. Reporting HPCI inoperability when there is no impact on the overall capability to fulfill the safety function (e.g., remove residual heat) will result in an overly conservative and detrimental assessment of this indicator.

SPECIFIC COMMENTS The following table provides specific comments and recommendations tied to the proposed rule or NUREG 1022 sections .

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Specific comments on 10 CFR 50. 72 SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS

50. 72(a)(l)(i) The declaration of any of the Emergency 50. 72(a)(l)(i) The declaration of any of the Emergency Classes specified in the licensee's approved Emergency Classes specified in the licensee's approved Emergency Plan; 8 Plan; 8 8

8 These Eme11geaey Glasses B:i'e aa:B::Pessea: m f..ppenclix These Emergency Classes are addressed in Appendix E E of thio pB:Ft.

of this part.

This section of the rule is clear as written and does not require the footnote for clarity.

50. 72(a)(4) The licensee shall activate the Emergency 50. 72(a)(4) The licensee shall activate the Emergency Response Data System (ERDS) l'i as soon as possible but Response Data System (ERDS) 5 as soon as possible but not later than one hour after declaring an emergency not later than one hour after declaring an emergency class of alert, site area emergency, or general emergency. class of alert, site area emergency, or general The ERDS may also be activated by the licensee during emergency. The ERDS may also be aetivatea: by the emergency drills or exercises if the licensee's computer lieeasee a:HHHg eme11geaey B::Pills 011 e~ereises if the system has the capability to transmit the exercise data. lieeasee's eomp:ateF system has the eapa:bility to tFB:D:sm.it the exe11eioe a:ata.

This last sentence should be deleted. It is superfluous information that adds no value to the discussion of reportability.

50. 72(b )(2) ... the licensee shall notify the NRC as soon as practical

... the licensee shall notify the NRC as soon as practical and in all cases, within eight hours of the occurrence or and in all cases, within eight hours of the occurrence of discovery of any of the following:

any of the following:

The addition of or discovery" provides for those events that are discovered to have occurred in the past, remained undetected for sometime, and presentl_y exist.

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50. 72(h)(2)(ii)(B) The nuclear power plant being in an 50. 72(h)(2)(ii)(B) The nuclear power plant being in an unanalyzed condition that significantly affects plant unanalyzed condition that significantly affects degrades safety. plant safety.

The phrase "significantly affects plant safety" has no positive or negative connotation. It is therefore recommended that the term degrades be substituted in place of affects.

50. 72(h)(2)(iv) Proposes a list of engineered safety As we have previously stated, we recommend that this feature systems: section be deleted and replaced with:

(A) Any event or condition that results in intentional Any event or condition that results in intentional manual actuation or valid automatic actuation of any of manual actuation or valid autom~tic actuation of any the systems listed in paragraph (b)(2)(iv)(B) of this enlrin.eered safetv feature (ESF), including actuation of section, except when the actuation results from and is the reactor :grotection system <RPS} from a critical part of a pre-planned sequence during testing or reactor condition, in res:gonse to actual :giant conditions that operation. (B) The systems to which the requirements of warrant ESF/RPS actuation, exce:gt when: (A} The paragraph (b)(2)(iv)(A) of this section apply are: (1) actuation results from and is :gart of a :gre-:glanned Reactor protection system (reactor scram, reactor trip). (2) seguence during testing or reactor o:geration; (B} The Emergency core cooling systems (ECCS) for pressurized actuation is invalid and: (l} Occurs while the system is water reactors (PWRs) including: high-head, :gro:gerly removed from service; (2} Occurs after the intermediate-head, and low-head injection systems and safetv function has been already com:gleted; or (3}

the low pressure injection function of residual (decay) Involves only the following s:gecific ESFs or their heat removal systems. (3) ECCS for boiling water reactors eguivalent systems: (i} Reactor water clean-u:g sy12tem; (BWRs) including: high-pressure and low-pressure core (ii} Control room emer@nm'.: ventilation system; (iii}

spray systems; high-pressure coolant injection system; Reactor building ventilation system; (iv} Fuel building feedwater coolant injection system; low pressure injection ventilation system; or (v} Auxiliarv building ventilation function of the residual heat removal system; and system.

automatic depressurization system. (4) BWR isolation condenser system and reactor core isolation cooling This option best meets the goal of clarity and simplicity.

system. (5) PWR auxiliary feedwater system. (6) This would require reporting of any ESF' as is defined Containment systems including: containment and reactor in each facilities FSAR. It would however be imDortant Page 9

vessel isolation systems (general containment isolation to include examples of non-reportable exceptions in the signals affecting numerous valves and main steam implementing guidance for systems that are considered isolation valve [MSIV] closure signals in BWRs) and to be ESFs, yet have lower levels of risk significance. As containment heat removal and depressurization systems, part of the rule change to risk inform Part 50, including containment spray and fan cooler systems. (7) consideration should be given to changing this section to Emergency ac electrical power systems, including: include a risk-informed list.

emergency diesel generators (EDGs) and their associated support systems; hydroelectric facilities used in lieu of EDGs at the Oconee Station; safety related gas turbine generators; BWR dedicated Division 3 EDGs and their associated support systems; and station blackout diesel generators (and black-start gas turbines that serve a similar purpose) which are started from the control room and included in the plant's operating and emergency procedures. (8) Anticipated transient without scram (ATWS) mitigating systems. (9) Service water (standby emergency service water systems that do not normally run).

50.72(b)(2)(v) Licensees shall report: "Any event or 50.72(b)(2)(v) Licensees shall report: "Any event or condition that at the time of discovery could have condition that at the time of discovery is eould have prevented the fulfillment of the safety function of preventinged the ability to fulfillment of the safety structures or systems that are needed to: ... function of structures or systems that are needed to: ...

This change is required to reflect the correct tense of the existence of an event or condition, rather than past speculation. Because of past confusion pertaining to the interpretation of this area, it is suggested that further discussion be included in the statements of consideration explaining that "is preventing" represents actual conditions and does not imply that further failures should be speculated.

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Specific comments on 10 CFR 50. 73 SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS

50. 73(a)(2)(i)(B) Any operation or condition occurring 50. 73(a)(2)(i)(B) Any operation or condition occurring within three years of the date of discovery which was within tm>ee- two years of the date of discovery which prohibited by the plant's Technical Specifications, except was prohibited by the plant's current Technical when: Specifications, except when:

(i) The technical specification is administrative in nature; (i) The technical specification is administrative in or nature; or (ii) The event consists solely of a case of a late (ii) The event consists solely of a case of a late surveillance test where the oversight is corrected, the test surveillance test where the oversight is corrected, the is performed, and the equipment is found to be capable of test is performed, and the equipment is found to be performing its specified safety functions. capable of performing its specified safety functione:-: or (iii) The event consists of a previously existing condition that does not comply with a newly implemented more restrictive Technical Specification.

As previously stated, we believe that two years is the appropriate time period for reporting histori!:al events.

The addition of "current" allows for plants that have recently converted to Improved Standard Technical Specifications to apply the current requirements to the identified condition, rather than considering Technical Specifications which are no longer applicable. "Safety function" should be singular to accommodate equipment with only one safety function. The third provision eliminates the need for a utility to do an unnecessary historical search for conditions that complied with a previous Technical Specification, but do not comply with a newly implemented, more restrictive Technical Specification.

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50. 73(a)(2)(ii)(B) The nuclear power plant being in an 50. 73(a)(2)(ii)(B) The nuclear power plant being in W unanalyzed condition that significantly affected plant an unanalyzed condition that significantly a:ffeeted safety; or degraded plant safety; or The phrase significantly affected plant safety" has no positive or negative connotation. It is therefore recommended that the term "degraded" be substituted in place of affected." (Similar to the change recommended or 10 CFR 50. 72 2 ii

50. 73(a)(2)(ii)(C) A component being in a degraded or As previously stated, we believe that this new criterion non-conforming condition such that the ability of the should be deleted. Following the controversy component to perform its specified safety function is surrounding the NMPC blowout panel violation, the significantly degraded and the condition could reasonably NRC staff acknowledged weaknesses in reporting be expected to affect other similar components in the requirements. The subsequent proposal to delete the plant. "outside the design basis of the plant" reporting criterion was a significant step towards improving the ambiguity associated with these issues. Collecting data on "degraded" components, as a part of this rule, is an unwarranted back/it even more confusing than "outside the design basis of the plant. " It represents a step in the wrong direction" with respect to correcting these identi

d weaknesses in the re orti re uirements.

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50. 73(a)(2)(iv)(A) Licensees shall report "Any event or As we have previously stated, we recommend that this condition that resulted in manual or automatic actuation section be deleted and replaced with:

of any of the systems listed in paragraph (a)(2)(iv)(B) of this section except when: (1) The actuation resulted from 50. 73(a)(2)(iv) Any event or condition that resulted in and was part of a pre-planned sequence during testing or a manual or automatic actuation of any engineered reactor operation; or (2) The actuation was invalid and; (i) safety feature (ESF} system, including the reactor Occurred while the system was properly removed from protection system (RPS), in response to actual plant service; or (ii) Occurred after the safety function had been conditions that warrant ESF/RPS actuation except already completed. when: (A) The actuation results from and is part of a

50. 73(a)(2)(iv)(B) "The systems to which the ... pre-planned sequence during testing or reactor operation: or (B) The actuation is invalid and: (1) Occurs while the system is properly removed from service: (2)

Occurs after the safety function has been already completed: or (3) Involves only the following specific ESFs or their equivalent systems: (i) Reactor water clean-up system: (ii) Control room emergency ventilation system: (iii) Reactor building ventilation system: (iv) Fuel building ventilation system: or (v)

Auxiliary building ventilation system.

Our proposed wording requires reporting only valid ESF actuations that address the response of the plant to actually challenges. It would however be important to include examples of non-reportable exceptions in the implementing guidance for systems that are considered to be ESFs, yet have lower levels of risk significance.

This would provide consistent wording between 10 CFR

50. 72 and 10 CFR 50. 73 and accomplish the intended change.

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50. 73(a)(2)(v) Licensees shall report: "Any event or 50. 73(a)(2)(v) Licensees shall report: "Any event or condition occurring within three years of the date of condition occurring within thi-ee- two years of the date of discovery that could have prevented the :fulfillment of the discovery that eewd would have prevented the safety function of structures or systems that are needed fulfillment of the safety function of structures or to: ... systems that are needed to: ...

As previously stated, we believe that two years is {he appropriate time period for reporting historical events.

This change reflects the existence of an event or condition, rather than speculation. It is suggested that further discussion be included in the statements of consideration explaining that "would" represents actual conditions and does not imply that further failures should be speculated. (Similar to 50. 72(bX2Xv)).

50. 73(b)(2)(ii)(J) For each human performance related 50. 73(b)(2)(ii)(J) For each root cause personnel error problem that contributed to the event, the licensee shall human peri'o:Pmanee rolated problem that eontributed to discuss ... the event, the licensee shall discuss ...

The shift from "personnel error" and the implied root cause" to human performance related problem" and "contributing factors" greatly increases the scope of investigation and burden to the licensee. It is only appropriate to require discussion of personnel error root causes.

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50. 73(b)(3)(ii) Are included in emergency or operating We recommend that this new criterion be deleted.

procedures and could have been used to recover from the Reporting submitted under this criterion should be based event in case of an additional failure in the systems on existing plant conditions. Emergency operating actually used for recovery. procedures provide direction for use of many plant systems. If an additional failure must be postulated for every event, multiple systems would be required to be included in the LER for each safety function. There exists an infinite combination of failures that could be postulated. This unbounded requirement would result in a large amount of additional information that would be of minimal use. The assessment of the safety consequences and implications of the event would become cluttered with hypothetical additional failures and possible plant resoonses.

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Specific comments on NUREG 1022, Revision 2 (draft)

SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS HPCS- High Pressure Core Spray, LPCI- Low Pressure Core Injection, and Abbreviations LPCS- Low Pressure Core Spray should be added to the list of abbreviations NPRDS arid SALP should be deleted since they are no lon}!er used and are not referenced in the NUREG.

2.5, Time Limits for Reporting These 10 CFR 50. 72 reporting times have some These 10 CFR 50. 72 reporting times have some flexibility flexibility because a licensee needs to ensure that because a licensee needs to ensure that reporting does not reporting does not interfere with plant operation.

interfere with plant operation. However, that does not However, that does not mean that a licensee should mean that a licensee should automatically wait until automatically wait until close to the time limit close to the time limit expiration before reporting. For expiration before reporting. For example, assume that a example, assume that a small radioactive release, that small Padioaotiv=e Peleaoe, that does not qualify for does not qualify for declaration of an emergency class, is deolB:Pation of an emePgenoy elaoo, is reported to the reported to the State. The rule requires reporting an State. The Fule Pe quires Feporting an event of this type event of this type as soon as practical and in all cases, as soon as praetieal and in all eases, :within eight hours within eight hours of occurrence. In this case, since it of oecunenee. In this ease, since it would be practical to would be practical to do so, the licensee should make the do so, the lieenoeo should make the ENS notification at ENS notification at about the same time as the report to about the same time as the Pepo:Ft to tho State.

the State.

We recommend deleting the example provided. One of the stated objectives for e:deruling the reporting time was to allow a more complete investigation of the event in order to avoid unproductive supplemental notifications.

It was agreed that 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was a sufficiently timely period to meet NRC's event reporting needs. The example could have the effect of eliminating this benefit, thereby forcing the licensee to make the notification. If delayed, yet reported within the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time period, a more accurate, complete report could be provided.

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3.2.1, Plant shutdown required by TS, Example (1)

An LER was not submitted under this criterion since the An LER was not submitted required under this criterion since the failed battery charger was corrected before the failed battery charger was corrected before the plant plant completed shutdown.

completed shutdown.

The proposed change would enhance the clarity of this provision by specifying that an LER was not required to be submitted, not merely whether or not one was submitted in this particular case.

3.2.1, Plant shutdown required by TS, Example (2) An LER was required submitted because of the An LER was submitted because of the completion of the completion of the TS-required plant shutdown.

TS-required plant shutdown.

The proposed change would enhance the clarity of this provision by specifying that an LER was actually required to be submitted, not merely whether or not one was submitted in this particular case..

3.2.1, Plant shutdown required by TS, Example (3) Isit an LER required?

It an LER required?

Answer:

Some judgment is :Pequired. An LER is not required.. if Some judgment is required. An LER is not required if the the situation eould have been oo:P:Peoted before the plant situation could have been corrected before the plant was was roquired to be shut down, and no otho:P oriteria in required to be shut down, and no other criteria in 50. 73 50.78 apply. The shut down is reportable, however, if apply. The shut down is reportable, however, if the the situation could not have been corrected before the situation could not have been corrected before the plant plant was required to be shut down, O:P if other eriteria was required to be shut down, or if other criteria of 50. 73 of 50.78 apply.

apply.

Minor editorial change (It was replaced with Is). Deleted superfluous information already stated in the question.

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3.2.2, Technical Specification Prohibited Operation Licensees are required to submit an LER for any or Condition operation or condition occurring within tlH.-ee- two years Licensees are required to submit an LER for any of the date of discovery which was prohibited by the operation or condition occurring within three years of the plant's Technical Specifications, except when: ...

date of discovery which was prohibited by the plant's Technical Specifications, except when: ... As previously stated, we believe that two years is the avvropriate time period for reportinR historical events.

3.2.2, Technical Specification Prohibited Operation Note that there would not be a two throe year or Condition, Safety Limit and LSSS Discussion limitation in this case because, in addition to the Note that there would not be a three-year limitation in requirements of §50. 73(a)(2)(i)(B), specific reporting this case because, in addition to the requirements of requirements are stated in §50.36(c)(l) and the TS.

§50. 73(a)(2)(i)(B), specific reporting requirements are stated in §50.36(c)(l) and the TS. As previously stated, we believe that two years is the appropriate time period for reporting historical events.

Reference to TS should be deleted. The NRG approved removal of this reporting requirement from ISTS in TSTF-5.

3.2.2, Technical Specification Prohibited Operation A missed or late surveillance test is reportable when it-or Condition, TS Surveillance Requirements the performance of the test indicates that equipment Discussion (i.e., one train of a multiple train system) was not A missed or late surveillance test is reportable when it capable of performing its specified safety functions (and indicates that equipment (i.e., one train of a multiple thus was inoperable) for a period of time longer than train system) was not capable of performing its specified allowed by TS (surveillance interval and extension plus safety functions (and thus was inoperable) for a period of the allowed outage time) ....

time longer than allowed by TS ...

Otherwise, the defioieney should be assumed to have Otherwise, the deficiency should be assumed to have oeoll.ffed halfway between the last suoeessful test or use occurred halfway between the last successful test or use and the untimely test that Pevealed the defieieney.W and the untimely test that revealed the deficiency.{5}

The proposed wording adds clarification concerning how it is determined that the system is inoperable and when it is determined to be reportable. It is recommended that the last sentence be deleted since this statement appears Page 18

to contain no basis. The deficiency should be assumed to have occurred at the time of discovery (as is the current practice) unless firm evidence exists to the contrary.

3.2.2, Technical Specification Prohibited Operation Immediately after the paragraph that starts out For a or Condition, TS Surveillance Requirements discrepancy discovered during a timely surveillance... "

Discussion insert the following:

ESF Ventilation System Technical Specifications revised per Generic Letter 83-13 require verification.

within 31 days after removal that a laboratory analysis of a representative carbon sample meets a testing limit of less than the plant-specific required percentage of methyl iodide penetration. For these ESF Ventilation Systems. the corresponding allowed outage time for a single train is less than 31 days (typically in the range from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 7 days). If the licensee is notified that the analysis results are unsatisfactory within the 31 days permitted by Tech Specs. but beyond the associated allowed outage time for the subject train.

and the ventilation system was capable of meeting its assumed design basis charcoal filter efficiency. the event would not be reportable.

This guidance clarifies one of the long-standing problems that have been associated with determining the reportability when an unsatisfactory ventilation sample result is obtained.

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3.2.2, Technical Specification Prohibited Operation STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent, or Condition, Entry into STS 3.0.3 Discussion establishes requirements for actions when an LCO is STS 3.0.3 (ISTS LCO 3.0.3), or its equivalent, establishes not met and no aotion statement is p:Po¥idod. Entry into requirements for actions when an LCO is not met and no STS 8.0.a is considered to indieato that a condition action statement is provided. Entry into STS 3.0.3 is existed longer than allowed by TS. Thus, ontey into considered to indicate that a condition existed longer than STS 8.0.a (ISTS LCO 3.0.a) for any reason o:P allowed by TS. Thus, entry into STS 3.0.3 (ISTS LCO juotifieation is Pepo:Ptablo.

3.0.3) for any reason or justification is reportable.

(1) an LCO is not met and the associated ACTIONS are not met. (2) an associated ACTION is not provided.

or (3) as directed by the associated ACTIONS themselves. Entry into STS 3.0.3 GSTS LCO 3.0.3) for either of the first two above reasons is generally reportable under this criterion. However. when the plant TS specify the entry into 3.0.3 as the required ACTION and that action and its completion time are met. the event is not reportable under this criterion.

Also, momentary Gess than approximately 5 minutes) entries into TS 3.0.3. regardless of the reason. are not reportable under this criterion. Any TS 3.0.3 entry involving actual plant shutdown should be reviewed for reporting under 50. 72(b)(l)(i)(A) and 50. 73(a)(2)(i)(A) for a plant shutdown required by TS.

The proposed wording is suggested to replace the existing outdated guidance. The proposed wording reflects the current operating philosophy associated with implementation of the Standard Technical Specifications.

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3.2.2, Technical Specification Prohibited Operation Sections 50.55a(g) and 50.55a(f) require the or Condition, Missed Tests Required by ASME implementation of 181 and 1ST programs in accordance Section XI Discussion with the applicable edition of the ASME Code for those Sections 50.55a(g) and 50.55a(f) require the pumps and valves whose function is required for safety.

implementation of lSl and 1ST programs in accordance STS Section 4.0.5 (or an equivalent) covers these testing with the applicable edition of the ASME Code for those requirements. (Gene1:1ally, there is no eompa.Pable ISTS pumps and valves whose function is required for safety. section.) A missed or late lST/ISI/ASME test (that is STS Section 4.0.5 (or an equivalent) covers these testing specifically required by Technical Specifications) is requirements. (Generally, there is no comparable ISTS reportable when it the performance of the test indicates section.) A missed or late IST/ISI/ASME test is that equipment (i.e., one train of a multiple train reportable when it indicates that equipment (i.e., one system) was not capable of performing its specified train of a multiple train system) was not capable of safety functions and, thus, was inoperable for a period performing its specified safety functions and, thus, was of time longer than allowed by TS (surveillance interval inoperable for a period of time longer than allowed by TS. and extension plus the allowed outage time).

This statement should be deleted since ISTS Section 5,

Programs and Manuals," has a section for these requirements in Inservice Testing Program". The proposed wording adds clarification concerning how it is determined that the system is not capable of performing the safety function and when it is determined to be reportable.

3.2.2, Technical Specification Prohibited Operation Also, if a fire protection deficiency results in the or Condition, Fire Protection Systems When inability to preserve at least one safe shutdown train in Required by TS Discussion the event of a fire, it should be 1:1eported evaluated for Also, if a fire protection deficiency results in the inability reporting as an unanalyzed condition that significantly to preserve a safe shutdown train in the event of a fire, it affeets degrades plant safety, as discussed in Section should be reported as an unanalyzed condition that 3.2.4 of this report.

significantly affects plant safety, as discussed in Section The guidance should be modified to indicate the need to 3.2.4 of this report. evaluate for reporting, rather than indicate that reporting is definitely required, since other trains or systems may be available to preserve safe shutdown.

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3.2.2, Example (3) Entering STS 3.0.3 (3) Entering STS 3.0.3 (3) Entering STS 3.0.3 due to lack of specific TS actions With essential water chillers (A) and (B) out of service, With essential water chillers (A) and (B) out of service, the only remaining operable chiller (A/B) tripped. This the only remaining operable chiller (A/B) tripped. This condition caused the plant to enter STS 3.0.3 condition caused the plant to enter STS 3.0.3 (equivalent (equivalent to ISTS LCO 3.0.3) for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , until chiller to ISTS LCO 3.0.3) for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , until chiller (A) was (A) was restored to service and the temperature was restored to service and the temperature was restored to restored to within TS limits.

within TS limits.

An LER is required for this event because STS 8.0.3 An LER is required for this event because STS 3.0.3 was was entered

in this case. there were no actions entered provided in the plant TS for that condition and STS 3.0.3 was entered for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The proposed wording reflects the current operating philosophy associated with implementation of the Standard Technical Specifications. An additional example is also included to clarify this position:

Entry into STS 3.0.3 when the plant TS specify 3.0.3 entry During a surveillance test on the A train of a two-train Standby Gas Treatment (SBGT) system, a condition was discovered on the B train that rendered it inoperable. The test was halted and steps taken to return the A train to a standby readiness condition.

During the restoration. switch manipulations momentarily rendered the A train inoperable. With both trains inoperable. the plant TS specify immediate entry into LCO 3.0.3. The entry into LCO 3.0.3 was logged and then exited within 1 minute once switch manioulation on the A train was comoleted.

Page 22

This event is not re:Qortable under this criterion because all the actions S:Qeci:fied by the :Qlant TS were com:Qleted within the reguired com:Qletion times. There was no O:Qeration or condition :QrOhibited by TS. Also, moment.arv entries into STS 3.0.3 are not renortable.

3.2.2 Technical Specification Prohibited Operation However, the existence of similar discrepancies in or Condition, Example (4), Multiple Test Failures multiple valves is- could be an indication that the However, the existence of i;iimilar discrepancies in discrepancies arose over a period of time, and the multiple valves is an indication that the discrepancies failure modes should be evaluated to make this arose over a period of time. determination.

Changed '~" to "could be" and added a statement that an evaluation should determine this conclusion. While in most cases this statement may be true, it should still be evaluated to reach this conclusion.

3.2.2, Example (5) Seismic Restraints i\ssamo that a:HPi:B:g an NRG e:r;alaatien ilt is found that Assume that during an NRC evaluation it is found that an exciter panel for one EDG had lacked appropriate an exciter panel for one EDG had lacked appropriate seismic restraints since the plant was constructed, seismic restraints since the plant was constructed, because of a design, analysis or construction because of a design, analysis or construction inadequacy. inadequacy. Aloe assume that, u l!pon evaluation, the Also assume that, upon evaluation, there is reasonable EDG is determined to be ino:Qerable thel'e is I'easenable doubt about the EDG's ability to perform its speci:fied a:euat aeeut the EI)G's aeility ta pe'Pfel'm its speemea:

safety functions during and after an SSE. safety funetiens a:uring ana: eftep an SSE.

The recommended wording del,etes superfluous information and equates the ability to perform a specified safety function to operabilit_-y.

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3.2.2, Example (6) Vulnerability to Loss of Offsite .Assume that d.During a design review it is found that a Power loss of offsite power could cause a loss of instrument air Assume that during a design review it is found that a loss and, as a result, Rmciliary feedwater (AFW) flow control of o:ffsite power could cause a loss of instrument air and, valves could fail open. Then for low steam generator as a result, auxiliary feedwater (AFW) flow control valves pressure, such as could occur for certain main steam could fail open. Then for low steam generator pressure, line breaks, high AFW flow rates could result in such as could occur for certain main steam line breaks, tripping the motor driven AFW pumps on thermal high AFW flow rates could result in tripping the motor overload. Therefore, the motor driven AFW pumps driven AFW pumps on thermal overload. The single were declared inoperable. The single turbine driven turbine driven AFW pump would not be affected. AFW pump would not be affected.

The recommended wording deletes superfluous information and clearly indicates that the equipment is determined to be inoperable (for a time greater than allowed by Technical Specifications).

3.2.3, TS Deviation per §50.54(x) Example When the action was completed the control room, When the action was completed the control room operators notified the NRC Operations Center, in operators notified the NRC Operations Center, in accordance with the reporting requirements of 10 CFR accordance with the reporting requirements of 10 CFR 50. 72, that they had exercised 10 CFR 50.54(x).

50. 72, that they had exercised 10 CFR 50.54(x). Subsequently, an LER was required submitted in Subsequently, an LER was submitted in accordance with accordance with 10 CFR 50. 73(a)(2)(i) {use of 10 CFR 10 CFR 50. 73(a)(2)(i) {use of 10 CFR 50.54(x)} as well as 50.54(x)} as well as 10 CFR 50. 73(a)(2)(v) {event or 10 CFR 50. 73(a)(2)(v) {event or condition that alone could condition that alone oould would have prevented ....}.

have prevented ....}.

illustrates clarity as to the appropriateness of the example. And replaces "alone could" with would" as was previously recommended.

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3.2.4, Degraded Condition If not reported as an emergency under 50. 72(a),

If not reported as an emergency under 50. 72(a), licensees licensees are required to report a seriously degraded are required to report a seriously degraded principal principal safety barrier or an unanalyzed condition that safety barrier or an unanalyzed condition that significantly affeets degrades plant safety ....

significantly affects plant safety ....

As previously stated, it is recommended that "affects" be replaced with "de!!rades".

3.2.4, Degraded Condition, (A) Nuclear power plant, It is proposed that this section be deleted since this including its principal safety barriers, being position is based on a Draft Regulatory Guide (DG-10 74._

seriously degraded, introduces a new section: Steam Generator Tube Integrity) that has not been approved. Discussions between the industry and the (3) Steam generator tube degradation in the following NRC are being held to define the steam generator circumstances ... program and Technical Specification requirements.

Some of the examples provided in the proposed section are contrary to agreements that have been made between the industry and the NRC staff. Recognizing that these agreements are still evolving, the proposed revisions to the rule(s) and NUREG 1022 must agree with the final positions on steam l!enerator issues.

3.2.4, Degraded Condition, (B) Unanalyzed As previously stated, it is recommended that affects" be condition that significantly affects plant safety replaced with "degrades". In addition, the discussion for Discussion unanalyzed condition that significantly degrades plant safety should include a memorandum from William T.

Russell to Thomas T. Martin dated April 1, 1991, that addresses the reportability of ASME Code Class ill defects under this criterion. This memorandum was in response to a request for clarification from Public Service Electric and Gas on reporting requirements regarding service water piping leaks.

Page 25

3.2.4, Degraded Condition, (C) Significantly As previously stated, it is recommended that this section degraded component(s) Discussion be deleted. Collecting data on components, as a part of Three examples are given that would be reportable under this rule, is an unwarranted backfit even more confusing the newly proposed criterion, summarized as follows: than "outside the design basis of the plant". The

1. Normally open valves are closed to support criterion lacks clarity in several respects:

surveillance testing. An incorrect determination was

What is a component?

made that valves remained operable while closed

What is degraded", yet operable?

(contrary to guidance in GL 89-10).

What is significant?

2. An incorrect maintenance procedure allowed

What is "reasonably expected?

overfilling of the valve operator grease box for several

What are similar components?

valves. Contamination of electrical components (not In addition, the examples that are provided, although qualified for exposure to grease) resulted from the possibly reportable under other existing criterion, do not overfilling. provide credible support to justify the need to report

3. Several valves located in containment (below the flood "degraded components". Specifically, the examples tend plane) were not qualified to be submerged. It was to imply, more directly, that the staff disagreed with the determined that the valves remained operable licensees' assessments of operability in each of these (pending replacement with qualified valves) after cases. Had the assessments in the examples indicated flooding occurred and that position indication could be inoperability, reporting would have been made under determined indirectly, using process parameters, if other existing criterion.

indication was subsequently lost.

3.2.4, Degraded Condition, Example (1) Ffollowing the end of cycle shutdown, ...

following the end of cycle shutdown, ... The event is reportable because the cladding failures The event is reportable because the cladding failures exceed expected values .fillll, are unique or widespread, exceed expected values, are unique or widespread, and and B:Pe oaused by unexpected faeto:PS.

are caused by unexpected factors.

Minor editorial change (capitalization). Whether the factors are unexpected or not is irrelevant to reportinR.

Page 26

3.2.4, Degraded Condition, Example (2)

The event is reportable beeauoe the degFade.tion eannot The event is reportable because the degradation cannot be eonsidePed aeooptable as is due to the generic be considered acceptable as-is. implications.

Many conditions identified by the licensee are not acceptable as-is. It is therefore more appropriate to indicate that the event is reportable due to the generic implications.

3.2.4, Degraded Condition, Example (3) Consequently, after discussion with the Regional Office, Consequently, after discussion with the Regional Office, the licensee appropriately retracted this event.

the licensee retracted this event.

fllustrates clarity as to the acceptability of the licensee's actions.

3.2.5, External Threat to Plant Safety, Discussion Usually, with the passage of time, it is apparent that an Usually, with the passage of time, it is apparent that an actual threat did not occur, and thus no LER is actual threat did not occur, and thus no LER is submitted submitted required (see Example 1). In some cases, (see Example 1). In some cases, with the passage of time, with the passage of time, it is judged that an actual it is judged that an actual threat did occur and, thus, an threat did occur and, thus, an LER is submitted LER is submitted (see Example 2). required (see Example 2).

The proposed change would enhance the clarity of this provision by specifying whether an LER was actually required to be submitted, not merely whether or not one was submitted.

3.2.5, External Threat to Plant Safety, Examples (2) It is proposed that these examples be del,eted. The and (3) licensee was required to make a report due to entry into (2) Hurricane the Emergency Plan, making reporting under this A licensee in southern Florida declared an Unusual Event category unnecessary.

after a hurricane warning was issued by the ...

(3) Fire With the unit at 100-percent power, the control room was notified that a forest fire was burning ...

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3.2. 7, Event or Condition That Could Prevent Fulfillment of a Safety Function, Discussion The intent of these criteria is to capture those events where there would have boon a failure of a safety 1'W indicated in tho Statement of Consido:Pe.ti:ons, The system to properly complete a safety function, intent of those criteria is to capture those events whore regardless of whether the system was needed at the there would have been a failure of a safety system to time of discovery. For example, if the high pressure properly complete a safety function, regardless of when safety injection system (both trains) failed, the event the failuroo wore clioooveFod OF whether tho oyotem was would be reportable even if there was no demand for the needed at the time.*For example. if the high pressure system's safety function.

safety injection system (both trains) failed. tho event would be reportable oven if there was no demand for tho If the event or condition 00\HEl- would prevent fulfillment system's safety function. of tho safety function at the time of discovery, it would be reportable under §50.72(b)(2)(v) (ENS notification).

If the event or condition could prevent fulfillment of the If it could have prevented fulfillment of the safety safety function at the time of discovery. it would be function at any time within three- NQ years of the date reportable under §50. 72(b)(2)(v) (ENS notification). If it of discovery, it would be reportable under §50. 73(b)(2)(v) could have prevented :fulfillment of tho safety function at (written LER).

any ti.mo within three years of tho date of discovery, it would be reportable under §50. 73(b)(2)(v) (written LER). Addition is needed to be consistent with the narrative in the subsequent paragraph. Replaces "could" with "would" as was previously recommended. As previously stated, we believe that two years is the appropriate time period for reporting historical events.

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8.2. 7, Event or Condition That Could Prevent This paragraph should be deleted. The interpretation Fulfillment of a Safety Function, Discussion that loss of offsite power is a condition that, of itself, could prevent fulfillment of a safety function is an overly Both offsite electrical power (transmission lines) and conservative and narrow view of a safety function. The onsite emergency power (usually diesel generators) are safety functions defined in the rule, (i.e., Shut down the considered to be separate functions by GDC 17. If either reactor and maintain it in a safe shutdown condition, offsite power or onsite emergency power is unavailable to remove residual heat, control the release of radioactive the plant, it is reportable regardless of whether the other material, or mitigate the consequences of an accident) system is available. GDC 17 defines the safety function can be fulfilled in the event of a loss of offsite power.

of each system as providing sufficient capacity and Loss of of/site power is governed by applicable Technical capability, etc., assuming that the other system is not Specifications, and actual losses are reported by ESF available. Loss of offsite power should be determined at actuation reporting (loss of offsite power results in the essential switchgear busses. actuation of Emertiency Diesels).

3.2.7, Event or Condition That Could Prevent These criteria cover an event or condition where Fulfillment of a Safety Function, Discussion structures, components, or trains of a safety system ee=ald would have failed to perform their intended These criteria cover an event or condition where functions ...

structures, components, or trains of a safety system could have failed to perform their intended functions ... If so, this would constitute an event that "effi:H0:- would have prevented" the fulfillment of a safety function, Last sentence of the paragraph following the paragraph on and, accordingly, must be reported.

GDC 17: If so, this would constitute an event that "could have prevented" the fulfillmAnt of a safety function, and, A design or analysis defect or deviation is reportable accordingly, must be reported. under this criterion if it ee-ald- would prevent fulfi11mfmt of the safety function of structures or systems defined in A design or analysis defect or deviation is reportable the rules.

under this criterion if it could prevent fulfillment of the safety function of structures or systems defined in the Replaces "could" with would" as was previously rules. recommended.

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3.2.7, Event or Condition That Could Prevent This portion of the NUREG contains guidance for Fulfillment of a Safety Function, Discussion determining whether a redundant train is operable and The staff believes that the conditions necessary to available. Operability determinations are performed by consider the redundant train operable and available, for the licensee using Generic letters 91-18 and 91-18 Rev. 1.

this purpose, should include the following: It is inappropriate to include operability determination

in cases where the redundant train should operate guidance in this reportability guideline. We recommend automatically, it is capable of timely and correct that the discussion be deleted, leaving only reference to automatic operation, or in cases where the redundant the generic letters in NUREG 1022. If this is not train should be operated manually, the operators acceptable, it is recommended that the guidance of the would detect the need for its operation and initiate generic letters be incorporated into NUREG 1022 and such operation, using established procedures for which that these generic letters be cancelled.

they are trained, within the needed time frame, without the need for troubleshooting and repair, and;

the redundant train is capable of performing its safety function for the duration required, and;

there is not a reasonable expectation of preventing fulfillment of the safety function by the redundant train 3.2. 7, Event or Condition That Could Prevent There are a limited number of single-train systems that Fulfillment of a Safety Function, Discussion perform safety functions (e.g., the High Pressure Coolant Injection System in BWRs). For such systems, There are a limited number of single-train systems that loss of the single train would only be reportable if it perform safety functions (e.g., the High Pressure Coolant resulted in an inability to fulfill pPev=eat the fulfillment Injection System in BWRs). For such systems, loss of the of the safety funotioa of that system aad, thePefoPe, is single train would prevent the fulfillment of the safety repol'te.ble evea though the plan.t teehnioal function of that system and, therefore, is reportable even speoifioatioas may allow sueh a eoaditioa to exist fop a though the plant technical specifications may allow such limited ti.me. a safety function (i.e., Shut down the a condition to exist for a limited time. reactor and maintain it in a safe shutdown condition, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.)

ReportinR HPCI inoverabilitv when it has no affect on Page 30

the ability to fulfill a safety function results in unnecessary burden to the licensee without a commensurate benefit in enhanced reactor safety. This is contrary to the cornerstone reasons stated for the revision to the rule and the NUREG.

If it can be shown that other safety systems, (e.g., RCIC, ADS) are available and capable of fulfilling the safety function without HPCI, the event should not be reportable.

3.2. 7, Event or Condition That Could Prevent No, a RCIC failure is not renortable. RCIC is included Fulfillment of a Safety Function, Example (2) in ISTS because it meets criterion 4 of 10CFR50.36 If the plant's safety analysis considered RCIC as a system based on its contribution to the reduction of overall needed to mitigate a rod ejection accident (e.g., it is nlant risk. RCIC is not credited in the nlant's safety included in the Technical Specifications) then its failure analysis. IT the plant's safety analysis eonsidered &CIC is reportable under this criterion; otherwise, it is not as a system B:eeded te BH:tigate a :Pod ajeetioB: aeeideB:t reportable under this section of the rule. (e.g., it is melaaed iB: the ~et1hnieal SpeemeatioB:S theB:

its faila:pe is :Pepo'Ptable nnae:11 this eatePieB:; othoPwi£le, it is B:ot reporiable nndo1:1 this seetioB: of tho l'\!le.

This clarification will correct problems associated with RCIC being included in TS, but not credited in a plant's sa{et_y anal_ysis.

Page 31

3.2. 7, Event or Condition That Could Prevent In addition to deleting the paragraph in the discussion Fulfillment of a Safety Function, Example (4) pertaining to GDC 17 criteria, we recommend deleting (4) Loss of Onsite Emergency Power by Multiple this example. It also is an overly conservative and Equipment Inoperability and Unavailability narrow view of a safety function considering that the TS allow for such a case and provide an AOT of two hours During refueling, one emergency diesel generator to restore this condition.

(EDG) in a two train system was out of service for maintenance. The second EDG was declared inoperable when it failed its surveillance test.

An ENS notification is required and an LER is required. As addressed in the Discussion section above, loss of either the onsite power system or the offsite power system is reportable under this criterion.

3.2.7, Event or Condition That Could Prevent This event is reportable because disabling the source Fulfillment of a Safety Function, Example (5) range detectors ooalEl would have prevented :fn)fillmAnt This event is reportable because disabling the source of the safety function to shut down the reactor.

range detectors could have prevented fulfillment of the safety function to shut down the reactor. Change "could" to "would" for reasons stated previously.

3.2. 7, Event or Condition That Could Prevent Thus, in this case, because of the use of the wrong Fulfillment of a Safety Function, Example (8) lubricant, the system "eould have" OP "would have~

Thus, in this case, because of the use of the wrong failed and is therefore reportable under 10 CFR lubricant, the system "could have" or "would have" failed. 50. 73(a)(2)(v).

Delete "could have" for reasons stated previously. Also should add that this is reportable; this appears to be an example of a 50. 73 report so it should be stated as such.

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3.2.7, Event or Condition That Could Prevent The independent failure (e.g., excessive set point drift)

Fulfillment of a Safety Function, Example (13) of a single pressure switch is not reportable unless it The independent failure (e.g., excessive set point drift) of alene- 00\liG would have caused a system to fail to fulfill a single pressure switch is not reportable unless it alone its safety function, or is indicative of a generic problem could have caused a system to fail to fulfill its safety that could have resulted in the failure of more than one function, or is indicative of a generic problem that could switch and thereby cause one or more systems to fail to have resulted in the failure of more than one switch and fulfill their safety function.

thereby cause one or more systems to fail to fulfill their safety function. Wording change is required to coincide with the proposed wording change (i.e. deleting "alone" and replacing could" with "would") .

3.2. 7, Event or Condition That Could Prevent . . . event or condition that alone oould would prevent Fulfillment of a Safety Function, Example (17) fulfillment of a safety function). (This repol'ting

... event or condition that alone could prevent fnlfil)mP.nt ei'iterion is dioouooed in Section 3.3.3 of this report.)

of a safety function). (This reporting criterion is discussed in Section 3.3.3 of this report.) Wording change is required to coincide with the proposed wording change (i.e. deleting "alone" and replacing "could" with "would'?. The line which refers to section 3. 3. 3 should be deleted. This section no longer exists in NUREG 1022.

3.2. 7, Event or Condition That Could Prevent Following evaluation of the condition, the event was Fulfillment of a Safety Function, Example (17- determined to be reportable because the HPCI eeula DUPLICATE) would have been prevented from performing its safety Following evaluation of the condition, the event was function ....

determined to be reportable because the HPCI could have been prevented from performing its safety function .... Change could" to "would" for reasoTtS stated previouslv.

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8.2.11, Contaminated Person Requiring Transport

A contract worker experienced a back injury lifting a Offsite, Example (1) tool while working in a contaminated area the roaetol' A contract worker experienced a back injury lifting a tool eonta.inment and was considered potentially while working in the reactor containment and was contaminated because his back could not be surveyed.

considered potentially contaminated because his back could not be surveyed. The example should be changed such that the empl.oyee was working in a contaminated area. In the given example, it is possible that the employee was working in a radiologically controlled area that was not contaminated. It could therefore be assumed that the worker was not contaminated, eliminating the need to make a report.

3.2.13, Loss of Emergency Preparedness We recommend that this example be deleted. It is Capabilities, Example (1) contrary to the guidance in the section on Loss of Offsite Response Capability that states: This does not apply in (1) Plant Access Roads Closed by Storm the case of routine traffic impediments such as fog, snow and ice which do not render the state and local The local sheriff notified the licensee that all roads to and from governments incapable of fulfilling their responsibilities.

the plant were closed because of a snow storm. The licensee It is intended to apply to more significant cases such as had two full-shift crews on site to support plant operations and no emergency declaration was made. The licensee notified the conditions around the Turkey Point plant after State and local authorities of the situation and made an ENS Hurricane Andrew struck in 1992 or the conditions notification. The licensee deactivated its station isolation around the Cooper station during the Midwest floods of procedures after the storm passed and the roads were 1993".

passable.

An ENS notification was made because the licensee determined that the road closing constituted a major loss of emerf!'encv offaite response capability. No LER is required.

Page 34

4.2.4 ENS Event Notification Worksheet (NRC The ENS Event Notification Worksheet (NRC Form Form 361) 361) is 8:B. attachment to IBfoHB:ation Notioe 89 89, The ENS Event Notification Worksheet (NRC Form 361) dated December 26, 1989, subjeet: Event Notifioation is an attachment to Information Notice 89-89, dated '3.lm~kshoots. Tho wo:Pkshoot provides the usual order of December 26, 1989, subject: Event Notification questions and discussion for easier communication and Worksheets. The worksheet provides the usual order of its use ...

questions and discussion for easier communication and its use ... Unless IN 89-89 is revised to incorporate a new NRG form 361, all reference to IN 89-89 should be deleted because the NRG Form 361 will be revised to reflect the channes to 10 GFR 50. 72 in this rulemakinn.

5.2.1, Narrative Description or Text (NRC Form For equipment that was inoperable at the start of the 366A, Item 1 7), (1) General event, provide an estimate of the time the equipment For equipment that was inoperable at the start of the became inoperable 8:B.d the last time tho equipment was event, provide an estimate of the time the equipment knoWB to be opePable. Indicate the basis for this became inoperable and the last time the equipment was conclusion (e.g., a test was successfully run or the known to be operable. Indicate the basis for this equipment was operating). For equipment that failed, conclusion (e.g., a test was successfully run or the provide the failure time 8:B.d the last time the equipment was operating). For equipment that failed, equipment was known to be ope:Pablo. Also provide the provide the failure time and the last time the equipment basis for the last time known operable.

was known to be operable. Also provide the basis for the last time known operable. Deletion of the last time equipment was known to be operable is necessary to conform with ISTS philosophy of equipment being operable since the last performance of the surveillance test (unless firm evidence exists to the contrary).

Page 35

Minor editorial changes to NUREG 1022, Revision 2 (draft)

SECTION/PROPOSED WORDING RECOMMENDED WORDING/COMMENTS 2.5, Time Limits for Reporting, For example, if a technician sees a problem, but a delay For example, if a technician sees a problem, but a delay occurs before an engineer or supervisor has a chance to occurs before an engineer or supervisor has a chance to review the situation, the discovery date (which starts review the situation, the discovery date (which starts the the 603Q..day clock) is the date that the technician sees 30-day clock) is the date that the technician sees a a problem.

problem.

Change is needed to align written guidance with the proposed chanee to required LER submission times.

2.5, Voluntary Reporting Instructions for voluntary ENS notifications and LERs Instructions for voluntary ENS notifications and LERs are discussed in Sections 4.2.2 and 5.1.5 of this report.

are discussed in Sections 4.2.2 and 5.1.5 this report.

Minor editorial chanee.

2.8, Retraction or Cancellation of Event Reports Example 3 in Section ~ 3.2.4 illustrates a case where Example 3 in Section 3.3.1 illustrates a case where there there were sound reasons for a retraction. The last were sound reasons for a retraction. The last event under ev=ent unde:r Example 1 in Seetion 3.3.2 illuotFates a Example" 1 in Section 3.3.2 illustrates a case where the ease where the roasono fop :rotFaoti:on we:re not reasons for retraction were not adequate. adequate.

Referenced sections need to be corrected to coincide with the appropriate sections in the Revision 2 of the NUREG.

In the latter example, the draft NUREG has deleted the discussion on the inappropriate retraction (see last event under Example 1 in Section 3.2.6; as such, either the strikeout that was done in that section needs to be restored OR the above second sentence needs to be deleted). NOTE: if the discussion on the inappropriate retraction is left in the discussion for the last event in Section 3.2. 6, then the above needs to reflect the appropriate section, i.e., ".... Example 1 in Section 3.2.6... "

Page 36

3.2.1, Plant shutdown required by TS If the shutdown is completed, licensees are required to If the shutdown is completed, licensees are required to submit an LER within 6000 days.

submit an LER within 30 days.

Change is needed to align written guidance with the proposed chanJte to required LER submission times.

3.2.1, Plant shutdown required by TS, Example (4) The plant lost one of two sources of offsite power due to The plant lost one of two sources of offsite power due to overheating in the main transformer. The TS allow 72 overheating in the main transformer. The TS allow 72 hours to restore the source or initiate a shutdown and hours to restore the source or initiate a shutdown and be be in HOT STANDBY...

HOT STANDBY...

Minor editorial chanl!e.

3.2.2, Technical Specification Prohibited Operation For a discrepancy discovered during a timely or Condition, TS Surveillance Requirements surveillance test (i.e., a surveillance test that was For a discrepancy discovered during a timely surveillance ~erformed within the TS reguired freguencv), it should test, it should be assumed that the discrepancy be assumed that the discrepancy occurred ...

occurred ...

Editorial addition for clarity. Although later in the NUREG the definition of "timely" is implied, it seems

,' prudent to specifically indicate what is meant by a "timely" surveillance test.

3.2.4, Degraded Condition In the ease sf a eempenent with a si{;Difiea.n.tly deg3caded In the case of a component with a significantly degraded ability te pel'fepm a safety fu:B:eti:en, whel:'e the eenditi:en ability to perform a safety function, where the condition eeuld afteettethel" similffi' eempeneB:ts m the plant, could affect/other similar components in the plant, lieensees Me Fequi.Ped te submit an. LER within 60 days.

licensees are required to submit an LER within 60 days.

As previously indicated, we recommend that this criterion be deleted. If retained, a minor editorial chanRe is required (extraneous "I '1.

3.2.7, Event or Condition That Could Prevent Examples (17), (18), (19) and (20) should be renumbered Fulfillment of a Safety Function, Examples (1 7), as (18), (19), (20) and (21) due to the duplication of (18), (19) and (20) example (17).

Page 37

3.2.10, Internal Threat to Plant Safety, Discussion In plant releases must be reported if they require In plant releases must be reported if they require evacuation of rooms or buildings and, as a result, the evacuation of rooms or buildings and, as a result, the ability of the operators to perform necessary duties for ability of the operators to perform necessary duties is safe operation of the plant is significantly hampered.

significantly hampered.

Editorial addition for safe operation of the plant" consistent with other parts of this section.

4.2.1, Timeliness See Section -2,.;-H 2.5 for further discussion of reporting See Section 2.11 for further discussion of reporting timeliness.

timeliness.

Minor editorial change due to renumbering of sections in theNUREG.

4.2.3, ENS Notification Retraction See section~ 2.8 for further discussion of retractions.

See section 2.10 for further discussion of retractions.

Minor editorial change due to renumbering of sections in theNUREG.

5.1.1, Submission of LERs See Section -2,.;-H 2.5 for further discussion of discovery See Section 2.11 for further discussion of discovery date. date.

Minor editorial change due to renumbering of sections in theNUREG.

5.1.4 Voluntary LERs See Section ~ 2. 7 for additional discussion of See Section 2.9 for additional discussion of voluntary voluntary LERs.

LERs.

Minor editorial change due to renumbering of sections in theNUREG.

5.2.1, Narrative Description or Text (NRC Form See Section -2,.;-H 2.5 for further discussion of discovery 366A, Item 17), (1) General date.

See Section 2.11 for further discussion of discovery date. Minor editorial change due to renumbering of sections in theNUREG.

Page 38

LFCREEK

'NUCLEAR OPERATING CORPORATION "99 EP 17 P2 :43 Otto L. Maynard President and Chief Executive Officer SEP 1 3 1999 WM 99-0051 Secretary DOCKET NUMBER U.S. Nuclear Regulatory Commission Washington, D.C . 20555-0001 PROPOSED RULE PR so J 7 :l-Attn.: Rulemaking and Adjudication's Staff ~l/ FR 31,,~c, I Reference : Federal Register Volume 64, No . 128, Pages 36291 - 36307, dated July 6, 1999

Subject:

Comments on Proposed Rule 10 CFR Parts 50 and 72, "Reporting Requirements for Nuclear Power Reactors,n and Draft NUREG-1022, Revision 2, "Event Reporting Guidelines - 10 CFR 50.72 and 50.73n Wolf Creek Nuclear Operating Corporation (WCNOC) appreciates the effort that the Nuclear Regulatory Commission (NRC) has put into revising 10 CFR 50.72 and 10 CFR 50.73 and allowing for industry involvement. Specifically, the table top exercises, held on November 13 , 1998, as discussed in the Reference page 36291 were of great benefit and allowed for an open exchange of information .

WCNOC supports the NRC's position to reduce or eliminate the reporting burden associated with events of little or no safety significance. The proposed rule has made progress toward this goal. However, there are specific portions of the proposed rule and associated NUREG-1022, Revision 2 that detract from this goal and if implemented may increase the reporting burden of licensee . This proposed rule was reviewed collectively by several utilities (ArnerenUE, WCNOC, Pacific Gas and Electric Company, STP Nuclear Operating Company, and TXU Electric) and our areas of concern include:

Significantly Degraded Components

Most significant is the last minute addition of a new reporting criterion for Significantly Degraded Component(s). WCNOC finds the new criterion as written to be unclear in focus and subject to widely varying interpretation. The attempt to capture items (components) which are seriously degraded , but not necessarily enough to render a system inoperable, is far below the current reporting threshold and represents a significant increase in licensee burden. Further , even if an operability statement is added to the criterion, as suggested at the tabletop, held on August 3, 1999, WCNOC contends that a component degradation significant enough to render a system inoperable would be captured by other reporting criteria , making this new criteria redundant and unnecessary.

UP 22-'---I___

~cknowtedqed by card ........................

191!

P.O. Box 411 / Burlington, KS 66839 / Phone: (316) 364-8831 An Equal Opportunity Employer M/F/HCNET

U.S. UCLEAR REGULATORY eotMSSION RULEMAKINGS &AOJlDCAlDfS STAFF OFFtCE OFHERETARY OFTHEC<WMISE~

WM 99-0051 Page 2 of 3 In addition, the new criterion as discussed by the staff at the table top exercise on August 3, 1999, appeared to have been added primarily as a data collection mechanism for motor operated valves. This is contrary to the stated objectives of the proposed rule which focuses on events of risk and safety significance. If component data is needed, there are other resources available to the NRC to obtain the information such as maintenance rule reports and the Equipment Performance Information Exchange (EPIX)

WCNOC recommends that the new criterion not be added to the rule. As discussed at the table top, held on August 3, 1999, significant component degradations that are risk and/or safety significant will be reported under other criteria such as loss of a function, common-mode failure, Part 21, or as a Technical Specification violation.

ESF Actuations

The addition of a specific list of systems which must be reported as Engineered Safety Features will increase .reporting by plants whose licensing basis does not include those specific systems. WCNOC recommends that we return to the pre-1998 practice of relying on each facility's Final Safety Analysis Report (FSARl and shift to a risk information approach when such criteria is fully developed. As part of the effort to risk inform Part 50, this section should be changed to be more risk informed

Invalid ESF actuation's are still included in the proposed rule change reporting requirements for the written LER. WCNOC recommends that the existing clarifications £0.r not .reporting certain Invalid actuations be retained in the guidance. Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event.

Specifically, actuations when the system is already properly removed from service, part of a planned evolution, that occur after the safety function has already been completed, or single component actuations of complex systems< which by themselves<< do not mitigate the consequences 0£ significant events, are not reportable.

LER Information

The scope of information requested for human performance events has increased by shifting from "personnel error" and the implied "root cause" to "human performance related problem" and "contributing factors."

more appropriate to require discussion of personnel error root causes.

It is WCNOC recommends deletion of the new criterion, 10 CFR 50. 73 (a) (3) (ii),

which requires a discussion of emergency operating procedures that could have been used to recover from an event be included in a LER. The proposed

.rule change would .result in a Lilrge amount of additional in£ormation that would be of minimal use. The safety consequences discussion would be cluttered with hypothetical failures and speculated plant responses.

Historical Limitations

WCNOC supports the new historical reporting criteria of three years for operations prohibited by the plant's Te.chnical Specification and conditions that could have prevented fulfillment of a safety function, but believes this limitation should be applied to the whole rule.

WM 99-0051 Page 3 of 3 In addition to the general comments above, WCNOC endorses the comments submitted to the NRC by the Nuclear Energy Institute.

Sincerely, OLM/jad cc: J. N. Donohew (NRC), w/a W. D. Johnson (NRC), w/a E.W. Merschoff (NRC)~ w/a Senior Resident Inspector (NRC), w/a

Florida Power & Light Compan P. 0. Box 14000, Juno Beach, FL 33408-0420 UCl{[,LO UStJ-

'99 SEP 14 P!2 :11 SEP - 7 1999 L-99-201 Ms. Annette L. Vietti-Cook DOCKET NUMBER Secretary PROPOSED RULE U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Rulemaking and Adjudication Staff Re: Florida Power & Light Company Comments Reporting Requirements for Nuclear Power Reactors (64 Fed. Reg. 36291 (July 6, 1999))

Dear Ms. Vietti-Cook:

Florida Power & Light Company (FPL), the entity licensed to operate the St. Lucie Nuclear Plant, Units 1 and 2, and the Turkey Point Nuclear Plant, Units 3 and 4, hereby submits the following comment on the above-referenced proposed rulemaking amending the event reporting requirements for nuclear power reactors.

In general, FPL agrees that the proposed change is intended to align reporting with needs, reduce burden where there is no safety significance and provide clarity to reporting. However, the rule as

presently proposed does not meet the stated objectives for the rulemaking. We have followed the development of the NEI comments on this rulemaking and endorse these comments, with particular emphasis on the significant concerns that are addressed.

FPL appreciates the opportunity to comment on the proposed rulemaking.

Sincerely yours,

?:a~g Manager Administrative Support and Special Projects an FPL Group company

UCLEAR R RlJLEMAKINGS OFFICEO OFTH CQMIMJSf;IQN Postma Copies 1'dr ~-- 1

-===-Entergy Entergy Operations, Inc.

P.O. Box 31995 Jackson , MS 39286-1995 (f)

99 UG 24 P3 :5 4 September 20, 1999 Of I Ms. Annette L. Vietti-Cook, Secretary U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn .: Rulemaking and Adjudications Staff

Subject:

Comments on Proposed Rule 10 CFR Parts 50 and 72, "Reporting Requirements for Nuclear Power Reactors," and Draft NUREG-1022, Rev. 2, "Event Reporting Guidelines - 10 CFR 50.72 and 50.73" Reference : Federal RegisterVol. 64, No. 128, Pages 36291 - 36307, dated July 6, 1999 CNRO-99/00022

Dear Ms. Vietti-Cook:

On July 6, 1999, the NRC published a proposed rule to amend the event reporting requirements for nuclear power reactors as contained in 10 CFR Sections 50.72, 50.73, and 72.216. Entergy Operations, Inc. (Entergy) appreciates the opportunity to comment on the proposed rule. Entergy supports the NRC's position to reduce or eliminate the reporting burden associated with events of little or no safety significance and to introduce risk-significance into the process. However, there are specific portions of the proposed rule and associated NUREG-1022, Revision 2 with which Entergy does not agree, either in whole or in part, as discussed in the accompanying attachments.

Although Entergy finds that certain changes will slightly reduce our reporting burden, we disagree with the NRC's assessment that the proposed rule represents an overall decrease in burden to the licensee based on the fo llowing points:

Although the proposed rule will decrease the number of phone-in reports pursuant to 10CFR50.72, Entergy believes this burden is very small when compared with the burden of processing and submitting Licensee Event Reports (LERs) pursuant to 10CFR50.73.

Per the proposed rule, systems that were excluded from reporting requirements via previous rulemaking because they represented little or no safety significance have been reinstated (e.g., Reactor Water Cleanup System). Such action will now lead to reporting all isolations, even those with no safety significance.

Entergy Operations, Inc.

1340 Echelon Parkway P.O. Box 31995 1999 Jackson, MS 39286-1995

.3, , 'UCLEAR GULATOR co M,Sb*~.

R LE AKINGS &ADJUOtCATI SSTAFF OFFICE OF THE SECREl OF THE COMMISS.

Pas Co.

Ad s

Ills

Comments on Proposed Rule 10 CFR Parts 50 and 72 September 20, 1999 CNR0-99/00022 Page 2 of 3

Systems that are not classified as Engineered Safety Features (ESF) will now be treated as ESF (e.g., Reactor Core Isolation Cooling System).

Invalid actuations are now included in the reporting requirements. The impact of this change is that the clarifications for what used to be reportable have been deleted.

Therefore, the proposed rule will treat all isolations or movements of a component as reportable regardless of safety significance.

The scope of information requested for human performance events has substantially increased, going well beyond previous direct root cause to now include associated contributing factors.

The imposition of a new requirement to report conditions involving degraded or non-conforming components with no regard to safety significance will result in a dramatic increase in burden adding no value to the process.

Entergy also finds that, contrary to the reasons given for the proposed rule, there is no indication of event reporting based on risk contained in the rule. Because of this, Entergy believes the proposed reporting process will convey a false elevated sense of problems to the general public causing undue alarm for situations that actually represent little or no safety or risk significance.

As expressed in the notice the NRC requested specific feedback in several areas. Our comments to these are presented below.

Proposed Reporting Times and Additional Levels (page 36295)

Entergy believes only three levels are needed. The designated three levels better define the significance of the initial report and should aid the NRC in responding to events.

Reporting Actuations of Prescribed Systems (page 36301)

Of the options listed in the notice, Entergy recommends Option 3 - return to the pre-1998 situation. This option, with the exclusions currently allowed, would be the easiest and most straightforward option to implement.

Reporting Historical Problems (page 36302)

Entergy believes reviews of historical events should be limited to 2 years or one operating cycle, whichever is shorter. This approach will enhance the importance of reported events while minimizing the burden to the licensee making the report. Entergy supports adding this limitation to all criteria that require a review of historical data.

Comments on Proposed Rule 10 CFR Parts 50 and 72 September 20, 1999 CNRO-99/00022 Page 3 of 3

Reporting Component Problems (page 36302)

Entergy has reviewed the proposed criterion and strongly recommends that it be deleted from the final rule. This criteria reinserts ambiguity and interpretation problems, lowers the significance level of the reporting criteria, adds a tremendous amount of burden back on the licensees and is redundant in many ways to other reporting mechanisms and criteria.

In addition to the general comments above and the specific comments provided in the attachments, Entergy supports the comments submitted to the NRG by the Nuclear Energy Institute (NEI) and the BWR Owners Group.

Again, thank you for the opportunity to provide our comments.

Sincerely, .

Director, Nuclear Safety &. Licensing MAK/GHD:ghd attachments cc: Mr. C. M. Dugger (VV-3)

Mr. W. A Eaton (GGNS)

Mr. R. K. Edington (RBS)

Mr. C. R. Hutchinson (ANO)

Mr.

Mr.

J. R. McGaha (ECH)

R. J. Fretz, NRR Project Manager, RBS N. D. Hilton, NRR Project Manager, ANO-1 M. C. Nolan, NRR Project Manager, AN0-2 C. P. Patel, NRR Project Mar:iager, Waterford-3 Mr. S. P. Sekerak, NRR Project Manager, GGNS

CNR0-99/00022 Attachment 1 Page 1 of 3 COMMENTS ON PROPOSED RULE 10 CFR Parts 50 and 72 SECTION / PAGE COMMENTS General Comment Licensees must use both the rule and NUREG-1022, Rev. 2 to determine reportability of conditions. The rule should be a stand-alone document written simple enough to be understood without the need for a 100+ page guidance document.

General Comment The terms "significantly affects" and "seriously degraded" are not defined anywhere in the proposed rule.

Page 36293, first column, In responding to Comment 7, the NRC states in part, "However, the Response to Comment 7 proposed amendments would not eliminate the requirement for a written report of an invalid actuation under 10 CFR 50. 73. There is still a need for reporting of invalid actuations because they are needed to make estimates of equipment reliability parameters, which in tum are needed to support the Commission's move towards risk-informed regulation.*

Contrary to the NRC's expectations, reporting invalid actuations will NOT provide the information needed to estimate equipment reliability parameters. Currently, and as proposed, licensees do not report system actuations that result from planned testing activities. However, these actuations must also be factored into reliability estimates to accurately reflect reliability. Also, this information may be collected via other less burdensome mechanisms, such as equipment performance and information exchange (EPIX) and Maintenance Rule reports.

In addition, reporting invalid actuations will convey a false elevated sense of problems to the general public causing undue alarm for situations that actually represent little or no safety or risk significance. Therefore, the new rule should NOT require invalid actuations to be reported.

Page 36295, third column, The NRC requested comments on the question of whether additional levels Response to Comment 26 should be introduced to better correspond to particular types of events.

Entergy believes only three levels are needed. The designated three levels better define the significance of the initial report and should aid the NRC in responding to events.

Page 36299, first column The example pertaining to missing or degraded fire barriers basically equates such conditions with degraded principal safety barriers (i.e., fuel cladding, reactor coolant pressure boundary, and containment). This is inappropriate and should be deleted.

CN RO-99/00022 Page 2 of 3 SECTION COMMENTS Page 36299, first and Two examples are given for steam generator tube degradation: (1)aburst second columns tube; and (2) a leaking tube.

In the first example, the guidance indicates that a reduction in safety factor margin is significant, yet there is no discussion or guidance for evaluating the system's ability to perform its safety function to determine safety significance. In addition, the 3.0 margin may not be the licensing basis for all PWRs.

The second example compares allowable leakage at power (1 gpm) to post-accident leakage. Expected post-accident leakage should be evaluated for impact on offsite dose. If the limits are not exceeded, no LER should be required.

Page 36301, first column The NRC requested comments pertaining to the proposed list of systems provided in §50.72(b)(2)(1v) and §50.73(a)(2)(Iv). Of the alternatives given, Entergy prefers option 3. This option with the exclusions currently allowed would be the easiest and most straightforward option to implement Page 36302, third column, The NRC is seeking comment on reporting historical problems Entergy

5. Reporting of Historical believes reviews of historical events should be limited to 2 years or one Problems operating cycle, whichever JS shorter. This approach will enhance the importance of reported events while minimizing the burden to the licensee making the report. Entergy supports adding this limitation to all criteria that require a review of historical data.

Page 36303,Section VI., The NRC claims 10CFR50.109 does not apply without giving any basis for Backfit Analysis the claim.

50. 72(a)(1 )(i) Delete footnote 8. §50.72(a)(1)(i) is clear as written.

50. 72(a)(4) Delete the last sentence. It is superfluous, adding no value to the discussion of reportability.

50. 72(b )(2)(ii)(B) The phrase "significantly affects plant safety" has no positive or negative connotation. Reword the section to read, "The nuclear power plant being in an unanalyzed condition that significantly degrades plant $afety."

50. 72(b)(2)(iv)(B) The system$ listed do not necessarily have a valid ESF logic actuation signal associated with them. Therefore, it could become confusing as to what should be reported as an actuation.

50. 72(b )(2)(iv)(B)(4) This section inappropriately assumes the Reactor Core Isolation Cooling system (RCIC) design basis is the same for all BWRs. This is not the case; for example, BWR/6 design does not credit RCIC for accident mitigation while earlier BWR designs may. A qualifying sentence should be added stating that listed systems should be credited in accident analyses or be risk-significant to be considered.

CNRO-99/00022 Attachment 1 Page 3 of 3 SECTION COMMENTS

50. 72(b )(2)(iv)(B)(6) This section should contain a qualifying statement that the listed systems are credited to mitigate accidents.

50. 72(b )(2)(iv)(B)(7) There is no benefit gained from reporting manual actuations of EOG support systems. Delete "and their associated support systems" from the section for EDGs and BWR dedicated Division 3 EDGs.

50. 72(b )(2)(iv)(B)(9) Revise this section to read, "Emergency service water function." The current wording would allow no reporting for plants that have emergency service water systems that run continuously.

50.73(a)(1) This section fails to recognize any risk significance factors.

50. 73(a)(2)(i)(B) Revise this section to read, "Any operation or condibon occurring within three years of the date of discovery which was prohibited by the plant's current Technical Specifications." This rewrite would direct plants that recently converted to Improved Standard Technical Specifications to apply the current requirements to the identified condition, rather than having to consider the previous requirements under old Technical Specifications which are no longer applicable.

50. 73(a)(2)(ii)(B) The phrase "significantly affects plant safety" has no positive or negative connotation. Reword the section to read, "The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety."

50. 73(a)(2)(ii)(C) This section should be deleted. This new requirement appears to be for data-collection only. The proposed section will result in a dramatic increase in burden with little or no value added. Component-based reporting contradicts system function-based reporting reflected in

50. 72(b )(2)(iv) and 50. 73(a)(2)(i)(B). At the very least, this section should be revised to require reporting conditions involving degraded or non-conforming components for which the associated system's safety function is significantly degraded, rather than the safety function of the component itself.

50. 73(a)(2)(iv)(B) The systems listed do not necessarily have a valid ESF logic actuation signal associated with them.

50. 73(a)(2)(ix)(J) This section greatly expands the scope of the current rule by requiring the licensee to investigate any human performance attributing cause. This new approach is a definite increase in burden.

50. 73(a)(3)(ii) This section requires the licensee to speculate on actions that could have been taken. This requirement adds significant burden with no added value.

CNRO-99/00022 Page 1 of 4 COMMENTS ON DRAFT NUREG-1022, Rev. 2 SECTION/ PAGE COMMENTS General Comment Most of the examples given in the NU REG involve very clear, specific conditions that are easily evaluated for reportability Unfortunately, these are not the ones licensees have trouble evaluating. More realistic examples should be provided. It would be beneficial to provide a bounding type explanation for each cntenon; i.e., a reportable condition that is almost not reportable and condition that is not reportable but almost could be.

ABBREVIATIONS Add "HPCS - High Pressure Core Spray," "LPCI - Low Pressure Coolant Injection," and "LPCS - Low Pressure Core Spray" to the list of abbreviations.

Remove NPRDS and SALP from the abbreviations list. They are not referenced or used in the text of the NU REG.

Page 1, Section 1. 1, last Revise the sentence to read, "Section 50. 73 requires wntten LERs to paragraph be submitted on reportable events within 60 days of their discovery."

The remainder of the sentence contains superfluous qualifiers.

Page 27, first paragraph The last sentence of the paragraph states, "Otherwise, the deficiency should be assumed to have occurred halfway between the last successful test or use and the untimely test that revealed the deficiency." What is the basis for the halfway time? The failure should be assumed to occur at the time the surveillance was due.

This condition should apply to violating Technical Specifications, only.

Revise the sentence to read, "Otherwise, the deficiency should be assumed to have occurred at the time the surveillance was originally due This condition applies to reporting Technical Specification violations only."

Page 28, Missed Tests Required Delete this example. It is confusing.

by ASME Section XI 1ST and ISi and.by TS 4 0.5, or Equivalent Page 29, Fire Protection Delete this example. It doesn't belong.

Systems When Required by TS Page 29, (2) Missed Surveillance This example is not consistent with the clarifications provided on Tests, second paragraph page 27. Revise the last sentence to read, "On the other hand, if the test showed the equipment was not capable of performing its specified safety functions and, thus, was inoperable in excess of the TS LCO allowed outage time, the event would be reportable.

CNRO-99/00022 Page 2 of 4 SECTION COMMENTS Page 30, (5) Seismic Restraints Delete this example. It adds no value.

Page 31, (6) Vulnerability to Make the example more definitive by revising the paragraph to read, Loss of Offsite Power "Assume that during a design review it is found that a loss of offsite power would cause a loss of instrument air and, as a result, auxiliary feedwater (AFW) flow control valves would fail open. Then for low steam generator pressure, such as would occur for certain main steam line breaks, high AFW flow rates would result in tripping the motor driven AFW pumps on them,al overload. The single turbine dnven AFW pump would not be affected.*

Page 33, Section 3.2.4, second Revise the paragraph to read, "In the case of a component with a paragraph significantly degraded ability that Inhibits the ability of Its associated system to perfonn its safety function, where the condition could affect/other similar components in the plant, licensees are required to submit an LER within 60 days.*

Conditions involving degraded or non-confom,lng components

should be reported only when they inh1b1t the ability of the associated system from performing its safety function.

Page 34, (3) Steam Generator Two examples are given for steam generator tube degradation: (1) a tube degradation in the following burst tube; and (2) a leaking tube.

circumstances In the first example, the guidar:ice indicates that a reduction in safety factor margin is significant, yet there is no discussion or guidance for evaluating the system's ability to perfom, its safety function to detem,ine safety significance. In addition, the 3.0 margin may not be the licensing basis for all PWRs.

The second example compares allowable leakage at power (1 gpm) to post-accident leakage. Expected post-accident leakage should be evaluated for impact on offsite dose If the limits are not exceeded, no LER should be required.

Page 37, (C) Significantly This requirement should be deleted. Imposing a new requirement to degraded components report conditions involving degraded or non-confom,ing components wtth no regard to safety significance will result in a dramatic increase in burden adding no value to the process.

Page 38, (1) Failures of Reactor Revise the sentence to read, "The event is reportable because the Fuel Rod Cladding Identified cladding failures exceed expected values, are unique, or During Testing of Fuel widespread." The factors, whether expected or unexpected, is Assemblies, second paragraph irrelevant to reportability.

Page 39, second paragraph Delete the paragraph. The "acceptableness" of the condition is not relevant to reportability. The degraded condition may need to be fixed but may be appropriately deferred for some period of time.

CNRO-99/00022 Page 3 of 4 SECTION COMMENTS Page 42, (2) Hurricane and (3) Delete these examples. They are not good examples because the Fire licensee would report the conditions because the Emergency Plan was entered Page 46, third remaining This paragraph states in part, "Invalid actuations are, by definition, paragraph those that do not meet the criteria for being valid. Thus, invalid actuations include actuations that are not the result of valid signals and are not intentional manual actuations."

What about loss of power to a trip unit (e g., a blown fuse)? In this situation, the trip unit would deactivate rather than receive and process an incoming signal. This condition should be construed as a non-safety response and, therefore, not reportable.

Page 47, first remaining Revise the paragraph to read, "If an im,<<alilil actuation reveals a paragraph defect in the system so the system failed or would fail to perform its intended safety function, the event continues to be reportable under other requirements of 10 CFR 50. 72 and 50. 73. When IR>.<<alld actuations excluded by the conditions described above occur as part of a reportable event, they should be described as part of the reportable event, in order to provide a complete, accurate and thorough description of the event."

A condition involving a system's failure to respond and perform its safety function when called upon should be reportable regardless of whether or not the signal is valid or invalid Page 50, (7) Actuation During Delete the entire paragraph. The remaining sentence provides Maintenance Activity, second guidance that bases reportab1hty solely on how the controlling paragraph procedure is written. This is an inappropriate approach.

Page 53, fourth paragraph This paragraph appears to contradict the rule as it applies to RCIC since RCIC is listed as a system but would fall out based on the guidance given in this paragraph. RCIC does not perform any of the identified functions (A) through (D).

Page 56, first paragraph The last sentence of the paragraph states, "For example, if a single RHR suction line valve should fail in such a way that RHR cooling cannot be initiated, the event would be reportable." This example is dependent on the specific mode of operation and Is too general as wntten. Delete the sentence.

Page 59, (5) Procedure Error This Is a bad example because the reactor is already shutdown in Prevents Reactor Shutdown Mode 5 and SRMs do not provide a trip function.

Function

CNRO-99/00022 Page 4 of 4 SECTION COMMENTS Page 61, (9) Oversized Breaker The basis for reportability is incorrect. Based on the information Wiring Lugs, Comment provided, only one (1) loose lug was found. Although the other lugs were of the improper size, they were securely fastened to the wire.

This event would NOT be reportable.

Page 64, (17) Multiple Control It is inappropriate to cite an NRC Information Notice to present a Rod Failures regulatory position since they are not enforceable.

Page 76, Section 3.2.11, (1) Revise the first sentence to read, "A contract worker experienced a Radioactively Contaminated back injury lifting a tool while working in tl:le i:ea&tgr GQRtaiRR18Rt !

Person Transported Offs1te for contaminated area and was considered potentially contaminated Medical Treatment because his back could not be surveyed.*

The worker must have been in a contaminated area to become contaminated or to be considered potentially contaminated.

Page 104 Section (1) begins by stating, "The cause(s), including any

[emphasis added] relation to the areas of:"

This is much too broad in scope to adequately address.

LER form Please provide the LER form in Microsoft Word as well as WordPerfect

DOCKET NUMBER PROPOSED RULE p 5 0 .,;. 1 ~

{pt/ FR 3G,~'1~

NUCLEAR REGULA TORY COMMISSION *99 JUL - 1 P4 :12 10 CFR Parts 50 and 72 RIN 3150-AF98 Reporting Requirements for Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission.

ACTION : Proposed rule .

SUMMARY

The Nuclear Regulatory Commission is proposing to amend the event reporting requirements for nuclear power reactors: to update the current rules, including reducing or eliminating the reporting burden associated with events of little or no safety significance; and to better align the rules with the NRC's needs for information to carry out its safety mission, including revising reporting requirements based on importance to risk and extending the required reporting times consistent with the time it is needed for prompt NRC action. Also, a draft report, NUREG-1022, Revision 2, is being made available for public comment concurrently with the proposed amendments.

~ i:101'199 DATES: Submit comments ~iAscrt dettc 76 days efter publieetior, if'I tFlc FeeleFel Register) .

Comments received after this date will be considered if it is practical to do so, but the Commission is able to ensure consideration only for comments received on or before this date.

ADDRESSES: Mail comments.to: Secretary, U.S. Nuclear Regulatory Commisl:>ion, Washington, DC 20555-0001. ATTN: Rulemakings and Adjudications Staff.

Deliver comments to: 11555 Rockville Pike, Rockville, Maryland, between 7:30 am and 4:15 p.m. Federal workdays.

Electronic comments may be provided via the NRC's interactive rulemaking website through the NRC home page (http://www.nrc.gov). From the home page, select "Rulemaking" from the tool bar at the bottom of the page. The interactive rulemaking website can then be accessed by selecting "Rulemaking Forum." This site provides the ability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, (301) 415-5905; e-mail CAG@nrc.gov.

Certain documents related to this rulemaking, including comments received, the transcripts of public meetings held, the draft regulatory analysis and the draft report NUREG-1022, Revision 2 may be examined at the NRC Public Document Room, 2120 L Street, N.W.,

(Lower Level), Washington, DC. These same documents also may be viewed and downloaded electronically via the interactive rulemaking web site established by NRC for this rulemaking.

FOR FURTHER INFORMATION CONTACT: Dennis P. Allison, Office of Nuclear Reactor Regulation, Washington, DC 20555-0001, telephone (301) 415-1178, e-mail dpa@nrc.gov.

SUPPLEMENTARY INFORMATION:

Contents I. Background 2

II. Rulemaking Initiation Ill. Analysis of Comments IV. Discussion

1. Objectives of Proposed Amendments

2. Discussion of Proposed Amendments

3. Revisions to Reporting Guidelines in NUREG-1022

4. Reactor Oversight

5. Reporting of Historical Problems.

8. Reporting of Component Problems.

7. Enforcement

8. Electronic Reporting

9. Schedule

10. State Input V. Environmental Impact Categorical Exclusion VI. Backfit Analysis VII. Regulatory Analysis VIII. Paperwork Reduction Act Statement IX. Regulatory Flexibility Certification X. Proposed Amendments I. Background Section 50.72 has been in effect, with minor modifications, since 1983. Its essential purpose is 11

- - - to provide the Commission with immediate reporting of .... significant events 3

where immediate Commission action to protect the public health and safety may be required or where the Commission needs timely and accurate information to respond to heightened public concern." (48 FR 39039; August 29, 1983).

r Section 50. 73 has also been in effect, with minor modification, since 1983. Its essential purpose is to identify" .... the types of reactor events and problems that are believed to be significant and useful to the NRC in its effort to identify and resolve threats to public safety. It is designed to proyide the information necessary for engineering studies of operational anomalies and trends and patterns analysis of operational occurrences. The same information can be used for other analytic procedures that will aid in identifying accident precursors."

(48 FR 33851; July 26, 1983).

II. Rulemaking Initiation Experience has shown a need for change in several areas. On July 23, 1998 (63 FR 39522) the NRC published in the Federal Register an advance notice of proposed rulemaking (ANPR) to announce a contemplated rulemaking that would modify reporting requirements for nuclear power reactors. Among other things, the ANPR requested public comments on whether the NRC should proceed with rulemaking to modify the event reporting requirements In 10 CFR 50.72, "Immediate notification requirements for operating nuclear power reactors,"

and 50.73, "Ucensee event report system," and several concrete proposals were provided for comment A public meeting was held to discuss the ANPR at NRC Headquarters on August 21, 1998. The ANPR was also discussed, along with other topics, at a public meeting on the role of industry in nuclear regulation in Rosemont, Illinois on September 1, 1998. The public 4

comment period on the ANPR closed on September 21, 1998. A comment from the Nuclear Energy Institute (NEI) proposed conducting "table top exercises" early in the development and review process to test key parts of the requirements and guidance for clarity and consistency.

That comment was accepted and a third public meeting was held on November 13, 1998 to discuss issues of clarity and consistency in the contemplated approach. Transcripts of these meetings are available for inspection in the NRG Public Document Room or they may be viewed and downloaded electronically via the interactive rulemaking web site established by NRG for this rulemaking, as discussed above under the heading "ADDRESSES." Single copies may be obtained from the contact listed above under the heading "For Further Information Contact" Ill. Analysis of Comments The comment period for the ANPR expired September 21, 1998. Twenty-one comment letters were received, representing comments from sixteen nuclear power plant licensees (utilities), two organizations of utilities, two States and one public interest group. A list of comment letters is provided below. The comment letters expressed support for amending the rules along the general lines of theobjectives discussed in the ANPR. Most of the letters also provided specific recommendations for changes to the contemplated amendments discussed in the ANPR. In addition to the written comments received, the ANPR has been the subject of three public meetings as discussed above under the heading "Background," and comments made at those meetings have also been considered.

The resolution of comments is summarized below. This summary addresses the principal comments (i.e., comments other than those that are: minor or editorial in 5

nature; supportive of the approach described in the ANPR; or applicable to another area or activity outside the scope of sections 50.72 and 50.73).

Comment 1: Several comments recommended amending 10 CFR 50. 73 to allow 60 days Onstead of the current 30 days) for submittal of Licensee Event Reports (LERs). They indicated that this would allow a more reasonable time to determine the root causes of events and lead to fewer amended reports.

Response: The comments are accepted for the reason stated above. The proposed rule would change the time limit to 60 days.

Comment 2: Two comments suggested a need to establish starting points for reporting time ciocks that are clear and not subject to varied interpretations.

Response: The reporting guidelines in this area have been reviewed for ciarity. Some editorial clarifications are proposed in Section 2.5 of the draft of Revision 2 to NUREG-1022, which is being made available for public comment concurrently with the proposed rule, as discussed below under the heading "Revisions to Reporting Guidelines in NUREG-1022."

Comment 3: Many comments opposed adopting a check the box approach for human performance and other information in LERs (as was proposed in the ANPR, with the objective of reducing reporting burden). They indicated that adopting a check the box approach would result in substantial implementation problems, and recommended continuing to rely on the narrative description which provides adequate information. One comment' opposed the idea of a check the box approach on the grounds that it would make LERs more difficult for the general public to understand. A few comments supported the check the box approach.

Response: The intent of the check the box approach was to reduce the effort required in reporting; however, the majority of comments indicate this would not be the case.

Accordingly, the proposed rule does not reflect adoption of a check the box approach.

6

Comment 4: Several comments opposed codifying the current guidelines for reporting human performance information in LERs (Le., adding the detailed guidelines to the rule, as was proposed in the ANPR). They recommended leaving the rule unchanged in this regard, indicating that sufficient information is being provided under the current rule and guidelines.

Response: The comments are partially accepted. The proposed rule would not codify the reporting guidelines (as proposed in the ANPR) for the reasons stated above.

However, the proposed rule would simplify the requirement. It is not necessary to specify the level of detail provided in the current rule. Accordingly, the amended paragraph would simply require a discussion of the causes and circumstances for any human performance related problems that contributed to the event. Details would continue to be provided in the reporting guidelines, as indicated in Section 5.2.1 of the draft of Revision 2 to NUREG-1022. This draft report is being made availac:s for public comment concurrently with the proposed rule, as discussed below under the heading "Revisions to Reporting Guidelines in NUREG-1022."

Comment 5: Several comments opposed codifying a list of specific systems for which actuation must be reported (by naming the systems in 10 CFR 50.72 and 50.73, as was proposed in the ANPR). They indicated that a system's contribution to risk can vary widely from plant to plant, which precludes construction of a valid universal list They recommended that, instead, actuation be reported only for those systems that are specified to be engineered safety features (ESFs) in the final safety analysis report (FSAR).

Response: The proposed rule would include a list of systems for which actuation would be reported. However, the concern is recognized and public comment will be specifically invited on several alternatives to the proposed rule.

7

Comment 6: Several comments opposed changing the criteria in 10 CFR 50. 72 and 50.73 which require reporting any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems .. .. The change proposed in the ANPR would have substituted the phrase "alone-or in combination with other existing conditions" for the word "alone" in this criterion. The comments indicated that this would add confusion, the rule as currently worded is sufficiently clear, and the need to consider other existing plant conditions in evaluating reportability is understood and uniformly implemented.

They recommended leaving the rule unchanged in this regard.

Response: The comments are partially accepted. The requirement would not be changed by substituting the phrase "alone *or in combination with other existing conditions" for the word "alone" in this criterion (as proposed in the ANPR).

However, the proposed amendments would change the rules by deleting the word "alone," so that they would require reporting "any event or condition that could have prevented fulfillment of the safety function of structures or systems ... " This would simplify the wording, rather than making it more complicated. It is not intended to change the meaning of the requirement, but to make the meaning more apparent in the wording of the rule. The following points, which are relevant to this question, would continue to be made clear in the reporting guidelines. See Section 3.2.7 of the draft of Revision 2 to NUREG-1022, which is being made available for public comment concurrently with the proposed rule, as discussed below under the heading "Revisions to Reporting Guidelines in NUREG-1022."

(1) It is not necessary to assume an additional random single failure in evaluating reportability. (If such an assumption were necessary, inoperability of a single train would generally be reportable *under this criterion.)

8

(2) It is necessary to consider other existing conditions in determining reportability. (For example, if Train A fails at a time when Train B is out of service for maintenance, the event is reportable.)

(3) The event is reportable regardless of whether or not a system was called upon to perform its safety function. (For example, if an emergency core cooling system [ECCSJ was incapable of performing its specified safety functions, the event is reportable even if there was no call for the ECCS function.)

(4) The event is reportable regardless of whether or not a different system was capable of performing the safety function. (For example, if the onsite power system failed, the event is reportable even if the offsite power system was available and capable of performing its ~afety

' *' ~

functions.)

Comment 7: Several comments recommended changing 10 CFR 50.72 and 50.73 to exclude reporting an invalid actuation of an ESF. (An invalid actuation is one that does not.

result from a plant condition that warrants ESF initiation.)

Response: The comments are partially accepted. The proposed amendments would eliminate the requirement for telephone notification of an invalid actuation under 10 CFR 50.72. Invalid actuations are generally less significant than valid actuations because they do not involve plant conditions (e.g., low reactor coolant system pressure} conditions that would warrant system a_ctuation. Instead, they result from other causes such as a dropped electrical lead during testing).

However, the proposed amendments would not eliminate the requirement for a written report of an invalid actuation under 10 CFR 50.73. There is still a need for reporting of invalid actuations because they are needed to make estimates of equipment reliability parameters, which in tum are needed to support the Commission's move towards risk-informed regulation.

9

This is discussed further in a May 7, 1997 Commission paper, SECY-97-101, Proposed Rule, 10 CFR 50.76, Reporting Reliability and Availability Information for Risk-significant Systems and Equipment," Attachment 3.

Comment 8: Several comments recommended changing 10 CFR 50.72 and 50.73 to limit certain reports to current events and conditions. That is, they recommended that an event or condition that could have prevented the fulfillment of the safety function of structures or systems .... be reported:

(1) by telephone under 10 CFR 50. 72(b)(2)(iii) only if it currently exists, and (2) by written LER under 10 CFR 50.73(a)(2)(v) only if it existed within the previous two years.

)

For a "historical" event or condition of this type (i.e., one which might have been significant at one time but has since been corrected) there is less significance than there is for a current event and, thus, immediate notification under 50.72{b)(2)(iii) is not warranted. With regard to 50.73(a)(2)(v), two years encompasses at least one operating cycle. Considerable resources are expended when it is necessary to search historical records older than this to make past operability determinations, and this is not warranted by the lesser significance of historical events older than two years.

Response: The comments are partially accepted, for the reasons stated above. That is, under the proposed rules,, an event or condition that could have prevented the fulfillment of the safety function of structures or systems ... would be reported by telephone under 10 CFR 50.72{b)(2)(iii) only if it exists at the time of discovery. An event or condition that could have prevented the fulfillment of the safety function of structures or systems .... would be reported by written LER under 10 CFR 50.73(a)(2)(v) only if it existed within the previous three years.

10

In addition, alL1ough not recommended in the comments, under the proposed rule an operation or condition prohibited by the plant's Technical Specifications would be reported under 50.73(a)(2)(i)(B) only if it existed within the previous three years. For this criterion as well, considerable resources are expended when it is necessary to search historical records older than three years to make past operability detem,inations, and this is not warranted by the lesser significance of historical events older than three years.

Three years is proposed, rather than two years as suggested in the comments, because the NRC staff trends plant perfom,ance indicators over a period of three years to ensure inclusion of penods of both shut down and operation.

Comment 9: Several comments _opposed using the tern, risk-significant (or significant) in the absence of a clear definition.

Response: The tern, "significant" would be used in two criteria in the proposed rules.

In the first criterion, sections 50.72 and 50.73 would require reporting an unanalyzed condition that significantly affects plant safety. In this context the tern, "significant would be defined by examples, five of which are discussed below under the heading "Condition that is outside the design basis of the plant." In the second criterion, section 50.73 would require reporting when a component's ability to perfom, its safety function is significantly degraded and the condition could reasonably be expected to affect other similar components in the plant. Again, the tern, "significant would be defined by examples, six of which are discussed below under the heading "Significantly degraded components."

Comment 10: Several comments recommended changing 10 CFR 50.72 and 50.73 to exclude reporting of an unanalyzed condition that significantly compromised plant safety on the basis that it is redundant to other reporting criteria.

11

Response: The comment is not accepted. Several types of wor.thwhile reports have been identified that could not readily be captured by other criteria as discussed further below under the heading "Condition that is outside the design basis of the plant."

Comment 11: Several comments recommended amending 10 CFR 50. 72 and 50. 73 to exclude reporting of a seriously degraded principal safety panier on the basis that it is redundant to other reporting criteria.

Response: The comments are not accepted. This criterion captures sor:ne worthwhile reports that would not be captured by other criteria, such as significant welding or material defects in the primary coolant system. However, some. clarifications are proposed in Section 3.2.4 of the draft reporting guidelines, to better indicate which events are serious enough to qualify for reporting under this criterion.

Comment 12: One comment recommended that, with regard to a condition or operation prohibited by the plant's Technical Specifications, reporting should be eliminated for violation of all administrative Technical Specifications.

Response: The comment is partially accepted. The proposed rule would eliminate

.reporting for Technical Specifications that are administrative in nature. The reporting guidelines would not change. As stated in the current reporting guidelines in NUREG-1022, Revision 1, failure to meet administrative Technical Specifications requirements is reportable only if it results in violations of equipment operability requirements, or had a similar detrimental effect on a licensee's ability to safely operate the plant For example, operation with less than the required number of people on shift would constitute operation prohibited by the Technical Specifications. However, a change in the plant's organizational structure that has not yet been J

approved as a Technical Specification change would not An administrative procedure violation or failure to implement a procedure, such as failure to lock a high radiation area door, 12

that does not have a direct impact on the safe operation of the plant, is generally not reportable under this criterion.

Comment 13: One comment recommended changing 10 CFR 50.73 to require that LERs identify: (1) how many opportunities to detect the problem were missed and (2) corrective actions to prevent future misses.

Response: No changes are proposed. If missed opportunities are identified and are significant to the event, they should be captured by the current requirements to provide a comprehensive description of the event and to describe corrective actions if they are significant to the event.

Comment 14: With regard to design issues, one comment recommended including language in the rules or their statements of considerations encouraging a voluntary report under 10 CFR 50.9 for a newly discovered de,:,;~i1 issue which is not otherwise reportable at the plant where first discovered (because the affected systems can still perform their specified safety functions) but which might have a significant impact on generic design issues at other plants.

Response: A statement encouraging submittal of voluntary LERs is included in the reporting guidelines. In addition, the guidelines would indicate that any significant degradation that could reasonably be expected to affect multiple similar components in the plant should be reported.

Comment 15: Several comments opposed placing a condition, related to systematic non-compliance, on the elimination of reporting of late surveillance tests (as proposed in the ANPR) under 10 CFR 50.73. The condition would be burdensome because licensees would n~ed to track instances of missed surveillance tests in given time periods.

13

Response: The proposed rule does not contain this condition. Reporting for the I

purpose of identifying systematic non-compliance is not needed because NRC resident inspectors routinely review plant problem lists, and thus would be aware of any systematic non-compliance in this area if it occurs.

Comment 16: One comment recommended changing the rules to allow licensees to rely on notifications made to resident inspectors, which could eliminate the need to make a telephone notification via the emergency notification system (ENS) and/or submit a written LER, at l~ast for some events or conditions. They indicated, for example, this should be adequate where the event is a decision to issue a news release.

Response: No changes are proposed. Telephone notifications to the NRC Operations Center, when required, are needed to ensure that the event can be promptly reviewed. This includes notification of the NRC Headquarters Emergency Officers and the Regional Duty Officer and consideration of whether to activate NRC incident response procedures. Written LERs, when required, are needed to ensure that events can be systematically reviewed for safety significance.

Comment 17: Some comments opposed amending 10 CFR 50.73 to require additional' information regarding equipment availability for shutdown events (as proposed in the ANPR) to support staff probabilistic risk assessments (PRAs). They indicated that it is rare that sufficient information is not available in an LER.

Response: The proposed rule would require such information. Frequently, when shutdown events are subjected to a probabilistic risk analysis, it is necessary to call the plant to determine the status of systems and equipment. The proposed rule would eliminate much of that need.

14

Comment 18: Several comments recommended deleting 10 CFR 50.72(b)(2)(i), "Any event found while the reactor is shut down, that, had rt been found while the reactor was in operation, would have resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded or being in an unanalyzed condition that significantly compromises plant safety." The comments Lndicated that because the plant would be shutdown, there is no need for immediate NRC action.

Response: The requirement for telephone reporting would not be entirely eliminated because, if a principal safety barrier is significantly degraded or a condition that significantly affects plant safety exists; the event may be significant enough that the NRC would need to initiate actions [such as contacting the plant to better understand the event and/or initiating a special inspection or investigation] within about a day even if the plant is shutdown.

However, in the proposed rule this specific criti;;non would be combined with 10 CFR 50.72(b)(1)(ii), "Any event or condition during plant operation that results in the condition of the nuclear power plant, including its principal.safety barriers, being seriously degraded or ... "

Also, the term "unanalyzed condition that significantly compromises plant safety" would be deleted. In combination with other changes, this would result in the folloyving crit~rion for telephone notification "Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded."

Comment 19: Some comments recommended that the NRC use enforcement discretion during the rulemaking process to provide early relief with regard to reporting a condition outside the design basis of the plant and/or a late surveillance test (condition or operation prohibited by Technical Specifications).

Response: The current rules will continue to apply until final revised rules are issued and become effective. However in dispositioning any violation, the risk- and safety-15

significance of the violation will be an important consideration. Establishing an interim enforcement discretion policy would involve the same critical elements as developing the revised rule and guidance including a provision for public comment This would complicate the rulemaking process, and essentially constitute a prediction of its final outcome, which may or may not tum out to be correct.

Comment 20: Several comment letters opposed the idea of tying enforcement criteria

{i.e., violation severity levels) to reporting criteria. They indicated this could have an unintended adverse effect on reporting and the resources consumed because in matching an event with a reporting criterion, a licensee would essentially be forced to make a preliminary determination of severity level.

) Response: The comments are not accepted. The proposed changes to the enforcement criteria, are discussed below under the heading "Enforcement."

Comment 21: As requested by the ANPR, a number of comments identified reactor reporting requirements other than sections 50.72 and 50.73 where changP-s are warranted.

Response: Comments regarding changes to reactor reporting requirements other than sections 50.72 and 50.73 will be addressed in a separate action. A Commission paper on that subject was submitted on January 20, 1999, SECY-99-022, "Rulemaking to Modify Reporting Requirements for Power Reactors" and the Commission issued a Staff Requirements Memorandum on March 19, 1999 directing the staff to proceed with planning and scheduling.

Comment 22: One comment recommended changing the required initial reporting time for some events to ".... within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or by the beginning of the next business day," instead of simply specifying ".... within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ." The comment indicated it does not appear that the NRG takes action on these events during non-business hours.

16

Response: The comment is not accepted. The NRC needs these reports in time to call the plant to find out more about the event and/or initiate a special inspection or an investigation, if warranted, within a day. Sometimes these actions are taken during non-business hours.

Comment 23: One comment recommended that an event or condition that could have prevented fulfillment of the safety function of structures or systems .... should be reportable only when the time limits of the TS are exceeded. It indicated that if the time limits are not exceeded the event is not significant enough to warrant reporting.

Response: The comment is not accepted. Generally, standard TS require commencement of shutdown within one hour if an important system, such as emergency ac power, is inoperable. However, the stated reason for allowing one hour before commencing the shutdown is to provide time to prepare for an orderly shutdown. Also, the condition might have lasted much longer than one hour before it was discovered. Finally, an event that results in a safety system failure (or inability to perform its function) is generally significant enough to j

warrant NRC review.

Comment 24: One comment from the State of Ohio recommended that, although rule changes are not necessary, efT!phasis should be place'd on positive notification of State and local agencies of emergency conditions before calling the NRC.

Response: The comment is accepted. It arose from a weakness in the N RC's response to an event at the Davis-Besse plant Because there were considerable difficulties in establishing telephone communications with the plant at the time of the event, NRC Operations Center personnel requested that the licensee remain on the line and said that the NRC would notify the State. However, the NRC did not do so in a timely manner. Training and procedure changes have been implemented to ensure this type of problem will not reoccur.

17

Comment 25: One comment letter, from the State of Illinois, stated the following: "In section 50. 72 of the advance notice of proposed rulemaking, seven non-emergency events listed as (f), are proposed to be reported in eight hours instead of one hour. Of those seven events, six {specifically, (ii), {iii), (iv), (v), (vi), and (vii)) would probably be classified as emergency events under existing emergency plans at an Illinois site .. .. This will cause reporting confusion during an event at a time when clarity is necessary. These six events should all be reported as emergency events, not non-emergency events. EAL thresholds in licensee emergency plans should be required to reflect them clearly. All of these events would affect the State of Illinois' response and our emergency plans. NRC must reconsider the categories of non-emergency events in the context of the current guidance to licensees for classifying EALs to ensure there is a clear distinction between emergency and non-emergency reportable events."

1 Response: Section 50.72 has been reviewed, and appears to be clear in this regard. It indicates the following:

(1) any declaration of an Emergency Class is reportable pursuant to 10 CFR 50.72(a)(1)(i) and (a)(3),

(2) the conditions listed in paragraph (b)(1), "One-hour reports," are reportable pursuant to paragraph (b)(1) if not reported as a declaration of an- Emergency Class under paragraph ffil, and (3) the conditions listed in paragraph (b){2), "Eight-hour reports, are reportable pursuant to paragraph (b)(2), if not reported under paragraphs (a) or (b}(1).

Comment 26: One comment letter, from the State of Illinois, opposed relaxing-the required initial reporting time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for the following types of events:

18

0) Airborne radioactive release that results in concentrations over 20 times allowable levels in an unrestricted area; (ii) _Liquid effluent in excess of 20 times allowable concentrations released to an unrestricted area; (iii) Radioactively contaminated person tran~ported to an offsite medical facili~ for 1treatment; (iv) News release or other government agency notification related to the health and safety of the public or onsite personnel, or protection of the environment.

The comment further indicated: "It is of paramount importance that those charged with regulating and monitoring the public impact of radiological releases are being kept informed of unplanned releases in a timely manner. Illinois law requires that we perform independent assessments, decide what actions may be necessary to protect the public, and assist in informing the public regarding any radiological risk. Should follow-up action to a release be necessary, then the less time that has elapsed, the better the state is able to respond in a timely and appropriate manner. We oppose any reduction in notification requirements for unplanned radiation releases from a site regardless of the source or quantity.

Timeliness is also important for items of obvious public interest. News of seemingly small events spreads quickly, particularly in local communities around the power plants.

Delayed reporting of such events means that we will be unprepared to respond to queries from local officials, or the media, with a resultant loss of public confidence. Therefore, we also oppose any reduction in notification requirements for newsworthy events."

Response: In the interest of simplicity, the proposed amendments would maintain just three basic levels of required reporting times in 10 CFR 50. 72 and 50. 73 (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and 60 days). However, the concern is recognized and public comment is specifically invited on 19

the question of whether additional levels should be introduced to better correspond to particular types of events, as discussed below under the heading "Required lnibal Reporting Times." Also, if in a final rule the NRC should relax the time limit to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , a State would not be precluded from obtaining reports earlier than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Comment 27: Two comment letters addressed coordination with States. The comment letter from Florida Power & Light Company stated The NRC's Public workshop on August 21, 1998, touched on a number of examples where opportunities exist to reduce reporting burdens. An industry representative commented that licensees sometimes have to report the same event to state agencies and the NRC provided one such example. FPL concurs with the recommendation that the time requirement for reporting an event to the NRC and to the state should be consistent wherever practical and possibly in some cases eliminated."

The com*ment letter from Northeast Nuclear Energy Company stated "Northeast Nuclear Energy Company agrees with extending the non-emergency prom.pt notifications to eight hours. This would help to eliminate unnecessary reports and retractions. However, it is necessary to have the individual states closely involved with the rule change since they may have requirements that are more restrictive or conflict with the proposed rulemaking. For example, in Connecticut all 10 CFR 50. 72 reports require notification of the state within one hour."

Response: The ANPR specifically requested State input. In addition, a letter requesting input was sent to each State. Written comments were received from the State of Ohio and the State of Illinois. In addition, representatives from several States attended one of the public meetings on the ANPR. The NRC will continue to solicit State input as the rulemaking process proceeds.

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Comment 28: One comment recommended eliminating two of the requirements for immediate followup notification dunng the course of an event, section 50.72(c)(2)(i), the results of ensuing evaluations or assessments of plant conditions, and section 50.72(c)(2)(ii), the effectiveness of response or protective measures taken. The comment indicated that the requirements continue to apply after the event and that they require reporting even if, for example, the result of a further analysis does not change the initial report.

Response: The comment is not accepted. The requirements for followup reporting apply only during. the course of the event. Followup reports~are needed while the event is

ongoing. For example, if an analysis is completed during an ongoing event, and it confirms an earlier estimate of how long it will take to uncover the reactor core if electric power is not restored, that information may very well be useful for the purpose of evaluating the need for protective measures (evacuation).

Comment 29: One comment recommended clarifying the reporting requirements for problems identified by NRC inspectors.

Response: No changes are proposed. The current.reporting guidelines include a paragraph making it clear that an event must be reported via telephone notification and/or written LER, as required. regardless of whether it had been discussed with NRC staff personnel or was identified by NRC personnel.

Comment 30: Several comments recommended changing the requirements in 50.46(a){iii)(2) for reporting errors in or corrections to ECCS analyses.

Response: These comments will be addressed in a separate action (along with other comments on reporting requirements other than sections 50. 72 and 50. 73).

Comment 31: Some comments raised issues regarding plant-specific reporting requirements contained in Technical Specifications (or other parts of the operating license).

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One suggestion was that 10 CFR 50. 72 and 50. 73 should be changed to address these issues. Another suggestion was that a Generic Letter be issued indicating that the NRC would be receptive to requests for license amendments to eliminate specific reporting requirements.

Response: No changes are proposed for sections 50.72 and 50.73, which identify generic reporting requirements. It is not feasible or appropriate to address the specific reporting requirements contained in individual operating licenses in this format.

The idea of issuing a generic communication to specific requests for license amendments wilf be addressed (i;il_<;mg with other comments on reporting requirements beyond the scope of sections 50,72 and 50.73) in a separate action.

Comment 32: One comment recommended that in section 50.72(b)(1){v), the word "offsite" be added before "communications capability" to make it clear that what must be reported is a loss of communications with outside agencies, not internal plant communications systems.

Response: The comment is accepted. In the proposed rule the word "offsite" would be added.

Comment 33: Several comments suggested that the NRC should define its needs relative to the information provided in LERs.

Response: The essential purpose of the LER rule is to identify the types of reactor events and problems that are believed to be significant and useful to the NRC in its effort to identify and resolve threats to public safety. The rule is designed to provide the information necessary for engineering studies of operational anomalies and trends, and patterns analysis of operational occurrences. To this end, the information required in LERs is generally needed to understand the event, its significance, and its causes in order to determine whether generic 22

or plant specific action is needed to preclude recurrence. Some further specific functions are discussed below.

It is necessary to identify and analyze events and conditions that are precursors to potential severe core damage, to discover emerging trends or patterns of potential safety significance, to identify events that are important to safety and their associated safety concerns and root causes, to determine the adequacy of corrective actions taken to address the safety concerns, and to assess the generic applicability of events.

The NRG staff reviews each LER to identify those individual events or generic situations that warrant additional analysis and evaluation. The staff identifies repetitive events and failures and situations where the frequency or the combined significance of reported events may be cause for concern. The NRG staff reviews past operating history for similar events and initiates a generic study, as appropriate, t:; focus upon the nature, cause, consequences and possible corrective actions for the particular situation or concern.

The NRG staff uses the information reported in LERs in confirming licensing bases, studying potentially generic safety problems, assessing trends and patterns of operational experience, monitoring performance, identifying precursors of more significant events, and providing operational experience to the industry.

The NRG determines whether events meet the criteria for reporting as an Abnormal Occurrence Report to Congress or for reporting to the European Nuclear Energy Agency (NEA).

The information from L!=Rs is widely used within the nuclear industry, both nationally and internationally. The industry's Institute of Nuclear Power Operation (INPO) uses LERs as a basis for providing operational safety experience feedback data to individual utilities through such documents as significant operating experience reports, significant event reports, 23

significant events notifications, and operations and maintenance reminders. U.S. vendors and nuclear steam system suppliers, as well as other countries and international organizations, use LER data as a source of operational experience data.

Comment 34: Some comments indicated that the licensing basis should be defined.

Response: No changes are proposed. The term "licensing basis" is not explicitly used in the event reporting rules or the draft reporting guidelines. It can come into play, via Generic Letter (GL) 91-18, "Information to Licensees Regarding two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability," in determining what the "specified safety function" of a system is. This relates to whether an event is reportable as an event or condition that could ha-ve prevented the fulfillment of the safety function of structures or systems .... and/or an operation or condition prohibited by the plant's technical specification (TS). However, any unsettled details regarding exactly which commitments are included in the licensing basis (for example because of differences between the definitions in GL 91-18 and 10 CFR 54.3) are not of a nature that wou 1':l change the determination of whether or not a system is capable of performing its specified safety functions (i.e., operable).

Comment 35: Several comments recommended conducting tabletop exercises (public meetings) early in the drafting process, involving licensees, inspectors, and headquarters personnel to discuss the draft amendments and associated and guidance.

Response: The Commission agrees. The recommended public meeting was held on November 13, 1998.

Comment 36: Several comments recommended conducting a workshop (public meeting) early during the public comment period to discuss the proposed rule and draft guidance.

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Responsh,: The Commission agrees. The recommended workshop has been added to the schedule.

Comment 37: Several comments recommended that the reporting guidelines be revised concurrently with the rules.

Response: The Commission agrees. Draft guidelines are being made available for comment concurrent with the proposed rules.

Comment 38: Several comment letters recommended reviewing enforcement criteria at the same time the rule is being developed to ensure consistent application of enforcement to reporting.

Response: The comment is accepted. The Enforcement Policy is being reviewed concurrently with development of the rule.

IV. Discussion

1. Objectives of Proposed Amendments

The purpose of sections 50.72 and 50.73 would remain the same because the basic needs rem~in the same. The objectives of the proposed amendments would be as follows:*

(1) To better align the reporting requirements with the NRC's current reporting needs.

An example is extending the required initial reporting times for some events, consistent with the need for timely NRC action. Another example is changing the criteria for reporting system actuations, to obtain reporting that is more consistent with the risk-significance of the systems involved.

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(2) To reduce the reporting burden, consistent with the NRG's reporting needs. An example is eliminating the reporting of design and analysis defects and deviations of little or no risk- or safety-significance.

(3) To clarify the reporting requirements where needed. An example is clarifying the cnteria for reporting design or analysis defects or deviations.

(4) To maintain consistency with NRG actions to improve integrated plant assessments.

For example, reports that are needed in the assessment process should not be eliminated.

2. Section by Section Discussion of Proposed Amendments General requirements [section 50. 72(a)(5)]. The requirement to inform the NRG of the type of report being made (i.e., emergency class declared, non-emergency 1-hour report, or non-emergency 8-hour report) would be revised to refer to paragraph (a)(1) instead of referring to paragraph (a)(3) to correct a typographical error.

Required initial reporting times [sections 50. 72(a)(5), (b)(1), (b)(2), and sections

50. 73(a)(1) and (d)]. In the proposed amendments, declaration of an emergency class would continue to be reported immediately after notification of appropriate State or local agencies not later than 1-hour after declaration. This includes declaration of an Unusual Event, the lowest emergency class.

Deviations from technical specifications authorized pursuant to 10 CFR 50.54(x) would continue to be reported as soon as practical and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of occurrence.

These two criteria capture those events where there may be a need for immediate action by the NRC.

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Non-emergency events that are reportable by telephone under 10 CFR 50. 72 would be reportable as soon as practical and in all cases within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (instead of within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as is currently required). This would reduce the burden of rapid reporting, while still capturing those events where there may be a need for the NRC to contact the plant to find out more about the event and/or initia_te a special inspection or investigation within about a day.

Written LERs would be due within 60 days after discovery of a reportable event or condition (instead of within 30 days as is currently required). Changing the time limit from 30 days to 60 days does not imply that licensees should take longer than they previously did to develop and implement corrective actions. They should continue to do so on a time scale commensurate with the safety significance of the issue. However, for those cases where it does take longer than thirty days to complete a root cause analysis, this change would result in fewer LERs that require amendment (by submittal of an additional report).

The Performance Indicator (Pl) program and the future risk-based performance indicator program provide valued input to regulatory decisions (e.g. Senior Management Meetings). Adding 30 days to the delivery of data supplying these programs would result in the reduction in the currency and value of these indicators to senior managers. With respect to the Accident Sequence Precursor program, the additional 30 days will add a commensurate amount of time to each individual event assessment since Licensee Event Reports (LERs) are the main source of data for these analyses. The delivery date for the annual Accident Sequence P'recursor report would also slip accordingly. The NRC staff would have to make

\

more extensive use of Immediate Notifications (10 CFR 50.72) and event followup to compensate in part for the Licensee Event Report (LER) reporting extension.

In the interest of simplicity, the proposed amendments would maintain just three basic

' \

levels of required reporting times in 10 CFR 50.72 and 50.73 (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and 60 days).

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However public comment is specifically invited on the question of whether additional levels should be introduced to better correspond or particular types of events. For example, 10 CFR 50. 72 currently requires reporting within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for events that involve low levels of radioactive releases, and events related to safety or environmental protection that involve a press release or notification of another government agency. These types of events could be maintained at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months so that information is available on a more timely basis to respond to heightened public concern about such events. In another example, events related to environmental protection are sometimes reportable to another agency, which is the lead

\

agency for the matter, with a different time limit, such as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . These types of events could be reported to the NRC at approximately the same time as they are reported to the other agency.

Operation or condition prohibited by TS [section 50. 73(a)(2)(i)(B)J. The term "during the previous three years" would be added to eliminate written LERs for conditions that have not existed during the previous three years. Such a historical event would now have less significance, and assessing reportability for earlier times can consume considerable resources.

For example, assume that a procedure is found to be unclear and, as a result, a question is raised as to whether the plant was ever operated in a prohibited condition. If operation in the prohibited condition is likely, the answer should be reasonably apparent based on the knowledge and experience of the plant's operators and/or a review of operating records for the past three years. The very considerable effort required to review all records older than three years, in order to rule out the possibility, would not be warranted.

In addition, this criterion would be modified to eliminate reporting if the technical specification is administrative in nature. Violation of administrative technical specifications have generally not been considered to warrant submittal of an LER, and since 1983 when the 28

rule was issued the staff's reporting guidance has excluded almost all cases of such reporting.

This change would make the plain wording of the rule consistent with that guidance.

Finally, this criterion would be modified to eliminate reporting if the event consisted solely of a case of a late surveillance test where the oversight is corrected, the test is performed, and the equipment is found to be functional. This type of event has not proven to be significant because the equipment remained functional.

Condition of the nuclear power plant, including its principal safety barriers, being seriously degraded [current sections 50. 72(b)(1}(it) and (b)(2)(i), replaced by new section 50.72(b)(2)(it), and section 50.73(a)(2)(ii)]. Currently, 10 CFR 50.72(b)(1)(ii) and (b)(2)(i) provide the following distinction: a qualifying event or condition during operation is initially reportable in one hour; a condition discovered while shutdown that would have qualified if it had it been discovered during operation is ini::.;::y ieportable in four hours. The new 10 CFR 50.72(b)(2)(ii) would eliminate the distinction because there would no longer be separate 1-hour and 4-hour categories of non-emergency reports for this criterion. There would only be 8-hour non-emergency reports for this criterion.

Unanalyzed condition that significantly compromises plant safety [sections

50. 72(b)(1)(ii)(A) and (b)(2}(t), and section 50. 73(a)(2)(ii)(A); replaced by new section

50. 72(b)(2)(it)(B), and section 50. 73(a)(2}(it)(B)]. Currently, 10 CFR 50.72(b)(1)(ii)(A) and (b)(2)(i) provide the following distinction: a qualifying event or condition during operation is initially reportable in one hour; a condition discovered while shutdown that would have qualified if it had it been discovered during operation is initially reportable in four hours. The new 10 CFR 50.72(b)(2)(ii)(B) would eliminate the distinction because there would no-longer be separate 1-hour and 4-hour categories of non-emergency reports for this reporting criterion.

There would only be 8-hour non-emergency reports for this criterion.

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In addition, the new 10 CFR 50.72(b)(2)(ii)(B) and 50.73(a)(2)(ii)(B) would refer to a condition that significantly affects plant safety rather than a condition that significantly compromises plant safety. This is an editorial change intended to better reflect the nature of the criterion.

Condition that is outside the design basis of the plant [current Section 50. 72(b)(2)(il)(B) and section 50. 73(a)(2)(it)(B)]. This criterion would be deleted. However, a condition outside the design basis of the plant would still be reported if it is significant enough to qualify under one or more of the following criteria.

If a design or analysis defect or deviation (or any other event or condition) is significant enough that, as a result, a structure or system would not be capable of performing its specified safety functions, the condition would be reportable under sections 50. 72(b)(2)(v) and,

50. 73(a)(2)(v) [i.e., an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down .. .].

F()r example, during testing of 480 volt safety-related breakers, one breaker would not trip electrically. The cause was a loose connection, due to a lug that was too large for a connecting wire. Other safety related breakers did not malfunction, but they had the same mismatch. The event would be reportable because the incompatible lugs and wires could have caused one or more safety systems to fail to perform their specified safety function(s).

Another example is as follows. An annual inspection indicated that some bearings were wiped or cracked on both emergency diesel generators (EDGs). Although the EDGs were running prior to the inspection, the event would be reportable because there was reasonable doubt about the ability of the EDGs to operate for an extended period of time, as required.

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If a design or analysis defect or deviation (or any other event or condition) is significant enough that, as a result, one train of a multiple train system controlled by the plant's TS is not capable of performing its specified safety functions, and thus the train is inoperable longer than allowed by the TS, the condition would be reportable under section 50. 73(a)(2)(i)(B) [i.e.,

an operation or condition prohibited by TS].

For example, if it is found that an exciter panel for one EOG lacks appropriate seismic restraints because of a design, analysis or construction inadequacy and, as a result, there is reasonable doubt about the EDG's ability to perform its specified safety functions during and after a Safe Shutdown Earthquake (SSE) the event would be reportable.

Or, for example, if it is found that a loss of offsite power could cause a loss of instrument air and, as a result, there is reasonable doubt about the ability of one train of the auxiliary feedwater system to perform its specified safe:Ly functions for a certain postulated steam line breaks, the event would be reportable.

If a condition outside the design basis of the plant (or any other unanalyzed condition) is significant enough that, as a result, plant safety is significantly affected, the condition would be reportable under sections 50. 72(b)(2)(it)(B) and 50. 73(a)(2)(it)(B) [i.e., an unanalyzed condition that significantly affects plant safety].

As was previously indicated in the 1983 Statements of Considerations for 10 CFR

50. 72 and 50. 73, with regard to an unanalyzed condffion that significantly compromises plant safety, "The Commission recognizes that the licensee may use engineering judgment and experience to determine whether an unanalyzed condition existed. It is not intended that this paragraph apply to minor variations in individual parameters, or to problems concerning single pieces of equipment. For example, at any time, one or more safety-related components may be out of service due to testing, maintenance, or a fault that has not yet been repaired. Any 31

trivial single failure or minor error in performing surveillance tests could produce a situation in which two or more often unrelated, safety-grade components are out-of-service. Technically, this is an unanalyzed condition. However, these events should be reported only if they involve functionally related components or if they significantly compromise plant safety."1 "When applying engineering judgment, and there is a doubt regarding whether to report or not, the Commission's policy is that licensees should make the report. "2 "For example, small voids in systems designed to remove heat from the reactor core which have been previously shown through analysis not to be safety significant need not be reported. However, the accumulation of voids that could inhibit the ability to adequately remove heat from the reactor core, particularly under natural circulation conditions, would constitute an unanalyzed condition and would be reportable. "3 "In addition, voiding in instrument lines that results in an erroneous indication causing the operator to misunderstand the true condition of the plant is also an unanalyzed condition and should be reported." 4 Furthermore, beyond the examples given in 1983, examples of reportable events would include discovery that a system required to meet the single failure criterion does not do so.

In another example, if fire barriers are found to be missing, such that the required degree of separation for redundant safe shutdown trains is lacking, the event would be reportable. On the other hand, if a fire wrap, to which the licensee has committed, is missing from a safe shutdown train but another safe shutdown train is available in a different fire area, 1

48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

2 48 FR 39042, August 29, 1983.

3 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

4 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

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protected such that the required separation for safe shutdown trains is still provided, the event .

would not be reportable.

If a condition outside the design basis of the plant (or any other event or condition) is significant enough that, as a result, a principal I

safety barrier is seriously degraded, it would be reportable under sections 50. 72(b)(2)(il)(A) and 50. 73(a)(2)(il)(A) [i.e., any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded]. This reporting criterion applies to material (e.g., metallurgical or chemical) problems that cause abnormal degradation of or stress upon the principal safety barriers (i.e., the fuel cladding, reactor coolant system pressure boundary, or the containment) such as:

(i) Fuel cladding failures in the reactor, or in the storage pool, that exceed expected values, or that are unique or widespread, or that are caused by unexpected factors.

(ii) Welding or material defects in the primary coolant system which cannot be found acceptable under ASME Section XI, IWB-3600, "Analytical Evaluation of Flaws" or ASME Section XI, Table IWB-3410-1, "Acceptance Standards."

(iii) Steam generator tube degradation in the following circumstances:

(1) The severity of degradation corresponds to failure to maintain structural safety factors. The structural safety factors implicit in the licensing basis are those described in Regulatory Guide 1.121. These safety factors include a margin of 3.0 against gross failure or burst under normal plant operating conditions, including startup, operation in the power range, hot standby, and cooldown, and all anticipated transients that are included in the plant design specification.

(2) The calculated potential primary-to-secondary leak rate is not consistent with the plant licensing basis. The licensing basis accident analyses typically assume [for accidents 33

other than a steam generator tube rupture (SGTR)] a 1 gpm primary-to-seconaary leak rate concurrent with the accident to demonstrate that the radiological consequences satisfy 10 CFR Part 100 and GDC-19. In these instances, degradation which may lead to leakage above 1 gpm under accident conditions, other than a SGTR, would exceed the threshold. For some units, the staff has approved accident leakages above 1 gpm subject to updating the licensing basis accident analyses to reflect this amount of leakage and subject to risk implications being acceptable. 5 (iv) Low temperature over pressure transients where the pressure-temperature relationship violates pressure-temperature limits derived ,from Appendix G to 10 CFR Part 50 (e.g., TS pressure-temperature curves).

(v) Loss of containment function or integrity, including containment leak rate tests where the total containment as-found, minimum-pathway leak rate exceeds the limiting condition for operation (LCO) in the facility's TS. 6

, Finally, a condition outside the design basis of the plant (or any other event or condition) would be reportable if a component is in a degraded or non-conforrrung* condition such that the ability of a component to perform _its specified safety function is significantly degraded and the condition could reasonably be expected to apply to other similar 5

In addition, if the extent of degradation is great (i.e., if many tubes are degraded or defective), a telephone notification and a written LER should be provided. The plant's TS typically provide specific requirements indicating when reporting is required (based on the number of tubes degraded or defective in terms of 'percent inspected') and those requirements should be used to determine reportability.

6 The LCO typically employs La, which is defined in Appendix J to 10 CFR Part 50 as the maximum allowable containment leak rate at pressure Pa, the calculated peak containment internal pressure related to the design basis accident. Minimum-pathway leak rate means the minimum leak rate that can be attributed to a penetration leakage path; for example, the smaller of either the inboard or outboard valve's individual leak rates.

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components in the plant This new criterion is contained in section 50.73(a)(2)(ii)(C) as discussed below.

As a result, these proposed amendments would focus the reporting of conditions outside the design basis of the plant to the safety significant issues while reducing the number of reports under the current rules in order to minimize the reporting of less significant issues.

In particular, the proposed amendments will help ensure that significant safety problems that could reasonably be expected to be applicable to similar components at the specific plant or at other plants will be identified and addressed although the specific licensee might determine that the system or structure remained operable, or that technical specification requirements were met. The proposed rules will provide that, consistent with the NRC's effort to obtain information for engineering studies of operational anomalies and trends and patterns analysis of operational occurrences, the NRC would be able to monitor the capability of safety-related compcments to perform their design-basis functions.

Significantly degraded component(s) [section 50. 73(a)(2)(ii)(C)]. This new reporting criterion would require reporting if a component is in a degraded or non-conforming condition such that the ability of the component to perform its specified safety function is significantly degraded and the condition could reasonably be expected to apply to other similar components in the plant. It would be added to ensure that design basis or other discrepancies would continue to be reported if the capability to perform a specified safety function is significantly degraded and the condition has generic implications. On the other hand, if the degradations are not significant or the condition does not have generic implications, reporting would not be required under this criterion.

For example, at one plant several normally open valves in the low pressure safety injection system were routinely closed to support quarterly surveillance testing of the system.

35

In reviewing the design basis and associated calculations, it was determined that the capability of the valves to open in the event of a large break loss-of-coolant accident (LOCA) combined with degraded grid voltage during a surveillance test was degraded. The licensee concluded that the valves would still be able to reopen under the postulated conditions and considered them operable. However, that conclusion could not be supported using the conservative standards established by Generic Letter 89-10. Pending determination of final corrective action, administrative procedures were implemented to preclude closing the valves. The event would be reportable because the capability of a component to perform its specified safety functions was significantly degraded and the same condition cquld reasonably be expected to apply to other similar components.

In another example, during a routine periodic inspection, jumper wires in the valve operators for three valves were found contaminated with grease which was leaking from the limit switch gear box. The cause was overfilling of the grease box, as a result of following a generic maintenance procedure. The leakage resulted in contamination and degradation of the electrical components which were not qualified for exposure to grease. This could result in valve malfunction(s). The conditions were corrected and the maintenance procedures were changed. The event would be reportable because the capability of several similar components

)

to perfonn their specified safety functions could be significantly degraded.

In a further example, while processing calculations it was determined that four motor operated valves within the reactor building were located below the accident flood level and were not qualified for that condition. Pending replacement with qualified equipment, the licensee determined that three of the valves had sufficiently short opening time that their safety function would be completed before they were submerged. The fourth valve was normally open and could remain open. After flooding, valve position indication could be lost, but valve 36

position could be established indirectly using process parameter indications. The event would be reportable because the capability of several similar components to perform their specified safety functions could be significantly degraded.

An example of an event that would not be reportable is as follows. The motor on a motor-operated valve (MOV) burned out after repeated cycling for testing. This event ~ould not be reportable because it is a single component failure, and while there might be similar MOVs in the plant, there is not a reasonable basis to think that other MOVs would be affected by this same condition. On the other hand, if several MOVs had been repeatedly cycled and then after some extended period of time one of the MOVs was found inoperable or significantly degraded because of that cyding, then the condition would be reportable.

Minor switch adjustments on MOVs would not be reported where they do not significantly affect the ability of the MOV to carry out its design-basis function and the cause of the adjustments is not a generic concern.

At one p!ant the switch on the radio trans'Tlitter for the auxiliary buildiri:: crane was used to handle a spent fuel cask while two protective features had been defeated by wiring errors.

A new radio control transmitter had been procured and placed in service. Because the new controller was wired differently than the old one, the drum overspeed protection and spent fuel pool roof slot limit switch were inadvertently defeated. While the crane was found to be outside its design basis, this condition would not be reportable because the switch wiring deficiency could not reasonably be expected to affect any other components at the plant Condition not covered by the plant's operating and emergency procedures {section

50. 72(b)(2)(il)(C), and section 50. 73(a)(2)(il)(C)J. This criterion would be deleted because it does not result 'in worthwhile reports aside from those that would be captured by other reporting criteria such as:

37

(1) an unanalyzed condition that significantly affects plant safety; (2) an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: shut down the reactor and maintain it in a safe shutdown condition; remove residual heat; control the release of radioactive material; or mitigate the consequences of an accident; (3) an event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; (4) an operation or condition prohibited by the plant's TS; (5) an event or condition that results in actuation of any of the systems listed in the rules, as amended; (6) an event that poses an actual threat to the safety of the nuclear power plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

Manual or autcmatic actuation of any engineered safety feature ESF [current sections

50. 72(b)(1)(iv) and (b)(2)(ii), replaced by new sections 50. 72(b)(2)(iv), and section 50.73(a)(2)(iv)]. Currently, sections 50.72(b)(1)(iv) and (b)(2)(ii) provide the following distinction: an event that results or should have resulted in ECCS discharge into the reactor coolant system is initially reportable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; other ESF actuations are initially reportable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The new 10 CFR 50.72(b)(2)(iv) would eliminate this distinction because there would no longer be separate 1-hour and 4-hour categories of non-emergency reports for this criterion. There would only be 8-hour non-emergency reports for this criterion.

The new section 50.72(b)(2)(iv) would eliminate telephone reporting for invalid automatic actuation or unintentional manual actuation. These events are not significant and thus telephone reporting is not needed. However, the proposed amendments would not 38

eliminate the requirement for a written report of an invalid actuation under 10 CFR 50. 73.

There is still a need for reporting of these events because they are used in making estimates of equipment reliability parameters, which in tum are needed to support the Commission's move towards risk-informed regulation. (See SECY-97-101, May 7, 1997, "Proposed Rule, 10 CFR 50.76, Reporting Reliability and Availability Information for Risk-significant Systems and Equipment," Attachment 3) .

.The term "any engineered safety feature (ESF), including the reactor protection system (RPS)," which currently defines the systems for which actuation must be reported in section 50.72(b)(2)(iv) and section 50.73{a)(2)(iv), would be replaced by a specific list of systems. The current definition has led to confusion and variability in reporting because there are varying definitions of what constitutes an ESF. For example, at some plants systems that are known to have high risk significance, such as emerge:,cy ac power, auxiliary feedwater, and reactor

core isolation cooling are not considered ESFs. Furthermore, in many cases systems with much lower levels 'if risk significance, such as control room ventilation systems, are considered to be ESFs.

In the proposed amendments actuation would be reportable for the specific systems named in sections 50.72(b)(2)(iv) and 50.73(a)(2)(iv). This would result in consistent reporting of events that result in actuation of these highly risk-significant systems. Reasonable consistency in reporting actuation of highly risk-significant systems is needed to support estimating equipment reliability parameters, which is important to several aspects of the move towards more risk-informed regulation, including more risk-informed monitoring of plant performance.

The specific list of systems in the proposed rule would also eliminate reporting for events of lesser significance, such as actuation of control room ventilation systems.

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The specific list of systems in the proposed rule is similar to the list of systems currently provided in the reporting guidelines in NUREG-1022, Revision 1, with some minor revisions. It is based on systems for which actuation is frequently reported, and systems with relatively high risk-significance based on a sampling of plant-specific PRAs (see Draft Regulatory Guide DG-1046, "Guidelines for Reporting Reliability and Availability Information for Risk-Significant Systems and Equipment in Nuclear Power Plants," particularly Tables C-1 through C-5).

This proposal to list the systems in the rule is controversial and public comment is specifically invited in this area. In particular, three principal alternatives to the proposed rule have been identified for comment:

(.1) Maintain the status quo. Under this alternative, the rule would continue to require reporting for actuation of uany ESF." The guidance would continue to indicate that reporting should include as a minimum the system on the list.

(2) Require use of a plant-specific , risk-informed list. Under this alternative, the list of systems would be risk-informed, and plant-specific. Licensees would develop the list based on existing PRA analyses, judgment, and specific plant design. No list would be provided in the rule.

(3) Return to the pre-1998 situation (i.e., before publication of the reporting guidance in NUREG-1022, Revision 1). Under this alternative, the rule would continue to require reporting for actuation of uany ESF." The guidance would indicate that reporting should include those systems identified as ESF's for each particular plant (e.g., in the FSAR).

With regard to this third alternative, it may be noted that this approach has the advantage of clarity and simplicity. There would be no need to develop a new list, and this is the practice that was followed from 1984-1997 without creating major problems. However, the lists of ESFs are not based on risk-significance. For example, emergency diesel generators 40

(EDGs) are known to be highly risk-significant; however, at six plants, the EDGs are not considered to be ESFs. Similarly, auxiliary feedwater (AFW), systems at pressurized water reactors (PWRs) are known to be highly risk-significant; however, at a number of plants these systems are not considered to be ESFs. Also, reactor core isolation cooling (RCIC) systems at boiling water reactors (BWRs) are known to be highly risk significant; however, at a number of plants these systems are not considered to be ESFs. In contrast, at many plants, systems with much lower levels of risk significance, such as control room ventilation systems, are considered to be ESF s.

Event or condition that could have prevented fulfillment of the safety function of structures or systems that ... {current sections 50. 72(b)(1)(il) and (b)(2}(i), replaced by new sections 50. 72(b)(2)(v} and (v,), and sections 50. 73(a)(2)(v} and (VI)] The phrase ."event or condition that alone could have prevented the fulfillment of the safety function of structures or systems .... " would be clarified by deleting the word "alone". This clarifies the requirements*by more clearly reflecting the principle that it is necessary to consider other existing plant conditions in determining the reportability of an event or condition under this criterion. For example, if one train of a two train system is incapable of performing its safety function for one reason, and the other train is incapable of performing its safety function for a different reason, the event is reportable.

The term "at the time of discovery" would be added to section 50.72(b)(2)(v) to eliminate telephone notification for a condition that no longer exists, or no longer has an effect on required safety functions. For example, it might be discovered that some time ago both trains of a two train system were incapable of performing their safety function, but the condition was subsequently corrected and no longer exists. In another example, while the-plant is shutdown, it might be discovered that during a previous period of operation a system 41

was incapable of performing its safety function, but the system is not currently required to be operable. These events are considered significant, and an LER would be required, but there would be no need for telephone notification.

The phrase "occurring within three years of the date of discovery" would be added to section 50.73(a)(2)(v) to eliminate written LERs for conditions that have not existed during the previous three years. Such a historical event would now have less significance, and assessing reportability for earlier times can consume considerable resources. For example, assume that during a design review a discrepancy is found that affects the ability of a system to perform its safety function in a given specific configuration. If it is likely that the safety function could have been prevented, the answer should be reasonably apparent based on the knowledge and experience of the plant's operators and/or a review of o~rating records for the past three years. The very considerable effort required to review all records older than three years, in order to rule out the possibility, would not be warranted.

A new paragraph, section 50.72(b)(2)(vi) would be added to 'clarify section 50.72. The new paragraph would explicitly state that telephone reporting is not required under section 50.72(b)(2)(v) for single failures if redundant equipment in the same system was operable and available to perform the required sa*fety function. That is, although one train of a system may be incapable of performing its safety function, reporting is not required under this criterion if that system is still capable of performing the safety function. This is the same principle that is currently stated explicitly in section 50. 73(a)(2)(vi) with regard to written LERs.

Major loss of emergency assessment capability, .offsfte response capability, or communication capability [current section 50. 72(b)(2)(v), new section 50. 72(b)(2)(xiii)]. The new section would be modified by adding the word "offsite" in front of the term 42

"communications capability" to make it clear that the requirement does not apply to internal plant communication systems.

Airborne radioactive release ... and liquid effluent release .. .[section 50. 72(b)(2)(viil) and sections 50. 73(a)(2)(viit) and 50. 73(a)(2)(ix)]. The statement indicating reporting under section 50.72(b)(2-)(viii) satisfies the requirements of section 20.2202 would be removed because it would not be correct. For example, some events captured by section 20.2202 would. not be captured by section 50.72(b)(2)(viii). Also, the statement indicating that reporting under section 50.73(a)(2)(viii) satisfies the requirements of section 20.2203(a)(3) would be deleted because it would not be correct. Some events captured by section 20.2203(a)(3) would not be captured by section 50. 73(a)(2)(viii).

The proposed extension of reporting deadlines to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months in section 50. 72 and 60 days in section 50. 73 raises questions about whether s1m1lar changes should be made to Parts 20, 30, 40, 70, 72 and 76. The merits of such changes, which may vary for different types of licensees, will be addressed in separate actions.

Contents of LERs [sections 50. 73(b)(2)(it)(F)and 50. 73(b)(2)(ii)(J)]. Paragraph (F) would be revised to correct the address of the NRC Ubrary.

Paragraph (J) currently requires that the narrative _section include the following specific information as appropriate for the particular event:

"(1) Operator actions that affected the course of the event, including operator errors, procedural deficiencies, or both, that contributed to the event (2) For each personnel error, the licensee shall discuss:

(i) Whether the error was a cognitive error (e.g., failure to recognize the actual plant condition, failure to realize which systems should be functioning, failure to recognize the true nature of the event) or a procedural error; 43

(ii) Whether the error was contrary to an approved procedure, was a direct result of an error in an approved procedure, or was associated with an activity or task that was not covered by an approved procedure; (iii) Any unusual characteristics of the work location (e.g., heat, noise) that directly contributed ta the error; and (iv) The type of personnel involved (i.e., contractor personnel, ufility-lic;:ensed operator, utility non-licensed operator, other utility personnel)."

The proposed amendment would change section 50.73(b)(2)(ii)(J) to simply require that the licensee discuss the causes and circumstances for each human performance related problem that contributed ta the event. It is not necessary to specify the level of detail provided in the current rule, which is more appropriate for guidance. Details would continue to be provided in the reporting guidelines, as indicated in section 5.2.1 of the draft of Revision 2 to NUREG-1022. This draft report is being made available far public comment concurrently with the pr0'."')Sed rule, as discussed below under the heading "Revisions to Reporting Guidelines in NUREG-1022."

Spent fuel storage cask problems {current sections 50. 72(b)(2)(vil) and 72.16(a)(1),

(a)(2), (b) and (c)l Section 50.72(b)(2)(vii) would be deleted because these reporting criteria are redundant to the reporting criteria contained in sections 72.216(a)(1), (a)(2), and (b).

Repetition of the same reporting criteria in different sections of the rules adds unnecessary complexity and is inconsistent with the current practice in other areas, such as reporting of safeguards events as required by section 73.71.

Also, a conforming amendment would be made to section 72.216. This is necessary because section 72.216(a) currently relies on section 50.72(b)(2)(vii}, which would be deleted, 44

to establish the time limit for initial notification. The amended section 72.216 would refer to sections 72. 74 and 72. 75 for initial notification and followup reporting requirements.

Assessment of Safety Consequences [section 50. 73(b)(3)]. This section currently requires that an LER include an assessment of the safety consequences and implications of the event. This assessment must include the availability of other systems or components that could have performed the same function as the components and systems that failed during the event. It would be modified by adding a requirement to also include the status of components and systems that "are included in emergency or operating procedures and could have been used to recover from the event in case of an additional failure in the systems actually used for recovery." This information is needed to better support the NRC's assessment of the risk-significance of reported events.

Exemptions [section 50. 73({)]. This provision w:>uld be deleted because the exemption provisions in section 50.12 provide for granting of exemptions as warranted. Thus, including another, section-specific exemption provision in section 50. 73 adds unnecessary complexity to the rules.

3. Revisions to Reporting Guidelines in NUREG-1022.

A draft report, NUREG-1022, Revision 2, "Event Reporting Guidelines, 10 CFR 50.72 and 50.73," is being made available for public comment concurrently with the proposed amendments to 10 CFR 50.72 and 50.73. The draft report is available for inspection in the NRC Public Document Room or it may be viewed and downloaded electronically via the interactive rulemaking web site established by NRC for this rulemaking, as discussed above under the heading "ADDRESSES." Single copies may be obtained from the contact listed 45

above under the heading "For Further Information Contact." In the draft report, guidance that is considered to be new or different is a meaningful way, relative to that provided in NUREG-1022, Revision 1, is indicated by redlining the appropriate text.

4. Reactor Oversight The NRC is developing revisions to process for oversight of operating reactors, including inspection, assessment and enforcement processes. In connection with this effort, the NRC has considered the kinds of event reports that would be eliminated by the proposed rules and believes that the changes would not have a deleterious effect on the oversight process. Public comment is invited on whether or not this is the case. In particular, it is requested that if any examples to the contrary are known they be identified.

5. Reporting of Historical Problems.

As discussed above, provisions would be added to sections 50. 73(a)(2)(i)(B) and

50. 73(a)(2)(v) to eliminate reporting of a condition or event that did not occur within three years of the date of discovery. (See the response to Comment 8, the discussion under the heading "Operation or condition prohibited by TS." and the discussion under the heading "Event or condition that could have prevented fulfillment of the safety function of structures or systems that ... ") Public comment is invited on whether such historical events and conditions should be reported (rather than being excluded from reporting, as proposed). Public comment is also invited on whether the three year exclusion of such historical events and conditions should be 46

extended to all written reports required by section 50.73(a) (rather than being limited to these two specific reporting criteria, as proposed).

6. Reporting of Component Problems.

As discussed above, a new reporting criterion would be added to require reporting if a component Is in a degraded or non-conforrrnng condition such that the ability of the component to perform its specified safety function is significantly degraded and the condition could reasonably be expected to apply to other similar components in the plant. (See the response to Comment 14 and the discussion under the heading "Significantly degraded component(s)

[section 50.73(a)(2)(ii)(C)].") Public comment is invited on whether this proposed new criterion would accomplish its stated purpose - to ensure that design basis or other discrepancies would continue to be reported if the capability to perform a specified safety function is significantly degraded and the condition has generic implications. Public comment is also invited on whether the proposed new criterion would be subject to varying interpretations by licensees and inspectors.

7. Enforcement.

The NRC intends to modify its existing enforcement policy in connection with the proposed amendments to sections 50.72 and 50.73. The philosophy of the proposed changes is to base the significance of the reporting violation on: (1) the reporting requirement, which will require reporting within time frames more commensurate with the significance of the underlying issues than the current rule, and (2) the impact that a late report may have on the 47

ability of the NRC to fulfill its obligations of fully understanding issues that are required to be reported in order to accomplish its public health and safety mission, which in many cases involves reacting to reportable issues or events. As such, the NRC intends to revise the Enforcement Policy, NUREG-1600, Rev. 1 as follows:

(1) Appendix 8, Supplement I.C - Examples of Severity Level Ill violations.

(a) Example 14 would be revised to read as follows -A failure to provide the required one hour telephone notification of an emergency action taken pursuant to 10 CFR 50.54(x).

(b) An additional example would be added that would read as follows -A failure to provide a required 1-hour or 8-hour non-emergency telephone notification pursuant to 10 CFR 50.72.

(c) An additional example would be added that would read as follows -A late 8-hour

\

notification that substantially impacts agency response.

(2) Appendix B, Supplement I.D -Examples of Severity Level IV violations.

(a) Example 4, would be revised to read as follows - A failure to provide a required 60-day written LER pursuant to 10 CFR 50. 73.

These changes in the Enforcement Policy would be consistent with the overall objective of the rule change of better-aligning the reporting requirements with the NRC's reporting needs. The Enforcement Policy changes would correlate the Severity Level of the infractions with the relative importance of the information needed by the NRC.

Section IV.D of the Enforcement Policy provides that the Severity Level of an untimely report may be redu~d depending on the individual circumstances. In deciding whether the Severity Level should be reduced for an untimely 1-hour or 8-hour non-emergency report the impact that the failure to report had on any agency response would be considered. For example, if a delayed 8-hour reportable event impacted the timing of a followup inspection that 48

was deemed necessary, then the Severity Level would not normally be reduced. Similarly, a late notification that delayed the NRC's ability to perform an engineering analysis of a condition to determine if additional regulatory action was necessary would generally not be considered for disposition at a reduced Severity Level. Additionally, late reports filed In cases where the NRC had to prompt the licensee to report would generally not be subject to disposition at reduced Severity Level and the Severity Level for failure to submit a timely Licensee Event Report (LER) would not be reduced to a minor violation.

In accordance with Appendix C of the Enforcement Policy, " Interim Enforcement Policy for Severity Level IV Violations Involving Activities of Power Reactor Licensees," the failure to file a 60-day LER would normally be dispositioned as a Non-Cited Violation (NCV). Repetitive failures to make LER reports indicative of a licensee's inability to recognize reportable* ..:

conditions, such that it is not likely that the NRC will be made aware of operational, design and configuration issues deemed reportable pursuant to 10 CFR 50. 73, will be considered for categorization at Severity Level Ill. This disposition may be warranted since such licensee performance impacts the ability of the NRC to fulfill its regulatory obligations .

8. Electronic Reporting.

The NRC is currently planning to implement an electronic document management and reporting program, known as the Agency-wide Document Access and Management System (ADAMS), that will in general provide for electronic submittal of many types of reports, including LERs. Accordingly, no separate rulemaking effort to provide for electronic submittal of LERs is contemplated.

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9. Schedule.

The current schedule is as follows:

08/99 - Conduct public workshop to discuss proposed rule and draft reporting guidelines (separate notice with workshop details will be published later this month)

(insert date 30 days after publication in the Federal Register) - Public comments due to 0MB (insert date 60 days after publication in the Federal Register) - Receive 0MB approval (insert date 75 days after publication in the Federal Register) - Public comments due to NRC 10/01/99 - Provide final rule and guidelines to NRC staff rulemaking group 11/05/99 - Provide final rule and guidelines to the formal concurrence chain 01/14/00 - Provide final rule and guidelines to CRGR and ACRS 02/11/00 - Complete briefings of CRGR and ACAS 03/10/00 - Provide final rule and guidelines to Commission 04/07/00- Publish final rule and guidelines

10. State Input.

Many States (Agreement States and Non-Agreement States) have agreements with power reactors to inform the States of plant issues. State reporting requirements are frequently triggered by NRC reporting requirements. Accordingly, the NRC seeks State comment on issues ,related to the proposed amendments to power reactor reporting requirements.

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Plain Language The President's Memorandum dated June 1, 1998, entitled, UPlain Language in Government Writing," directed that the Federal government's writing be in plain language. The NRC requests comments on this proposed rule specifically with respect to the clarity and effectiveness of the language used. Comments should be sent to the address listed above.

V. Environmental Impact: Categorical Exclusion.

The NRC has determined that this proposed regulation is the type of action described in categorical exclusion 10 CFR 51.22(c)(3)(iiQ. Therefore neither an environmental impact statement nor an environmental assessment ~..;.; been prepared for this proposed regulation.

VI. Backfit Analysis.

The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to information collection and reporting requirements such as those contained in the proposed rule. Therefore, a backfit analysis has not been prepared. However, as discussed below, the NRC has prepared a regulatory analysis for the proposed rule, which examines the costs and benefits of the proposed requirements in this rule. The Commission regards the regulatory analysis as a disciplined process for assessing information collection and reporting requirements to determine that the burden imposed is justified in light of the potential safety significance of the information to be collected.

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VII. Regulatory Analysis.

The Commission has prepared a draft regulatory analysis on this proposed rule. The analysis examines the costs and benefits of the alternatives considered by the Commission.

The draft analysis is available for inspection in the NRC Public Document Room or it may be viewed and downloaded electronically via the interactive rulemaking web site established by NRC for this rulemaking, as discussed above under the heading "ADDRESSES." Single copies may be obtained from the contact listed above r under the heading "For Further Information Contact."

The Commission requests public comment on this draft analysis. Comments on the draft analysis may be submitted to the NRC as discussed above under the heading "ADDRESSES."

VIII. Paoerwork Reduction Act Statement This proposed rule would amend information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This rule has been submitted to the Office of Management and Budget for review and approval of the information collection requirements.

The public reporting burden for the currently existing reporting requirements in 10 CFR 50.72 and 50.73 is estimated to average about 790 hours0.00914 days 0.219 hours 0.00131 weeks 3.00595e-4 months per response (i.e., per commercial nuclear power reactor- per year) including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collection. It is estimated that the proposed amendments would 52

impose a one time implementation burden of about 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months per reactor, after which there would be a recurring annual burden reduction of about 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months per reactor per year. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collection contained in the proposed rule and on the following issues:

, Is the proposed information collection necessary for the proper performance of the NRC, including whether the information will have practical utility?

Is the estimate of burden accurate?

Is there a way to enhance the quality, utility, and clarity of the information to be collected?

How can the burden of the information collection be minimized, including the use of automated collection techniques? .

Send comments on any aspect of this proposEd information collection, including suggestions for reducing this burden, to the Information and Records Management Branch (T-5 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 or by Internet electronic mail to BJS1@NRC.GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150AF98), Office of Management and Budget, Washington,, DC 20503.

Comments to 0MB on the information collections or on the above issues should be submitted by (insert date 30 days after publication in the Federal Register). Comments received after this date will be considered if it is practical to do so, but consideration cannot be ensured for comments received after this date.

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Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, an information collection unless it displays a currently valid 0MB control number.

IX. Regulatory Flexibility Certification.

In accordance with the Regulatory Flexibility Act (5 U.S.C. 605(b)), the Commission certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number of small entities. This proposed rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the size standards established by the NRG (10 CFR 2.810).

X. Proposed Amendments.

List of Subjects 10 CFR Part 50: Antitrust, Classified information, Criminal penalties, Fire prevention, Intergovernmental rela(ions, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

10 CFR Part 72: Criminal penalties, Manpower training programs, Nuclear materials, Occupational safety and health, Reporting and recordkeeping requirements, Security measures, and Spent fuel.

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For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C.

553, the NRG is proposing to adopt the following amendments to 10 CFR Part 50 and 10 CFR Part 70.

PART 50- DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

1. The authority citation for Part 50 continues to read as follows:

AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936,937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.

2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C.

5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 955 as amended (42 U.S.C.

2131, 2235), sec. 102, Pub. L.91-190, 83 Stat 853 (42 U.S.C. 4332). Sections 50.13, 50.54(D.D.), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S,C.

2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 184, 68 Stat 954, as amended (42 U.S.C. 2234). Appendix Falso issued under sec. 187, 68 Stat 955 (42 U.S.C. 2237).

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2. Section 50.72 is revised by amending paragraphs (a) and (b) to read as follows:

§ 50. 72 Immediate notification requirements for operating nuclear power reactors.

(a) General requirements. 7 (1) Each nuclear power reactor licensee licensed under

§ 50.21(b) or§ 50.22 of this part shall notify the NRC Operations Center via the Emergency Notification System of:

(i) The declaration of any of the Emergency Classes specified in the licensee's approved Emergency Plan; 8 or (ii) Of those non-Emergency events specified in paragraph (b) of this section.

(2) If the Emergency Notification System is inoperative, the licensee shall make the required notifications via commercial telephone service, other dedicated telephone system, or any other method which will ensure that a report is made as soon as practical to the NRC Operations Center. 9

10 (3) The licensee shall notify the NRC immediately after notification of the appropriate State or local agencies and not later than one hour after the time the licensee declares one of the Emergency Classes.

(4) The licensee shall activate the Emergency Response Data System (ERDS) 11 as soon as possible but not later than one hour after declaring an emergency class of alert, site 7

Other requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained elsewhere in this chapter, in particular §§ 20.1906, 20.2202, 50.36, 72.74, 72.75, and 73.71.

8 These Emergency Classes are addressed in Appendix E of this part.

9 Commercial telephone number of the NRC Operations Center is (301) 816-5100.

10

[Reserved]

11 Requirements for EROS are addressed in Appendix E,Section VI.

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area emergency, or general emergency. The EROS may also be activated by the licensee during emergency drills or exercises if the licensee's computer system has the capability to transmit the exercise data. .

(5) When making a report under paragraph (a)(1) of this section, the licensee shall identify:

0) The Emergency Class declared; or (ii} Either paragraph (b)(1), "One-Hour Report," or paragraph (b)(2} "Eight-Hour Report,"

as the paragraph of this section requiring notification of the Non-Emergency Event.

(b) Non-emergency events- (1) One-Hour reports. If not reported as a declaration of the Emergency Class under paragraph (a) of this section, the licensee shall notify the NRC as soon as practical and in all cases within one hour of the occurrence of any deviation from the plant's Technical Specifications authorized pursuant to§ 50.54(x) of this part.

(2) Eight-hour reports. If not reported under paragraphs (a) or (b)(1) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following:

(i) The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.

(ii) Any event or condition that results in:

(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (8) The nuclear power plant being in an unanalyzed condition that significantly affects plant safety.

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(iii) Any natural phenomenon or other external condition that poses an actual threat to the safety of the nuclear power plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the plant.

(iv)(A) Any event or condition that results in intentional manual actuation or valid automatic actuation of any of the systems listed in paragraph (b)(2)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

(B) The systems to which the requirements of paragraph (b)(2)(iv)(A) of this section apply are:

(1) Reactor protection system (reactor scram, reactor trip).

(2) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay)

I heat removal systems.

(3) ECCS for boiling water reactors (BWRs) inclu~ing: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; feedwater coolant injection system; low pressure injection function of the residual heat removal system; and automatic depressurization system.

(4) BWR isolation condenser system and reactor core isolation cooling system.

(5) PWR auxiliary feedwater system.

(6) Containment systems including: containment and reactor vessel isolation systems (general containment isolation signals affecting numerous valves and main steam isolation valve [MSIV] closure signals in SWRs) and containment heat removal and depressurization systems, including containment spray and fan cooler systems.

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(7) Emergency ac electrical power systems, including: emergency diesel generators (EDGs) and their associated support systems; *hydroelectric facilities used in lieu of EDGs at the Oconee Station; safety related gas turbine generators; BWR dedicated Division 3 EDGs and their associated support systems; and station blackout diesel generators (and black-start gas turbines that serve a similar purpose) which are started from the control room and included in the plant's operating and emergency procedures.

(8) Anticipated transient without scram (ATWS) mitigating systems.

(9) Service water (standby emergency service water systems that do not normally run) .

(v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (8) Remove residual heat; (C) Control the release of radioactive material, or (0) Mitigate the consequences of an accident.

(vi) Events covered in paragraph (b)(2)(v) of this section may include one or more procedural errors, equipment failures, c1:nd/or discovery of design, analysis, fabrication, ,

construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to this paragraph if redundant equipment in the same system was operable and available to perform the required safety function.

(vii) Reserved.

(viii) (A) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , results in concentrations in an unrestricted area that exceed 20 times the applicable concentration specified in appendix B to part 20, table 2, column 1.

59

(B) Any liquid effluent release that, when averaged over a time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds 20 times the applicable concentration specified in appendix B to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.

{ix) Any event that poses an actual threat to the safety of the nuclear power plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

(x) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.

{xi) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.

(xii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).

3. Section 50.73 is revised by amending sections (a), (b)(2){ii)(F), (b)(2)(ii)(J), (b)(3),

(d), and (f) to read as follows:

§ 50.73 Licensee event report system.

60

\

(a} Reportable events. (1) The holder of an operating license for a nuclear power plant (licensee) shall submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 60 days after the discovery of the event. Unless otherwise specified in this section, the licensee shall report an event regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event.

(2} The licensee shall report:

(i)(A) The completion of any nuclear plant shutdown required by the plant's Technical Specifications.

(B) Any operation or condition occurring within three years of the date of discovery which was prohibited by the plant's Technical Specifications, except when:

(i) The technical specification is administrative ii, nature; or (it) The event consists solely of a case of a late surveillance test where the oversight is ,

correrted, the test is performed, and the equipment is found to be capable of oerforming its specified safety functions.

(C) Any deviation from the plant's Technical Specifications* authorized pursuant to

§ 50.54(x) of this part (ii) Any event or condition that resulted in:

(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; (B) The nuclear power plant being in an unanalyzed condition that significantly affects plant safety; or 61

(G) A component being in a degraded or non-conforming condition such that the ability of the component to perform its specified safety function is significantly degraded and the condition could reasonably be expected to affect other similar components in the plant.

(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the.nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant (iv)(A) Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section, except when:

(1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or (2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed.

(8) The systems to which the requirements of paragraph (a)(2)(iv)(A) of this section apply ar.e:

(1) Reactor protection system (reactor scram, reactor trip).

\

(2) Emergency core cooling systems (EGGS) for pressurized water reactors (PWRs)

' including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(3) EGGS for boiling water reactors (BWRs) including: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; feedwater coolant injection system; low pressure injection function of the residual heat removal system; and automatic depressurization system.

(4) BWR isolation condenser system and reactor core isolation cooling system.

62

(5) PWR auxiliary feedwater system.

(6) Containment systems including: containment arid reactor vessel isolation systems (general containment isolation signals affecting numerous valves and main steam isolation valve [MSIV] closure signals in BWRs) and containment heat removal and depressurization systems, including containment spray and fan cooler systems.

(7) Emergency ac electrical power systems, including: emergency diesel generators (EDGs) and their associated support systems; hydroelectric facilities used in lieu of EDGs at the Oconee Station; safety related gas turbine generators; BWR dedicated Division 3 EDGs and their associated support systems; and station blackout diesel generators (and black-start gas turbines that serve a similar purpose) which are started from the control room and included in the plant's operating and emergency procedures.

(8) Anticipated transient without scram (ATWS) mitigating systems.

(9) Service water (standby emergency service water systems that do not normally run).

(v) Any event or condition occurring within three years of the date of discovery that could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

(vi) Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need 63

not be reported pursuant to this paragraph if redundant equipment in the samb system was operable and available to perform the required safety function.

(vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

(viii)(A) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in appendix B to part 20, table 2, column 1.

(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds 20 times the applicable concentrations specified in appendix B to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.

(Ix) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

(b) * * *

(2) * * *

(ii) * * *

(F) The Energy Industry Identification System component function identifier and system name of each component or system referred to in the LER.

64

(1) The Energy Industry Identification System is defined in: IEEE Std 803-1983 (May*

16, 1983) Recommended Practice for Unique Identification in Power Plants and Related Facilities-Principles and Definitions.

(2) IEEE Std 803-1983 has been approved for incorporation by reference by the Director of the Federal Register.

A notice of any changes made to the material incorporated by reference will be published in the Federal Register. Copies may be obtained from the Institute of Electrical and Electronics Engineers, 345 East 47th Street, New York, NY 10017. IEEE Std 803-1983 is available for inspection at the NRC's Technical Library, which is located in the Two White Flint North building, 11545 Rock.ville Pike, Rockville, Maryland; and at the Office of the Federal Register, 1100 L Street, NW, Washington, DC (J) For each human performance related problem that contributed to the event, the licensee shall discuss the cause(s) and circumstances.

(3) An assessment of the safety consequences and implications of the event. This assessment must include the availability of systems or components that:

(i) Could have performed the same function as the components and systems that failed during the event, or 65

(ii) Are included in emergency or operating procedures and could have been used to recover from the event in case of an additional failure in the systems actually used for recovery.

(d) Submission of reports. Licensee Event Reports must be prepared on Form NRC 366 and submitted within 60 days of discovery of a reportable event or situation to the U.S.

Nuclear Regulatory Commission, as specified in§ 50.4.

(e) Report legibility. The reports and copies that licensees are required to submit to the Commission under the provisions of'this section must be of sufficient quality to permit legible reproduction and micrographic processing.

(f) Reserved.

PART 72 - LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE

4. The authority citation for Part 72 continues to read as follows:

AUTHORITY: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 189, 68 Stal 929, 930, 932, 933, 934, 935, 954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42U.S.C.2071,2073,2077,209~,2093,2095,2099,2111,2201,2232,2233, 2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L.86-373, 73 Stat. 688, as amended {42 66

U.S.C. 584 t, 5842, 5846); Pub. L.95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-486, sec. 7902, 106 Stat. 3123 (42 U.S.C. 5851); sec. 102, Pub. L.91-190, 83 Stat 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135, 137, 141, Pub. L.97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153, 10155, 10157, 10161, 10168).

Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat 955 (42 U.S.C. 2239); sec. 134, Pub. L.97-425, 96 Stat. 2230 (42 U.S.C. 10154). Section 72.96(d) also issued under sec. 145(9), Pub. L. 100-203, 101 Stat.

1330-235 (42 U.S.C. 10165(9)). SubpartJ also issued under secs. 2(2), 2(15), 2(19), 117(a),

141(h), Pub. L.97-425, 96 Stat. 2202, 2203, 2204, 2222, 2224, (42 U.S.C. 10101, 10137,(a),

10161(h)). Subparts Kand Lare also issued 1_*..,-::!er sec. 133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 (42 U.S.C. 10198).

I 67

5. Section 72.216 is revised by amending paragraphs (a), (b), and (c) to read as follows:

§ 72.216 Reports.

(a) Reserved.

(b) Reserved.

(c) The general licensee shall make initial and written reports in accordance with

§§ 72. 7 4 and 72. 75.

Dated at Rockville, Maryland, this J.5 ~ day of June, 1999.

For the Nuclear Regulatory Commission.

~~~-~

Secretary of the Commission.

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