ML23151A484

PR-050 - 59FR48180 - Technical Specifications

ML23151A484
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Issue date:	09/20/1994
From:	Hoyle J NRC/SECY
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PR-050, 59FR48180
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ADAMS Template: SECY-067 09/20/1994 PR-050 - 59FR48180 - TECHNICAL SPECIFICATIONS PR-050 59FR48180 RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

STATUS OF RULEMAKING PROPOSED RULE:

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

TECHNICAL SPECIFICATIONS PROPOSED RULE FED REG CITE:

59FR48180 PROPOSED RULE PUBLICATION DATE:

09/20/94 ORIGINAL DATE FOR COMMENTS: 12/05/94 NUMBER OF COMMENTS:

3 EXTENSION DATE:

I I

FINAL RULE FED. REG. CITE: 60FR36953 FINAL RULE PUBLICATION DATE: 07/19/95 NOTES ON: SEE ALSO FINAL POLICY STATEMENT ON TECHNICAL SPECIFICATIONS PUBLIS STATUS HEDON 7/23/93 AT 58 FR 39132.

FILE LOCATED ON Pl.

OF RULE:

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

TECHNICAL SPECIFICATIONS PROPOSED RULE PROPOSED RULE DATE PROPOSED RULE SECY PAPER: 94-156 SRM DATE:

07/28/94 SIGNED BY SECRETARY:

09/14/94 FINAL RULE FINAL RULE DATE FINAL RULE SECY PAPER: 95-128 SRM DATE:

06/22/95 SIGNED BY SECRETARY:

08/24/95 STAFF CONTACTS ON THE RULE CONTACT!: CHRISTOPHER I. GRIMES CONTACT2:

MAIL STOP: O-11E22 PHONE: 504-1161 MAIL STOP:

PHONE:

DOCKET NO. PR-050 (59FR48180)

DATE DOCKETED DATE OF DOCUMENT In the Matter of TECHNICAL SPECIFICATIONS TITLE OR DESCRIPTION OF DOCUMENT

---

s---------------------------------------------

09/15/94 12/05/94 12/07/94 12/08/94 07/18/95 04/16/96 09/14/94 12/05/94 12/07/94 12/02/94 07/13/95 04/05/96 FEDERAL REGISTER NOTICE - PROPOSED RULE COMMENT OF NUCLEAR ENERGY INSTITUTE (THOMAS E. TIPTON) (

1)

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

2)

COMMENT OF UNION ELECTRIC (A.C. PASSWATER) (

FEDERAL REGISTER NOTICE - FINAL RULE LTR FM J.W. DAVIS, NEI, TO C. GRIMES, NRC, REGARDING IMPLEMENTATION OF IMPROVED TECHNICAL SPECIFICATIONS

3)

DOCKETED USNRC I

NUCLEAR ENERGY INSTITUTE

'96 APR 16 P 3 :58 OFF IC[ c1::- SECPE TARY OOCKt. 1 1 ~~~ &.:,tf<VICE 8fU NCH April 5, 1996 Mr. Christopher I. Grimes Branch Chief, Technical Specifications Branch Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 PROJECT NUMBER: 689

Dear Mr. Grimes:

James W. Davis PROJECT MANAGER OPERATIONS & MANAGEMENT DOCKET NUMBER PR PROPOSED RULE 5 0

( 59 FR-4~ \-Be)

The purpose of this letter is to express industry concern regarding the NRC staffs ability to review the large number of expected Improved Technical Specifications (ITS) submittals by the end of 1996. We are concerned that costs will be higher and review schedules extended beyond that seen in the pilot process, which were used by utilities in their cost-benefit studies for implementation of the ITS.

The lead plants, as well as a few of the early follow-on plants, have received NRC Safety Evaluation Reports (SERs) for their conversion submittals. Each of these SERs was received on a schedule consistent with the utility's anticipated implementation date. The NRC review process used for these plants included the dedication of a lead reviewer from the Technical Specifications Branch, supplemented by limited use of contractor personnel. This experience has provided an industry expectation on the cost of conversion to the ITS.

It is expected that the NRC will receive ITS conversion submittals from approximately 20 plants before the end of calendar year 1996. Many of these submittals are the result of strong NRC encouragement to pursue conversion to ITS. In recent discussions, you have indicated the NRC will be unable to support the upcoming submittals with the same process used for the lead and early follow-on plants. Some possible alternatives to the review process may include increasing reliance on contractor support, reassigning Project Managers (PMs) to coordinate ITS reviews and directing review responsibilities to NRR technical review branches.

I 776 I STREET, NW SUITE 400 WASHINGTON, DC 20006-3708 PHONE 202.739.8000 FAX 202.785 4019

U.S. NUCLEAR REGULATORY COMMISSIOt-.

DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date'?\<<:' & ~

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Coples Received_l------

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Mr. Christopher I. Grimes April 5, 1996 Page 2 There are three issues that must be considered in evaluating any alternate approach to the review process for conversion to the ITS:

1. Lack of familiarity with the ITS process by some NRC staff reviewers and many contractors will increase review time;

2. The potential for reopening previously disputed and resolved technical issues contained in the ITS NUREGs; and

3. Competition for PM resources from other licensing actions and the impact on timely completion of the SERs.

Most utility decisions to convert to ITS are based on a comparison of the costs associated with developing, obtaining NRC approval of, and implementing plant-specific ITS versus the benefits associated with operation under ITS.

The cost assumptions used in these comparisons could be adversely impacted if the continuity of the review process is not maintained. The efficiency of standardization is lost if reviewers or contractors are not familiar with the ITS, leading to unnecessary rounds of questions or requests for additional information.

Similarly, re-opening previously settled technical issues may lead to increased reviews of these issues. Both these circumstances would directly impact the amount of licensee resources involved in the conversion as well as NRC resources and review fees charged for this licensing action. In addition, initial cost estimates for conversion did not assume any costs associated with diverting NRC PM resources away from other licensing actions needed to support operation of the plants.

With respect to the benefits associated with ITS implementation, most utilities are estimating 6 months of NRC review time, as measured from the date of submittal.

This estimate is based on previous statements from the NRC staff and senior management, specifically Bill Russell's comments at the 1995 Regulatory Information Conference stating that a conversion could be reviewed by the NRC in four to six months. We acknowledge that plant-specific technical issues could add some time to the review of the submittal, but these issues should be able to be resolved on a parallel path, and should add no more than two months to the NRC review schedule.

Our other concern is that for the majority of licensees converting to ITS, the timing of the implementation of the ITS was tied to a particular evolution or scheduled activity. If implementation is delayed, this could result in a reduction in the

Mr. Christopher I. Grimes April 5, 1996 Page 3 benefits associated with ITS as well as significant hardship associated with advance planning of an outage.

We are providing these concerns for your consideration in your proposed changes to the conversion process to handle increased ITS conversion submittals. We request a meeting with you to discuss the NRC and industry ITS conversion process. In addition, we believe there will be some additional comments on this issue at the upcoming Regulatory Information Conference breakout session on Improved Technical Specifications.

4I Sincerely, JHE/rs:ec

DOCKET NUMBER PROPOSED RULE 5 o (5q&4-~ l~D}

NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AFO&

Technical Specifications AGENCY:

Nuclear Regulatory Comission.

ACTION:

Final rule.

DOCKETED US RC

[75go-ol-P]

• 95 JJL 18 PJ :11 OFFICE Of S[C~ET RY r "" ' /

~. *....., I,* I -~..

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations pertaining to technical specifications for nuclear power reactors.

The rule codifies criteria for determining the content of technical specifications.

Each licensee covered by these regulations may voluntarily use the criteria as a basis to propose the relocation of existing technical specifications that do not meet any of the criteria from the facility license to licensee-controlled documents.

The voluntary conversion of current technical specifications in this manner is expected to produce an improvement in the safety of nuclear power plants through a reduction in unnecessary plant transients and more efficient use of NRC and industry resources.

EFFECTIVE DATE:

1 loOrR..~CoC\53

'-f>~ '"1 \ l9 \C\5°

FOR FURTHER INFORMATION CONTACT:

Christopher I. Grimes, Chief, Technical Specifications Branch, Division of Project Support, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Connission, Washington, DC 20555-0001, Telephone:

(301) 415-1161.

SUPPLEMENTARY INFORMATION:

Background

Section 182a. of the Atomic Energy Act of 1954 (Act), as amended (42 U.S.C. 2232), mandates the inclusion of technical specifications in licenses for the operation of production and utilization facilities. The Act requires that technical specifications include information concerning the illlOunt, kind, and source of special nuclear material; the place of use; and the specific characteristics of the facility. That section also states that technical specifications shall contain information the Comission requires through regulation to enable it to find that the utilization of special nuclear material will be in accord with the co111non defense and security and will provide adequate protection of public health and safety. Finally, that section requires technical specifications to be made a part of any license issued.

The Comission promulgated§ 50.36, *Technical Specifications,* which implements Section 182a. of the Atomic Energy Act on December 17, 1968 (33 FR 18610). This rule delineates requirements for determining the contents of technical specifications. Technical specifications, at a minimum, must set 2

forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection of the health and safety of the public. Specifically,§ 50.36 requires the following:

Each license authorizing operation of a production or utilization facility of a type described in§ 50.21 or§ 50.22 will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to

§ 50.34. The C0111J1ission may include such additional technical specifications as the Commission finds appropriate.

Technical specifications cannot be changed by licensees without prior NRC approval.

However, since 1969, there has been a trend toward including in technical specifications not only those requirements derived from the analyses and evaluation in the safety analysis report but also essentially all other Comission requirements governing the operation of nuclear power reactors.

This extensive use of technical specifications was due in part to a lack of well-defined criteria (in either the body of the rule or in some other regulatory document) for what should be included in technical specifications.

Since 1969, this use has contributed to the volume of technical specifications and to the several-fold increase in the number of license amendment applications to effect changes to the technical specifications. It has diverted both NRC staff and licensee attention from the more important require111ents in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety.

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On ~rch 30, 1982 (47 FR 13369), the NRC published in the Federal Register I proposed amendment to Part 50.

The proposed rule would have revised§ 50.36, *Technical Specifications,* to establish a new system of specifications divided into two general categories. Only those specifications contained in the first general category as technical specifications would have become part of the operating license and would have required prior NRC approval for any changes. Those specifications contained in the second general category would have become supplemental specifications and would not have required prior NRC approval for 110st changes. The NRC review of the first general category of specifications would have been the same as that currently performed for technical specification c}Janges, which are amendments to the operating license. For the second category, *supplemental specifications,* the licensee would have been allowed to make changes within specified conditions without prior NRC approval.

The NRC would have reviewed these changes when they were made and would have done so in a manner similar to that currently used for reviewing design changes, tests, and experiments performed under the provisions of§ 50.59. Because of difficulties with defining the criteria for dividing the technical specifications into the two categories of the proposed rule and because of other higher priority licensing work, the proposed amendment was deferred.

In the early 1980s, the nuclear industry and the NRC staff began studying whether the existing system of establishing technical specification requirements for nuclear power plants needed improvement. During this period, an NRC task group known as the Technical Specifications Improvement Project (TSIP) and a Subcommittee of the Atomic Industrial Forum's (Alf's} Comittee 4

OR Reactor Licensing and Safety performed two studies of this issue. 1 The overall conclusion of these studies was that many improvements in the scope and content of technical specifications were needed and that a joint NRC and industry program should be initiated to implement these improvements.

Both groups mda specific recOIJIDQndations; these are s1J11111arized as follows:

(1) The NRC should adopt the criteria for defining the scope of technical specifications proposed in the Alf and TSIP reports. Those criteria should then be used by the NRC and each of the nuclear steam supply system vendor owners groups to completely rewrite and streamline the existing standard technical specifications (STS). This process would result in the transfer of many requirements from control by technical specification requirements to control by other 110chanisms [e.g., the final safety analysis report (FSAR),

operating procedures, quality assurance (QA) plan] that would not require a license amendment or prior NRC approval when changes were needed.

The new STS should place greater emphasis on human factors principles in order to make the text of the STS clearer and easier to understand.

The new STS should also improve the bases section of technical specifications, which gives the purpose for each requirement in the specification.

1SECY-86-10, *Recommendations for Improving Technical Specifications,*

January 13, 1986, contains both *Reco11111endations for Improving Technical Specifications,* NRC Technical Specifications Improvement Project, September 30, 1985, and *Technical Specifications Improvements,* Alf Subcommittee on Technical Specifications Improvements, October 1, 1985.

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(2) A parallel program of short-tera improvements in both the scope and substance of the existing technical specifications should be initiated in addition to developing new STS as stated in recOIIID8ndation 1.

On February 6, 1987 (52 FR 3788), the NRC published in the Federal Register for public coanent an *interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors* (interim policy statement) containing proposed criteria in response to recoanendation 1.

These criteria were generally derived from the criteria proposed in the Alf and TSIP reports and were modified slightly on the basis of discussions between the NRC staff and the industry. The public co1J111ent period for the interim policy statement expired on March 23, 1987.

The criteria were developed with the intention that they would apply to limiting conditions for operation (LCOs).

The NRC staff believed that the safety li its needed to remain unchanged in the technical specifications because of their 1110re direct link to protection of the physical barriers that guard against the uncontrolled release of radioactivity. At the time the criteria were developed, the industry did not wish to address administrative controls and design features in the effort to improve the STS.

Later, however, both the industry and the NRC staff realized that it would be beneficial to include upgraded administrative controls and design features in the improved STS, and these were handled separately fr011 the application of the criteria to the LCOs.

6

The NRC has developed a progra11 for short-tara improvements as described in recOIAlll8ndation 2 (above). These are known as *1ine-ite11* improvements and are generic improvements developed and promulgated by the NRC staff for voluntary adoption by licensees.

Subsequently, improved vendor-specific STS were developed and issued by the NRC in September 1992. The improved STS were published as the following NRC reports:

NUREG-1430, *standard Technical Specifications, Babcock and Wilcox Pl ants*

NUREG-1431, *standard Technical Specifications, Westinghouse Plants*

NUREG-1432, *standard Technical Specifications, Combustion Engineering Plants*

NUREG-1433, *standard Technical Specifications, General Electric Plants, BWR/4*

NUREG-1434, *standard Technical Specifications, General Electric Plants, BWR/6*

Copies of these NUREGs, as revised, may be purchased from the Superintendent of Documents, U.S. Government Printing Office, by calling (202) 275-2060 or by writing to the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082. Copies are also available fr0111 the National Technical Information Service, 5825 Port Royal Road, Springfield, VA 22161.

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These iaproved STS were the result of extensive technical meetings and discussions~ the NRC staff, industry owners groups, vendors, and the Nuclear Management and Resources Council (NUMA.RC).

On July 22, 1993 (58 FR 39132), the Commission published a *final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors*

(final policy statement), which incorporated experience and lessons learned since publication of the interim policy statement. The C011111ission has decided not to withdraw the final policy statement because it contains detailed discussions of the four criteria and guidance on how the NRC staff and licensees should apply the criteria.

The interim policy statement identified three criteria to be used to define which of the current technical specification requirements should be retained or included in technical specifications and which LCOs could be relocated to licensee-controlled documents, as follows:

Criterion 1:

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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Criterion 3: A structure, syste11, or component that is part of the primary success path and which functions or actuates to itigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The interim policy statement also stated that, in addition to structures, systems, and components captured by the three criteria, it was the Coaaission's policy that licensees retain in the technical specifications LCOs for a specified list of systems that operating experience and probabilistic risk assessment (PRA) had generally shown to be important to public health and safety. In the final policy statement, the COlllUission retained this thought as a fourth criterion as follows:

Criterion 4:

A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

As stated in the final policy statement, if a requirement meets any~ of the four criteria, it should be retained or included in technical specifications.

The final policy statement also addressed co111Dents received on the interim policy statement and described the CoD111ission's intent with regard to use of the criteria and their codification through rulemaking.

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This final rule codifies the four criteria contained in the final policy stateaent fer defining the scope of LCOs in technical specifications. These criteria are intended to be consistent with the scope of technical specifications as stated in the Statement of Consideration for the final rule issuing§ 50.36 (33 FR 18610, December 17, 1968). The Statement of Consideration discussed the scope of technical specifications as including the following:

In the revised system, emphasis is placed on two general classes of technical matters:

(1) those related to prevention of accidents, and (2) those related to mitigation of the consequences of accidents.

By systematic analysis and evaluation of a particular facility, each applicant is required to identify at the construction permit stage, those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity. Such items are expected to be the subjects of Technical Specifications in the operating license.

The first of these two general classes of technical matters to be included in technical specifications is captured by Criteria 1, 4, and, to some extent, Criterion 2, in that they address systems and process variables that alert the operator to a situation when accident initiation is more likely. The second general class of technical matters is explicitly addressed and captured by Criteria 2, 3, and 4.

By applying the four criteria contained in this rule, a licensee should capture the conditions for operation of its facility that are required to meet the principal operative standard in Section 182a. of the 10

Atomic Energy Act, thit is, that idequata protection is provided to the health and sifety of the public.

The C01111ission recognizes that the four criterii carry a theme of focusing on the technical requirements for features of controlling importance to Sifety. Since many of the requirements ire of significance to the health ind safety of the public, this rule reflects the subjective statement of the purpose of technical specificitions expressed by the Atomic Safety and Licensing Appeal Board in Portland General Electric Company (Trojan Nuclear Plant), ALAB-531, 9 NRC 263 (1979). There, the Appeal Board interpreted technical specifications as being reserved for those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

The Coa:aission wishes to emphasize that this rule is intended to be consistent with the language of Section 182a. of the Atomic Energy Act, the current§ 50.36 rule, and previous interpretations of the regulations.

This rule merely clarifies the scope and purpose of technical specifications by identifying criteria which can be used to establish, more clearly, the framework for LCOs in technical specifications.

The Commission believes that amending§ 50.36 to include the four criteria contained in the final policy statement will codify a viable, potentially safety-enhancing and cost-saving method for technical specification improvement.

The Comission continues to encourage licensees to use the 11

i111proved STS as the ~sis for plant-specific technical specifications.

As stated in the final policy statement, the C011111ission will place the highest priority on requests based on the criteria for individual license amendments that are used to evaluate all of the LCOs for an individual plant to determine which LCOs should be included in the technical specifications. Related surveillance requirements and actions would be retained for each LCO that remains in the technical specifications. Each LCO, action, and surveillance requirement should have supporting bases. Such requests would constitute c0111plete conversions to the improved STS.

In addition, the Comission will also entertain requests to adopt portions of the improved STS, even if the licensee does not adopt all STS improvements.

These portions will include all related requirements and will be developed as line-itea improvements by the NRC staff when they are clearly generic in nature, when there is evidence that a significant number of licensees could benefit fro111 the improvement, and when the industry expresses interest in the iinprovement.

The COD111ission encourages all licensees who submit technical specification related submittals based on these criteria to emphasize human factors principles to the extent practical consistent with the fomat and content of their current technical specifications.

LCOs that do not meet any of the criteria, and their associated actions and surveillance requirements, may be proposed for relocation fr011 the technical specifications to licensee-controlled documents, such as the FSAR.

The criteria may be applied to either standard or custoa technical 12

specifications. The Coaaission will also consider the criteria in evaluating future generic requirements for inclusion in technical specifications.

The C01111ission expects that licensees, in preparing their technical specification sublllittals, will ~tilize any plant-specific PRA or risk survey and any available literature on risk insights and PRAs.

This material should be employed to strengthen the technical bases for those provisions that remain in technical specifications,. when applicable, and to indicate whether the provisions to be relocated contain constraints of importance in limiting the likelihood or severity of the accident sequences that are co111DOnly found to dominate risk. Similarly, the NRC staff has and will continue to employ risk insights in evaluating technical specifications submittals.

In addition to the use of PRA in Criterion 4 to determine the scope of technical specifications, PRA has been used as a basis for a number of improvements to the content of technical specifications over the last several years. The NRC staff has approved several relaxations in technical specification allowed outage times and surveillance test intervals which were based on PRA.

In addition, the NRC staff used PRA to develop screening criteria to evaluate all of the changes in allowed outage times and surveillance test intervals that were made during the development of the improved STS.

The industry and the NRC staff have used PRA to an even greater extent in the development and review of the technical specifications for advanced reactor designs.

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The industry and the NRC staff are currently exploring several new approaches to utilizing PRA for technical specification improvements including the use of on-line risk assessment tools. In addition, the industry and the NRC staff are using PRA to explore further improvements in technical specifications by examining the risks during shutdown and during the transition between IIOdes of operation.

As a part of this ongoing program of improving technical specifications, the C0111Dission will continue to consider methods to aake batter use of risk and reliability information for defining future generic technical specification requirements.

During technical specification conversions, the staff will apply the backfit rule(§ 50.109) when adding new requirements from the improved STS to individual plant technical specifications, provided the licensee does not voluntarily accept the new requiremen~s.

If, however, the staff suggested additional changes are needed to make the licensee requested changes acceptable from the standpoint of adequate protection or compliance with NRC regulations, § 50.109(a)(2) and§ 50.109(a)(3) do not apply and the request may be denied without the additional items.

Summary of Public Connents The C011111ission received three letters co11111enting on the proposed rule.

Each letter contained several comments.

One C01111118nter representing the connercial nuclear industry expressed concern that there is insufficient regulatory guidance on how the NRC staff 14

intends to iaplement this rule with respect to the fourth criterion

[§ 50.36(c)(2)(ii)(D)]. The commenter believes that this rule should not be IIIOdified until the NRC and the industry have reached a c0111110n understanding of the application, threshold, and intent of Criterion 4.

The coaaenter stated,

• it is our view, and the Commission apparently recognizes, that this criterion goes beyond the 'adequate protection' standard for public health and safety and license compliance purposes embodied in the first three criteria.*

Similar to this conwent on the proposed rule, the Advisory C0111Dittee on Reactor Safeguards (ACRS) c0f1111ented in a June 18, 1993, letter to the Chairman that the NRC staff needs to provide more detailed guidance on the definition of "significant to public health and safety,* as it is used in Criterion 4.

Criterion 4 is intended to capture those constraints that probabilistic risk assessment or operating experience show to be significant to public health and safety, consistent with the Co1m1ission*s PRA Policies.* The level of significance either would need to be such that it justified including the constraints in the technical specifications to ensure adequate protection of the public health and safety or that the addition of such constraints provides substantial additional protection to the public health and safety.

The C01111ission identified four systems that meet Criterion 4 in the final policy statement based on previous qualitative reviews of operating experience and risk. They are reactor core isolation cooling/isolation condenser, residual heat removal, standby liquid control, and recirculation pump trip.

The Commission recognizes, however, that other structures, systems, or 15

COlllpOA&Ats uy aeet tbis criterion. Plant-and design-specific PRAs have yielded valuible insight to unique plant vulnerabilities not fully recognized in the safety, design basis accident, or transient analyses.

The NRC's current regulator¥ requirements are largely based on detanainistic engineering criteria involving the use of multiple barriers and defense in depth. Recently, the NRC staff has formulated a comprehensive plan for the application of PRA technology and insights throughout the agency. It is expected that the PRA Implementation Plan will serve as the framework for continued and future applications of PRA at the NRC.

Implementation of this plan will increase the systematic use of risk assessment techniques.

To ensure consistent and appropriate decision-making that incorporates PRA methods and results, it is important that coherent and clear application guidelines are applied.

As part of the PRA Implementation Plan, such guidelines will be established (incorporating safety goals and backfit rule considerations) that address the interdependence of probabilistic risk and deterministic engineering principles. The process of developing these guidelines will involve c011111Unications among the NRC staff, the nuclear industry, and the public to ensure that all parties understand the role of PRA methods and results in NRC's risk management efforts. The NRC staff anticipates that, as it gains experience with the development and use of such PRA application guidelines, it will be better able to refine such phrases as

• significant to public health and safety,* and other phrases that are used in many of the COD111ission's regulations.

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The Colllllission could delay publication of this final rule until the PRA application guidelines are in place. However, the Coaaission believes that the experience gained while using the criteria under the interi and final policy statements combined with the limitations imposed on the NRC staff by the backfit rule provide assurance that, in the interim, the staff's use of Criterion 4 to apply PRA to technical specification content will be properly controlled. The Conaission has concluded that it is appropriate to publish this final rule, which provides the framework for technical specifications, at this time.

One c0111118nter stated that the PRA portion of the fourth criterion should Plant be clarified to include only those equipment items ~mportant to risk-significant sequences as defined in Generic Letter 88-20, *Individual Examination for Severe Accident Vulnerabilities,* Appendix 2, and reported in licensees' individual plant examination (IPE) reports.

The IPE program has resulted in connercial reactor licensees using risk-assessment methods to identify plant-specific severe accident vulnerabilities.

Since submittal of their IPE reports, any licensees have enhanced their plant-specific PRAs and have gained additional insights into unique plant vulnerabilities. These additional insights from PRAs are being used by licensees in such areas as implementation of the maintenance rule.

As stated in the Commission's *Proposed Policy Statement on the Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities,* the use of PRA technology should be increased in all regulatory matters to the 17

'<1 4

extent supported by the state of the art in PRA 11athods and data and in a aarmer t~t COlll)lemants the NRC's detenainistic approach and supports the NRC's traditional defense-in-depth philosophy. The C01111ission will continue to apply PRA to technical specifications in accordance with its proposed policy statement on the use of PRA.

In addition, guidance for specific applications or classes of applications will be developed under the PRA Implementation Plan. The C0111111ss1on believes this 1s a 110re appropriate aeans to define how Criterion 4 will be used in practice, rather than to limit the structures, systems, and components captured by Criterion 4 to those items important to risk-significant sequences as defined in Generic Letter 88-20, Appendix 2, and reported in licensees' IPE reports. The Comnission believes that this process will provide the NRC staff and the industry with additional risk insights, beyond those identified through the IPE program.

The same c0111Denter said that the operating experience portion of the fourth criterion should be deleted because it is subjective and because no equipment would satisfy only that portion of the fourth criterion and none of the other criteria.

While operating experience is an important part of PRA, not all PRA models are sophisticated enough to capture all operating experience. The COR111ission believes that operating experience can play an important role in determining the safety significance of structures, systems, and components and that there will be no adverse impact by including operating experience as part of Criterion 4.

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One comenter emphasized that the development of implementation guidance, especially with respect to Criterion 4, should be consistent with the impleaentation guidance of the aaintenance rule.

As stated previously, the Cpmmission believes that the improved STS, the final policy statement, the backfit rule(§ 50.109), and the statement of consideration for this rule contain sufficient guidance on implementation of the criteria to proceed with rulemaking. Supplementary guidance will continue to be provided to the NRC staff that will support the process for implementing the four criteria on both a generic and plant-specific basis, and will be publicly available. The NRC staff will ensure that any guidance documents that relate to the implementation of the four criteria will be consistint with the implementation guidance of the maintenance rule along with the guidance for other rules promulgated by the Commission.

One COIIIIK!nter expressed a concern with respect to the level of PRA information necessary to support the relocation of existing technical specifications which do not meet the first three criteria.

If a technical specification provision does not meet any of the first three criteria, and if the current PRA knowledge or operating experience does not identify the structure, system, or component as risk significant, the NRC staff will not preclude relocating such technical specifications. The level of PRA information necessary to support relocation would be considered as part of the overall review of the supporting basis for the proposed change.

The C0111Rission expects that licensees will utilize PRA insights to indicate 19

whether the provisions to be relocated contain constraints of importance in li iting the likelihood or severity of the accident sequences that are coanonly found to doainate risk.

One cOlllll8nter stated that the implementing guidance needs to be clear on how the proposed criteria would be used to detennine if new requireaents are to be incorporated into technical specifications.

The C01111ission believes that the improved STS, the final policy statement, the backfit rule(§ 50.109), and the statement of consideration for this rule contain sufficient guidance on implementation of the criteria. The staff will also ensure that application of the criteria to new requirements is consistent with the guidance in the draft *Regulatory Analysis Guidelines,* Revision 2, published in August 1993 (NUREG/BR-0058), and the final version of Revision 2 when it is approved by the Comnission.

In addition, the NRC has recently published NUREG/CR-6141, *Handbook of Methods for Risk-Based Analyses of Technical Specifications,* December 1994, which sUJ1111arizes systematic risk-based methods to improve various aspects of technical specification requirements. The handbook was developed through research sponsored by the NRC and will be used as a reference document to assist the NRC staff in reviewing licensees' risk-based analyses submitted as part of the bases for proposed changes in facility technical specifications. This guidance will be updated periodically to incorporate lessons learned and changes in the state of the art, will help ensure the criteria are applied in a consistent and controlled manner, and will be publicly available. As stated above, as part of the PRA Implementation Plan, PRA application guidelines will be established 20

(incorporating safety goals and backfit rule considerations) that address the interdependence of probabilistic risk and detenainistic engineering principles. As these application guidelines develop, they will progressively be used to provide guidance to the NRC staff on the use of the criteria contained in this rule and the ~pplication of the backfit rule to new regulatory requirements.

Ona coamenter stated that the same or similar criteria to those in the rule should also be applied to 10 CFR 50.36(c)(3), (4), and (5), so that surveillance requirements, design features, and administrative controls which do not provide the necessary *adequate protection of the health and safety of

~

the public* can be relocated to other licensee-controlled documents.

With respect to§ 50.36 (c)(3), *surveillance Requirements,* the C01D111ission stated in the final policy statement that appropriate surveillance requirements and actions should be retained for each LCO which remains or is included in the technical specifications.

The criteria in§ 50.36(c)(2) apply to safety functions. Therefore, the C011111ission does not believe that these criteria can be appropriately applied to the types of requirements found in the *design features* and

• administrative controls* sections of the technical specifications. The NRC staff has, however, been pursuing separate improvements to these requirements, in cooperation with industry, using the intent of the criteria to identify the optimum set of requirements in each of these areas and to eliminate redundancy 21

to ether regulations consistent with the inillWI requirements of§ 50.36 and the At0111c Energy Act, as amended.

One c011Denter stated that the removal of items fr011 plant technical specifications may decrease enf~rceability and licensee attention to safety.

The Coaaission does not agree that the removal of items from plant technical specifications will decrease licensee attention to safety.

On the contrary, the Coaission believes thit implementation of the criteria contained in this rule will produce an improvement in the safety of nuclear power plants through the use of more operator-oriented technical specifications, improved technical specification bases, reduced action stitement induced plant transients, ind more efficient use of NRC and industry resources. Clirificition of the scope and purpose of technical specifications has provided useful guidance to both the NRC and industry and has resulted in improved technical specifications that ire intended to focus licensee and plant operator attention on those plant conditions ost important to safety.

The Conaission also does not igree that the removal of items froa plant technicil specifications will hive iny adverse impact on the NRC's ibility to take enforcement action on safety-significant issues. The improved STS ire intended specifically to focus on the operiting plant parameters and associated surveillance criteria of safety significance. The Co11111ission requires compliance with technical specifications, and expects adherence to comR1itments contiined in licensee-controlled documents. Violations and deviations will, as in the past, be handled in accordance with the NRC 22

enforcement policy in 10 CFR Part 2, Appendix C.

Any changes to a licensee's technical specifications to apply these criteria will be aade by the license amendment process prior to implementation.

When a licensee elects to apply these criteria, some requirements are relocated froa technical specifications to the FSAR or to other licensee-controlled documents. Licensees are to operate their facilities in conformance with the descriptions of their facilities and procedures in their FSAR.

Changes to the facility or to procedures described in the FSAR are to be aade in accordance with 10 CFR 50.59.

The Commission will take appropriate enforcement action to ensure that licensees comply with 10 CFR 50.59. Changes made in accordance with the provisions of other licensee-controlled documents (e.g., QA plan, security plan) are subject to the specific requirements for those documents.

Nothing in this rule limits the authority of the NRC to conduct necessary inspections and to take appropriate enforcement action when regulatory requirements or co11111itments are not met.

The same coninenter stated that the removal of items froa plant technical specifications will diminish public participation rights in the regulation of operating nuclear power plants by diminishing the universe of potential operating license amendment cases.

Any changes to a licensee's technical specifications to apply these criteria will be made by the license amendment process before implementation.

The review of each license amendment will involve an opportunity for public_

participation. One of the goals of the technical specifications improvement 23

progr111 wu to aake aore efficient use of NRC and industry resources by focusing attention on those plant conditions most important to safety and, in turn, reducing the number of license uaendment requests. Since 1969, there has been a trend toward including in technical specifications not only those requirements derived froa tha analyses and evaluations included in the safety analysis report but also essentially all other C0111Rission requirements governing the operation of nuclear power reactors. This extensive use of technical specifications is due in part to a lack of well-defined criteria (in either the body of tha rule or in some other regulatory document) for what should be included in technical specifications. This has contributed to the volume of technical specifications and to the several-fold increase, since 1969, in the number of license amendment applications to effect changes to the technical specifications. It has diverted both NRC staff and licensee attention froa the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety.

The c0111J1enter found it curious that an industry and an agency that claim to be able to quantify the risks of nuclear power are unable to quantify this illlJ)act on safety, and stated, *Perhaps if it is unquantifiable, the alleged adverse impact does not really exist.*

The Colllllission agrees that there are li itations and uncertainties in the ability to quantify the impact on safety described above.

Uncertainties exist in any regulatory approach and these uncertainties are derived froa knowledge limitations. A probabilistic approach has exposed some of these limitations and yielded an improved framework to better focus and assess their 24

significance and 1ssist in developing I strategy to accOIIIDOdate thea in the regulatory process.

Tb& Commission does not intend, however, to let these li itations prevent it froa taking steps to improve the regulations in a manner that will have substantial safety benefits. The C011111ission believes the public will be better serve4 by focusing both NRC and industry attention on the ost safety-significant items.

The NRC staff has made three changes to this rule since it was published in its proposed fora.

The first change was made in order to maintain consistency with other NRC staff and Comission documents that have been issued since this rule was published in its proposed form.

In

§ 50.36(c)(2)(ii)(D), the term *probabilistic safety assessment* has been changed to *probabilistic risk assessment.*

The second and third changes are in§ 50.36(c)(2)(iii). The beginning of the first sentence was changed to read, *A licensee is not required to propose to 1110dify technical specifications.**

• rather than *A licensee is not required to IAOdify technical specifications ***

• This change was made to clarify that a licensee would be required to modify their technical specifications if the Commission determined that a new requirement was necessary in accordance with the backfit rule and the new requirement met one of the four criteria contained in§ 50.36(c)(2)(ii).

The third change is the deletion of the last sentence in

§ 50.36(c)(2)(iii). The sentence read, *However, for technical specification amendments a licensee proposes after [insert the effective date of this 25

docUIIIRAt], tbe criteria in paragraph (c)(2)(ii) of this section provide an acceptable scope for limiting conditions for operation.* This sentence was deleted because it did not add or modify any requirements and the thought is adequately expressed in this statement of consideration.

Finding of No Significant Environmental Impact: Availability The C0111ission has determined under the National Environmental Policy Act of 1969, as amended, and the C011111ission regulations in Subpart A of Part 51, that this final rule is not a major Federal action significantly affecting the quality of the human environment and will not degrade the environment in any way. Therefore, the Commission concludes that there will be no significant impact on the environment from this rule. This discussion constitutes the environmental assessment and finding of no significant impact for this rule; a separate assessment has not been prepared.

Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget, approval number 3150-0011.

26

Regulatory Analysis The Comission has determined that a regulatory analysis is not required for this rule. The Colllmission believes that the intent of the regulatory analysis has been met through t~ extensive consideration given to the development of the *final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors* and the improved STS, both of which gave the public an opportunity for coaaent.

In addition, the detenaination that no regulatory analysis is necessary was noted in the Federal Register Notice for the proposed rule, and the NRC received no coaments on this issue.

The criteria being added to§ 50.36 are the same as those contained*in the final policy statement and have been used by the NRC and the nuclear power industry to define the content of technical specifications since September 1992.

The rule does not impose any requirements but, rather, allows nuclear power reactor licensees to voluntarily use the criteria to relocate existing technical specifications that do not meet any of the criteria to licensee-controlled documents.

The NRC staff also uses these criteria to determine whether technical specifications are appropriate to provide regulatory control over new requirements or positions that have been justified consistent with the backfit rule.

The COA11ission considered the need for and consequences of this action when it made the decision not only to publish the criteria in the final policy statement but also to codify the criteria through rulemaking. Appropriate alternative approaches to this action have been identified and analyzed over 27

the life of the Technical Specifications Improvement Prograa, beginning with an earlier attempt to define the content of technical specifications through rulemaking.

As described in the background discussion, the Coaaission published a proposed amendment to§ 50.36 (47 FR 13369) on March 30, 1982.

However, because of difficultie~ with defining criteria for technical specifications and because of other higher priority licensing work, the rule change was deferred. In February 1987, the Connission published an *1nteria Policy Statement on Technical Specification Improvements f~r Nuclear *Power Reactors,* and in July 1993, published the final policy statement. During its review of the final policy statement, the Comission concluded that the four criteria should be codified in a rule. Thus, alternative approaches to regulatory objectives have been identified and analyzed, and the C0111Dission has decided that there is no preferable alternative to codifying the four criteria in a rule. With regard to evaluation of values and impacts of alternatives, the Comission believes there is no difference in the values or impacts of applying the criteria under the final policy statement or through a rule, except that the criteria are more readily available to future users in a rule rather than in a policy statement.

Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 [5 U.S.C.

605(b)], the Comuission certifies that this final rule does not have a significant economic impact on a substantial number of small entities. This 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 28

of *saall antitias* 1s given in the Regulatory Flexibility Act or the Small Business Size Standards in regulations issued by the Small Business Administration at 13 CFR Part 121.

~ackfit Analysis The NRC us detarained that the backfit rule,§ 50.109, does not apply to this final rule and, therefore, a backfit Ulalysis is not required for this final rule because these amendments do not involve any provisions that would impose backfits as defined in§ 50.109(a)(l).

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

For the reasons given 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. 552 and 553, the NRC is adopting the following amendment to Part 50.

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.

29

936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); sacs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 u.s.c. 5841, 5842, 5846).

Section 50.7 also issued un~er 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(dd), 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. 122, 68 Stat. 939 (42 U.S.C. 2152).

Sections 50.80-50.81 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. In§ 50.36, paragraphs (c)(2) and (3) are revised to read as follows:

§ 50,36 Technical specifications, (c)*

(2) Limiting conditions for operation, (1) Limiting conditions for operation are the lowest functional capability or perfonuance levels of equipment required for safe operation of the 30

facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action per11itted by the technical specifications until the condition can be met.

When a limiting condition for operation of any process step in the systea of a fuel reprocessing plant is not !R9t, the licensee shall shut down that part of the operation or follow any remedial action permitted by the technical specifications until the condition can be met.

In the case of a nuclear reactor not licensed under§ 50.2l(b) or i 50.22 of this part or fuel reprocessing plant, the licensee shall notify the C011111ission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence.

The licensee shall retain the record of the results of each review unt1l the CORDission terminates the license for the nuclear reactor or the fuel reprocessing plant.

In the case of nuclear power reactors licensed under

§ 50.2l(b) or§ 50.22, the licensee shall notify the Commission if required by

§ 50.72 and shall submit a licensee Event Report to the Commission as required by§ 50.73.

In this case, licensees shall retain records associated with*

preparation of a licensee Event Report for a period of three years following issuance of the report. For events which do not require a licensee Event Report, the licensee shall retain each record as required by the technical specifications.

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1.

Installed instrumentati'on that is used to detect, and indicate in the control room, a significant abnormal degradation of the 31

ructor coelant pressure boundary.

(B) Criterion 2.

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary swccess path and which functions or actuates to itigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4.

A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

(iii) A licensee is not required to propose to modify technical specifications that are included in any license issued before [insert the effective date of this document] to satisfy the criteria in paragraph (c){2){ii) of this section.

(3) Suryejllance requirements, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will 32

be within safety limits, and that the limiting conditions for operation will be 111et.

Dated at Rockville, Maryland, this

{ J -0_ day of

• -E-1] -, 1995.

e,7 For the Nuclear Regulatory C01111ission *

/

7 John..,t'. Hoyle, Secretary of the Couaission.

33

UMON ELECTRIC

~;z 1901 Chouteau Avenue Post Office Box 149 St. Louis, Missouri 63166 314-621-3222 DOCKET NUMBER Pl PROPOSED RULE S a_

( S-'1 FR LJ frl 'ro)

OOCr~ETEO USHRC

• 94 DEC -g Al: :33 December 2, 1994 Secretary U.S. Nuclear Regulatory Commission ATTN: Docketing and Service Branch Washington, D.C. 20555-0001 Gentlemen:

ULNRC-3109 PROPOSED RULE ON TECHNICAL SPECIFICATION IMPROVEMENTS

References:

1) 59FR4818O dated September 20, 1994

2)

ULNRC-3O23 dated May 20, 1994

3) 58FR39132 dated July 22, 1993

4)

NRC letter to Westinghouse Owners Group (T. Murley to R. Newton), "NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications,"

dated May 9, 1988 Union Electric concurs with the changes to 10CFR5O.36(c) as proposed in Reference 1. The following comments are in response to the Commission's specific request for input on the fourth criterion added to 10CFR5O.36(c) as subpart (c) (2) (ii) (D) regarding those guidelines that the Commission should use in defining "significant to public health and safety."

Criterion 4 would retain in the Technical Specifications those structures, systems, and components (SSCs) that are safety significant based on probabilistic safety assessment (PSA) or operating experience.

Union Electric has submitted (Reference 2) a package of Technical Specification changes based on the Final Policy Statement on Technical Specification Improvements (Reference 3). The four criteria Acknowledged by card..............................,

. NUC~':~'- r :>_,_£\TORY COrviMISSION DOC-<E: 11, {; & SEAVlCE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Jiastmarx Date /:t-

/1~1 ICopie,;.w::ived ___

/ ____ _

Ad!! C:-", r; Reprc-0uced ~ ----

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Secretary USNRC -3109 Page 2 addressed in Reference 3 are the same.as those proposed for codification in Reference 1. The guidelines used to answer the PSA portion of the fourth criterion were based on the reporting criteria of Appendix 2 to Generic Letter 88-20. Technical Specifications covering equipment of prime importance in limiting the likelihood or severity of the risk-dominant sequences were retained. Risk-dominant sequences were reported in Section 3.4 of the Callaway IPE (if the sequence frequency was :::_ lE-06 per reactor year for core melt or:::_ lE-07 per reactor year for containment bypass). Screening forms addressing each of the four criteria (with yes/no answers and discussion) were developed for every LCO in support of Reference 2. In almost every case where an LCO was retained, the basis for retention was an affirmative answer to one or more of the first three criteria. If an LCO screening form contained an affirmative answer to the fourth criterion, in most cases the form also answered one or more of the first three criteria affirmatively. The only exceptions to the above were based on NRC positions in References 3 and 4, i.e. Technical Specifications 3.3.3.5 (fourth criterion answered affirmatively for remote shutdown instrumentation based on NRC concerns with fire risk expressed in Reference 4; currently being addressed in the IPEEE), and 3.9.8.1 and 3.9.8.2 (fourth criterion answered affirmatively for RHR in Mode 6 based on References 3 and 4).

The PSA portion of the fourth criterion should be clarified to include only those equipment items important to risk-significant sequences as defined in Generic Letter 88-20 Appendix 2 and reported in licensees' IPE reports. The operating experience portion of the fourth criterion should be deleted since it is subjective and since no equipment would satisfy only that portion of the fourth criterion and none of the other criteria.

We appreciate the opportunity to provide these comments.

Should you have any questions on the above, please contact us.

Sincerely, A. C. Passwater Manager, Licensing & Fuels GGY/kea

cc:

T. A. Baxter, Esq.

Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.

Washington, D.C.

20037 M. H. Fletcher Professional Nuclear Consulting, Inc.

18225-A Flower Hill Way Gaithersburg, MD 20879-5334 L. Robert Greger Chief, Reactor Project Branch 1 U.S. Nuclear Regulatory Commission Region III 801 Warrenville Road Lisle, IL 60532-4351 Bruce Bartlett Callaway Resident Office U.S. Regulatory Commission RR#l Steedman, MO 65077 L. R. Wharton (2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E21 11555 Rockville Pike Rockville, MD 20852 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102

December 7, 1994 DOCKET NUMBER PR 5 O PR POSED RULE~-=--..--

~ <I/ FR '-1 i'-J~O COMMENTS OF OHIO CITIZENS FOR RESPONSIBLE ENERGY, INC.

ON PROPOSED RULE, "TECHNICAL SPECIFICATIONS," 59 FED.

(SEPTEMBER 20, 1994)

OCRE's concerns about this proposed rule are:

oocKETlt~G &

SE VICE BR.~NCH I

1.

the removal of i terns from plant technical Y-NRC

~

decrease enforceability and licensee attention to safety. ~

, a

~ ~~

licensee violates the provisions of its license, of which t

}1'.rr1-cal specifications are a part, then there is clear cause

~

enforcement action.

Licensees may for this reason pay more attention to technical specification requirements than to such provisions when they are removed from the technical specifica-tions.

For example, see NUREG-1275, Vol. 4, "Operating Experi-ence Feedback Report - Technical Specifications":

"From our experience, we have found that the reporting of opera-tional events, even those of relatively low safety significance, does provide a

stimulus for licensees to conduct root cause analyses and to effect corrective action to prevent recurrence of such events.

This consideration must be carefully weighed when contemplating modification of reporting requirements for opera-tional events with relatively low safety significance."

p.

xviii.

Removing requirements from plant technical specifications will also remove the associated reporting requirements (10 CFR 50.73(a) (2)).

Diminished reporting requirements will also reduce the flow of data on plant safety and operations to the NRC, which will degrade the NRC's ability to monitor plant safety.

2.

the removal of items from plant technical specifications will diminish public participation rights in the regulation of operating nuclear power plants by diminishing the universe of potential operating license amendment cases.

A licensee cannot make any change to the tech specs (not even the correction of a

typographical error) without seeking a formal operating license amendment in which members of the public have the right to inter-vene.

A license amendment hearing is a full, formal evidentiary hearing held pursuant to Section 189a of the AEA and Subpart G of 10 CFR

2.

However, once provisions are removed from the tech specs to internal plant documents, they can be changed by the licensee under 10 CFR 50.59 with no public input.

The Federal Register notice states that the overly broad use of tech specs to impose requirements has resulted in an adverse but unquantifiable impact on safety.

OCRE finds it curious that an industry and an agency that claim to be able to quantify the risks of nuclear power are unable to quantify this impact.

Perhaps, if it is unquantifiable, the alleged adverse impact does 1

Acknowledged by card...... ~.~... ~.. ~... :.~~~... ~

. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date L1-/ j'-/ 4 Copies Received

/

Add'I Cop;es Rep.-ro-du-c-ed_:3 ____ _

Special Distribution fl Z: J{JS, POI(

f:;_,fl,,_,.

not really exist.

The Federal Register notice requests comment on what guidelines the NRC should use in defining "significant to public health and safety" in Criterion 4.

OCRE would recommend leaving this phrase undefined so as not to restrict the NRC's discretion as the regulator.

Respectfully submitted, Susan L. Hiatt Director, OCRE 8275 Munson Road Mentor, OH 44060-2406 (216) 255-3158 2

~~~~1:~~t~ Pl 5 o DOCKETED {f"ci F R 4 9--J ~O) lie, *ir, NU CLEAR ENERGY I NST ITUTE J

December 5, 1994 Mr. John C. Hoyle, Acting Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTN: Docketing and Service Branch

• 94 O[C -S Al 1 :S~homos E. Tipton VICE PRESIDENT, OP~RATIONS & ENGINEERING OFFICE:-, - :fC~~ *~~-r DOCKt Tlr '.3 &

. ;,r oH Hd

SUBJECT:

Proposed Amendment to 10 CFR 50.36, Technical Specifications

Dear Mr. Hoyle:

These comments are submitted on behalf of the Nuclear Energy Institute (NEI) 1 in response to the September 20, 1994, Federal Register notice (59 FR 48180 through 48183) concerning proposed changes to 10 CFR 50.36, Technical Specifications, that codify criteria for determining the content of technical specifications. NEI and its utility members support the proposed regulation that codifies the Final Policy Statement on Technical Specifications, but with reservations as noted below.

In the Final Policy Statement on Technical Specifications (58 FR 39132, July 22, 1993), the NRC staff provided these same criteria as guidance for determining which regulatory requirements and operating restrictions should be included in technical specifications.

These criteria were key to reaching agreement on the purpose and content of the Improved Technical Specifications (ITS) (published as NUREG-1430 through 1434).

In addition to providing criteria for the voluntary conversion of plant-specific technical specifications, we have encouraged the use of the criteria for other technical specification related activities, such as future generic requirements or specific changes to existing technical specifications. The rule would provide added stability in determining the scope and content of technical specifications for previously licensed plants as well as plants licensed for operation after the effective date of the rule. There is a brief discussion in the Statements of Consideration that the Commission will consider the four criteria in evaluating future generic requirements for inclusion in technical specifications. The 1NEI 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. NEI is the successor organization to the Nuclear Management and Resources Council (NUMARC).

1776 I STREET, NW SUITE 400 WASHINGTON, DC 20006-3708 PHONE 202 739.8107 FAX 202 785. I 898 r)

Acknowledged by card.... ?I~... ?..L.!P.QS.... ~

\J.S. NUC!..r,;;:',.. __ ~: '.1;-_;;,y C 1,,, 'll~SION ooc:..*< *; i ~c, (:..,::.d !CE SECTION OFFlr;f C,f* 1 HE SECRETARY OF fHE COMMiSSION Docume t Statistics Postma Dcte H~ cLe./ w v-4 Copies Recs\11.1d ___

....:../ ____ _

Add'I Copie~ Reoroduced....,,3..__..,,.... __

Special Dtstripution fl[ ns PO!I.,

L

Mr. John C. Hoyle December 5, 1994 Page 2 implementing guidance needs to be clear on how the proposed criteria would be used to determine if new requirements are to be incorporated into technical specifications.

We support the concept and the discussions describing the application of the first three criteria. Their use in creating the ITS NUREGs has led to a common understanding between the NRC staff and the industry as to the intended content of technical specifications. We feel that the codification of these criteria is an important step in maintaining the technical specifications as a document that identifies those primary features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval.

Our most significant concern is that there is insufficient regulatory guidance or policy on how the NRC staff intends to implement the regulation with respect to the fourth criterion of the proposed rule [§ 50.36(c)(ii)(D)]. While the first three criteria are relatively clear and concise requirements, the fourth criterion requires limiting condition of operation for... A structure, system or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

The Discussion of Criterion 4 in the Final Policy Statement on Technical Specifications states that the intent of this criterion is to retain or include in the technical specifications those requirements that probabilistic safety assessment (PSA) or operational experience expose as significant to public health and safety, consistent with the Commission's Safety Goal and Severe Accident Policies. It is our view, and the Commission apparently recognizes, that this criterion goes beyond the "adequate protection" standard for public health and safety and license compliance purposes embodied in the first three criteria.

The Statements of Consideration specifically request comment on Criterion 4 and what guidelines the Commission should use in defining the term "significant to public health and safety." We support the use of PSA techniques and applications but are concerned since the industry and NRC staff have not yet achieved a clear, consistent understanding regarding the use of PSA technology. In the past year, NEI has had several discussions with the NRC staff concerning the regulatory use of PSA. The industry's PSA Application Guide will provide a framework for using PSA insights as an aid in decision making. We believe the guide will effectively complement the Commission's PSA Policy Statement and agency implementation plan. We also provided extensive comments on the NRC's draft Regulatory Analysis Guidelines in December 1993.

Taken together, these documents should converge the techniques and methods for determining a threshold for evaluating current and future regulatory activities, including technical specifications. It is imperative that the discussion surrounding the intent and use of Criterion 4 for technical specifications be consistent with the overall policy discussions in the PSA Policy Statement. As such, we believe that more thought and explanation of the NRC staffs intent is required with respect to Criterion 4 prior to proceeding with rulemaking. NEI welcomes the opportunity to continue to work with the NRC staff to develop the additional guidance necessary to support implementation of Criterion 4. Once the NRC and the industry have reached a common understanding of the

Mr. John C. Hoyle December 5, 1994 Page 3 application, threshold and intent of Criterion 4, then 10 CFR 50.36 would be modified to reflect that understanding.

We also have a specific concern with respect to the level of PSA information necessary to support the relocation of existing technical specifications which do not meet the first three criteria. Neither Criterion 4 nor the accompanying discussion state whether qualitative PSA results based on a generic vendor/NSSS plant type or qualitative, plant-specific results would be acceptable. Requiring plant-specific results to support relocating technical specifications raises questions as to what is the appropriate level of detail, the acceptability, and the timeliness of the PSA. These issues are being addressed in ongoing NRC/industry interactions, and again a common understanding should be agreed to prior to rule making.

We note that the Maintenance Rule (10 CFR 50.65) and the implementation guidance created by NEI (NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," May 1993) provide requirements and guidance with respect to identifying risk-significant systems, structures and components (SSCs).

There is also a provision to perform a periodic assessment of the effectiveness of maintenance actions, which provides reasonable assurance that SSCs are capable of performing their intended function. In addition, there is a requirement to assess the impact on overall plant safety function upon removal of SSCs from service, to ensure that overall plant safety function are maintained. The development of implementation guidance, especially with respect to Criterion 4, should be consistent with the implementation guidance in these areas of the Maintenance Rule.

We also note that this proposed rule change only applies to 10 CFR 50.36(c)(2), "Limiting Conditions for Operation." These same or similar criteria should also be applied to 10 CFR 50.36(c)(3), (4), and (5), such that surveillance requirements, design features, and administrative controls which do not provide the necessary "adequate protection of the health and safety of the public" can be relocated to other licensee-controlled documents.

We appreciate the opportunity to comment on this important proposed rulemaking. If there are any questions, please contact Jim Eaton or Dave Modeen of the NEI staff.

JHE/rs

DOCKET NUMBER P~OPQSED RULE PR 5 0

( 5C/,FI< '1<J'-J';-O)

NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AF06 Technical Specifications AGENCY:

Nuclear Regulatory Commission.

ACTION:

Proposed rule.

'94 SEP 15 P 4 :4 8

SUMMARY

The Nuclear Regulatory Commission (NRC) is proposing to amend its regulations pertaining to technical specifications for nuclear power reactors.

The proposed rule would codify criteria for determining the content of technical specifications. These criteria were developed in recognition of the overly broad use of technical specifications to impose requirements, diverting both NRC and licensee attention from the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety.

Each licensee covered by these regulations may voluntarily use the criteria as a basis to propose the relocation of existing technical specifications that do not meet any of the criteria from the facility license to licensee-controlled documents.

The voluntary conversion of current technical specifications in this manner is expected to produce an improvement in the safety of nuclear power plants through a reduction in unnecessary plant transients and more efficient use of NRC and industry resources.

1

DATE:

.. /~~-/ 1'-/

Comment period e~pires (75 dafs after publication in the Federal 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.

ADDRESSEES : Mail written comments to:

Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN:

Docketing and Service Branch.

Deliver comments to: 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm on Federal workdays.

Copies of comments received may be examined and copied for a fee at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT:

Christopher I. Grimes, Chief, Technical Specifications Branch, Division of Operating Reactor Support, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:

(301) 504-1161.

SUPPLEMENTARY INFORMATION :

Background

Section 182a. of the Atomic Energy Act of 1954 (Act), as amended (42 U.S.C. 2232), mandates the inclusion of technical specifications in li~enses for the operation of production and utilization facilities.

The Act 2

requires that technical specifications 'include information concerning the amount, kind, and source of special nuclear material, the place of use, and the specific characteristics of the facility. That section also states that technical specifications shall contain information the Convnission requires through regulation to enable it to find that the utilization of special nuclear material will be in accord with the co11111on defense and security and will provide adequate protection of public health and safety.

Finally, that section requires technical specifications to be made a part of any license issued.

The Commission promulgated§ 50.36, *Technical Specifications," which implements Section 182a. of the Atomic Energy Act.on December 17, 1968 (33 FR 18610). This rule delineates requirements for determining the contents of technical specifications. Technical specifications set forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection of the health and safety of the public. Specifically, § 50.36 requires the following:

Each license authorizing operation of~ production or utilization facility of a type aescribed in§ 50.21 or§ 50.22 will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to

§ 50.34.

The Commission may include such additional technical specifications as the Commission finds appropriate.

3

Technical specifications cannot be changed-by licensees without prior NRC approval.

However, sinLe 1969, there has been a trend toward including in technical specifications not only those requirements derived from the analyses and evaluation included in the safety analysis report but also essentially all other Co11111ission requirements governing the operation of nuclear power reactors. This extensive use of technical specifications was due in part to a lack of well-defined criteria (in either the body of the rule or in some other regulatory document) for what should.be included in technical specifications.

This use has contributed to the volume of technical specifications and to the several-fold increase in the number of license amendment applications to effect changes to the technical specifications since 1969.

It has diverted both NRC staff and licensee* attention from the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety.

On March* 30, 1982 (47 FR 13369), the NRC published in the Federal Register a proposed amendment to Part 50.

The proposed rule would have revised§ 50.36, "Technical Specifications," to establish a new system of specifications divided into two general categories. Only those specifications contained in the first general. category as technical specifications would have become part of the operating license and would have required prior NRC approval for any changes.

Those specifications contained in the second general category would have become supplemental specifications and would not h~ve required prior NRC approval for most changes.

The NRC review of the first general category of specifications would have been the same as that currently performed for technical specification changes, which are amendments 4

to the operating license.

For the second category, supplemental specifications, the licensee would have been allowed to make changes within specified conditions without prior NRC approval.

The NRC would have reviewed these changes when they were made and would have done so in a manner similar to that currently used for reviewing design changes, tests, and experiments performed under the provisions of§ 50.59.

Because of difficulties with defining the criteria for dividing the technical specifications into the two categories of the proposed rule and because of other higher priority licensing work, the proposed amendment was deferred.

In the early 1980s, the nuclear industry and the NRC staff began s~udying whether the existing system of establishing technical specification requirements for nuclear power plants needed improvement.

During this time frame, an NRC task group known as the Technical Specifications Improvement Project (TSIP) and a Subcommittee of the Atomic Industrial Forum's (Alf)

Committee on Reactor Licensing and Safety performed two studies of this issue. 1 The overall conclusion of these studies was that many improvements in the scope and content of technical specifications were needed and that a joint NRC and industry program should be initiated to implement these improvements.

1SECY-86-10, "Recommendations for Improving Technical Specifications," dated January 13, 1986, contains both "Recommendations for Improving Technical Specifications," NRC Technical Specifications Improvement Project, September 30, 1985, and "Technical Specifications Improvements," Alf Subcommittee on Technical Specifications Improvements, October 1, 1985.

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Both groups made specific recommendations which are summarized as follows:

(1) The NRC should adopt the criteria for defining the scope of technical specifications proposed in the AIF and TSIP reports. Those criteria should then be used by the NRC and each of the nuclear steam supply system vendor owners groups to completely rewrite and streamline the existing Standard Technical Specifications (STS).

This process would result in the transfer of many requirements from control by technical specification requirements to control by other mechanisms [e.g., the final safety analysis report (FSAR),

operating procedures, quality assurance (QA) plan] that would not require a license amendment or prior NRC approval when changes were needed.

The new STS should include greater emphasis on human factors principles in order to make the text of the STS clearer and easier to understand.

The new STS should also provide improvements to the bases section of technical specifications, which gives the purpose for each requirement in the specification.

(2) A parallel program of short-term improvements in both the scope and substance of the existing technical specifications should be initiated in addition to developing new STS as stated in Recommendation (1).

On February 6, 1987 (52 FR 3788), the NRC published in the Federal Register for public comment an Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors containing proposed criteria in response to Recommendation (1). These criteria were generally derived from the criteria proposed in the AIF and TSIP reports and were modified slightly on the basis of discussions between the NRC staff and the 6

industry.

The public comment period for the interim policy statement expired on March 23, 1987.

The criteria were developed with the intention that they would apply to limiting conditions for operation {LCOs).

The NRC staff believed that the safety limits needed to remain as is in the technical specifications because of their more direct link to protection of the physical barriers that guard against the uncontrolled release of radioactivity.

At the time the criteria were developed, the industry did not wish to address administrative controls and design features in the effort to improve the STS.

Later, however, both the industry and the NRC staff realized that it would be beneficial to include upgraded administrative controls and design features in the improved STS, and these were handled separately from the application of the criteria to the LCOs.

The NRC has developed a program for short-term improvements as described

(

~ '

in Reconvnendation (2). These are known as "line-item" improvements and are generic improvements developed and promulgated by the NRC staff for voluntary adoption by licensees.

Subsequently, improved vendor-specific STS were developed and issued by the NRC in September 1992.

The improved STS were published as the following NRC reports:

NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants" 7

• NUREG-1431, "Standard Technical Specifications, Westinghouse Plants"

• NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants"

• NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4"

• NUREG-1434, "Standard Technical Specifications, General Electric Plants, BWR/6" Copies of NUREGs may be purchased from the Superintendent of Documents, U.S. Government Printing Office, by calling (202) 275-2060 or by writing to the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082.

Copies are also available from the National Technical Information Service, 5825 Port Royal Road, Springfield, VA 22161.

These improved STS were the result of extensive technical meetings and discussions among the NRC staff, industry owners groups, vendors, and the Nuclear Management and Resources Council (NUMARC).

Finally, on July 22, 1993 (58 FR 39132), the Commission published a Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, which incorporated experience and lessons learned since publication of the interim policy statement.

The interim policy statement identified three criteria to be used to define which of the current technical specification requirements should be retained or included in technical specifications and which LCOs could be relocated to licensee-controlled documents, as follows:

8

Criterion 1:

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2:

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a desi_gn

.... ' i: ~

basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The interim policy statement also stated that, in addition to structures, systems, and components captured by the three criteria, it was the Commission's policy that licensees retain in the technical specifications LCOs for a specified list of systems that operating experience and probabalistic safety assessment had generally ~hown to be important to public health and safety.

In the final policy statement, the Commission retained this thought as a fourth criterion to capture those requirements that operating experience or probabilisti( safety assessment show to be significant to public health and safety.

The final policy statement also addressed corrments received on the interim policy statement and described the Commission's intent with regard to use of the criteria and their codification through rulemaking.

9 l

The Commission believes that amending§ 50.36 to include the four criteria contained in the final policy statement could codify a viable, potentially safety-enhancing and cost-saving method for technical specification improvement.

The Commission encourages licensees to use the improved STS as the basis for plant-specific technical specifications.

As stated in the final policy statement, the Commission will place the highest priority on *requests based on the criteria for individual license amendments that are used to evaluate all of the LCOs for an individual plant to determine *which LCOs should be included in the technical specifications. Related surveillance requirements and actions would be retained for each LCO that remains in the technical specifications.

Each LCO, action, and surveillance requirement should have. supporting bases.

In addition, the Commission will also entertain requests to adopt portions of the improved STS, even if the licensee does not adopt all STS improvements.

These portions. w-ill include, all related.-requirements and wJll normally be developed as line-item improvements by the NRC staff. The-C'onnnission encourages all licensees who submit technical specification related submittals based on these criteria to emphasize human factors principles.

LCOs that do not meet any of the criteria, and their associated actions and surveillance requirements, may be proposed for relocation from the technical specifications to licensee-controlled documents, such as the FSAR.

The criteria may be applied to either standard or custom technical specifications.

The Commission will also consider the criteria in evaluating future generic requirements for inclusion in technical specifications.

10

During individual technical specification conversions, a backfit analysis will be performed in cases of nonvoluntary addition of new requirements from the improved STS to individual plant technical specifications, unless the staff-suggested additional changes are needed to make the changes requested by the licensee acceptable from the standpoint of adequate protection or compliance with NRC regulations, in which case the request may be denied without the additional items.

The Commission requests comments on the criteria being proposed for inclusion in§ 50.36 and, particularly, on Criterion 4 and what guidelines the Commission should use in defining "significant to public health and safety."

Finding of No Significant Environmental Impact:

Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission regulations in Subpart A of Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and would not degrade the environment in any way.

Therefore, the Commission concludes that there will be no significant impact on the environment from this proposed rule. This discussion constitutes the environmental assessment and finding of no significant impact for this proposed rule; a separate assessment has not been prepared.

11

Paperwork Reduction Act Statement This proposed rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget, approval number 3150-0011.

Regulatory Analysis The Commission has determined that a regulatory analysis is not required for this proposed rule.

The Convnission believes the intent of the regulatory analysis has been met through the extensive* consideration given to the*

development of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors and the improved STS, both. of which involved an opportunity for public comment.

The criteria being added to

§ 50.36 are.identical to thos~ contatned:Jn th~ final policy statement and.

have been used by 'the NRC and the nuclear power -industry to define the content of technical specifications since September 1992.

The criteria will continue to be used even if this proposed rule is not adopted.

The proposed rule does not impose any requirements but, rather, allows nuclear power reactor licensees to voluntarily use the criteria to relocate existing technical specifications that do not meet any of the criteria to licensee-controlled documents.

The NRC staff also uses these criteria to determine whether technical specifications are appropriate to provide continued regulatory control over new requirements or positions that have been justified consistent with the backfit rule.

12

The Commission considered the need for and consequences of this proposed action when it made the decision to not only publish the criteria in the final policy statement but also to codify the criteria through rulemaking.

Appropriate alternative approaches to this action have been identified and analyzed over the life of the Technical Specifications Improvement Program, beginning with an earlier attempt to define the content of technical specifications through rulemaking.

As described in the background discussion, the Co11111ission published a proposed amendment to§ 50.36 (47 FR 13369) on March 30, 1982.

However, because of difficulties with defining criteria for technical specifications and because of other higher priority licensing work, the rule change was deferred.

In February 1987, the Commission published an interim policy statement on Technical Specification Improvements and in July 1993, published the final policy statement. During review of the final policy statement, the Commission concluded that the four criteria should be codified in a rule. Thus, alternative approaches to regulatory objectives have been identified and analyze~, and the Cormnission has decided that there is no clearly preferable alternative to codifying the four criteria in a rule.

With regard to evaluation of values and impacts of alternatives, the Commission believes there is no difference in the values or impacts of implementing the criteria through use of the final policy statement or through a rule, except that the criteria are more readily available to future users in a rule than in a policy statement.

13

Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 [5 U.S.C.

605(b)], the Commission certifies that, if promulgated, this rule will 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 nsmall entities" as given in the Regulatory Flexibility Act or the Small Business Size Standards in regulations issued by the Small Business Administration at 13 CFR Part 121.

Backfit Analysis The NRC has determined that the backfit rule, § 50.109, does not apply to this proposed rule and, therefore, a backfit analysis is not required because these amendments do not involve any provisions that would impose backfits as defined in§ 50.109(a)(l).

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

For the reasons given in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, 14

as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the following amendment to Part 50.

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. 1244, 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(dd), 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. 122, 68 Stat. 939 (42 U.S.C. 2152).

Sections 50.80-50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234).

Appendix Falso issued under SP~. 187, 68 Stat. 955 (42 u.s.c. 2237).

15

2. In§ 50.36, paragraphs (c)(2) and (3) are revised to read as follows:

§ 50.36 Technical specifications.

(c)*

(2) Limiting conditions for operation.

(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee. shall shut down the reactor or follow, any remedial action permitted by the technical specifications until the condition can be met.

When-a limiting condition for operation of any process step in the system of a fuel reprocessing plant is not met, the licensee shall shu.t_ __ dq_wn that part of the operation or follow any remedial action permitted by the technical specifications until the condition can be met.

In the case of a nuclear reactor not licensed under§ 50.2l(b) or§ 50.22 of this part or fuel reprocessing plant, the licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence.

The licensee shall retain the record of the results of each review until the Convnission terminates the license for the nuclear reactor or the fuel reprocessing plant.

In the case of nuclear power reactors licensed under 16

§ 50.2l(b) or§ 50.22, the licensee shall notify the Co111T1ission if required by

§ 50.72 and shall submit a Licensee Event Report to the Commission as required by§ 50.73.

In this case, licensees shall retain records associated with preparation of a Licensee Event Report for a period of three years following issuance of the report.

For events which do not require a Licensee Event Report, the licensee shall retain each record as required by the technical specifications.

(ii) A*technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1.

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2.

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

17

(D) Crjterjon 4.

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

(iii) A licensee is not required to modify technical specifications that are included in any license issued before [insert the _effective date of this document] to satisfy the criteria in paragraph (c)(2)(ii) of this section.

However, for technical specification amendments a licensee proposes after

[insert the effective date of this document], the criteri-a in pa~agraph (c)(2)(ii) of this section provide an acceptable scope for limiting conditions n

for operation.

(3) Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operati9n will be within safety limits, and that the limiting conditions for operation will be met.

rL Dated at Rockville, Maryland, this /c/ - day of 199~.

FOR THE NUCLEAR REGULATORY COMMISSION.

Joh C. Hoyle, Acting Secretary of the Convnissfon.

18