ML19263D804

Responds to Questions Contained in Re HR 2608,on Consolidation of NRC Into One Location

ML19263D804
Person / Time
Issue date:	03/23/1979
From:	Hendrie J NRC COMMISSION (OCM)
To:	Udall M HOUSE OF REP., INTERIOR & INSULAR AFFAIRS
References
NUDOCS 7904130270
Download: ML19263D804 (51)

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UNITED sTAik; -

NUCLEAR REGULATORY CDMMISSION L o ^^ "'

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WASHINGTON, D. C. 20555 gg p N2d[/

March 23, 1979 s, u OFFICE OF THE CH AI RM AN The Honorable Morris Udall, Chairman Subcommittee on Energy and the Environment Comittee on Interior and Irsular Affairs United States House of Representatives Washing.on, D.C. 20515

Dear Mr. Chairman:

I ar pleased to respond to the questions raised in your letter of March 7,1979 for use in the Subcomittee's mark-up of H.R. 2608, iSincerely,

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I GENERAL QUESTIONS

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March 13, 1979 i

Question:

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What is the current estimate of the cost (in dollars and in inefficiencies of operation) of the dispersed location of NRC offices?

Answer, The current estimate of the dollar cost of the dispersed location of NRC headquarters offices is 55.82 million, an 18.3% increase since March 1977.

A functional breakdown of these costs is attached. Although increased efficiency of operation has helped to compensate for the effect of inflation in most categories, the principal increase in costs (potential savings) is attributable to contract security guard s.rvices for our dispersed office locations.

There is a much greater cost associated with our dispersed posture, a cost, or penalty, that I cannot measure for you in dollar terms, but I assure you it is a very severe penalty and one I feel Feenly as chief executive officer of the agency. That penalty has to do with the inefficiencies of having substantial numbers of our staff at all levels in transit between locations all the time, with the difficulties of convening the many interoffice meetings we have to have, with having staff managers at all levels spending-their-time and energy getting --- ~~

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downtown to Commission meetings-and to other locations for other-business instead of being able to concentrate ~on effective management, and finally with the devastating effect on the morale and working efficiency of the staff generally from being located in all these out-lying locations, separated from the Commission and the rest of the s taff.

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March 13, 1979 POTENTIAL SAVINGS FROM CENTRALIZING NRC HEADQUARTERS INTO ONE LOCATION ANNUAL FUNCTION DOLLARS MANYEARS/YR

. SATELLITE MAIL ROOM MANNING 5

. MAIL-MESSENGER ORIVERS 3

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MAIL ROOM EQUIPMENT COSTS

$ 13,000

. MAIL-MESSENGER VEHICLE OPERATING COSTS 10,490

. SUPPLY INVENTORY AVOIDANCE 200,000

. TRUCK LEASE (0 $195/ month) 2.'

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. WAREHOUSEMEN & DRIVERS 10

. SHUTTLE BUS CONTRACT COSTS 146,525

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. LOST TIME IN TERMS OF PASSENGER TRAVEL BETWEENBUILDINGS'(0NSHUTTLE) 18

. SAVINGS IN SECURITY RELATED SERVICES:

GUARDS, COMMUNICATIONS, RADIO-DISPATCHED VEHICLE, ETC.

1,451,783 SAVINGS IN TELECOMMUNICATIONS EQUIPMENT AND SERVICES:

FACSIMILE, INTERNAL TELEPHONE, ETC.

77,400 l

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1 l March 13, 1979 POTENTIAL SAVINGS FROM CENTRALIZING NRC HEADQUARTERS INT 0 ONE LOCATION ANNUAL FUNCTION DOLLARS MANYEARS/YR

. SAVINGS IN ADP EQUIPMENT AND MAINTENANCE 58,000

. ADP OPERATORS 2

. DOCUMENT CONTROL SYSTEM CONSOLIDATION

.25

. DCS DUPLICATING COSTS 24,000

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. OFFICE COPYING EQUIPMENT 192,000

. DUPLICATE TECHNICAL LIBRARY FACILITY:

69,600 BOOKS, PERIODICALS, MICROFICHE, SHELVING, FURNITURE, ETC.

REFERENCE BOOKS, PERIODICALS 20,000 LIBRARIAN SERVICES 1.75 DUPLICATE LAW LIBRARY 25,000 WORD PROCESSING EQUIPMENT RENTAL 50,000 WORD PROCESSING OPERATORS / SUPERVISORS 3

SAVINGS IN MAINTENANCE OF DUPLICATE DOCKET FACILITIES 400,000 DOCKET PERSONNEL SERVICES 60 CONSOLIDATION OF PUBLIC DOCUMENT ROOMS 1

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$2,740,138 104.5 l

(Approx.$3,075,958)

GRAND TOTAL $5,816,096 (18.3% increase since March 1977) l l

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March 13, 1979 Question:

2.

Does the Commission believe that consolidation of HRC headquarters at a site downtown in the District of Columbia has significant advantages over similar consolidation in suburban Maryland? Does the Commission have any data which demonstrates the relative advantages of consolidation in downtown Washington?

Answer:

I am convinced from my experience as Chairman of the NRC that the agency will operate most efficiently and effectively from a location in downtown Washington, Both of my predecessor Chairmen came to the same conclusion as a result of their experience.

It is the decision of the Commission that we should consolidate the headquarters staff and Comnission Offices in one location downtown.

Let me note that when the NRC was first formed, in January 1975, it considered establishing its principal base of operations in the Bethesda area in order to place the Commission in closer proximity to the rest of the headquarters staff. The Commission's experience during its first year of operation, however, convinced it that the Commission itself must be located in downtown Washington to properly discharge its responsibilities.

The basic reason is that the heightened interest in nuclear energy matters has increased tremendously the number and frequency of meetings -

and discussions taking place between the Commissioners-and staff officers and members of Congress, Congressional Committees, the Executive Office, the Departments of State, Defense, Commerce, Justice and Energy, the Environmental Protection Agency, and other agencies, both governmental and non-governmental.

Moreover, NRC's activities with international agencies and state government organizations have expanded greatly, and numerous diplomatic and state representatives meet frequently with NRC personnel.

From the standpoint of location then, experience has convinced us that being in downtowr Washington facilitates these interagency relationships, while, as we have discovered, a suburban location has the opposite effect.

I have so testified on April 11, 1978 before the Subcommittee on Public Buildings and Grounds of the Senate Comnittee on Environment and Public Works.

Commissioner Bradford notes that the final GSA Environmental Impact Statement, which provides a comparison of consolidation in downtown Washington and sur-burban Maryland, is expecten to be published in final form in early April.

He feels that the resu!?s of their evaluation and the responses to public comments shculd be considered in reaching a decision in the future location of NRC.

Commissioner Ahearne believes that arguments for a downtown Washington location are not convincing.

QUESTION 3.

What is the rationale for keeping distinct the physical security function of the Office of Nuclear Material Safety and Safeguards and the Office of Reactor Regulation?

ANSWER.

The Office of Nuclear Material Safety and safeguards (NMSS) is now responsible for the safeguards regulation of fuel cycle facilities and transportation activities, and the Office of Nuclear Reactor Regulation (NRR) is responsible for safeguards regulation of reactors. The main reason for organizationally separating these functions is to facilitate coordination of regulatory actions involving a particular facility; for example, NRR now has responsibility for all licensing activity, including both safety and safeguards, involving reactors.

The Commission has asked the Executive Director for Operations (ED0) to undertake a review of the organization of safeguaras activities within the agency. This review will be completed and presented to the Commission shortly.

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Does the Commission maintain a consolidated file that classifies reactor abnormal occurrences by type and reactor? If the Commission does not keep such a file, why does it not do so?

ANSWER.

The NRC requires nuclear power plant licensees to report incidents or events which involve a variance from regulations or license conditions (Technical Specifications).

Because of the broad scope of the reporting requirements and the conservative attitude towards safety, several thousand events per year are reported to the NRC by all nuclear power plant licensees combined.

The reports are called Licensee Event Reports (LERs).

The NRC evaluates each LER to determine the safety implications, any generic aspects, corrective actions, etc.

To facilitate evaluation of the events and to facilitate dissemination of abstracts of the information in the reports to the nuclear industry, the public, and other interested groups, the LERs are coded into a computer file.

The file is also useful for retrieval of information and for generating information such as the number of certain types of events or event causes versus reactor name, facility, reactor type, etc.

LERs are also reviewed to determine whether the event had an actual impact on public health or safety.

Such events are designated abnormal occurrences and are-evaluated against criteria which were set forth in an NRC policy statement published in the Federal Register (42 FR 10950) on February 24, 1977.

These abnormal occurrences, together with the criteria, are published in quarterly reports to Congress as required by Section 208 of the Energy Reorganization Act of 1974 The abnormal occurrences, and their updates of information, generally list the reactors (and their type) affected.

The number of abnormal occurrences has been relatively low, generally less than 10 per calendar year.

Due to the relatively few abnomal occurrences reporud to date and the fact that the affected reactors and their type can generally be found in the quarterly abnormal occurrence reports, a conslidated file or computer program for abnormal occurrences has not been developed to date.

However, we have recognized that as the number of abnomal occurrences (and the associated reactors affected) increase, it would be desirable to develop a file for abnormal occurrences, similar to that presently available for LERs. Accordingly, such a computer program is under development.

QUESTION 5.

What is the rationale for the Commission risk assessment activities being conducted under the auspices of the Office of Nuclear Regulatory Research?

ANSWER.

The field of quantitative risk assessment is still in an early stage of development. NRC needs to perform research to develop improvements in these techniques to make them operationally useful for assessing nuclear facilities and assisting regulatory decisionmaking.

Thus it is logical at this stage of the development of risk assessment to place prime responsibility for this activity in the Office of Nuclear Regu-latory Research.

It should be noted, however, that other NRC program offices are undertaking projects to prepare for the utilization of risk assessment, when it is further developed.

For example, the Office of Inspection and Enforcement has undertaken a study to determine how risk assessment can be used in the inspection process, and other pro-gram offices are monitoring the research effort for additional risk assessment developments.

QUESTION 6.

Please indicate what steps the NRC has taken to upgrade its International Safeguards Frugram in FY 1979, including a status repo:t on the expenditure of funds specifically authorized in Public Law 35-601 for this purpose.

ANSWER.

The Commission has placed increased emphasis on international safeguards matters in FY 1979, requiring additional staff efforts and knowledge of

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developments in this area. Work has focused on the following:

in-depth technical support to a number of international safeguards efforts, including (a) the development and implementation of the U.S. Action Plan to Strengthen IAEA Safeguards; (b) the international review of fuel cycle alternatives (INFCE); and (c) bilateral training and consultation to IAEA staff and member countries in developing their own safeguards control programs.

Also, NRC is seeking more information on IAEA safeguards capabilities and implementatio activities as part of the export licensing review process and is developing and amending federal regulations necessary to implement the NNPA of 1978 and the US/IAEA Safeguards Agreemat. The Agreement was established to encourage other nations to place their nuclear activities under international (IAEA) safeguards and to generally encoucage upgrading of safeguards arrangements. NRC staff has been actively work'ng with the Executive Branch agencies, U.S. industry and IAEA staff to preoare for implementation of the Agreement in NRC-licensed nuclear facilities. NRC is also providing the computer base.necessar.y_to. track.. foreign origin ~

materials in the U.S. as well as data necesstry to satisfy the IAEA report -

ing requirements; is participating in the U.S.. Program of Technical Assis-tance for IAEA safeguards (POTAS and TASTEX projects); and is providing significant expert participation in a variety of international conferences, symposia, working and advisory groups convened by IAEA to evaluate, develop, or improve specific international safeguards programs.

NRC currently plans to obligate $210,000 to contracts in support of inter-national safeguards projects in FY 1979. To date, $188,000 is already committed; formal obligation of funds will occur shortly as contract documents are processed.

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e QUESTION 7.

The following concerns the Commission's procedure for review and concurrence on proposed research and technical assistance contracts:

(a) Over a year ago, the Commission approved a policy statement regarding procedures for user office concurrences prior to the letting of research contracts by the Office of Nuclear Regulatory Research ("SECY-77-1308-Procedures for Processing User Office Research 99quirements" dated January 19, 1978). Have all res arch contracts awarded since January 19, 1978 been approved according to the procedures spelled-out in the Commission's policy statement?

If not, why not?

ANSWER.*

At the time SECY-77-1303 Procedures for processing User Office Research Requirements, was approved by the Commission, a significant ongoing research program was underway.

The NRC Research Panel in 1976 had reviewed and endorsed the RES safety research program, which included most of the program efforts underway in January 1978. The panel had members from each relevant program office (NRR, RES, MMSS, SD) and from several EDO offices (MPA, COM, EDO-TA). This report was ultimately endorsed by each Office Director. _The ACRS report.

to Congress on the Reactor Safety Research Program, December 31, 1977 also generally endorsed.RES program plans. for FY.1978.

SECY-77-1303 is silent on ongoing research; however, it was the general understanding that the intent was to apply these procecures to new programs. During the remainder of FY 1978, RES operated on the foregoing understanding.

In this context, 213 separate actions were processed by RES from February 1 to September 30, 1978 obligating funds alloted to RES.

These actions include initiation, renewal, interim obligations and modifications of contracts, laboratory work orders and interagency a greemen ts.

During October and November 1978, RES reviewed all of these FY 1978 as well as planned FY 1979 actions. User office, ED0 or Commission approval was obtained for all ongoing and new tasks prior to obligation of funds or contract action, except for three tasks which are currently under review approval in accordance with SECY-77-130B procedures.

• Commissioner Bradford and Commissioner Gilinsky feel that the answers to question 7 are not as responsive as they could be.

QUESTI0M 7.

(b)

In addition to the NRC's user office concurrence procedures, does the Commission or staff have any other established procedures for reviewing proposed research and technical assistance contracts to de-termine if they are consistent with Agency-wide priorities and research needs?

If not, do you think it would be useful to develop a centralized review mechanism to review and approve all research and technical assistance contracts?

ANSUER.*

Current procedures call for Commission review of all research projects over $1,000,000, all technical assistance contracts in excess of

$250,000, and all safeguards projects over $20,000. All projects re-viewed by the Commission take into consideration the position in the agency-wide priorities and the potential to fill a stated research need. All research and technical assistance contracts of $100,000 are reviewed for duplication by one of the two interoffice review processes (one handles commercial and the other interagency work).

At the present time there is a recommendation being prepared that calls for the creation of a central review group that will verify that sufficient justification exists for all contract placements in excess of $100,000.

Further managerial reviews by the EDO (and the Commission in some cases) will also be made under this proposal.

For t'he special area of' Safeguards, a separate interagency group- - -

composed of the technical. managers. across the program office staff.

examines every project in a similar manner, irrespective of-the dollar value.

A second interagency group, covering the area of Waste Management, is currently being organized.

QUESTION 7.

(c) Do the Commission's concurrence procedures apply to the approval of a contract renewal when a significant change in funding level and/or project scope is proposed?

ANSWER.*

Yes. All contracts which are reviewed by the Commission and which have a significant change in funding level and/or project scope require new coordination within the staff and additional approval from the Commission.

• Commissioner Bradford and Commissioner Gilinsky feel that the answers to question 7 are not as responsive as they could be.

Question 8: Please provide a complete list of those reactors in which the replacement of steam generators may be necessary. What does the NRC estimate the cost of each af these replacements to be?

Answer

_ Replacement (Repair) Proarams Steam generator replacement (repair) is currently under way at Surry 2 and has also been appmved by the NRC for Surry 1.

The NRC staff is currently reviewing ti.e licensee's~ proposed replacement programs for i

Turkey Point 3 and 4.

Consumers Power Company has stated that they l

are considering replacement of the Palisades steam generators because over 20 percent of the Palisades steam generator tubes have been plugged due to an earlier tube wastage problem which has generally arrested. We are not aware of any other plants considering steam generator replace-ment (repair). However, as indicated below, approximately half of the operating plants have experienced some fonn of degradation and unless these problems arrest, other plants may have to consider replacement.

Operating Degraded Plants Severely Manufacturer Plants Plants Affected*

Westinghouse 26 13 4

Combustion Engineering 8

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Babcock & Wilcox.

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• Severely Effected Plants are those with more than 10 percent of all tubes plugged, i.e., Surry 1, Surry 2, Turkey Point 3, Turkey Point 4 and Pal i sades~.

Steam Generator Degradation The steam generator tube degradation mechanisms fall into three general categories:

(1) wastage or tube wall thinning due to corrosive chemical attack or vibration and metal-to-metal contact between tubes and support structure (bars), (2) tube denting caused by build up of corrosion pro-ducts between tube support structures (plates) and outside tube walls (and varicus associated problems), and (3) tube cracking due to metal fatigue resulting from flow-induced vibration.

Steam generators manufactured by Westinghouse and Corbustion Engineering are of the recirculation type (tubes are "U" shaped) and have experienced both wastage and denting degradation. The NRC staff has recently issued a report titled "Sumary of Operating Experience with Recirculating Steam Generators," NUREG 0523, which discusses tube degradation mechanisms, the programs for resolving these problems, actions being taken by the NRC and for licensing new plants.

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.g Steam generators manufactured by Babcock and Wilcox (B&W) are of the once-through type (tubes are straight) and have experienced flow-induced cracking and have not experienced problems with either denting or wastage.

5 Far fewer tubes have been plugged in B&W plants.

Cost of Steam Generator Replacement (Repair)

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The total cost of the repair at Surry 1 and 2 will be $133 million.

E This consists of the cost of $66 million for the repair itself (material and labor), plus the $66 million differential fuel cost that will be i

experienced while the plants are down for repair, plus $1 million for disposal of removed steam generators (assuming $0.014/Kw. hr. cost

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differential).

i Occupational Exposure Associated with Steam Generater Repair Steam generator problems are a major contributor to work exposure (occupa-tional exposure).

For example, at Surry 1 ud 2 in 1977 steam generator inspections resulted in a combined exposure of about 1400 man-rem while total other occupational exposure was about 1000 man-rem. The 1977 U.S.

industry-wide average occupational exposure per reactor was between 500-600 man-rem.

Following steam generator repair, the NRC staff estimates a significant reduction in occupational exposure.from steam generator inspec-tions (less than about 100 man-rem / year).

It.is estimated that a total of about 2000 man-rem occupational exposure will result from replacement' of the three steam generators at each of the Surry units.

Question 9.

Congressman Vento expressed concern at the March 2 hearing about who bears the cost of the NRC's regulatory activities. Please answer the following related questions:

(a)

From the time an applicant publicly announces the intent to build a nuclear power plant, what is the total cost of the NRC's licensing process leading to a final decision on a construction permit application? How much of this total is paid by the applicant?

(b) From the time an NRC construction permit is issued, what is the total cost of the NRC's licensing activities leading to a final decision on an operating license? How much of this total is paid by the applicant?

(c) What portion of the NRC's reactor inspection costs are borne by the licensee curing construction? During operation? What portion of NRC's vendor and contractors inspectiors are borne by the vendors or contractors?

What is the NRC's rationale for not requiring the licensees to bear the full costs of these inspections?

Answer (a)

In the NRC -cost recovery program the cost incurred in the review of an application for a construction permit was--

determined and thisccost~was used-as an uppere.limi.t.fori.the: ",rm fee schedule developed.by the Commission in.FY 1977.

Under the fee program a licensee would be. assessed a. charge. based.

on actuzl cost up to $1,069,000.

Since the current fee schedule was developed costs have risen substantially, e.g.,

approximately 50% for the Office of Nuclear Reactor Regulation; however, recovery is still limited to $1,069,000.

It is the NRC's intent to reassess the costs of licensing and inspection in the near future and where appropriate adjustments will be made.

It should be noted that none of the Commission's costs associated with generic licensing, standards development and research are included in fees.

(b) The average cost of reviewing an application for an operating license is $1,024,500. Licensees are assessed fees based on actual review costs up to a limit of $1,024,500.

Operating license fees were developed on the same basis as fees for construction permits.

Costs in NRR for these reviews haya approximately doubled.

(c) At the time tne current fee senedule was developed in FY 1977, inspection costs asscciated with a construction permit were approximately $160,000 and $290,000 for an operating license.

These costs were used as the upper limit in the NRC schedule of license fees.

Licensees pay inspection costs up to these limits.

l Question & Answers Vendors and contractors pay none of the costs incurred in inspections. The Commission's schedule of fees is based on guidance from Court decisions involving the FCC's license fee program.

Both the FCC and NRC derive their statutory authority for cost recovery from the Independent Offices Appropriation Act of 1952.

Court guidance provided that fees must be assessed to an applicant for specific measurable services.

Since vendors in the Licensee Contractor and Vendor Inspection Program (LCVIP) do not file applications or requests for licenses or approvals, the NRC is unable to assess the vendor for inspections.

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QUESTION *

10. As Congressman Weaver requested at the Subcommittee hearings, please provide a list of all post-Construction permit design changes that have been required by the NRC for Washington Public Power i

Supply System's (WPPSS) units 1, 2, 3, 4, and 5.

ANSWER.

To place the request by the Subcommittee in proper context,,

it should be noted that there are numerous changes in the preliminary design and in the design criteria of a proposed nuclear power plant

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during the staff's construction permit (CD) review.

From the discussion in the transcript of the hearing held on February 22, 1979, we under-stand that the Subecmittee's concern involves those changes made to a facility design after issuance of the CP.

In particular, as discussed by Chairman Hendrie in the February 22nd hearing, the issuance of a CP by the staff attests to the technical feasibility of a proposed facility based on the staff's review of an applicant's preliminary design and design criteria.

In the course of establishing the detailed design after issuance of the CP, changes may be made by the applicant to arrive at the final design. Changes of this type made by the applicant after CP issuance are not reportable to the Comission, even though numerous, unless they are reportable in accordance with Section 50.55(e)(fi) of 10 CFR Part 50 (design deficiencies). Howe'.ar, even those cMnges not reportable to the staff will be presented in the OL application and will be evaluated by the staff prior to issuance of an operating license.

Those requirements imposed by the staff on the five WPPSS reactors, following issuance of the CP's,are presented in Attachment 1.

In addition to these staff requirements, the staff has issued letters for WNP Nos.1, 3, 4 and 5 (Attachment 2) stating additional requirements related to new or revised staff positions. While the WPPSS response to some of these items may well be that the additional staff requirements can be satisfied by the present design, it is anticipated that some changes to the design of a particular facility may be necessary. The staff will evaluate the applicant's response to all items in Attachment 2 during the OL review phase.

It should be noted that a single staff requirement for a facility can generate numerous individual change notices due to the inethod of awarding contracts; i.e., changes may be generated in a civil structures contract, an electrical con ract and a mechanical contract.

There may be additional staff raquirements for the WNP-2 facility I

as a result of the program to evaluate the dynamic pool loads in the Mark II containment LISTIllG 0F SAFETY-RELATED MATTERS TO BE RESOLVED AFTER ISSUAtlCE OF Tile CP Facility Item Description Basis All Fire Design features to enhance the capability to Appendix A to Protection protect safety components from fire.

Section 9.5.1 of the SRP All Industrial Design features which will minimize the potential 10 CFR Part 73 Security of industrial sabotage of safety-related systems.

WilP-3&5 Dewater-tionitor the dewatering system during construction Verify parameters ing system to verify parameters used in the design of the used in the design system.

WilP-2 Drywell Seal design to minimize bypass leakage in the Staff criteria bypass containment.

and ACRS recomend-leakage ation WilP-2 Pipe whip Pipe. restraints to prevent damage to safety-related Staff criteria and components due to postulated pipe breaks ACRS recomendation WilP-2 Criteria Stress and deformation limits for inactive pumps Staff criteria and for pumps and valves.

ACRS recommendation and valves WitP-2 Main steam Design of a closed sealing system for the main Staff criteria and line steam line isolation valves.

ACRS recommendation WilP-2 Control Design of a system to minimize the probability Staff criteria and rod drop of an accidental control rod drop.

ACRS recommendation WNP-2 ATWS Design of a system to minimize reactor transients Staff criteria and in the event of a postulated failure of the control ACRS recomendation rod c.ystem.

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i attacnment z

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NOV 21197B Docket No. 50-460, 50-513,50-508 l

and 50-509 I

l Mr. N. O. Strand Managing Director I

Washington Public Power Supply System P. O. Box 968 l'

3000 George kashington Way Richland, Washington 99352 i

SUBJECT:

IMPLEMENTATION OF STAFF REVIEW REQUIREMENTS - WASHINGTON PUPLIC POWER SUPPLY SYSTEM (WPPSS) NUCLEAR PROJECT N05.1, 3, 4 AND 5 - OPERATING LICENSE REVIEW During the last several years we have reviewed and approved several new regulatory guides and branch technical positions or other modifications to existing staff positions. Our practice is that substantive changes in staff positions be considered by the NRC's Regulatory Requirements Review Coa:2ittee (RRRC) which then recommends a course of sction to the Director.

Office of Nuclear Reactor Regulation (NRR). The recn:tnended action includes an it:plementation schedule.

The Director's approval then is used by the NRR staff as review guidance on individual licensing matters. Some of these actions will affect your application.

This letter is intended to bring you up to date on these changes in staff positions so that you -

cay consider the:n in your Final Safety Analysis Report ]FSAR) preparation.

The RRRC applies a categorization nomenclature to each of its actions.

(A copy of the svunary of RRRC Meeting No. 31 concerning this'categoriza-tion is attached as Enclosure 1.)

Category I natters are those to be applied to applications in accordance with the innplementation section of the published guide. We have enclosed lists of actions which are either Category 2 or Category 3, which are defined as follows:

Category 2: A new position whoe' oplicability is to be determined on a case-by-case ba!

You should describe the extent to

>:hich your design aforms, or you should describe an acceptable alterna.e, or you should de.nonstrate why confor-mance is not necessary.

Category 3: Conformance or an acceptable alternative is required.

If you do not conform. or do not has e an acceptable alternate. -

the9 staff-approvec design revis1ons will be required.

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,y NOV ' 1 1978 Mr. 14. O. Strand 4 i

k'e believe that providing you with a list of the Category 2 and 3 matters ar.': seu w wa*.; will be useful in your FSAR preparation, and they will De an essential part of our operating license review. Enclosure 2 is a

' list of the Category 2 matters. is a list of {he Category 3 matters In addition to the RRP.C categories, there also exists an i;RR Category 4 list which are those matters not yet reviewed by the RRRC, but which a

!j the Director, !!RR, has deemed to have sufficient attributes to warrant l

their being addressed and considered in ongoing reviews. These matters will be treated like Category 2 matters until such time as they are reviewed by the RRRC, and a definite implementation program is developed..

I A current list of Category 4 matters is attached (Enclosure 4). These also should be considered in your FSAR.

In some instances the items in the enclosures may not be applicable to

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your applicatica. Also, we recognize that your application may, in some l

instances, already conform to the stated staff positions.

In your FSAR you should note such compliance.

If you have any questions please let us know.

Sincerely, I

j C4.n 1ig=ed br..

Eger S. Boyd Roger S. Boyd, Director Division of Project Management Office of fluclear Reactor Regulation

Enclosures:

As stated p

cc:

See next page I

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September 15, 1978

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CATEGORY 2 MATTERS l}'

Document 1

Number Revision Date Title RG 1.27 2

1/76 Ultimate Heat Sink for Nuclear Power Plants RG 1.52 1

7/76 Design, Testing, and Maintenance Criteria for Engineered-Safety-l Feature Atmosphere Cleanup System j

Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants (Revision 2 has been published but the changes from Revision 1 to Revision 2 may, but need not, be considered.

RG 1.59 2

8/77 Design Basis Floods for Nuclear Power Plants RG 1.63 2

7/78 Electric Penetration Assemblies in Containment Structures for Light Water Cooled Nuclear Power Plants RG 1.91 1

2/78 Evaluation of Explosions Postulated to Occur on Transportation Rottes Near Nuclear Power Plant Sites RG 1.102 1

9/76 Flood Protection for Nuclear Power Plants RG 1.105 1

11/76 Instrument Setpoints RG 1.108 1

8/77 Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants RG 1.115 1

7/77 Protection Against Low-Trajectory Turbine Missiles RG 1.117 1

4/78 Tornado Design Classification RG 1.124 1

1/78 Service Limits ar.d Loading Combinations for Class 1 Linear Type Component Supports RG 1.130 0

7/77 Design Limits and Loading Combinations for Class 1 Plate-and Shell-Type Component Supports (Continued) k--

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CATEGORY 2 MATTERS (CONT'D).

Continued Document Number Revision Date Title RG 1.137 0

1/78 Fuel Oil Systems for Standby Diesel Generators (Paragraph C.2)

RG 8.8 2

3/77 Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will be as Low as is Reasonably Achievable (Nuclear Power Reactors)

BTP ASB Guidelines for Fire Protection for 9.5-1 1

Nuclear Power Plants (See Implementation Section, Section D)

BTP MTEB 5-7 4/77 Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping RG 1.141 0

4/78 Containment Isolation : Provisions-for Fluid Systems

. _N_

t_._._-

/

September 15, 1978 CATEGORY 3 MATTERS Document Number Revision Date Title RG 1.99 1

4/77 Effects of Residual Elements on Predicted Radiation Damage to C

Reactor Vessel Materials (Paragraphs C.1 and C.2.

RG 1.101 1

3/77 Emergency Planning for Nuclear Power Plants RG 1.114 1

11/76 Guidance on Being Operator at the Controls of a Nuclear Power Plant RG 1.121 0

8/76 Bases for Plugging Degraded PWR Steam Generator Tubes RG 1.127 1

3/78 Inspection of-Water-Control Structures Associated with Nuclear Power Plants RSB 5-1 1

1/78 Branch Technical-Position:-Design Requir ments of.the Residual Heat:Remova'l Systi RSB 5-2 0

3/78 Branch Technical _ Position:

Reactor -

Coolant System Overpressurization Protection (Draft copy attached)

RG 1.97 1

8/77 Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During -

arid Following an Accident (Paragraph C.3 - with additional guidance on paragraph C.3.d to be provided later)

RG 1.68.7 1

7/78 Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants RG 1.56 1

7/78 Maintenance of Water Purity in Boiling Water Reactors l

Attachment:

r STP RSB 5-2 (Draft)

__...._...__._m__a._____..-

y

,t BRAttCH TECH!l! CAL POSITIO!! RSB 5-2 OVERPRESSURIZATION PROTECT!0ti 0F PRESSURIZED WATER REACTORS WHILE OPERATIt!G AT LOW TEMPERATURES A.

Backcround General Design Criterion 15 of Appendix A,10 CFR 50, requires that "the Reactor Coolant System and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences."

Anticipated operational occurrences, as defined in Appendix A of 10 CFR 50, are "those conditions of normal operation which are expected-to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power."

Aopendix G of 10 CFR 50 provides the fracture toughness requirements for reactor pressure vessels under all conditions.

To-assure that the Accendix G 1imits of the-reactor coolant.oressure boundary:are nota 7. u: '

exceeded during any anticipated operational occurrences,- Tecnnical Specification pressure-temperature limits are provided. for cperating the plant.

The primary concern of this position is that during startup and shutdown conditions at icw temperature, especially in a water-solid condition, the reactor coolant system pressure might exceed the reactor vessel pressure-tem::erature limitations in the Technical Specifications established for protection against brittle ' fracture.

This inadvertent overpressurization could be generated by any one of a variety of mal-functions or operato" errors. Many incidents have occurred in operating plants as described in Reference 1.

Adcitional discussion on the background of +,his position is contained in Reference 1.

lft'I~'-M iM)

B.

Branch Position 1.'

A system should be designed and installed which will prevent exceeding the applicable Technical Specifications and Appendix G limits for the reactor coolant system while operation at low, temperatures.

The system shoul be capable of relieving pressure during all anticipated overpressurization events at a rate sufficient to satisfy the Technical Specification limits, particularly while the reactor coolant system is in a water-solid condition.

2.

The system must be able to perform its function assuming any sing active compocent failure.

Analyses using appropriate calculational techniques must be provided which demonstrate that the system will provide the required pressure relief capacity assuming the most limiting single active failure.

The cause for initiation of the event, e.g., operator error, component malfunction, will not be considered as the single active failure.

the most limiting allowable ooerating conditions and systemsThe configuration at the time of the costulateo cause of the overoressure All ootential overpressurization events must'be ' considered event.

when establishino the worst case event.

Snnie events may be per' vented by protective interlocks or by locking out power ila se events should be reviewed on an individual basis

~

mterinck/ power lockout is acceptable, it car. be excluderi from

'If the the analyses peuv'ided the controls to prevent the event a in the plant Technical Specifications.

re- -

3.

The 5y5 tern #iust' asst the ~ design requirements of-!CEE leolementation).

279'(566"

the electrical'insfrumentation anc control system must pr alarms to alert the operator to:

- ~~

a.

properly enaale the system at the correct plant condition during cooldown, b.

indicate if a pressure transient is occurring.

4 To assure operational readiness, the overpressure protection system must be tested in the following manner:

A test must be performed to assure operability of the system a.

electronics prior to each shutdown.

b.

A test for valve operability must, as a minimum be conducted

_ as specified in the ASME Code Section XI.

Subsequent to system, valve, or electronics mainten;nce, a test c.

on that portion (s) of the system must be performed prior to declaring the system operational.

, /.,.;' '

y~ /.

. 5.

The system must meet the requirements of Regulatory Guide 1.26,

~

" Quality Group Classifications and Standards for Water, Steam,

and Radioactive-Waste-Containing Components of Nuclear Power Plants" and Section !!! of the ASME. Code.-

l l

6.

The overpressure protection system must be designed to function during an Operating Basis Earthquake.

It must not compromise the design criteria of any other safety-grade system with which it would interface, such that the requirements of Regulatory Guide 1.29, ", Seismic Design Classification" are met.

7.

The overpressure protection system must not depend on the availability of offsite power to perform its function.

8.

Overpressure protection systems which take credit for an active component (s) to mitigate the consequences of an overpressurization event must include additional analyses considering inadvertent system initiation / actuation or provide -justification-'to show that '

existing analyses bound such an event.

C.

Imolementation The Branch Technical Position...as specified in Section B, will be -used --

in the review of-all-Preliminary Design Approval (PDA)~, Final Design

~ - - -

Approval (FDA), Manufacturing License (ML),' Operating License '(OL), and Construction Permit (CP) applications involving plant. designs incorporating-pressurized water reactors.

All ascects of..the position will be -applicable to all applications;-including-CP applications utilizing the replication ~

oction of.the Cami55ica!s.istandardizstirpecgram,rthat arrdo'cketed" 3'"

af ter 14 arch 14,- 1978. _ A.1 aspects of-the positionr with the exception fir

of reasonable 2nd. justified _ deviations from-IEEE 279 requirements, will be applicable to CP, OL, ML, PDA, and FDA applications docketed prior to March 14, 1978 but for which the licensing action has not been completed as of March 14, 1978.

Holders of appropriate PDA's will be informed by letter that all aspects of the position with the exception of IEEE 279 will be applicable to their approved standard designs and that such designs should be modified, as necessary, to conform to the position.

Staff approval of proposed modifications can be applied for either by application by the PDA-holder on the PDA-docket or by each CP applicant referencing the standard design on its docket.

The following guidelines may be used, if necessary, to alleviate impacts on licensing schedules for plants involved in licensing proceedings nearing completion on March 14, 1978:

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1.

Those applicants issued an OL during tihe period between March 14,

^

1978 and a date 12 months thereafter may merely commit to meeting the position prior to OL issuance but shall, by license condition

^

be required to install all required staff-approved modification 5',

prior to plant startup following the first scheduled refueling outage.

2.

Those applicants issued an OL beyond March 14, 1979 shall install all required staff-approved modifications prior to initial plant startup.

3.

Those applicants issued a CP, PDA, or ML during the period between March 14, 1978 and a date 6 months thereafter may merely comit to meeting the position but shall, i,y license condition, be required to amend the application, within 6 months of the date of issuance of the CP, pDA, or ML, to include a description of the proposed modifications and the bases for their design, and a request for staff approval.

4 Those applicants issued a CP, PDA, or ML after September 14, 1978 shall have staff approval of proposed modifications prior to issuance of the CP, PDA, or ML.

D References l.

NUREG-0138, Staff Discussion of Fif teen Technical Issues Listed in Attachment to November 3,1976 Memorandum from Director, NRR, ^

to NRR Staff.

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OUESTION 11.

As Congressman Huckaby requested at the Subcommittee hearing, please provide a brief status report on nuclear waste manage-ment programs in other countries (e.g., France, Japan, Germany, and the Soviet Union).

In doing so, indicate if there is anything those countries have learned that the United States could adopt to help solve our nuclear waste problems.

ANSWER.

The current status of. nuclear waste. management programs in_ _various countries throughout the yprid was recently summarized by the International Atomic Energy Agency. --

Excerpts of that assessment for various countries follow.

Austria Investigations concern the potential _ sites. for waste disposal in. Austrian territory and also the possibilities of obtaining the services.of_.other countries for disposal of nuclear waste.

Crystalline rock formations, especially of the Southern Bohemian massif, are being studied.

An initial design has been made.for.a repository..for..the. waste _which_is expected.to._ __ __.

be returned from. foreign spent. fuel _ reprocessing _in _the.1990.'s.. Assurance ____.._

of waste disposal is a condition for operating the first Austrian nuclear power station. - - ~ -

Belgium and Italy In Belgium and Italy, the utility of argillaceous (clayey) formations for waste disposal is being investigated.

In Belgium, prospecting and laboratory investigations of the properties of a clay bed near Mol are under way.

Studies also concern radionuclide migration, heating experiments, and conceptual design, including underground excavations in the clay bed.

Field teste in an experimental cavern will also be carried out.

Similar studies in Italy concentrate on theoretical studies and laboratory tests, and deal with the hydro-geological, geotechnical and thermal characteristics of clay formations in an area in southern Italy.

Canada Canada is exploring the potential of the plutons (large bodies of igneous rock) in the Canadian Shield for waste repositories.

Salt formations are II nternational Atomic Eneroy Agency Bulletin, Volume 20, Number 4, I

August 1978.

l l

. also being studied. The program covers geological reconnaissance, conceptual design and engineering for a model repository.

Laboratory and field tests being carried out include heating experiments in a part of an operating mine and studies on groundwater behavior.

Czechoslovakia Czechoslovakia has surveyed its territory to select a repository at shallow depth for low-and-intermediate-waste from nuclear power plant operations.

For high-level waste disposal, interest is focused on the use of crystalline rocks.

France Activities concentrate on the potential for waste disposal in crystalline rocks, in particular, granite.

Consideration is also given to the possibilities of waste disposal in salt, shale or clay. The characteristics of deep granite formations are being studied.

Investigations also include the modeling of radionuclide transfer by groundwater, for which a mathematical model. has been developed, and studies of geochemical barriers such.as layers of clay.

German Democratic Reput.l_ic, The potential for. adioattiva waste disposal.on_its<. territory..was studied and a former salt mine was selected for adaptation as a central repository for the disposal of low-ald-intermediate-ievel vaste from ruclear power- -

plant operations. The problems connected with the disposal of high-level waste into rock salt are also being examined.

Federal Republic of Germany The Federal Republic of Germany has been using the former Asse salt mine as a research and development facility.

Preparations for test disposal of solidified, high-level waste are under way.

The research program includes a rock cavern project at the Asse salt mine for demonstrating the emplacement of low-and-intermediate-level waste directly from the surrace into a large deep underground cavity.

Research also covers theoretical, laboratory and field investigations concerning heat dissipation and rock mechanics, risk analysis, repository design, the storage of conditioned spent-fuel elements from the AVR reactor, and investigations on the possibilities of in-situ solidification of liquid low-and intermediate-level waste.

Reconnaissance work has been carried out at several salt domes in the northern part of the country for a potential repository that could take the waste from a fuel cycle center planned for the late 1980's. The suitability of a specific site is currently being investigated.

p

. India Various geological formations are being evaluated, in particular, igneous rocks and selected sedimentary deposits in non-seismic zones.

Detailed site investigations are expected to follow.

The Netherlands The possibilities of waste disposal in one of the salt domes in the north-eastern part of the country are being explored. Work is being carried out on a repositcry design, thermal impacts and safety assessment, including sorption studies in saline groundwater.

Sweden i

An all-out effort was devoted to meet the requirements of a law passed in 1977 which sets out conditions for establishing any new nuclear power plant.

The conditions require the reactor operator to have a contract providing for the reprocessing of spent fuel and to demonstiate how and where the final disposition (disposal) of the highly radioactive waste. resulting.from.

reprocessing the fuel can be accomplished with absolute safety.

As a result of the law, besides long-term studies sponsored by the Government, the Swedish utility companies. started a special project '.' Nuclear Fuel. Safety" (KBS) to demonstrate the possibil_ities.of. disposing.high-level waste.and/or spent fuels in a deep geological repository in crystalline rocks.

The work encompasses area surveys hydro-geological' site -investigationrand -field-tests; mechanical and hydrological characteristics of the rock; special investigations on low groundwater flow and radionuclide migration in fractured rocks; conditioning and emplacement techniques; repository design and safety analyses. The project also includes field tests of a granite formation in the former Stripa iron mine in central Sweden, where in-situ heater experiments and special hydro-geologic investigatiens are being performed.

Switzerland Studi concern the potential of anhydrite caverns for the disposal of low-and-intermediate-level waste, and the possibility of high-level waste disposal in certain salt or crystalline basement rocks.

United Kinadom Research is being carried out on the suitability of crystalline rocks for the disposal of high-level waste and, to a lesser extant, on argillaceous and salt formations. The research efforts encompass laboratory and theoretical studies of rock and groundwater behavior, interim field studies at sites having equivalent characteristics to those which eventually may prove i

suitable, and actual site investigations. They involve fluid / rock and fluid / waste interactions, groundwater research, modeling of thermal stressing and scaled heater experiments, repository design studies, detailed surface and sub-surface geological surveying.

U.S.S.R.

The USSR has been studying the technical, geological, hydrological physico-chemical, and thermal aspects connected with the disposal of low-intermediate-and high-level liquid waste by injection into deep, isolated porous strata.

Comprehensive field experience is available and tests for the injection of high-level liquid waste have been performed.

Research and development work l

1s also being done on the disposal of solidified high-level waste.

It includes investigations and field experiments of long-term storage in engineered, dry-storage facilities at shallow depths, and studies of the use of deep, continental rocks, in particular, of salt formations.

The Nuclear Regulatory Commission (NRC) plans to make greater use of the results of the development efforts being conducted in various countries in the near future. To date, the NRC has not made a concerted effort to integrate the results of international waste management programs in our waste management activities such as regulatory development and licensing.

It would appear that the activities on the international level would be particularly valuable to the United States in formulating decisions on matters relating to the selection of appropriate geologic media for waste disposal.

6

QUESTION 12.

Congressman Vento has requested that the NRC provide answers for the Subcommittee record to the following nuclear waste questions.

(a)

NRC staff has said that repositories for high-level waste should be subject to NRC review and/or rarmitting before major drilling is conducted on the site.

Large-scale drilling or sinking of shafts may permanently damage the safety of a repository site if improperly done, according to the staff.

DOE has plans to begin such major drilling activities at at least one site, Carlsbad, New Mexico, perhaps within a year. Will NRC have a regulatory review program in place before major drilling 5"rts?

ANSWER.

The NRC is unaware of any plans by DOE to begin any type of large scale drilling or sinking of shafts at the WIPP site within one year.

Based on our current schedule, we propose to have our regulations, both procedural and technical, in place in draft form by early 1980 and in final form by early 1981.

EPA's schedule indicates that.they will also have their regulations in place on this time frame.

QUESTION.

(b)

NRC staff regularly advises prospective applicants for reactor and uranium mill licenses whether their preliminary plans seem to adcluately consider licensing (i.e., health and safety) issues.

Does NRC now perform such a role with respect to DOE's plans for high level waste repositories? Has DOE solicited such advice?

ANSWER.

Several meetings between DOE and NRC were held during the spring and early summer of 1978 to discuss the licensing aspects of the WIPP facility.

However, the Ptzblic Works Appropriations Bill, H.R.12928, prohibited DOE from making expenditures for any purpose related to the obtaining or issuing of a license to operate the WIPP facility. Therefore, these meetings were terminated.

On February 6, 1979, NRC and DOE initiated what is intended to be a continuing series of technical information exchange meetings relating to the radioactive waste program.

The first meeting focused on the overall DOE waste management effort. Topics for future meetings include siting and engineering criteria for waste repositories, geologic explorations, environmental surveys, waste immobilization, in-situ test programs, repository conceptual designs, and quality assurance.

QUESTION.

(c) What could be the benefits and/or problems associated with such a role for NRC vis-a-vis DOE?

i ANSWER.

i The NRC views these types of meetings as having a beneficial effect on both NRC and DOE.

For DOE, the meetings will provide their staff with the NRC's point of view on procedural and technical licensing issues.

NRC will benefit by obtaining up-to-date and first-hand data and information on repository issues which would enable the NRC to develop better regulc tions, standards, and criteria.

Few, if any problems, should result from this exchange of data and information if the meetings are condur.ted in an open and appropriate manner.

i

0 0

e b

9 e

II QUESTIONS FOR MR. DENTON

Ql[ESTIONII.l. On February 22 and March 2 there was a discussion of the potential impact of the LOFT experiments.upon regulations.

Please elaborate for the record what you believe the impact of these experiments will be.

In particular:

(a) Do you believe that it is probable that the LOFT experiments will lead to a loosening of regulations so that existing reactors will be allowed to operate at higher power levels?

ANSWER.

We do not believe that the LOFT experimental results will lead to a loosening of regulations such that operating reactors could operate at higher power levels. The intent of the LOFT tests is to provide integral experimental data for the verification of computer programs used to analyze reactor behavior.

In addition, the basic phenomena and system behavior under postulated loss-of-coolant conditions can be evaluated.

Subsequently, this information would be helpful to confirm the margins in safety analyses.

QUESTION.

(b)

Do you have confidence in the ability.

to scale the LOFT data to 1000 megawatt olants?

ANSUER.

It is not intended that the LOFT experimental results be scaled directly to 1000 megawatt plants.

Rather, the LOFT data will be used to assess and refine current computer programs used for reactor analysis. The extrapolation to large scale reactors will be made through application of the computer programs.

We are confident that the extrapolation can be made, albeit with some uncertainty which will be accounted for.

QUESTION.

(c) Do you believe that the LOFT program can be reduced in scope? If so, please indicate the kinds of experiment that you believe might be eliminated.

ANSWER.

The nuclear portion of the LOFT experimental program has just begun; test L2-2 is the only nuclear test performed thus far.

Test L2-2 was performed at an intermediate power level.

The test series is designed to run through FY-86 and will provide information at higher power levels for postulated hot and cold leg breaks, alternate ECCS systems, concurrent steam generator tube failures, operational transients, and anticipated transients without scram. These tests are all important to the regulatory program. At this time, we do not believe that the planned test schedule should be reduced in scope. As additional tests are performed and the data evaluated, the need to perform certain tests in the scheduled program can be reassessed.

QUESTION II.2. The following relate to the Semiscale experiments:

a.

What will be the likely impact of these experiments upon the NRC's reactor regulations and licensing conditions?

ANSWER.

The Semiscale experiments are used to provide integral system behavior test data primarily for LOCA analysis code assessment. These experiments provide confirmatory data for NRC regulatiens and licensing conditions.

The regulations are judged to provide a conservative margin above statistical uncertainties and the experiments are designed to gain insight as to the nature and size of this margin.

In this regard, Semiscale results help to establish and evaluate the adequacy (i.e., uncertainty) and conservative margins in the computer codes used to calculate ECCS performance.

This code uncertainty assessment will be used as part of the overall assessment of the conservatism in Appendix K to 10 CFR 50 to support proposed changes to 10 CFR 50.46 and Appendix K to 10 CFR 50 Regulations.

The Semiscale program is also used to establish uncer.tainty margins against which vendor licensing models can he judged.

Two recent examples of this are the small break LOCA-models and the upper head. injection (UHI) model.

In order to address :oncerns-associated with the capabilityrof licensing -

codes to conserva'..vely prenict -the.thermarl-hydraul-ic behavior _ assoc.iated.

with small piping breaks, a small break experiment was recently performed in the Semiscale facility.

Each PWR reactor vendor was requested to predict the experiment. When all of the predictions are completed, the data will be compared and the prediction capability of present computer models assessed.

UHI experiments are scheduled to be performed in Semiscale from which uncertainties in the performance evaluation of PWR plants equipped with upper head injection ECC system will be better defined.

The Semiscale program is also used to establish.a data base against which calculational techniques can be checked to determine if observed phenomena are properly predicted. : A recent example of this is Semiscale test 5-07-6, in which hyrdaulic oscillations in the vessel downcomer were observed.

Since this phenomenon was not predicted to occur by the analysis codes, the staff is pertor.inng a detailed evaluation to determine if the same phenomenon would occur in large prototype reactors and if the analysis codes properly account for this phenomenon.

Safety margins in light water reactors will be evaluated in

part by comparing best estimate calculations to those obtained from conservative licensing evaluation model codes. The Semiscale program provides a major source of data to assess the models in the best estimate code being developed under flRC's Research Program.

QUESTION 2.

b.

What would be the impact upon the f1RC's regulatory program of termination of Semiscale after fiscal year 1979?

~

ANSWER.

The termination of the Semiscale program after fiscal year 1979 would eliminate c major source of experimental integral systems data used for licensing code assessment.

Some of the more important examples of this are as follows:

1.

The review of the new Westinghouse ECC upper head injection-(UHI) system indicated large uncertainties in the performance evaluation of UHI systems. This was to a large extent due..to the lack cf experimental data.

It is the NRC goal to experimentally investigate the new UHI ECC systems ' prior to power operation mf the first UHI-equipped plant.

Semiscale provides the experimental vehicle to obtain the ne' assary data.

2.

To date, all integral LOCA tests performed in Semiscale-MOD-1 and LOFT utilize a short Es foot core.

There are outstanding questions about the applicability of these data to 12 foot cores, typical of PWRs.

It is the goal of the Semiscale MOD-2 and MOD-3 systems, which will have 12 foot cores, to answer these questions.

3.

Semiscale test performed to date have identified physical phenomena not accounted for in LOCA licensing models.

Should these phenomena be important for PWR evaluation, experimental data are needed to support development of revised models.

4.

More data are needed to evaluate uncertainties in small break calculations and improve small break analyses, and semiscale is presently the only source of this data.

5.

Two aspects of LOCA analyses presently being questioned are'the effects of accumulator water volume and contrinrrrnt backpressure on LOCA.

Semiscale testing was planned to obtain this needed data.

The elimination of the Semiscale facility at the end of FY 79 would also require the staff to rely almost completely on the vendors and applicants for confirmatory research data in support of their proposed models. This has not been an entirely satisfactory situation in the past.

OUESTION 2.

c.

Assuming Semiscale is not terminated, what modifications would you suggest in this program?

ANSWER.

NRR and RES have worked closely on the Semiscale program, and the current program has already been modified to accomodate NRR needs.

These include small break testing and upper head injection testing.

In addition to the presently planned testing, NRR is considering requesting that the following additional tests be included in the program scope.

e Perform additional testing in support of UHI tests.

This testing includes upper head / plenum draining characterization tests, and two-pipe flooding / bypass characterization tests.

e Perform testing of upper plenum injection.

a Perform testing on the effect of accumulator water volume on LOCA e Perform testing on the effect of containment backpressure on LOCA.

QUESTION II.3.

The following concerns research conducted by the Office of Regulatory Research related to security at nuclear power plants.

a.

How has the safeguards research conducted by the Office of Regulatory Research been used by the Office of Reactor Regulation?

ANSWER.

The safeguards research applicable to reactors falls into three

~

categories:

1.

Vulnerability analyses, including fault tree analyses of systens vital for safe shutdown, barrier penetration data, etc.

2.

Evaluation of safeguards effectiveness of reactor design alternatives.

3.

Evaluation models for the quantitative evaluation of the effectiveness of physical security measures.

NRR uses the fault tree vulnerability analyses-to assess-the adequacy and completeness of licensee submitted vital area lists and has made use of data concerning tarriers' and security-hardwarer~ "

~ ~ ~

The physical security evaluation models' have-limited applicability to~

nuclear NRR has used simplified versions of these models (power reactors.i.e., "EASI" and "TS0") as a tool to aid in the qualita assessment of physical security by security experts.

QUESTION.

b.

Since tha NRC was established, what changes in reactor security regulations have been instituted as the result of research sponsored by the Office of Nuclear Regulatory Research?

ANSWER.

The reactor vulnerability data identified as a result of research has contributed to the NRC's basis for issuing a 10 CFR 73.55, which requires a substantial upgrading of security at power reactors, and has influenced the development of the basic approach to physical security at power reactors.

QUESTION.

c.

How does the Office of Reactor Regulation intend to use the results of ongoing security related research being conducted by the Office of Nuclear Regulation Research?

ANSWER; NRR is explo-ing the need for further research in the following areas:

2_

1.

Detection of modern explosives and instrumental personnel surveillance methods at nuclear power reactors.

2.

Development of evaluation criteria and technical data base in the area of guard training and contingency planning.

3.

Continued improvement of vulnerability analysis tools.

4.

Evaluation of design alternatives to reduce vulnerability of power reactors to sabotage.

NRR plans to use the first 3 of the above items to improve the effectiveness, completeness, and efficiency of the reviews and evaluation of licensee submitted security, contingency and guard training plans.

The 4th item is intended to identify long term improvements in power reactor design.

NRR plans to initiate changes and/or additions-in-the regulations, reg. guides, and review plans, as appropriate to implement any design modifications or improvements identified by this program.

QUESTION.

d.

If the Office of Reactor Regulation. controlled-the funds expended upon reactor related security by the Office of Regulatory ~Research would 'these funds have been expended upon the. programs which have been conducted to date?_ What-changes-would you suggest witn respect to the nature of these programs?

ANSWER.

If NRR had control over the funds expended on reactor-related safeguards, NRR would expand research related to vulnerability analyses and reactor design evaluations, identified as items 1 and 2 of question 3.a above.

NRR would deemphasize the development of complex physical security evaluation models (identified as item 3 in 3.a above).

Funds currently committed to these models could be re-programmed to expand research on matters related to physical security hardware and procedures used for intrusion detection, access control and personnel activity monitoring at power reactors.

III QUESTIONS FOR MR. DIRCKS

e

' Question - Mr. Dircks (Director, Office of Nuclear Material Safety and Safeguards):

III-1-a.

How does ONMSS use safeguard research conducted by the Office of Nuclear Regulatory Research?

Answer:

The safeguards research program is expected to be beneficial in assisting with the formulation of regulations and guides, assisting in assessing the effectiveness of licensee safeguards systems with respect to threats and safeguards performance and supporting inhouse studies and-other associated-aspects. -

of the regulatory program. Specific exampley are:

1) the products from the " Effectiveness Evaluation Methods for* -> -

Material. Control and Accounting" research project will-be-technical-inputs considered in the' development:of the proposed MC&A upgrade rule and associated. guidance;.2)_it_is. hoped..

that the Safeguards Ne.twork Analysis. Program.wi<l.inbe usefu bin:

formulating insights ~into guard: tactics and; strategies planned

~

as responses to possible terrorist attack; as such, it may be a

used as a tool to assist in field evaluations of proposed changes in licensee safeguards made in response to the proposed Physical Protection Upgrade Rule; 3) the communicated threat credibility study provides advice to the DOE, the NRC, the FBI, and other appropriate agencies during an actual or perceived emergency relating to nuclear extortion threats; 4) a project entitled " Insider Crime Analogous to the Potential Threat to Nuclear Programs" is expected to aid in the formulation of prudent standards and regulations by NRC; 5) the results of the trans-portation safeguards evaluation methodology research program has served as a basis for evaluating some aspects of the guard levels

.-' III-1-a (cont.)

required for an in-transit physical protection system, as established in the Physical Protection Upgrade Rule; 6) the research program for investigating the vulnerabilities of spent LWR fuel shipping containers is required to assist in the formu-lation of policy concerning safeguarding shipments of spent LWR fuel and as a basis for supporting development of safeguards regulations which may be required for protection of sper.r fuel

~

shipments.

f O

e

Question - Mr. Dircks (Director, Office of fluclear Material Safety andSafeguards):

III-1-b.

To what extent is the safeguards research conducted by the Office of Nuclear Regulatory Research geared toward facilities and activities related to the use s

of recycled plutonium as a reactor fuel?

Answer:

Following the Presidential announcement of April 7, 1977, to defer indefinitely U.S. commercial reprocessing and recycling of plutonium, the safeguards research work that was related directly to recycle (mainly reprocessing pla t_ safeguards).,

was terminated, except for documenting the results of work done to that point.

The current safeguards research program applies to existing programs and does not assu.e that the use.-

of plutonium will be widespread. Some of the results of current research -projects,-although -notsdirected toward

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plutonium recycle, could be applied, in whole or in part,- to plutonium programs. The Administration's policy of defer ~ ring reprocessing and recycle clearly means that there will be no widespread use of plutonium within the U.S. during an indefinite period.

III-1-c Question - Mr. Dircks (Director, Office of Nuclear Material Safety and Safeguards):

III-1-c.

To what extent does the safeguards research con-ducted by the Office of Nuclear Regulatory Research apply to existing fuel cycle facilities and activities?

Answer:

The safeguards research program endorsed by NMSS is expected to be applicable to existing fuel cycle facilities and activities.

These projects, identified in response to question III-1-a above, will be monitored by NMSS at appropriate checkpoints for de-termining that the work being performed continues to be relevant to our program.

In similar fashion, t&GS will review the FY80 program, endorse those portions relevant to our needs, and monitor performance. As the products of research projects are completed,.they are subjected to. technology; transfer.ctesting sse by NMSS to establish those which demonstrate appl.icabil.ity;to_--

existing fuel cycle facilities and activities.

Question - Mr. Dircks (Director, Office of Nuclear Material Safety and Safeguards):

III-1-d.

Since the NRC was established what regulatory changes in safeguards regulations have been instituted as the result of research sponsored by the Office of Nuclear Regulatory Research?

Answer:

By its nature, research has a long time constant from project initiation to adoption of results.

Consequently, none of the changes in safeguards regulations which have been prepared since NRC was established until now have been insti,tuted as a result of research sponsored by the Office of Nuclear Regulatory Research. However, it should be noted that the safeguards research program is expected to produce products -in the near - --

future which will be beneficial in assisting with the formula-tion of regulations and-other aspects of the regulatory program. --~

Specific examples which have been discussed in answering-question-III-1-a include:

1) " Effectiveness Evaluation Methods for Material Control and Accounting;" 2) Safeguards Network Analysis Program (SHAP); 3) Transportation Safeguards Methodology Research Program; 4) the program for investigating the vulnerabilities of spent LWR fuel shipping containers.

III-1-e Question - Mr. Dircks (Director, Office of Nuclear heterial Safety and Safeguards):

III-1-e.

What would be the impact of a 50% reduction in the total amount requested for safeguards research and technical assistance contracts? Would this ' lead to a diminution in sscurity in the domestic nuclear' industry? If so, in what way?

Answer:

Should the FY80 funding requested by NMSS to fund technical assistance contracts be reduced by 50%, the following projects would not be supported:

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- A project to develop and demonstrate the use of process monitoring techniques for timely detection of material losses;

- A contract to support the use of a developed and accepted method. (Matrix Analysis of Insider Threat) for evaluation -

of collusion possibilities;

- Development of the Improved Safeguards Information System (ISIS), which is to Ecovide NRC with more accurate and timely information needed to monitor and audit SNM transactions; A project to develop improved-computer simulation to permit modeling of licensee ~ material. flows: and measurement 17 systems in order to gain better understanding of sources ---

of Inventory Differences and the validity of licensee Limit of Error calculations;

- Contracts to provide technical training support to NRC staff in the areas of security lighting evaluation, closed-circuit television (CCTV) systems evaluation, and evaluation of other security sensors and alarm systems; A contract to investigate how physical security and-material control and accounting performance requirements can be integrated so as to improve licensee flexibility and initiative.

2 Should the funding requested be reduced by 50% in FY 80, the following research programs proposed in support of t&tSS requirements would not be supported:

- Evolving research projects whose task definition has not yet been finalized.

- A project investigating the technological feasibility and safeguards benefit to be derived from applying thermite SSNM use denial technology to licensee activities involving transport of sensitive quanti-ties of SSNM.

- A project to develop methods for operationally evaluating the' effectiveness of physical protection----

of special nuclear material (SNM) in transit.

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- A project to develop a multi-discipline capability for comprehensive assessment of nuclear extortion threats.

A 50% red'uction in the total amount requested for -safeguards --

research contracts would not. be. expected to have an immediate effect on current-domestic-security. - However,-over-the-~~-

long term-the ' impact of such a reduction-could-jeopardize ~%

future security -by curtailing our ability -to remain abreast ~~~

of developing technologies, techniques, systems and method-ologies in the field of safeguards. Technical assistance contracts, on the other hand, are directed toward more immedi ate requi rements ;. consequently,. a..curtailt.ent.. of_. funds in this area would be expected to show a more direct impact.

Moreover, the impact of a 50% reduction in the total amount requested in FY 80 for safeguards research and technical assistance could be expected to diminish the NRC's future capability to assess safeguards effectiveness, to identify the most promising approaches for improved effectiveness, and to perform value impact analysis of proposed changes.

In addition, it could be expected to slow the development of improved regulations.

III-1-e (cont.)

• Public confidence in the itRC regulatory program and the adequacy and independent confirmation of its information base is an ingredient of the domestic nuclear industry's security. This confidence could be adversely affected by such a significant reduction in the safeguards research and technical assistance contracts.

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e a

III-1-f Question - Mr. Dircks (Director, Office of Nuclear Material Safety and Safeguards):

III-1-f, In the NRC's proposed FY80 budget, what is the total arount requested for safeguards research and technical assistance contracts?

Answer:

FY80 BUDGET (Dollars in thousands)

Safeguards Research 5,100

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Safeguards Technical Assistance SD 3,100 -

Nb5S 2,530 NRR 970 IE 1,490 ADM 1,120 9,210 Functionally, the budgeted FY 80 NRC safeguards research and'"

technical assistance contractual program is apportioned as follows:

57% - Routine Safeguards Regulation - This category of activities includes those actions taken to regulate safeguards of licensed nuclear reactors, fuel cycle facilities, and the transportation of special nuclear material.

30% - Nuclear Explosive Device Safeguards Program Developr.ent -

This category inc_ludes those actions required to assess the need for changes in physical security and material control and accounting regulations, and to evaluate the changes, in the interest of deterring, protecting against, detecting and responding to the thef t of materials which

e III-1-f(cont.)

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_2_

can result in a nucle 2r explosive device.

About 1%

of this is to evaluate and establish safeguards for alternate fuel cycles and to review and evaluate the need for integrated safeguards regulations.

9% - Radiological Safeguards Program Development - This category includes those actions required-'to' assess'the -' -

need Rn changes in physical security regulations, and to evaluate the changes, with respect to safeguarding against theft of potentially radiologically hazardous-quantities of materials and against sabotage-of activities involving such materials.7:

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1% - Safeguards" Assurance

'This cat'egoryrincWdes contractual.-

actions needed to assure the-Commission that' safeguards' requirements are met and obtain independent assurance of the effectiveness of licensees' safeguards programs through an f1RC inspection and enforcement program.

3% - International Safeguards

-This category includes actions -

required to implement the US/IAEA safeguards agreement and to provide assistance to strengthen international safeguards.