ML20154K285

Reactor Safety in Us:Regulatory Perspective. Paper Presented on 861106 at Italian Committee on Nuclear Risk & Intl Scientific Culture & World Lab in Rome,Italy

ML20154K285
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Issue date:	11/06/1986
From:	Harold Denton Office of Nuclear Reactor Regulation
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c REACTOR SAFETY IN THE UNITED STATES: A REGULATORY PERSPECTIVE

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't Presented To ITA(IkN COMMITTEE ON NUCLEAR

, RISK AND INTERNATIONAL SCIENTIFIC CULTURE AND WORLD LAB. -

November 6, 1986 i

Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission

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SUMMARY

The current status of licensing in the V. S. nuclear industry 10 presented and the regulatory process applicable to the U.S. nuclear power industry is examined, including past, present, and planned future actions and activities necessary to provide for the protection of the public health and safety.

Finally, a brief review of recent operating experience in the V. S.

is presented and the ootential implications of the Chernobyl accident are examined vis-a-vis V. S. reactors.

a U.S. POWER REACTOR LICENSING Currently, these are 105 nuclear power reactors licensed in the United States, representing an installed capacity of about 90,000 MWe. These reactors provide about 13% of the electricity produced in the U.S. Although several other countries produce more than 50% of their electric power from nuclear fission, the United States has, by far, the largest installed nuclear generating capacity of any country.

In addition to these 105 licensed reactors, another 16 are under active construction, with an additional five plants deferred but not cancelled.

Over the last several years, the focus of the mission of the U.S. Nuclear Regulatory Comission (NRC) has shifted from the review of construction permit and operating license applications to the regulation of a maturing operational i ndus try. This slowdown in licensing activity (no new construction pent't applications have been received since 1978) has enabled the agency (and the industry) to think more carefully about the insights to be gained from the evaluation of reactor operating experience as well as the knowledge gained from the extensive ongoing severe accident research programs.

Many changes have occurred in the conduct of NRC business over the years as experience has been gained with the regulatory process applicable to the U.S. nuclear industry. These changes, and pending changes, should improve, streamline, and strengthen the regulatory process while placing an increased emphasis on meaningful improvements in reactor safety.

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2 From the outset the United States has played a leadership role in its cornitment to the utilization of nuclear energy and to the research and development efforts which are necessary for the safe utilization of nuclear reactor technology.

Two major disruptions to the U.S. nuclear program occurred in the 1970s which led to financial dif ficulties, excess reserve margins, and major cost disruption for most utilities involved in the licensing and construction of nuclear power plants.

The 1973 Mid-East oil embargo and the 1979 reactor accident at the Three Mile Island Unit 2 (TMI-2) facility caused, respectively, sharp increases in oil and energy prices resulting in excess reserve margins due to conservation and a large number of expensive backfit requirements for improved reactor safety.

The disruptions to the U.S. nuclear industry from the oil embargos are largely behind it; at the same time, the majority of the improvements resulting from the TMI-2 accident are complete and the number of new backfit measures for improved reactor safety has been drastically reduced. In fact, the most cost-effective measures to reduce the risks of severe reactor accidents in U.S. power reactors have already been identified. Still under consideration, however, are measures to improve reactor decay heat rcmoval systems and measures to upgrade containment performance for certain reactor designs.

The following presents some of the major features of the regulatory process in the U.S. This process his, in many respects, been changed over the last several years in an effort to add an element of stability and certainty to the licensing process, while assurinq that no undue risk to the public health and safety is preranted.

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. j Defense-in-Depth One of the primary tools in assuring that the risk to public health and safety from power reactor operation is acceptably small is the use of the defense-in-depth concept. This concept incorporates three levels of safety in nuclear plants. The first level requires design against failures, the second level requires measures to protect against failures and the third level requires features to mitigate consequences of failure. Application of the defense-in-depth concept also resulted in the provision of multiple physical barriers between the radioactivity contained in the reactor fuel and the environ-ment outside the plant. The defense-in-depth concept is also applied widely in the design and review of many of the individual systems of the plant, leading to requirements for redundant and independent subsystems and backup systems.

l These requirements are embodied in NRC reculations, the Standard Review Plan, and regulatory guides that are based on sound engineering practices established over the past decades, and that undergo continued review and improvement as operating experience accrues. Our comprehensive research program provides the tehncial bases for the confirmation of NRC's safety decisions and for needed improvements.

Design and Acceptance Criteria in the U.S. Licensing Process No specific nuclear power plant design is required by the NRC; the design submitted by the applicant must only conform to broad-based statutory and regulatory standards in order to obtain approval. The General Design Criteria (GDC) which are contained in Appendix A to Title 10, Part 50 of the Code of Federal Regulations set forth minimum requirements for principal design criteria

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for water cooled nuclear power plants of similar plants of similar design and '

location previously licensed by the Commission. These criteria constitute "engineering goals" rather than "precise tests or methodologies by which reactor safety can be fully gauged." The General Design Criteria are supplemented by the staff's internal standards for measuring compliance with the GDC. However, neither the Standard Review Plan nor the Regulatory Guides are embodied in NRC regulations. Thus: although compliance with these interpretive materials generally can be expected to facilitate staff approval, they are not, strictly speaking, mandatory. Applicants are free to select other methods for complying with the General Design Criteria which then must be found acceptable by the staff.

These GDC constitute only general statements of design objectives or principles; specific details of satisfactory means of implementing these criteria are contained in other documents such as the Standard Review Plan (SRP) and in Regulatory Guides.

Operating Reactor Safety Assessment With 105 reactors licensed, we continue to gain knowledge and insights from the evaluation of reactor operating experience. Events which occur at U.S.

operating reactors that could potentially lead to an ordered shutdown (or derating), facility license amendment, or new generic concern, are reviewed by both NRC Region-based and Headquarters staff. The reviews of these events involve the evaluation of events against existing licensing analyses, evaluation of plant and operator performance during events, review of licensee analyses, and evaluation of corrective action prior e plant restart.

J Additionally, the U.S. reviews reactor operating events that occur in other countries to incorporate that experience into our data base. The review of all of this operational experience provides a basis for making changes to improve the safety of our facilities.

Probabilistic Risk Assessment Another tool used in the analysis of potential safety issues is Probabilistic Risk Assessments (PRAs). PRAs continue to provide valuable insight into the importance of certain potential safety issues and a mechanism for the systematic identification of strengths and weaknesses in nuclear power plant design and operation.

Probabilistic Assessments are used routinely in setting priorities for new issues according to their safety significance and for allocating resources.

Similarly, probabilistic assessments are used by the NRC in weighing alternative approaches to generic safety issues. In addition, PRAs can provide an important basis, along with deterministic engineering judgement, for the regulatory analyses of new requirements proposed by the staff.

In general, it appears that PRAs are a useful tool for predicting risks from operation of a nuclear power plant. They are valuable as a method for estimating the ir.portance of safety issues and for providing insights on ways to irrprove reactor safety. At the same time, PRAs are not a replacement for determininistic engineering methods; rather, they are a useful supplement to sound engineering practice. The NRC staff believes that plant-specific PRAs should be developed for all nuclear power reactors to identify potential severe accident vulnerabilities.

Relative Risks of Nuclear Power As an Energy Source in the United States As for the case of any energy production technology, there is indeed an element of risk for the activities necessary to construct and operate a nuclear power plant. However these risks are not large when compared with the risks of other forms of energy production. Nuclear power and natural gas have been found to have the lowest overall risk of present day energy production technologies.

In addressing the matter of risk the USNRC has established quantitative safety objectives which support the NRC's qualitative safety goals.

In adopting its safety goal, the NRC established its regulatory objective of "reasonable assurance" that a severe core damage accident will not occur at a U.S. nuclear power plant.

The NRC's safety goals also serve to place the risk from reactor operation in perspective with respect to other risks. Specifically, the Comission has established the following quantitative safety objectives for nuclear plant operation:

• The risk of prompt fatalities from reactor operation should not exceed 0.1% of the total risk of accidental fatality.

• The risk of latent cancer should be no greater than 0.1% of total cancer risk.

l While no deat5 itributable to reactor operation is acceptable, the goal that the NRC is seeking to establish is that individual and societal risks from nuclear power plants are to be considerably less than the risks from other l

activities.

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POLICY STATEMENT ON SEVERE REACTOR ACCIDENTS A major regulatory action on the part of the U.S. NRC was taken on August 8,1985 with the issuance of the "Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plant." Agency activity to at; dress severe reactor accident concerns was first contained in an October 2, 1980 advance notice of proposed rulemaking; subsequently, a "Proposed Comission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation" was published for public coment on April 13, 1983.

The main focus of NRC's August 8,1985, Policy Statement on Severe Reactor Accidents, which reflected coments received on the April,1983 proposed policy statement and announced withdrawal of the October 1980 advance notice of proposed rulemaking, is on the decision process involving staff approval or, optionally, Connission certification of new standard designs for nuclear power plants. At the same time, this Policy Statement also provides guidance on decisions and analytical procedures for the resolution of severe accident issues for other classes of future plants and for existing plants.

The Comission strongly endorsed the design certification process for future plants in this Policy Statement, while affirming the Comission's belief that a new design for a nuclear reactor can be shown to be acceptable for severe accidents if the following criteria and procedural requirements are met:

• Demonstration of compliance with the procedural requirements and criteria of current Comission regulations.

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• Demonstration of technical resolution of all applicable Unresolved Safety Issues and the medium and high priority Generic Safety Issues, including a special focus on assuring the reliability of decay heat removal systems and the reliability of both AC and DC electrical supply systems.

• Completion of a Probabilistic Risk Assessment (PRA) and consideration of the severe accident vulnerabilities the PRA exposes along with the insights that it may add to the assurance of no undue risk to public health and safety and

• Completion of a staff review of the design with a conclusion of safety acceptability using an approach that stresses deterministic engineering analysis and judgement complimented by PRA.

Published at the same time as this Policy Statement was NUREG-1070 "NRC Policy on Future Reactor Designs." In addition to the Severe Accident Policy Statement, this report also discusses how the policy statement relates to a number of other ongoing NRC programs. The Severe Accident Policy Statement helped to achieve improved stability and predictability of reactor regulation in a manner that merits improved public confidence in the regulatory decision making process.

A number of ongoing NRC programs are related to the severe accident policy statement, including the Severe Accident Research Program, the implementation of safety measures resulting from lessons learned in the TMI accident, safety goal development, the resolution of Unresolved Safety issues and other generic safety issues, and rulemaking for source term revision that are important to future licensing of advanced designs and continued operation of existing designs.

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-g-The NRC staff has prepared an "Implementation Plan for the Severe Accident Policy Statement end the Regulatory Use of New Source Term Informa-tion". This implementation plan has been developed to take into consideration cr. going severe accident research and uncertainties in PRA numerical assessments.

Ir. this regard, NRC is considering its policy on the use of PRA in new reactor plant license applications.

In addition, based on results from ongoing research on source terms and insights gained from evaluations of severe reactor accidents, a number of changes in NRC rules and regulations will likely be in order (for future plants) to resolve generic severe reactor accident related vulnerabilities.

The NRC believes that in light of the many changes that have been ir.plemented in existing plants in recent years, existing plants pose no undue risk to public health, safety and property. Further, no additional regulatory action is proposed to address severe accidents for these reactors unless significant new safety information arises.

Standardization in the U.S.

The NRC staff has proposed revisions to the Comission's 1978 Standardiza-l tion Policy Statement which would reflect the experience the NRC has acquired in implementing the 1978 Policy Statement, the applicable provisions of the Severe Accident Policy Statement and of the proposed standardization legislation,

l and the current views of the Commission ano industry on standardization. The focus of the revised policy is on Reference System Design Certification which should be used to satisfy the ultimate licensing goal of constructing plants of certified designs on pre-approved sites. The staff is in the process of developing implementation guidance for the proposed revised Standardization Policy Statement. NUREG-1225, "Implementation of NRC Policy on Nuclear Power Plant Standardization" and the Standardization Policy Statement are expected to be issued for comment in the first quarter of fiscal year 1987.

The Three Mile Island Unit 2 (TMI) accident in March,1979 resulted in a hiatus in NRC licensing activities including those related to standardization.

Recently, renewed interest in the standardization desiga concept has led to a number of actions both within NRC and in the industry. The Commission's policy on severe accidents (accidents not considered within the design bases) began to change soon after the TMI accident. A large number of changes in design and operating procedures for nuclear power plants were mandated by the Commission following numerous investigations of the causes of the TMI accident along with a general probing of vulnerabilities of severe accident risk in other types of plants.

The staff has proposed a revised Policy Statement on Standardization to supersede the aforementioned 1978 Standardization Policy Statement. The focus of the revised policy is the reference system design certification, i

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The staff believes that implementation can benefit the public health and safety by:

(1) Concentrating the resources of designers, engineers and vendors.

(2) Stimulating standardized programs of contruction practices and quality assurance.

(3) Improving the training of personnel.

(4) Fostering more effective maintenance and improved operation.

The NRC staff believes that its proposed standardization policy will reduce the current large variety of design, construction, operational and c.aintenance practices. Consistent with this belief the staff has strongly endorsed the utilization of essentially complete standardized plant designs.

In the proposed standardization policy statement, the staff proposes that a design would be certified by the Comission following issuance of a Final Design Approval (FDA) by the staff and the completion of a rulemaking pro-ceeding. The Design Certification would certify the design as acceptable for incorporation by reference in individual license applications. A certified design would not be subject to litigation in individual licensing proceedings on those issues which are covered by the design certification rulemaking. However, the adequacy of interface requirements may be 1 :igated in subsequent individual licensing proceedings.

CHERN0BYL ACCIDENT FOLLOW-UP Following the April 26, 1986 accident at the Soviet Union's Chernobyl Nuclear Power Plant, Unit 4, President Reagan established an Interagency Task Force to assess the impact of the accident on the U.S. and to ascertain the implications of the Chernobyl accident for U.S. reactors.

Although evaluation of the implications of the Chernobyl accident on U.S. reactors is not complete, reviews of the accident scenario and the RBMX design to date have not identified any aspects of the accident which show a clear-cut nexus to comercial power plants in the U.S. However, work is continuing on a priority basis to either confinn that current regulatory practices and policies are sound or to identify needed improvements.

e e." g, i CONCLUSION s

Nuclear power is making an important contribution to the energy needs of the United States and, I believe, will continue to do so for a long time.

4 The U.S. nuclear experience has clearly denonstrated that with proper attention to design, construction, and operation, nuclear power can be a safe, clean, and reliable source of electricity.

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ITfNERARY ROME, ITALY 11/4-8/86 Tuesday, November 4 11:30 p Courier from MNBB 1: 05 p LV National via TWA 410 Business Class 2:03 p AR Kennedy airport 6:00 p LV Kennedy via Alitalia 611 Business Class Wednesday, November 5 S:30 a AR Rotre, Italy Take taxi from airport to Hotel (Porter at Hotel will pay taxi Hotel LeGrand Hotel 83 V. E. Orlando Rome Phone 6-4709 Country code 39- City code 6-Hotel number 4709 Dr. Pilarski will be in Rome Thursday, November 6 9:00 a Conference Starts at Vila Madonna Italian Government Comittee on Nuclear Risk International Center for Scientific Culture and World Lab

Contact:

Mr. Pilarski Your Speech - Reactor Safety in the U.S.: A Regulatory Perspective Friday, November 7 Conference continues Saturday, November 8 12 Noon LV Rome via Alitalia 610 Business Class 3:15 p AR Kennedy Airport 5:40 p LV Kennedy via Pan Am 585 Business Class 7:00 p AR National Courier from National to MNBB My Contact in Geneva, Switzerland was Dr. Pilarski or Secy. Miss Maria 7aini 41-22-83-3196 - If Mr. Stello's needs to be reached and cannot be reached at Hotel call Switzerland. I have been told conference number cannot be given out.

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NOTE 1 was told by Dr. Pilarsk1 thtt the date, time, place, reason, for the open ticket could not be given to us. The ticket ray not have to be used, it is only if Dr. Zichichi feels it necessary that the trip be taken.

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LEAVING ROME ON SATURDAY. NOVEMBER 8,1986 Lv Rome via Pan AM 8310:45 am AR Kennedy 3:35 pm tv- Kennedy 5:40 pm via Pan AM 585 Ar Natonal 7:00pm

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- " tina 10s1Kam 1 no iv State Dput via taxi for Mational 12 :00pm 6 20 i., Ne i,iv=si AF vis AA M 5 i:00pm Ar hsM]}e. TX 1:45pm 61 neanviiiu. in via e o6i c:ispu Ar thattanooga. TM 3:51pm Lodging 28.66 + 1/2 of 125Maf E = 112.50 i a1 is 11/18 TVA Meetian Lv Chattanoogs via M 927 6:00pm AP Mathvilla. TM R. h _ 1 Lv Nashville via AA 156 6:32'mp )

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1/2 of 12$ Nafh e 112 EO . M an Kuw v ta tarkville fm Ma t t aul Metre Station 1 110 PGA fu Matre ta saurgin3 l Rockville residence 4a1 41 i

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1 NOVEMBER -17, 1986 Lv Naticnal AP via AA 945 ,1:00pm Ar Nashville, TN 1:45pm  ;

Lv Nashville via AA 821  ? :15pn i Ar Chattanooga, TN 3:51cm .

NOVEMBER 18, 1986

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Ar Nashville, TN 5:35pn

Lv Nashville via AA 156 6

32pm Ar National 9:00pm LODGING AT BEST WESTERN  :

(615) 266-7331 ,

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Bethesda, Md. E. 1.-4 son-use :1 -

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IS NUCLE.AR REGULA7074Y COMMIS5+ON ac.*ons Ae6

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-cu we' . REQUEST FOR APPROVAL OF OFFICIAL FOREIGN TRAVEL

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Victor Stello. Mr- W2 a e S ~

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Executive Director for Operations g 3, 49 g 9)

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Office of Nuclear Reactor Regulation l yg,,,g gg, l Cit, 13 SwlssetSS ADD *ill Washington, DC 20555 13 ,ectogit it tis aa, e re ew s CLASS 8 'iO e n, e Se CLASS s e t a $ s;aw ,,,w . , a cto e s e,,, w., =,. e.g.a.e, w owrg as as e re+ n LOCa tC4 8' W Ova 6b 145<iCT 08 0'lCwiSC% COIS6 OAfil 0,Il 5 emissemoa rat se' 840 18 iO 11/24 Lv Dulles Participating in CSN! mtg X 11/25 Ar Paris 11/29 Lv Paris i

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re. seostate . Patricia E. Schem naoito ano

' ' ' ' ' " ' ' Travel Assistant

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• 960*00 Travel Services Staff RM/AFO

~ awov%f cats rotu . 2,100.00 .

October 21, 1986 D COneCLeetieCIS ah0'on atwaar 8 *e se s**ee se er bet aseano.e'ers-23 088<I 08 mi t tee n atoga, enocaaws Co Cunt = Cts ano c"avaan 54*a' ** LL 14 Ron D. Hauber Q9d <J i6 e smy %u A/O for IC Mfice nf International proer m

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Qaft L feng across - 088sCt Of twl EstCbht D8&CTCn pos Ottaatcht M a5 0yttMO *i p an t A it a we sic,% a t , a g songcntanett I Oa=caes+utttoetwanis Oesa..2,ie

. esi: Victor Stello, Jr.

a=ot='a'o=5 "a

4.a..i a s e s ... t at c=s Executive Director for Operations Oa'i 4.

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Participation in CSN! meeting

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This trip has been coordinated with the Office of International Programs I cati -

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Regulation

'aeoid R. Denton, Director. Of fice of Nuclear Peactor .

etinanas, i

PRIVACY ACT STATEMENT Pmnt to S U $ C. 552 <ex3t enactw eto iew by section 3 of the Pnvaey Act of 1974 iPwtwie Law S3 579). the tonoweg a; 20 state W suppN information to the Nwctar Regulatory Comm.ss.on on NRC Form 445 Th s mf ormaton is maints.ned m a syrem ano oesenbed at 40 Feders: Reg ster 45341 (October 1.19751 een 31 U S C. 21. 22. 24. 49. R 66a. and 952. 5 U.S C. 5701; Feder at Travel Regwtatens and Federal Property Manag

1. AUTHORITY 101 7.

informaton entered on this f orm e used to see we e the reawwed NRC approval for omc41 foreign vave'

2. PRINCIPAL PURPOSE (S)

Mformaten on this foem may be used for transmittal to the State Department to secure pauports. The eformaten me, e '

3. ROUT 1NE USES escesed to an apprwate Fede'a!. State, ce local agency m the esent the information ino. cates a votaten or po aa admin.strative or pad.c.at proceedeg la ado. ten, th.s mformaten may ce transferred to an apprope. ate Feders State the entent re e aet and necesbaS 'or that agene) s cec s,oa abo.' ecw F

te event and neCeuary for en NRC oec+on 0+ to an approchate edeer agency to Dra. e is

4. WHETHER OtSCLOSURt t$ WANDATORY OR VOLUNT ARY AND EFFICT ON INDMOUAL O

.o w etary if the reevested Wormaton 4 not provw3ed honeser. accroval may be demed i 5 $v$TLM M AN AGintSI AND ADDRES$

I Coeteouer 0**.ce of the Contro aer a S Nwc4ar Reg #etory Commason A est.egton. O C 2265 l **

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11/2h Lv Rathanda af flea via PBA  ? ! 10 nia En in 24 Ar Dulles 3:15pm Lv Dullas via Pan Am 590 heltne Ar New York 5:53pm Lv hw Yark via Pan Am 114 (ekgna 11/1C Ar Paris 7:45 am Tant to hatal (<nfI 3 9e 11/?A Of ficial Business e,,w. en -A fra- Amt.1 f e ri i g 11/27 offic'lal Business e n y,, en --A fre== hatal fSri i y 11/73 Trip to St Laurent reactor t af e hatal e nn-i

$ pent day at St. Leursat reactor tatura Parla haeal C.nham #

11/14 - Taxi to airport fa hotel (50F) 3 ,

23 Iv P ela wla Paa A= 115 11 3n =

Ar b York 1:35pm iw -

vark v t . **a A- $31 5. ;a.,

Ar Dulles 6:50pm Iw Bullan 7.in Ar Rockville residence vla PGA 50 10 b5 P A R E t M*. AT Md i f t ITIAVIl l FE l_ntT TICKET) THE FEE FOR PARKING 15 u ,t-3t nn PER DIEM 12 en l iT/9k 1/9 4*w ! $95 -

11/251/2 day 4 $6 3 00 l .

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period and the above cost was Incurred.

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. TRIP TO PARIS ON NOVEMBER 24-29, 1986 November 24, 1986 Lv Dulles via PAM 550 4:15pn Ar New York 5:29pn Lv New York via PAft 114 6:45pm NOVEMBER 25, 1986 Ar Paris 7:45am NOVEMBER 29, 1986 i

, Lv Paris via PA!1115 ll:30am

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VISIT TO SHEARON HARRIS U/LRubenstein, l I

EChristenbury LV NATIONAL VIA PI 233 7:25am AR RALEIGH, N.C. 8:20am

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Vic tor Stello, Jr. , Executive Director uyn t a p non.nn n < ,. , e t n ,. .m tea %;a espcaaw Oss CI;e C> v %C %

for Operations MNBB u n, p.3m pg334,1 3n w a3 I S ."4 D '% G socw gswega I g ,,,3.g 4 3 Is aggcg aoou i,9wega .

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3/14 Return USA l

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Director for Operations

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"'"** ' 3,175.00 Patricia E. Schem ua

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Travel Assistant wscia*%i:.s i

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AD FOR IC, IP Oa'i 2/ 5/87

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WNETHER DISC 106Um!is MANDAfomy OR VOLUNTARY ANO EFFECT ON IN0fv10 VAL OF NOT PROV

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..evuCHER(PA272)

SCHEDULE OF EXPENSES

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Ar Dulles 3: 20pm

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2/27 Ar Frankfurt 9:30am Ly r rmak f u r t vi a Pia i= ?k !2:3rp i

Ar Moscow 5:45pm 2/28 (saturdavl *'2h la "^^ f ~ - (Of'!?!:! l.

Business) IICKnNGE RATE $1.00 USA

• 64.14 RUBLES pasinart l 3/l ra*I*'ri'!^- at M (Sunday) stay in Moscow - of ficial business

! (! .50 ruble s) g y 3/2 ( Pbn day I ' I" "f" * ^f'!!'0! ML en 3/3 (Tuesday) stay In Moscow - of ficial business 3/4 (Vednesdayl Iw Ma*** 'l y '^ Eep^rn5' p (late due to weather) ..

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3/s (Th u r< dayi 1,rmy g e. 3544;;; 33;;7,33, 3,gg p Lv Energodar bust to Zaporoshlye then fl en ri.w 1/6 Kiew official business train to Moscow (h5141-= q ef :::! t -P ettecl .J) 1/7 ,

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PAYEE COPY

TRAvtl. VOUCHER (PART C SCHEDULE OF EXPEN$tt **'*/**"*

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2/26 - 1/2 day 9 $IstT -J O $ E 00 f 2/27 - 3/4 day 6 $6 4.50 1/! day 9 $120 30.00 2/28 - 1 day 0 $120 120.00 120.00 l 3/1 - 1 day 9 $120 3/2 - I day 9 $120 120.00 3/3 - 1 day 6 $120 120.00 3/4 - I day 6 $120 120.00 3/5 - 1 day 9 $200 120.00 3/6 - 1 day e $120 120.00 3/7 - 1 day 9 $120 120.0S 3/8 - 1 day e $120 120.00 3/9 - 1 day G $120 120.00 3/10 - I day e $120 120.00 ,

3/11 - 1 day G $120 120.00 3M 2. M day. t $430 120.00 Via 4 & t'81M ' '84 s. 120.00

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We appreciate your patronage and always welcome your candid critique of our performance.

FOR FUTL*RE RESER VA TIONS t 1 800 228 9898 i

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STAYING AT THE RAMADA INN, in SPRINGFIELD 217-544-8800 $40 + Tax . :.

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a TOPICS FOR APRIL 13 NRR MANAGEMENT MEETING

1. Administrative Matters

- Working hours for supervisors

- Position descriptions (SES and Non-SES)

- SES performance contracts

- Philosophy on travel (Domestic and Foreign); allocation of travel funds

- Philosophy on promotions and awards

- Transfer of work (continuity for major projects)

Technical assistance contracts

2. Regional interactions and focal points

- Inspectica issues (degree of involverrent, direction, oversight)

- Licensing issues (degree of involvernent, cirection, oversight)

- Resource allocation and decisions (FTE and contractor)

- Regional / site visits

- Regional assessment

- Promulgation of above policies to regions

3. Interface with AE0D/ Events evaluation process

- Daily review process

- Events follow-up

- Events Reports: PN's; Daily Reports; 50.72; etc.

- Mechanism for disseminating event infonnation 4 Interface with RES

5. Interface between NRR divisions

- Division by division discussion

6. What specific changes will be implemented to emphasize NRR and regional focus on operational safety?

7. Degree of and rechanism for region and technical review branch involvement in reviews of technical specification improvement submittals.

8. Generic Conxnications Branch scope

- Comunications on events or all generic connunications - (generic letters)

- Need for new connunication instruments to provide event information to regions

- Continued use of Information Notices ano Bulletins to provide event information to utilities and vendors vs. generic letter?

9. Process for review of enforcement cases.

10. How will Headquarters' inspections be coordinated?

- DRis focal point?

- What constitutes inspections (e.g. Maintenance Program Reviews?)

11. Who has respctsibility for the following:

- Technical lead on EQ

- Generic Letter 83-8 connitments

- SIMS (priorities and inspect vs. not decisions)