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NRC Staff Response to Orange County First Set of Discovery Requests to NRC Staff.* Staff Is Now Voluntarily Providing Responses to Orange County'S Request for Production of Documents.With Certificate of Svc.Related Correspondence
	ML20212L144
Person / Time
Site:	Harris
Issue date:	10/05/1999
From:	Uttal S; NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:	; AFFILIATION NOT ASSIGNED
References
	CON-#499-20871 99-762-02-LA, LA, NUDOCS 9910070173
	Download: ML20212L144 (101)
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WTED CORRespCNDENCE l

' DOCKETED I qgneg i October 5,1999 9 OCT -6 All :07 1

UNITED STATES OF AMERICA Oi, J NUCLEAR REGULATORY COMMISSION A'i BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of .) ,.,.

) Docket No. 50-400-LA CAROLINA POWER & LIGHT )

COMPANY ) ASLBP No. 99-762-02-LA j

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(Shearon Harris Nuclear Power Plant) )

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NRC STAFF'S RESPONSE TO ORANGE COUNTY'S FIRST SET OF l DISCOVERY REOUESTS TO NRC STAFF j 1

The Nuclear Regulatory Commission staff (Staff) hereby responds to Orange County's First Set of Discovery Requests to NRC Staff, filed September 20,1999.

The Staff notes that 10 C.F.R. il 2.744 and 2.790, which govern the production of NRC records and documents, contemplate that most NRC docums nts will be available for inspection and copying L the public document room, and,if they have been withheld from the pubF- document room pursuant to i 2.790, a request to the Executive Director for Operations for the production of such a document is required by 6 2.744, which must state, among other things, why the requested record or document is relevant to the proceeding.

Notwithstanding these regulations, and in accordance with a September 23,1999 agreement of counsel, without waiving any objections or privileges, and except as specified below, the Staff is now voluntarily providing responses to Orange County's request for production of documents. In doing so, the Staff is not waiving its right to require full 9910070173 991005 gDR ADOCK 0Lc00400 h

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l 4-l l .' 2-compliance with the Commission's regulations regarding any future discovery requests made by Orange County in this matter.

i I. GENERALOBJECTIONS

1. The Staf objects to Orange County's discovery requests to the extent that they call for disclosure oflit gation strategy and other material protected under 10 C.F.R. I 2.740 or 8

other protection provided by law, attorney work prodr.ct, privileged attorney-client materials, and other privileged materials such as draft agency documents protected by executive privilege.

2. The Staff objects to Orange County's discovery requests to the extent that they request information or documents relating to licensees and/or entities other than Carolina Power & Light's Shearon Harris Nuclear Plant. Such discovery requests call for information which is irrelevant, immaterial, and not calculated to lead to the discovery of admissible evidence, and are over-broad and unduly burdensome.

3. The Staff objects to Orange County's discovery requests to the extent that they require identification of the home addresses and telephone numbers of Staff employees or contractors, which are protected from disclosure by the Privacy Act,5 U.S.C i 552a(b) and 10 C.F.R. I 2.790(a)(6). The disclosure of such information is irrelevant and unnecessary.

4. The Staff objects to Orange County's discovery requests to the extent that they seek discovery which is beyond the scope of the two contentions admitted by the Board in this proceeding. Orange County is only permitted to obtain discovery of matters that pertain to the subject matter within the scope of this proceeding.

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i II. GENEPAL DISCOVERY REQUESTS A. GENERALINTERROGATORIES Pursuant to agreement between Orange County and the Staff, l these general interrogatories apply to both Orange County .

! admitted contentions; are in addition to thp fifteen interrogatories per contention allowed by the Board's July 29, 1999, Memorandum and Order; and are continuing in i accordance with 10 CFR i 2.740(e) through the end of the l discovery period, October 31, 1999, as established in the Board's July 29,1999 Memorandum and Order.

GENERALINTERROGATORY NO.1. State the name, business address, and job title of each person who j l supplied information for responding to these interrogatories, l requests for admission, and requests for the production of l

documents. Specifically note for which interrogatories and requests for admissions each such person supplied i

. information. For requests for production, note for which 1 contention each such person supplied information.

STAFF'S RESPONSE: The following persons supplied information for responding l to Orange County's First Discovery requests:

Richard Laufer ,

Project Manager, Shearon Harris Nuclear Power Plant .

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commissb Washington, D.C. 20555 Document Requests for both contentions Lawrence Kopp Senior Reactor Engineer Reactor Systems Branch, Division of Systems Safety Analysis Of5ce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

s-Document Request for contention 2 Xenneth C. Heck j Quality Operations Engineer .j Office of Nuclear Reactor Regulation 1 U.S. Nuclear Regulatory Commission I Washington,D.C. 20555

- Document request for contention 3 ,.

Donald G. Naujock Technical Reviewer Materials and Chemical Engineering Branch, Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555' Document Request for contention 3 James A. Davis Technical Reviewer, l Materials and Chemical Engineering Branch, Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Document request for contention 3 GENERAL INTERROGATORY NO. 2. For each admitted Orange County contention, give the name, address, ,

profession, employer, area of professional expertise, and l educational and scientific experience of each person whom l the NRC Staff expects to provide sworn affidavits and l declarations in the written filing for the Subpart K proceeding described in the Board's July 29,1999, Memorandum and i Order and the general subject matter on which each person is ' .

l expected to provide sworn affidavits and declarations for the l written filing. For purposes of answering this interrogatory, '

the educational and scientific experience of expected affiants and declarants may be provided by a resume of the person attached to the response.

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f l The Staff has not yet made a final determination regarding who will provide sworn affidavits, but provides the following as persons who are likely to provide a sworn affidavit or declaration in the written filing for the subpart K proceeding:

, Richard Laufer Project Manager, Shearon Harris Nuclear Power Pfant l

General subject matter: The overall project Lawrence Kopp Senior Reactor Engineer General subject matter: Contention 2 Kenneth C. Heck Quality Operations Engineer General subject matter: Contention 3 Donald G. Naujock Technical Reviewer General subject matter: Contention 3

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James A. Davis Technical Reviewer General subject matter: Contention 3 Anthony Ulses Nuclear Engineer Reactor Systems Branch, Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation l U.S. Nuclear Regulatory Commission Washington,D.C. 20555 ;

l General subject matter: Contention 2 l l A copy of the resume of each person named in this answer is annexed hereto as l

attachment 1. The Staff reserves the right to amend this answer as discovery continues, j GENERAL INTERROGATORY NO. 3. For each admitted Orange County contention, identify each expert on whom the NRC Staff intends to rely on in its written filing l

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for the Subpart K proceeding described in the Board's July 29,1999 Memorandum and Order, the general subject matter on which each expert is expected to provide sworn affidavits and declarations for the written filing, the qualifications of each expert whom the NRC Staff expects to provide swom affidavits and declarations for the written filing, a list of all publications authored by the expert within the preceding ten years, and a listing of any other cases in which the expert has testified as an expert at a trial, hearing or by deposition within the preceding four years.

Lawrence Kopp List of publications is contained in attached resume General subject matter: Contention 2; criticality Kenneth C. Heck General subject matter: Contention 3 l Donald G. Naujock Publication listed in attached resume.

General subject matter: Contention 3 James A Davis List of publications attached.

General subject matter: Contention 3 The Staff reserves the right to amend this answer as discovery continues.

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i B. GENERAL DOCUMENT REQUESTS

. The County requests the Staff to produce the following documents directly or indirectly within its possession, custody or control.

REOUEST NO 1. All documents in your possession, custody or control that are identified, referred to or used in any way in responding to all of the aboys general interrogatories and the following interrogatories and requests for admissions relating to specific contentions.

STAFF'S RESPONSE: No documents, other than the attached resumes, were used in answering the general interrogatories.

EEQUEST NO. 2. All documents in your possession, custody or control relevant to each Orange County admitted ;

contention, and .to the extent possible, segregated by contention and separated from already produced documents.

STAFF RESPONSE: All available, non-objectionable, responsive documents that are relevant to the two contentions, as admitted into the proceeding by the Board, which are

- not in the Public Document Room (PDR) or have not been previously produced will be provided in response to this request either with this document or within 30 days of the date of Ore.nge County's first Discovery Request to the NRC Staff. Documents which are i generally available to the public will not be produced. A list of responsive documents served together with this response is annexed hereto as attachment 2. The Staff reserves the l

right to amend this answer as discovery continues.

REOUEST NO. 3. All documents (including experts' opinions, workpapers, affidavits, and other materials used to render such opinion) supporting or otherwise relating to 1

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testimony or evidence that you intend to use in your Subpan K presentation and/or the hearing on each Orange County admitted contention.

STAFF RESPONSE: The Staff objects to this document request as being overly broad, unduly burdensome, seeking pre-decisional, trial preparati'on or privileged material or material exempted from disclosure by 10 C.F.R. Il 2.N.; and 2.790. Without waiving these objections, any available, non-objectionable documents that are relevant to the two contentions, as admitted into the proceeding by the Board, which are not in the Public Document Room (PDR) or have not been previously produced will be provided in response l

to this request within 30 days of the date of Orange County's first Discovery Request to the NRC Staff. The Staff reserves the right to amend this answer as discovery continues. !

IV. SPECIFIC DOCUMENT REQUESTS Subject to the limitations specified on page 1 of this document, the Staff will respond to Orange County's specific document requests within 30 days of receipt of Orange County's First Discovery Requests.

Respectfull sub dtted,

/-

Susan L. ttal l

Counsel for NRC staff

. Dated at Rockville, Maryland this 5* day of October 1999 l

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!t RESUME James A. Davis Security Clearance: O (Top Secret)

Education:

High School Worthington High School, Worthington, Ohio College B. Met.E., The Ohio State University,1965, Metallurgical Engineering M.S., The Ohio State University,1965, Metallurgical Engineeting Ph.D, The Ohio State University,1968, Metallurgical Engineering 39/45 Credits towards an MBA, Canisius College, 1972-1976.

Experience:

11/11/1990 to Present: U.S. Nuclear Regulatory Commission, GG-14-10 i Supervisor: E. Sullivan !

Mailing Address: U.S. NRC Mail Stop 07-D4 ',

Washington, D.C. 20555 l

l am a technical reviewer in the Chemical Engineering and Metallurgy Section of the Division of Engineering, the Office of Nuclear Reactor Regulation. Areas of responsibility include coatings for nuclear power plants, license renewal for Calvert Cliffs and Oconee, all threaded fastener issues (such as stress corrosion cracking, boric acid corrosion, and fatigue), chemical decontamination, Boiling Water Reactor intemals cracking, pump and valve internals cracking, pipe integrity issues, and corrosion behavior for dry cask storage, and interaction of coatings with spent fuel water, Currently, I am coordinating the responses to a generic letter on containment coatings for nuclear power plants. I am the NRC representative to ASTM D-33 on coatings for power generation facilities. I am also a member of the Board of Directors for the National Board of Registration for Nuclear Safety Related Coating Engineers & Specialists. I am also a member of ASME B1 on threaded fasteners. I was a member of an Augmented Inspection Team at Palisades on fuel handling problems and Point Beach on the hydrogen burn as a result of interactions between borated water and the inorganic Zinc coating during dry cask loading operations. I was Contract Technical Monitor and Project Officer for numerous contracts at Brookhaven National Labs. I was a technical reviewer for the design of the Navy Seawolf Submarine and on the DOE project to produce tritium in a commercial reactor (Watts Bar). I was acting section chief on numerous occasions, and for several months at a time. I have made numerous presentations to senior NRC management including the Chairman, the Executive Director for Operations, the Committee to Resolve Generic issues, and the Advisory Committee on Reactor Safety and Safeguards. I testified before Representative Dingle's staff on the safety of fasteners in nuclear power plants as a result of concerns raised by a private citizen. I convinced his staff that there is no safety issue because of the redundant design of mechanicaljoints, the fact that the joints will leak before they break, and that the joints are inspected every refueling outage.

w 8/1981 to 4/1990: Polyken Technologies /Kendall Co., Senior Research Associate Supervisor: Jordan Kellr.,r Address: Polyken Technologies /Kendall Co.

17 Hartwell Avenue Lexington, MA 02173 I was responsible for domestic and intemational technical marketing for Polyken Pipeline Coatings.- As part of my job, I made technical presentations on Polyken Pipeline Coatings in North America, USSR, Egypt, India, Iraq, Japan, Australia, Bolivia, France, England, Germany, Czechoslovakia, Italy, Switzerland, Algeria, Singapore, and Jakarta to ministries and high ievel government officials. I coordinated joint research between Polyken and the VNilGAS and VNilST Technical Institutes in Moscow on the development of high temperature pipeline coatings. I contracted with independent laboratories to certify Polyken products for international customers. I was the Polyken representative to National Association of Corrosion Engineers (NACE) committees on Underground Pipeline Coatings,- Arctic Corrosim, and Cathodic Protection. I was appointed by the President of National Association of Corrosion Engineers to the international Relations Committee. Also, I was the company representative to American Society of Testing and Materials and the American Water Works Association technical committees. I was responsible for analyzing competitive coatings using fast Fourier transform infrared spectroscopy, gas chromatography, mass spectroscopy, scanning electron microscopy, differential scanning calorimetry, thermal gravimetric analysis, moisture vapor transmission apparatus, and mechanical test equipment. I conducted slow strain rate stress corrosion cracking tests on line pipe steel to develop inhibitors that could be added to coating primers to control stress corrosion cracking. I conducted slow strain rate tests on various blends of polypropoline to develop blends that are resistant to environmental cracking. I was acting section chief for extended periods of time.

11/1979 to 8/1981: Arthur D. Little, Senior Consultant Supervisor: William Lee Address: Artt,ur D. Little i 15 Acorn Park Cambridge, MA 02138 I was a consultant to DOE on the metallurgical and mechanical condition of defense nuclear waste tanks that were damaged during stress relieving and that were damaged by sulfate reducing bacteria. This included a review of the structuralintegrity of the waste tanks and a review of DuPont's program to control stress corrosion cracking of the tanks. I was a consultant to DOE on the Savannah River Defense Waste Form program.

Twelve contractors were developing defense waste forms including borosilicate glass,

high silica glass, synrock, coated borosilicate glass, HIP rock, and stoichiometric concrete. A pilot plant borosilicate glass facility was constructed at Savannah River. My job was to visit with each contractor twice a year and review their programs. I then reviewed progress reports and prepared an assessment of each contractor's work and recommended that individual programs be expanded, contracted, or canceled. I was also a member of a team that developed models for long term storage of commercial nuclear waste sponsored by Battelle. I consulted to numerous commercial customers on corrosion, fracture mechanics, coating, metallurgical, and plating issues.

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t. 11/197e.,,/1979: Allied Tube and Conduit Corp., Director of Research Supervisor: L. Volmuth Address: Allied Tube and Conduit 16100 South Lathrop Harvey, IL 60426 I was responsible for research and development in the areas of metallurgical tube forming, low frequency electric resistance welding, chemical cleaning of steel, high speed in-line galvanizing, surface treating, and coating of electrical conduit, fence posts, and specialty tubing. I was also responsible for the Quality Control Department and the Process Control Laboratory.

6/1976-11/1978: Allegheny Ludlum Steel Corp., Research Specialist-Supervisor: George Aggen Address: Allegheny Ludlum Steel Corp.

Research Center Brackenridge, PA 15014 1 determined the level of columbium and vanadium stabilizers required to avoid sensitization of ferritic stainless steels. I developed intergranular sensitization tests for low chromium ferritic stainless steels. I used electrochemical and electron-optical including scanning electron microscopy, auger, and transmission electron microscopy techniques for failure analyses. I used sensitive analytical techniques to determine carbon, nitrogen, columbium, and titanium contents of stainless steel. I provided customer service by recommending specific grades of stainless steel for corrosive applications.

11/1970-6/1976: Bell Aerospace: Senior Research Scientist Supervisor: A. Watts Address: Bell Aerospace Wheatfield, NY l was program Manager on numerous Navy sponsored programs involving the corrosion of aluminum alloys, stainless steels, and titanium alloys in high velocity sea water for the Navy's high performance ships program. I examined the influence of mean stress intensity and stress intensity amplitude on the corrosion fatigue behavior of aluminum alloys and titanium alloys in sea water using fracture mechanics specimens. I determined the comparability of various materials with j cooling fluids for the hypersonic airplane. I examined the comparability of rocket fuels and oxidizers with fuel handling equipment. I managed the fracture mechanics group. I developed microelectrodes for measuring potential and pH inside of ,

growing stress corrosion cracks of aluminum alloys and alloy steels. !

1/1968-11/1970: U.S. Steel Corporation, Senior Research Engineer Supervisor: Brian Wilde Address: U.S. Steel Corp l

l Jamison Lane Monroeville, PA 1 developed steel with improved corrosion resistance using linear polarization, anodic l

I polarization, transmission electron microscopy, and scanning electron microscopy. l .

I conducted fundamental studies on the mechanism of pitting, stress corrosion cracking, I hydrogen embrittlement, and intergranular corrosion uslag electrochemical techniques, ,

static and dynamic straining techniques, hydrogen permeation cells, and optical and

scanning microscopy techniques. I was the group leader of the electrochemistry group.

I NRC Awards:

Performance Award - December 21,1994 l Certificate of Appreciation - February 1,1995 - Seawolf Class Submarine design review Group Award - December 10,1996 - AIT Team at Point Beach i Pe:Tormance Award - January 10,1997 NRC Training: / ,

Power Plant Engineering (2 week course) l Boiling Water Reactors-General Electric Design (1 week course)

Pressurized Water Reactors-Westinghouse Design (1 week course)

Non Destructive Testing (2 week course)

Sexual Harassment Awareness (2 day course)

AIDS Awareness (1 day course) i Allegation Training (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months annually)

Procurement Training (1 day) l Ethics Training (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months annually)

Security Training for the Seawolf Design Review (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months )

Security Training for the DOE tritium Project (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) !

NUDOCS (NRC document retrieval system) (1 day) dBase ill (2 days)

Autos LAN Training (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) !

Introduction to Netscape

. Introduction to Microsoft NT !

Introduction to Wordperfect 8.1 ,

Other Relevant Training Public Speaking-Kendall-3 Days, Allegheny Ludlum-5 days Effective Writing-U.S. Steel (3 days)

PUBLICATIONS-PRESENTATIONS 1 J. A. Davis and J. D. Kellner, " Electrochemical Principais Applied to Operating Pipelines," Presented at Corr /89, Underground Corrosion Symposium, New Orleans, Louisiana, April 17-21,1989

2. F. J. Witt, J. A. Davis and R. A. Hermann,"Second EPRI Balance-of Plant Heat Exchanger NDE Workshop, The U. S. Nuclear Regulatory Commission Perspective," Second EPRI Balance-of-Plant Heat Exchange NDE Workshop, Key West, Florida, May 26-29,1992

3. J. A. Davis, " Full Reactor Coolant System Chemical Decontamination, NRC Approval of the Westinghouse Owners Group Topical Report" Fifth EPRI

Workshop on Chemical Decontamination, Charlotte, North Carolina, June 8-9, ,

! 1993 I

4. James A. Davis and Richard E. Johnson, "The Regulatory Approach to Fastener i integrity in the Nuclear Industry," Symposium on Structural Integrity of Fasteners, Sept. 8-19,1993, to be published as a Special Technical Publication.

5. James A. Davis, " Nuclear Power Plant Service Water Systems, NRC Staff l Perspective," international Joint Power Generation Conference, ASME Power Division, Heat Exchange Committee, Kansas City, Mo., October 17-21,1993

6. J. A. Davis, G. P. Hornseth, and R. A. Hermann, " Third EPRI Balance of Plant Heat Exchanges NDE Workshop, A Regulatory Perspective," Third EPRI Balance of l Plant Heat Exchange Workshop, Myrtle Beach, SC., June 6-8,1994 '

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7. R. A. Hermann, M. Banic, and J. A. Davis, " Primary Water Stress Corrosion l Cracking of Alloy 600," Specialists Meeting on Cracking in LWR RPV Head Penetrations, International Atomic Energy Agency, Philadelphia, PA, May 2-4, ;

1995 ;

8. Robert A. Hermann, James A. Davis, and Merrilee J. Banic, " Age Related Degradation in Operating Nuclear Plants," presented at ASME International Vessel '

and Piping Conference, Honolulu, Hawaii, July,1995 l

9. James A. Davis, "A Regulatory Prespective on Service Water Problems," )

Invited Lead Speaker, Fourth EPRI Balance-of-Plant Heat Exchanger NDE !

Symposium, June 10-12,1996, Jackson Hole, Wyoming. !

10. James A. Davis, " Nuclear Power Plant Fastener Thread Gaging - NRC Staff j Perspective," presented at Eight International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 10-14,1997, Amelia Island, Florida.

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11. James A. Davis,"A Regulatory Prespective on Service Water Problems,"

Invited Lead Speaker, Fifth EPRI Balance-of-Plarit Heat Exchanger NDE Symposium, June 15-17,1998, Lake Tahoe, Nevada.

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4 Anthony P. Ulses

"; 301 415 1194 apu@nrc. gov Education N

Master's of Science (MS) in Nuclear Engineering University of Maryland at College Park, May 1999.

Bachelor's of Science and Engineenng (BSE)in Nuclear Engineering, Cum laude University of Michigan at Ann Arbor, August 1992.

. Experience United States Nuclear Regulatory Conunission 5/93-present Nuclear Engineer, Reactor Systems Branch, Division of Systems Safety and Analysis United States Nuclear Regulatory Conunission 9/92 5/93 fgglggte Fellow. Advanced Reactors Project Directorate Computer Code Development

' Maintained and upgraded legacy physics codes on UNIX workstations -

Developed VIKTORIA code for fuel channel analysis

. Coupled TRAC and NESTLE codes

: Assisted in NEWT development I Computer Code Analysis -

SCALE- Fuel lattice criticality studies, depletion and collapsed cross section preparation e MCBEND Reactor pressure vessel fluence studies

NEWT ' Fuel power distributions, depletion and collapsed cross sections

. . DOORS 3.2 _ 3D Transport Calculations

TRAC / NESTLE. 3D BWR A1WS Studies

.- DRAGON 3.2 ~ Fuel power distributions, depletion and collapsed omss sections

. Computer Codes

~ DANfSYS 3.1, MONK Computer Languages and Operating Systems

UNIX, Fortran 90/95, Fortran 77, C, Windows, DOS, PVM Regulato y Esperience:

License amendment evaluations

'. Fuelmanufacturerinspections

Performing Audit Calculations of Licensee Analyses GeneralExperience:.

Mr.naging High Performance Cornputer Networks

Digital UNIX System's Administration l

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h LCurence 1. Kopp i Senior Reactor Engineer Ec.ucation Ph.D., Nuclear Engineering, University of Maryland,1968 M.S., Physics, Stevens institute of Technology,1959 B.S., Physics, Fairleigh Dickinson College,1956.

Employment U.S. Nuclear Regulatory Commission, Senior Reactor Engineer,1965 - present Performs safety evaluations of reactor license applications, technical. specifications, core reloads,' spent fuel storage facilities, and topical reports. Developed regulatory guides, information notices, generic letters, rulemaking related to reactor physics, safety ankysis, and fuel storage. Assisted in development of improved technical specifications in areas of reactivity control, power distribution limits, and fuel storage.

Westinghouse Astronuclear Laboratory, Senior Scientist,19631965 Evaluated nuclear analytical methods to be used in the design of NERVA rocket reactors.

- Analyzed experiments performed in the Los Alamos zero power reactor.

Martin-Marietta Nuclear Division, Senior Engineer,19591963 Performed core physics calculations on fluidized bed and PM-1 reactors. Performed parametric studies of reactors applicable to nuclear rocket applications. Programmed several

~ FORTRAN computer codes.

Federal Electric Corporation, Senior Programmer,1957-1959 Curtiss-Wright Research Division, Programmer / physicist,19561957 Professional Societies American Nuclear Society ANS-10 Mathematics and Computations Standards Committee ANSI N-17 Standards Committee on Research Reactors, Reactor Physics & Radiation Shielding Publications "The NRC Activities Concerning Boraflex Use in Spent-Fuel Storage Racks," invited paper, American Nuclear Society Annual Meeting, June 1996.

" Potential Loss of Required Shutdown Margin During Refueling Operations," invited paper, American Nuclear Society Annual Meeting, June 1990.

" Recommended Programming Practices to Facilitate the Portability of Scientific Computer Programs," ANS Proceedings of the Topical Meeting on Computational Methods in Nuclear Engineering, April 1979.

l "The N:utron R:sonance Integrcl of N turci Dysprosium," Ph.D. th: sis, University of Mtrylind, 4

1968.

" Pool Reactor Experiments with Contrc! Rods," Transactions of the American Nuclear Society, Vol.10, Pg.16,1967 (co author).

" Procedures for Obtaining Few-Group Constants for Systems Having Rapid Spectral Variation With Position," Transactions of the American Nuclear Society, Vol. 8, pg. 303,1965 (co-author).

lmproved Nuclear Design Method for NERVA Calculations - NSDM 11, WANL-TME-1091, Westinghouse Astronuclear Laboratory,1965 (co-author).

" Analysis of Experiments Performed in Los Alamos 75PO Reactor' WANL TME 273 Westinghouse Astronuclear Laboratory,1963.

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9. ' KENNETH C. HECK -

735 University Avenue Sewanee, Tennessee 37383 -

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Tel:(301) 415-2682 email: kehignrc. gov

SUMMARY

OF SKILLS Technical, supervisory, arid management experience in the electric power industry and with assignments in engineering, project I management, project engineering, plant start-up, plant and program evaluation, quality assurance, and licensing. ,

Proficiencies include design engineering, control systems, electronics, accounting, and computer applications.

EXPERIENCE

- i Nuclear Renulatory Commisalon (May 1997 - Present) ' l Quality Operations Engineer (Headquarters), inspection Program Management

. Review, evaluate, audrt quality assurance programs and other administrative control aspects for nuclear power plants. ;

e' Perform program development functions related to all aspects of the agency's quality assurance programs. I e Conduct inspections of vendors who provide products and services to the nuclear industry.

Tennessee Vallev Authority Lead Auditor, Quality Services (June 1995-October 1996)

Provided staff augmentation services in the areas of quality assurance and licensing. !

- Developed audit / consultation services for implementing intomational(180-9000) quality standards.

Principal Evalustor, Nuclear Assurance & Licensing (October 1988 June 1995) - l e Conducted independent audits / evaluations of nuclear power programs, processes, and plant events. ;

e Served as Technical Secretary for the Nuclear Safety Review Board (senior safety oversight body) from shutdown of l TVA's nuclear program through recovery of the Sequoyah and Browns Ferry nuclear plants.

. . Conducted independent verifications of the effectiveness of completed corrected actions through successful startup of the Watts Bar nuclear plant.

Senior Evalustor, Nuclear Managers Review Group (March 1987 to October 1988) e Developed and implemented a review program to assess activities associated with the design, conetruction and ,

operation of TVA nuclear plants. Findings were reported directly to the Manager, Nuclear Power with recommendations !

i for improvements.

independent Contractor (December 1985-March 1987)

Design Engineer / System Engineer, Engineering Department

. Modified the integrated control system and non-nuclear instrumentation following shutdown of the Davis Besse nuclear plant.

- Developed engineering designs, implemented modifications, and tested control systems at power through successful program recovery.

Sabcock & Wilcox (March 1970-November 1985)

Project Engineer, Plant Services (September 1984-November 1985) e Developed and deployed hardware and inspection services for repair and maintenance of steam generators and

.. pressure vessels.

e- ' Managed field installation of fuel handling bridge in Kumatori, Japan.

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Project Manager, international Business (June 1982-September 1984)

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Developed markets for B&Wtechnology services in Europe and the Pacific Basin in partnership with intemational companies such as Brown Boveri(Germany), Framatome (France), Sumitomo (Japan) and McDermott intemational (Hong Kong).

Principal Engineer, Plant Performance (January 1980-June 1982) e Supervised 9 member team developing operator guidelines for ariticipated reactor transients.

Specialized in original control system analysis and design, principal accomplishments including:

e Developed course on plant control systems, a Consulted onsite on steam generator performance problems, a Completed operational / accident transient analyses for several nuclear contracts, a Performed failure modes and effects analysis for the integrated reactor, control system,

. m Extended methods for reactor power determination, s Developed original analyses and conceptual control schemes for steam generator overfill, water hammer transients, anticipated transients without reactor scram, two-phase natural circulation cooling, and reactor vessel embrittlement.

Technical Advisor, Plant Design (January 1976-December 1980,)

e On loan to Brown Boveri, Germany, through licensing of the reactor safety systems for the Muehlheim Kaartich i nuclear plant, to consult on technical licensing issues and oversee the development of complex, nonproprietary computer codes for reactor safety analyses.

Senior Engineer Technical Staff (March 1970-January 1976) e Applied intemal and industry research to nuclear plant design, provided technical assistance to the engineering department, and developed computer codes licensed for performing transient thermal-hydraulic analyses.

= On loan to Duke Power as test engineer during hot functional testing at Oconee nuclear power station.

EDUCATION Master of Science / Bachelor of Science, Mechanical Engineering; Lehigh University Master of Engineering Administration; George Washington University Eschelor of Applied Accounting; Tennessee Wesley College Associate of Computer Science; Chattanooga State Associate of Electronics; U.S. Naval Electronics School ,

l CERTIFICATIONS )

Registered Professional Engineer (#20668, VA); Certified Quality Systems Auditor,1S0-9000 (#005630); Certified Manager ;

(#02929); Toastmasters International (Able Toastmaster) l I

PROFESSIONAL ASSOCIATIONS ;

Amencan Society of Mechanical Engineers, American Nuclear Society, American Society for Quality Control, institute of l Electrical and Electronic Engineers l

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m.c w Richard J. Laufer Experience-2/99 - Present: NRC Project Manger- Shearon Harris Nuclear Power Plant o

Serve as the Headquarters Focal Point for information and Communication on all issues concoming the Shearon Harris Nuclear Power Plant. Maintain nearly daily communication with the licensee, the resident inspectors, and the regional staff. Participate in all significant licensee meetings in the region and on-site. Serve as Back-up Project Manager (PM) for another plant in the Project Directorate (currently H.B. Robinson),

e Prepare and coordinate the numerous documents generated to support the licensing activities of the assigned plant. These documents include license amendments and exemptions and their associated environmental assessments and Federal Register Notice, Task Interface Agreement Responses, controlled correspondence, and numerous letters to the licensee asso( M with closing out Generic Letters, relief requests, and requests for additionalinformation.

Coordinate, participate, and manage meetings and briefings by ensuring that the appropriate NRC

! contacts are informed, that meeting notices are prepared, and by preparing an accurate and concise meeting summary in a timely manner, 2/98 - 2/99: NRC Project Manager- Duane Amold Energy Center

-7/93 - 2/98: NRC Project Manager - Kewaunee Nuclear Power Plant

,2/93 - 7/93: NRC Project Engineer - Division of Reactor Projects 5/89 - 2/93: NRC Operator Licensing Examiner - Operator Licensing Branch

- Certified NRC Operator Licensing Examiner on Westinghouse pressurized water reactors and non-power reactors 3/86 - 5/89: Engineering Division Officer on Navy nuclear submarine USS Vallejo (SSBN 658)

(Qualified as Engineering Officer of the Watch, Engineering Duty Officer)

Trainino:

1/90 Completed NRC's Westinghouse Technology Full Series Course 5/84- 3/86: Navy nuclear power training Education:

5/84: B.S. Degree in Systems Engineering; U. S. Naval Academy, Annapolis, MD

RESUME Donald G. Naujock College Education:

BS, The University of Wisconsin-Milwaukee,1971, Material Science Engineering i MS, The University of Wisconsin-Milwaukee,1972, Metallurgical l

Experience:

August 1991 to Present: U.S. Nuclear Regulatory Commission, Mail Stop O-7D4, Washington, D.C. 20555 i e,.

Employed as a technical reviewer for the Materials and Chemical Engineering Branch of the Division of Engineering of the Office of Nuclear Reactor Regulation. My accomplishments are: j Assessed the Performance Demonstration Initiative (PDI) program developed by the nuclear ;

utilities for implementing Appendix Vill to Section XI of the American Society of Mechanical Engineers Code. Developed and coordinated the staffs effort to reference Appendix Vill in the final rule that was issued in the Federal Register on September 22,1999. Developed the l chemical ranges in Draft Regulatory Guide DG 1070 for verification of product check analysis l used in the commercial dedication process of steels. Prepared information Notice 92-60," Valve Stem Failure Caused By Embrittlement,

August 20,1992, and Proposed Generic Communication; " Effectiveness of Ultrasonic Testing Systems in Inservice inspection Programs," Federal Register, Volume 61, No. 252, Pages 69120 - 69124. Reviewed over 30 ASME Code cases for endorsement by the NRC; the Code cases covered subjects, such as, nondestructive testing, nondestructive techniques, inservice inspections, welding, and materials. Reviewed over 50 submittals requesting attematives to the 10 CFR 50.55a ,

paragraphs (c),(d),(e),(f),(g), and (h); the submittals covered subjects, such as, inspections of i reactor vessels using performance demonstrated ultrasonic techniques, use of wire penetrameters in radiography, changes in hydrostatic testing, inspection coverage of welds, inspections of control rod drive welds, ultrasunic testing qualification for intergranular stress-corrosion cracking, ultrasonic testing to determine water level in piping, ultrasonic testing in lieu of radiographic testing of dissimilar metal welds, and weld repairs / overlay of non-structural seal welds. Participated in over 10 ventor inspections to examine issues on securing test specimens i using electrical discharge machining, ultrasonic testing for cold cracks at cladding-to-basemetal interface, processing forged material, chemical analysis of commercial grade material, demonstration the phase array ultrasonic testing system, manufacture of small diameter fasteners, process and ultrasonic testing of zirconium alloy fuel rod assemblies, and heat treating of commercial dedicated material. Technical monitor of over 7 contracts with national laboratories on participating in the evaluation of ultrasonic techniques, assisting in reviewing public comments and topical reports, and developing specifications for a mobile nondestructive testing facility, Co-Authored, "U.S. Nuclear Regulatory Commission Perspective on Performance Demonstration of Ultrasonic Testing Systems," Presented at the 22* MPA-Seminar, " Safety and Reliability of Plant Technology," October 10 and 11,1996, University of Stuttgart, Germany.

I December 1988 to June 1991, Tennessee Valley Steel Corporation, Chief Metallurgist

l'.a- , _

f 1

My accomplishment was establishing a quality control facility and operation to the restoration of an abandon mini-steel mill. Responsible for all company quality control and customer service functions. Maintained steel tonnage records and both chemical and physical testing labs for mini-steel mill. Trained technicians to test cooling water pH and hardness, to operate optical emission spectrometer, to calculate furnace and ladie alloy additions, and to coordinate plant l electrical power demand with local utility. Wrote safe operating gocedures for equipment, refurbished test preparation equipment, and participated in acc. dent reviews. Resolved slag entrapment, porosity, chemical variations, and off dimensional billets. Solved customer processing and material selection problems and coordinated shipping, testing, and storage of hazardous furnace emission by-products.

March 1986 to August 1988, Tennessee Valley Authority, Metallurhi6al Engineer (Contractor from 3/86 to 8/87 and employee from 10/87 to 8/88)

My accomplishments are establishing the technical justification for selecting Type 347 modified

- stainless steel for nuclear piping applications, developing heat treatment and welding techniques for enhancing corrosion resistance and reducing manufacturing costs. Resolved material related employee concerns, fabrication inconsistencies, and intergranular stress corrosion cracking issues. Investigated the use of counterfeit fasteners, interfaced with vendors, and upgraded procurement procedures. Provided guidance in the application of ASME Codes and ASTM Specifications for construction personnel.

August 1982 to January 1986, Brush Wellman inc, Senior Manufacturing Engineer My accomplishments are a 50 % increase in induction molting capacity and a 50% increase in l

- new electric are furnace uptime. Coordinated $3.5 million dollar cast shop expansion; justified, :

~

selected, debugged, and evaluated performance of new induction fumaces, heat treat furnace, I planer mill, band saws, gun drill, and ventilation equipment. Redesigned tap hole configuration, i charge material handling bins, and molds for electric are fumace. Responsible for product l quality, productivity, and yield for beryllium-copper alloys processed through electric arc i fumaces, induction fumaces, direct chill casting machines, and billet conditioning equipment. l Wrote routings, job descriptions, process procedures, operating standards, and equipment specifications. Maintained variances and 5-year expansion plan. Investigated fumace failures and established preventive measures to reduce reoccurrences. ,

l l

4

E l

ATTACHMENT 2 LIST OF DOCUMENTS SERVED ON ORANGE COUNTY IN RESPONSE TO FIRST DISCOVERY REQUESTS TO THE NRC STAFF Contention 2 - Criticality

1. Letter from Brian Grimes, NRC, to ll Power Reactor Licensees, dated April 14, 1978.

2. Letter from Brian Grimes, NRC, to All Power Reactor Licensees, dated January 18, 1979, modifying letter of April 14,1978.

2. Draft 1, Regulatory Guide 1.13, Revision 2, " Spent Fuel Strorage Facility Design Basis," dated September 23,1981.

Contention 3

1. X MET, Portable XRF Analyzer brochure.

2. Request for Additional Information, dated December 23,1998, faxed copy

3. Request for Additional Information, dated March 24,1999.

4. Request for Additional Information, dated September 10,1999, faxed copy.

5. Request for Additional Information, dated April 21,1999, faxed copy.

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^ ENCLOSURE 2

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/ e UNIT E D ! T ATES

! *1 g, [,j tJUCLE AR REGULATORY COMMisslON

~ s' l W ASHINGTON ") C.205$5 j

..oril 14, 1978 l

To All Power Reactor Licensees Gentlemen:

Enclosed for your infonnation and possible future use,is the NRC l guidance on spent fuel pool modifications, entitled " Review and Acceptance of Spent Fuel Storage and Handling Applications". This document provides (1) additional guidance for the type and extent. i of information needed by the NRC Staff to perfonn the review of ,

licensee proposed modifications of an operating reactor spent fuel storage pool and (2) the acceptance criteria to be used by the NRC Staff in authorizing such modifications. This includes the information needed to make the findings called for by the Commission in the Federal Register Notice dated September 16,1975 (copy enclosed)

, with regard to authorization of fuel pool modifications prior to the ,

completion of the Generic Environmental Impact Statement, " Handling and Storage of Spent Fuel from Light Water Nuclear Power Reactors".

The overall design objectives of a fuel storage facility at a reactor complex are governed by various Regulatory Guides, the Standard j Review Plan (NUREG-75/087), and various industry standards. This i guidance provides a compilation in a single document of the pertinent -

portions of these applicable references tnat are needed in addressing spent fuel pool modifications. No additional regulatory requirements i are imposed or implied by this document.

Based on a review of license applications to date requesting authorization to increase spent fuel storage capacity, the staff has had to request ,

additional information that could have been included in an adequately documented initial submittal. If in the future you find it necessary to apply for authorization to modify onsite spent fuel storage capacity, the enclosed guidance provides the necessary information and acceptance criteria utilized by the NRC staff in evaluating these applications. Providing the informaticn needed to evaluate the ;

matters covered by this document would likely avoid the r.ecessity for NRC questions and thus significantly shorten the time required to process a fuel pool modification amendment.

Sincerely, s

' ~ %., %W Brian K. Grimes, Assistant Director for Engineering and Projects Division of Operating Reactors

Enclosures:

1. NRC Gui, dance ,

2. Notice i

5l m ,

ENCLOSURE NO. 1 cue b"'W M'

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1 OT POSITION FOR REVIEW AND ACCEPTANCE OF SPENT FUEL STORAGE AND HANDLING APPLICATIONS r I. BACKGROUND Prior to 1975, low density spent fuel storage racks Nere designed with a large pitch, to prevent fuel pool criticality even if the pool contained the highest enrichment uranium in the light. water reactor fuel assemblies. Due to an increased demand on storage space for i spent fuel assemblies, the more recent approach is to use high density I storage racks and to better ut'ilize available space. In the case of l operating plants the new rack system interfaces with the old fuel pool structure. A proposal for installation of high density storage racks The .

may involve a plant in the licensing stage or an operating plant. !

requirements of this position do not apply to spent fuel storage and handling facilities away from the nuclear reactor complex. '

On September 16, 1975, the Commission announced (40 F. R. 42801) its l intent to prepare a generic environmental impact statement onInhandling this and storage of spent fuel from light water power reactors.

notice, the Commission also announced its conclusion that it would not '

be in the public interest to defer all licensing actions intended to ameliorate a possible shortage of spent fuel. storage capacity panding completion of the generic environmental impact statement.

The Commission directed that in the consideration of any such proposed licensing action, an environmental impact statement or environmental impact appraisal shall be prepared in which five specific factors in ,

addition to the normal cost / benefit balance and environmental stresses should be applied, balanced and weighed.

The overall design objectives of a fuel storage facility at the reactor complex are governed by various Regulatory Guides, ,

Based on the reviews of such applications to date it is obv.ious that '

the staff had to request additional information that couldItbeis easily the ,

included in an adequately documented initial submittal.

intent of this document to provide guidance for the type and extent of information needed to perform the review, and to indicate the acceptance l .-

criteria where applicable.

4 t I-1 t

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II. REVIEW DISCIPLINES The objective of the staff review is to prepare (1)TheSafety broadEvaluation staff Report, and (2) Environmental Impact Appraisal.

disciplines involved are nuclear, mechanical, material, structural, and environmental.

Nuclear and thermal-hydraulic aspects of the review include the poten-tial for inadvertant criticality in the normal storage and~ handling of the spent fuel, and the consequences of credible accidents with respect to criticality and the ability of the heat removal system

' to maintain sufficient cooling.

Mechanical, material and structural aspects of the review concern the capability of the fuel assembly, storage racks, and spent fuel pool system to withstand the effects of natural phenomena such as earth-quakes, tornadoes, flood, effects of external and internal missiles, thermal loading, and also other service loading conditions.

The environmental aspects of the review concern the increased thermal and radiological releases from the facility under normal as well as accident conditions, the occupational radiation exposures, the genera-tio,n of radioactive waste, the need for expansion, the commitment of-material and nonmaterial resources, realistic accidents, alternatives to the proposed action and the cost-benefit balance.

The information related to nuclear and thermal-hydraulic type of analyses is discussed in Section III.

The mechanical, material, and structural related aspects of informa- :

tion are discussed in Section IV.

The information required to complete an environmental impact assess-ment, including the five f actors specified by the Commission, is ,

provided in Section V.

. f 11-1 c ,

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.III. NUCLEAR AND THERMAL-HYORAULIC CONSIDERATIONS

~

-1.-.

Neutron Multiplication Factor

- To--include all credible conditions,: the licensee shall calculate the effective neutron multiplication factor, k in the fuel

storagepoolundetthefollowingsetsofassumI$f,orgiitions:

c 1.1 ~ Normal-Storage a.- .The racks shall be designed.to contain the most r.eactive fuel authorized to be stored in the facility without any control rods or any noncontained' burnable poison and.the fuel shall be assumed to be at the most reactive point in its life.

b .- .The moderator shall be assumed to be pure water at the temperature within the fuel pool limits which yields the j largest reactivity.

c. ~.The array shall be assumed to be infinite in lateral extent ,

or to be surrounded by an infinitely thick water reflector -

and' thick concrete,"* as appropriate to the design,

d. Mechanical uncertainties may be treated by assuming " worst case" conditions or by performing sensitivity studies and obtaining. appropriate uncertainties.

e.- Credit may be taken for the neutron absorption in structural materials and in solid materials added specifically for .

neutron absorption, provided a means of inspection is estab- * -

lished (refer to Section 1.5). ;

1.2 Postulated Accidents The double contingency principle of' ANSI N 16.1-1975 shal.1 be ',

applied. It shall require two unlikely, independent, concurrent '

l.

events to produce a' criticality accident.

I Reali.stic initial conditions (e.g. , the presence of soluble

- boron) may.be' assumed for the fuel pool and fuel assemblies. The

""Noncontained" burnable poison is that which is not an integral part of the fuel assembly.- l

- It should'be noted that'under certain conditions concrete may be a more "

l affective reflector than water. l

,}.1 f { , 1

i (1) dropping of a fuel postulated accidents shall include: element on top of the racks and location of a fuel assembly in the pool; (2) a dropping or tip- , ping of the fuel cask or other heavy objects into the tive position of_ the fuel racks; and (4) loss of all cooling systems or flow under the accident conditions, unless the cooling system is_ single failure proof. 1.3 Calculation Methods The calculation method and cross-section values shall be veri by comparison with critical experiment data for assemblies similar Sufficiently diverse to those for which the racks are designed. configurations shall be calculated to render improbable theSo far as practi-

                      " cancellation of error" in the calculations.

cable the ability to correctly account for heterogeneities (e.g., thin slabs of absorber between storage locations) shall be demonstrated. A calculational bias, including the effect of wide spacing between assemblies shall be determined from the comparison between calcu-lation and experiment. A calculat. ion uncertainity shall be determined such that the true multiplication factor will be less than the calculated value with a 95 percent probability at a 95 The total uncertainity factor on k ff percent confidence level.shall be obtained valuebyforathe statistical combina The k tional and mechanical uncertainties. racks shall be obtained ! calculational bias, and the total uncertainty. 1.4 Rack Modification For modification to existing racks in operating reac > review: The overall size of the fuel assembly which is to be stored (a) in the racks and the fraction of the total cell' area which represents the overall fuel assembly in the model of the nominal storage lattice cell; 0 + stainless steel flux trap lattices; the nominal (b) For H.3 thickhess and type of stainless steel used in the storage racks and the thermal (.025 ev) macroscopic neutron absorp-tion cross section that is used in the calculation method

                 '              for this stainless steel; Also, for the H,0 + stainless steel flux trap lattices, the (c) change of the calculated neutron multiplication factor of 111-7.

l

                                     ~     m 1

l i infinitely long fuel assemblies in infinitely large arrays l inthestoragerack(i.e.,thegofthenominalfuelstorage j f lattice cell and the changed g) for: 2 (1) A change in fuel leading in grams of U ss, or . assumed that this change is made by increasing the . enrichment of the U2ss; and, , (2) A change in the thickness of stainless steel in the storage racks assuming that a decreasq.in stainless ; I steel thickness is taken up by an increase in water thickness and vice versa; For lattices which use boron or other strong neutron absorb-

(d) l ers provide: ! i (1) The effective areal density of the boron-ten atoms 2 or the (i.e., B10 atoms /cm . ten atoms for other neutron absorbers) between fuel assemblies. I j Similar to Item C, above, provide the sensitivity of '

                       .      (2)      thestore.gelatticecellgto:.                                                   ;c The fuel loading in grams of U235, or equivalent,                      j (a)    per axial centimeter of fuel assembly, l

j (b) The storage lattice pitch; and, ! i; The areal density of the boron-ten atoms between (c) fuel assemblies, i 4 1.5 Acceptance Criteria for Criticality , The neutron multiplication factor in spent fuel pools shalll 1ess than or equal to 0.95, includino all uncertainties, under

l l all conditions ;

l I l (1) For those facilities which employ a strong l. l

       .-                       storage pool, the licensee shall                            The provide t I

and retention of the strong absorber in the racks. results of an initial, onsite verification test shallfi-show within 95 percent confidence limits that there is a suf cient the neutron amount of neutron multiplication factor atabsorber in or below 0.95. In the racks h ll addition, coupon or other type of surveill j 111-3

                              .~ n        ,

l periodic basis throughout the life of the racks to verify the continued presence of a sufficient amount of neutron absorber in the racks to maintain the neutron multiplication factor at or below 0.95. (2) Decay Heat Calculations for the Spent Fuel The calculations for the amount of thermal energy that will have to be removed by the spent fuel pool cooling system shall be made in accordance with Branch Technical Position APCSB 9-2 entitled, " Residual Decay Energy fpr Light Water. Reactors for Long Term Cooling." This Branch Technical Position is part of the Standard Review Plan (NUREG 75/087). . Thermal-Hydraulic Analyses for Spent Fuel Cooling (3) Conservative methods should be used to calculate the maximum fuel temperature and the increase in temperature of theThe maxim water in the pool. assembly and between fuel assemblies should also be calculated. Ordinarily, in order not to exceed the design heat load for

                . the spent fuel cooling system it will be necessary to do shutdown prior to moving fuel assemblies into the spent fuel pool. The bases for the analyses should include the estab-lished e.coling times for both the usual refueling case and the full core off load case.

A potential for a large increase in the reactivity in an 2H O flux trap storage lattice exists if, somehow, the water is kept out or forced out of the space between For the this fuelreason, assem-blies, conceivably by trapped air or steam.it is necessary to sh is such that this will not occur and that these spaces will Also, in some cases, di. rect always have water in them. gamma heating of the fuel storage cell It iswalls and oftothe necessary intercell water may be significant. consider direct gamma heating of the fuel storage cell walls , and of the intercell water to show that boiling will not occur in the water channels between the fuel assemblies. Under postulated accident conditions where all non-Category I spent fuel poci cooling systems become inoperative, it is , necessary to show that there is an alternate method forWhen this alte cooling the spent pool water. requires the installation of alternate components or signifi-cant physical alteration of the cooling system, the detailed steps shall be described, along with the time required for each. Also, the average amount of water in the fuel pool and the expected heat up rate of this water assuming loss of all cooling systems shall be specified. III-4 I'

- 1 Potential Fuel and Rack Handling Accidents ' (4) t The method for moving the racks Also,to forand plantsfrom and into and ou of the fuel pool, should be described. i different where the spent fuel pool modification requ res l fuel handling procedures than that described d in the Fina Safety Analysis Report, the differences should be discusse . If potential fuel and rack handling accidents ll not occur, the neutron multiplication factor accidents These postulated in the fuelshali poolnotshabe the exceed cause0.95. of the loss of cooling for either the spent fuel or the reactor. Technical Specifications i (5) To insure against criticality, the following technical spec - fications are needed on fuel storage in high density racks: . 1. The neutron multiplication factor in the fuel pool shall be less than or equal to 0.95 at all times. 2. The fuel loading (i.e. , grams of uranium-235, or equivalent, perTheaxial centimeter of assem number of grams of racks should be limited. h uranium-235, or equivalent, put in the plan fuel pool. Excessive pool water temperatures may Analyses i lead to excessive of water due to evaporation and/or cause fogging. of thermal load should consider loss systems. ating of all pool c a' temperatures, consideration shall For be given to ture that wpuld resolve the concerns described above. . limiting values of pool water temperatures refer toentit ANSI-N210-1976 Reactor Spent Fuel Storage Facilities at Nuclear Power ' Stations," except that the requirements of th in the maximum heat load with normal cooling systems operation. e se .

111-5

TIONS IV. MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERA

(1) Description of the Spent Fuel Pool and Racks h ing the Descriptive information including plans es shall andand be sections s ow l aspects spent fuel pool in relation to other plant structur d functions of provided in order The to define the primary structur I the pool and the racks. main safety function h of the spentspen , b mal fuel pool and the racks is to maintain t e fuel cask in a safe configuration through all environmental and a no loadings, such as earthquake, and impact due to spentor drop, drop of. a spent fuel assembly,dling. object during routine spent fuel han f the The major structural elements reviewed below.- and the extent o descriptive information required are indicated > The general arrangements Support of the Spent Fuel Racks: tical . (a) and principal features of the horizontald and indi-cks the to ver supports to the spent fuel racks All gaps t ts should the fuel potl wall and the foundation slab. (clearance or expansion allowance) d base slab and sliding con should be indicated.new' rack system and the old fuel poo

                            .should be discussed, i.e., inter a tra, etc.                                                           to the be If sideconnections walls of the pool      of.the such that racksthe pool    are f radio-   made liner   may to t perforated, the provisions for avoiding leakage o active water of the pool should be indicated.

Postulation of a drop accident, h and quanti-Fuel Handling: (b) fication of the drop parameters aret reviewed ight unde environmental discipline. include a straight drop on the top of a bottom rack, In-ofa s ra drop through an fidividual cell all f a the rack.way to the postulated l the rack, and an inclined drop on the tcp o i l ' I e tegrity of the racks and the fuel poel cu ld be provided l material, and structural disciplines.cient det

               '                 to facilitate this review.

IV-1 i s o. go

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            -   (2) Applicable Codes, Standards and specifications Construction materials shoeld conform to Section III, Subsec-tion NF of the ASME* Code. All Materials should be selected to be compatible with the fuel pool environment to minimize corro-sion and galvanic effects.

Design,' fabrication, and installation of spent fuel racks of stainless steel material may be performed based.upon the AISC** specification or Subsection NF requirements of Section III of the Once a code is ASME B&PV Code for Class 3 component supports. c M sen its provisions must be followed in entirety. When the AISC specification procedures are adopted, the yield stress values for stainless steel base metal may be obtained from the Section III of the ASME B&PV Code, and the design. stresses de-fined in the AISC specifications as percentages of. the yield stress may be used. Permissible stresses for sta.inisss steel welds used in accordance with the AISC Code may be obtained from Table NF-3292.1-1 of ASME Section III Code. Other materials, design procedures, and f abrication techniques will be reviewed on a case by case basis.

                .(3) Seismic and Impact Loads For plants where dynamic input data such as floor response spec-tra or ground response spectra are not available, necessary dynamic analyses may be performed using        the criteria described in The ground response Section 3.7 of the Standard Review Plan.

spectra and damping values should correspond to Regulatory Guide 1.60 and 1.61 respectively. For plants where dynamic data are available, e.g., ground response spectra for a fuel pool sup-ported by the ground, floor response spectra for fuel pools supported on soil where soil-structure interaction was considered in the pool design or a floor' response spectra for a fuel pool supported by the reactor building, the design and. analysis of the new rack system may be performed by using either the existing input parameters including the old damping values or new param-The use eters in accordance with Regulatory Guide 1.60 and 1.61. of existing input with new damping values in Regulatory Guide 1.61 is not acceptable. Seismic excitation along three orthogonal directions should be imposed simultaneously for the design of the new rack system.

               "American Society'of Mechanical Engineers Boiler and Pressure Vessel Codes, Latest Edition.
              **American Institate of Steel Construction, Latest Edition.

IV-2

f' - The peak response from each direction shouid be combined by square root of the sum of the squares. l the same hori-available for a vertical and horizontal directions o zontal direction. The effect of submergence of the rack system NRC.

on the dam l the mass of the fuel racks has been under study by the Submergence in water may introduce damping d from two s l viscous drag, and (b) radiation of energy awayi f. hrom the subm body in those cases where the confining Viscous boundaries are fa away to prevent reflection of waves at the boundaries. Based up , , damping is generally negligible. l current study for a typical'high density rackl damping configuration, wa reflections occur at the boundaries so that no addit should be taken into account. h A report on the NRC study is to be published shortly under ! title " Effective Mass and Damping of Submerged StructuresThe , , (UCRL-52342)," by R. G. Dong. ble basis ' l this report on the added mass effect provide d/or an acceptaI

               .for the staff review. water is not acceptable without applicable tes detailed analytical results.

Due to gaps between fuel assemblies and f fuel the walls of t tubes, additional loads will be generatedAdditional by thei impact o assemblies during a postulatedl velocity seismic excita the kinetic energy of the fuel assembly. the fuel assembly may be estimated to be thed fuel spectra associated assembly. with the natural frequency d the sup- of the subm well as overall effects on the walls of the rackl anIt should porting framework. loads on the fuel assembly cept- do not le t Loads generated from other postulated i locity at the total mass of the impacting missile, the maximum vetarget the time of impact, and the ductility ratio of the utilized to absorb the kinetic energy. 4 (4) Loads and Load Combinations: d Any change in the temperature distribution f should due to the modification should be identified. f the applicable design loads and'various combl l walls and base l maximum temperature distribution through the poo IV-3

                     -                                       r                               .

l' .. cue to Temperature gradientinacrc:s the rack structurell a the oesign of the rackne snould slab. differential heating effect between a fu should be indicated and incorporatedMaximum xisting pool floor, - upli structure. be indicated including the consideration o if applicable. acceptable if they y l l The specific loads and load conbinations aretions of'Se I are in conformity with the applicable por 3.8.4-II.3 of the Standard Review Plan. Design and Analysis Procedures description of how (5) Details of the mathematical model including " aps ah uld the important parameters are obtained s othe methods ing the following: lump the masses of the between the support systems and gaps be w d used to account for and the guide tubes; the l walls;methods and, the effect used t the effect of sloshing water on the poodistribution and of submergence on the mass, the massd the fuel racks. tive damping of the fuel bundle an dance with Section ceptable. The The design and analysis procedures id in acco 3.8.4-11.4 effect on gaps, sloshing water,in water should be quan and damping due to submergence h lateral restraint atflexibil , When wails and the pool walls capability areh floor utilized to sustain such loads of the walls response tl should be provided. traint point 4t the fundamental frequency less than than 33a,t those Hertz), te the baseppro spectra corresponding to the latera , of the pool.two separate analyses should i responsebe perform spectra ation provided that A spectrum analysis of lthe rack system (a) corresponding to the highest support e evi there is not significant peak frequen bjectingittothe A static analysis of the rack t system by su f (b) maximum relative support ldisplacemen . s above l The resulting stresses from the two ana yse combined by the absolute sum method.

IV-4

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                                                                                                    !        l 5

In order to determine the flexibility of the pool wall it is i . acceptable for the licensee to use equivalent mass and stiffness properties obtained from calculations similar to t l Should the fundamental frequency of McGraw Hill Book Company. the pool wall model be higher than or equal to 33 Hertz, it may be assumed that the response of the pool wall and the corres-pending lateral support to the new rack system are identical to those of the base slab, for which appropriate floor response spectra or ground response spectra may already' exist.

     *   (6) Structural Acceptance Criteria                                                             i When AISC Code procedures are adopted, the structural acceptance criteria are those given in Section 3.8.4.II.5For         of the   Standard stainless Review steel the   Plan  for steel and acceptance          concrete criteria       structures.

expressed as a percent' age of yield I stress Plan. should satisfy Section 3.8.4.II.5 of the Stan ( used for the racks, the structural acceptance criteria are those !, given in the Table below. i' For impact loading the ductility ratios utilized to absorb kinetic energy in the tensile, flexural, When compressive, considering the effects of and shearing modes seismic should be quantified. loads, factors of safety against gross slidin racks and rack modules under all probable 3.8.5.11-5 service of the conditions Stand-shall be in accordance with the SectionThis position on factors o ard Review Plan. sliding and tilting need not be met provided any one of the following conditions is met: it can be shown by detailed nonlinear dynamic analyses that , (a) the amplitudes of sliding motion are minimal, and impact i between adjacent rack modules or between a rack module and the pool walls is prevented provided that th - Section 3.8.5.11.5 of the Standard Review Plan.

     '                    it can be shown that any sliding and tilting motion will be (b) contained within suitable geometric constraint ances is incorporated.

Haterials, Quality Control, and Special Construction Techniques (7) The materials, quality control procedures, and any The sequence special con-of in-struction techniques should be described. stallation of the new fuel racks, and a descrip IV-5

l ,. , TABLE Load Combination Acceptance Limit i Elastic Analysis _ Normal limits of NF 3231.la D+L Normal limits of NF 3231.la D + L + E. 1.5 times normal limit's or the D + L + To lesser of 2 Sy and Su

                                                                                                     ,I 1.5 times normal limits or the -

D + L + To + E leser of 2 Sy and Su 1.6 times normal limits or the lesser of 2 Sy or Su D + L + Ta + E l f aulted condition limits of D + L + Ta + E NF 3231.1c Limit Analysis j Limits of XVII-4000 of Appendix XVII a 1.7 (D + L) of ASME Code Section III 1.7 (D + L + E) 1.3 (D + L + To) 1.3 (D + L + E + To) l 1.1 (D + L + Ta + E) Notes: 1. The abbreviations in the table above are those l Section 3.8.4 of the Standard Review i Plan whe ! is deti,.ed except for Ta which is definel conditions. f Defomation limits specified by the Design Specification

2. limits shall be satisfied, and such deformation limits should preclude damage to the fuel assemblies.

3. The provisions of NF 3231.1 shall be amended by the requirements of the paragraphs c.2, 3, and 4 of the Regulatory Guide 1.124 entitled " Desi l IV-6

e . ,

                                                                                               .          i ided.      Methods for.struc-                 !

the construct 3cn phase should beterials utilized to prov The ibed. material for tural qualification of special poison ma the fuel pool tibility inside of the absorb neutron radiation should be descr the fuel rack is reviewed for compaThe quality of the fu fluorides, baron, heavy environment. pH value and the available chlorides, long-term integrity of the

                      metals should be indicated                       ch asso       that poison          therack materials Acceptance criteria for special materian         cedures. :uualification pr should be based upon the results of the       <                                   ,

If connections between the lding rackprocedure and theforpoo the welding, the welder as well as the wei d in accordan welding assembly shall be qualif e cable code. terial is used for If precipitation hardened stainless steel ma testing l racks, hardness ; the construction of the spent fuel poo In addition, l t d properly. should be performed on each rack componen to verify that each part is heat trea t eheatresist,ance. corrosion treatment shou the surface film resulting from theremoved-l Testing and Inservice Surveillance ial stability and (8) Methods for verification ofi long-term l tests. matermaterial u . 1 mechanical integrity of special pothesonne fuel racks ano Inservice surveillance requirements for by casethedepen poison material, if applicable, t areThese features w and integrity of the features. t l basis to determine the type and the ex pool and the fuel rack system.

            ~                                                                                             4 E

e IV-7 l e , I

        '.                        ,-                      e

V. COST / BENEFIT ASSESSMENT i ntal 1.

Following is a list of information needed for the env ro Cost / Benefit Assessment: d storage 1.1 What are the specific needs that require increaseInclude in th

      .                  capacity in the , spent fuel pool (SFP)?                     fuel-status of contractual arrangements,.if any, with (a)      storage or fuel-reprocessing facilities, htheexpectednumber (b) propMed refueling schedule,        includin i the         SFP at is reached, of fuel assemblies that willd be                   in the transferr number of spent fuel assemblies presently store (c)                                                             '

SFP, in the control rod assemblies or other components stored (d) SFP, and blies would i and (e) the additional time period that spent fu ' l d with the (f) the estimated date that the SFP will proposed increase in storage capacity. osed be fil e 1.2 Discuss the total construction associated tion.with the p ts (direct and modification, including engineering, ity of capi Discuss the alternative to increasing the storage c , 1.3 the SFP. The alternatives considerediTable), should incl shipment to a fuel reprocessing facility (if ava (a) facility, shipment to an independent spent fuel storage (b) I shipment to another reactor site, ~~ i (c) l L shutting down the reactor. (d) include a cost

The discussion of options (a), (b) and (c) idingshouldd l comparison in terms of dollars per KgU s ;

I replacement power either from within or outs , generating system. e V-1 U_

I l %: _ _ 1.4 Discuss w..atherg the commitment of material resou foreclose the alternatives available with respect to Describe the material resources spent fuel storage capacity.that would be consumed by the propos l 1.5 Discuss the additional heat load and the anticipated maximum temperature of water in the SFP which would result from the proposed expansion, the resulting increase in eva t systems and whether there will be any significant increase in the amount of heat released to the environment. . V.2. RADIOLOGICAL EVALUATION 2. Following is a list of information needed for radiological evaluation: 2.1 The present annual quantity of solid radioactive wastes gen-Discuss the e erated by the SFP purificat Rn system. increase in solid the capacity cf the SFP. 2.2. Data regardihg krypton-85 measured from the fuel If. data building ven-are not

      ~                tilation system by year for the last two years.

available from the fuel building ventilation syst 2.3 The increases in the doses to personnel from the SFP, including the following: Provide a table showing the most recent gamma isotopic (a) analysis of SFP water identifying the principal radio-nuclides and their respective concentrations. The models used to determine the external Consider dose equivalent the dose equiva-(b) rate frcm these radionuclides. 1ent rate at some distance above the center and edge of the (Use relevant experience if necessary). pool respectively. A table of recent analysis performed to determine the (c) principal airborne radionuclides and their respective concentrations in the SFP area. / The model and assumptions used to determ (d) (c) above in the SFP area and at the site boundary. I V-2

                                                                                 ?                     .

I

         ~                                                                     '

... r -

't    .,                                                       ,                                   ,

(e) An estimate of the increase in the annual man-rem burden from more frequent changing of the deminerali:er resin and , filter media. ' (f) The buildup of crud (e.g. , 58Co, SOCo) along the sides of the pool and the removal methods that will be used to reduce radiation levels at the pool edge to as low as reasonably achievable. (g) The expected total man-rem to be received by personnel ' occupying the fuel pool area based on all operations in ' that area including the doses resulting.from (e) and (f) above. , A discussion of the radiation protection program as it affects (a) through (g) should be provided. , 2.4 Indicate the weight of the present spent fuel racks that will be removed from the SFP due to the modification and discuss what - will be done with these racks. V.3 ACCIDENT EVALUATION 3.1 The accident review shall consider: (a) cask drop /tip analysis, and (b) evaluation of the overhead handling system with respect to Regulatory Guide 1.104. 3.2 If the accident aspects of review do not establish acceptability with respect to either (a) or (b) above, then technical _ specifica- l tions may be required that prohibit cask movement in the spent l fuel building. ' 3.3 If the accident review does not establish acceptability wi,th i! respect to (b) above, then technical specifications may be required that: . (1) define cask transfer path including control of (a) cask height during transfer, and (b) cask lateral position during transfer f indicate the minimum age of fuel in pool In sections specialduring cases (2) movement of heavy loads near the pool. evaluation of consequences-limiting engineered safety features such as isolation systems and filter systems may be required. V-3

F l . '

- l l

3.4 If the cask drop /tip analysis as in 3.1(a) above is promised for future submittal, the staff evaluation will include a conclusion l on the feasibility of a specification of minimum age of fuel based on previous evaluations. 3.5 The maximum weight of loads which may be transported over spent fuel may not be substantially in excess of that of a single fuel assembly. A technical specification will be required to this effect. 3.6 Conclusions that determination of previous Safety Evaluation i l Reports and Final Environmental Statements have 'not changed significantly or impacts are not significant are made so that a negative declaration with an Environmental Impact issued. This will involve checking realistic as well as con-servative accident analyses. l 4 i i V-4 e

c. ,

1 VI. REFERENCES

1. Regulatory Guides

                                '1.13     -

Design Objectives for Light Water Reacter Spent F

     -                                           ' Storage Facilities at Nuclear Power Stations-
                                           -        Seismic Design Classification 1.29 1.60
                                            -         Design Response Spectra for Seismic Design of Nu        ,

Power Plants Damping Values for Seismic Design of Nuclear Powe 1.61 Plants

                                              -        Design Basis Tornado for Nuclear Power Plants
                                  .1.76                                                                         '

Combining Modal Responses and Spatial Compon 1.92 Seismic Response Analysis

                             ~                          Overhead Crane Handling Systems for Nuclear. Powe 1.104 -

Plants 1.124 - Design Limits and Loading Combinations for C1t Linear-Type Components Supports

2. Standard Review Plan Seismic Design 3.7 Other Category I Structures 3.8.4 - - -

                                                  -       Fuel Storage and Handling 9.1                                                          "

Fire Protection System 9.5.1 - - Industry Codes and Standards

.3.

1. American Society of Mechanical Engineers, Bo sure Vessel Code Section III, Division 1 2. American Institute of Steel Construction Specifi 3. American National Standards Institute, N210-76 i American Society of Civil Engineers, Sugges

4. for Structures of Aluminium Alloys 6061-T6 a .

VI-1 l

a ,- . .

5. The Aluminium Association, Specification for Aluminium Structures f

e h 4 4 l j i l - I I

         .                                                                            I VI-2

http#hnvw.ntc.govMtXGLelMMflMcKoh/WW/gMZard.txt

11 1 as; y

   \'          GL79004 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20565 January 18, 1979
             'To All' Power Reactor Licensees
             -Gentlemen:                                                                                                      )

Our-letter of April 14, 1978, provided NRC Guidance entitled, " Review and

             -Acceptance of Spent Fuel Storage and Handling Applications." Enclosed are modifications to this document for your information and use. These involve-

pages IV-5 and IV-6 of the document and comprise modified rationale and corrections.

Sincerely, Brian K. Grimes, Assistant Director

                                                                  .for Engineering and Projects Division of Operating Reactors

Enclosure:

Pages IV-5 and IV-6 cc w/ enclosure: Service List 7903080173 l I In order to-determine the flexibility of the pool wall it is acceptable for the licensee to use equivalent mass and stiffness properties obtained from calculations similar to those described in " Introduction

                      - to Structural Dynamics" by J. M. Biggs published by McGraw Hill Book                                  i Company. Should the fundamental frequency of the pool wall model be                                  J higher than or equal to 33 Hertz, it may be assumed that the response                                l of the pool wall'and the corresponding lateral support to the new rack                               l system are identical to those of the base slab, for which appropriate                                i floor response spectra or ground response spectra may already exist.                                 l (6)f Structural Acceptance Criteria
                                                                                                                            ~

When AISC Code procedures are adopted, the structural acceptance criteria are those given in Section 3.8.4.II.5 of the Standard Review Plan for steel and concrete structures. For stainless steel the

                      . acceptance criteria expressed as a percentage of yield stress should                                  i satisfy Sectien 3.8.4.II.5 of the Standard Review Plan. When subsection
                      . NF,~Section III, of the ASME B&PV Code is used for the racks, the structural acceptance criteria are those given in the Table below. When buckling loads are considered in.the design, the structural acceptance 4                    criteria shall be limited by the requirements of Appendix XVII-2110(b) of the ASME' Boiler and Pressure Vessel Code.

For impact loading the ductility ratios utilized to absorb kinetic energy in the tensile, _ flexural, compressive, and shearing modes should

                      - be quantified. When'considering the effects of seismic loads, factors of safety against gross sliding and overturning of racks and rack modules under all' probable service conditions shall be in accordance with the Section 3.8.5.II-5 of the Standard Review Plan. This position i

l of 3 10/55912:36 F%4 Lt

y_-~_-- http //www.nrc. gov /NRC/G EN ACT/G C/G L/1979/gl79004.txt 3

   ,                                                                                                                         \

, , on factors of safety against sliding and tilting need not be met provided any one of the following conditions is met: (a) it can be shown by detailed nonlinear dynamic analyses that the i amplitudes of sliding motion are minimal, and impact between adjacent rack modules or between a rack module and the pool walls is prevented provided that the . factors of safety against tilting are within the values permitted by Section 3.8.5.II.5 of the Standard Review Plan. (b) it can be shown that any sliding and tilting motion will be contained within suitable geometric constraints such as thermal clearances, and that any impact due to the clearances is incorporated. (7) Materials, Quality Control, and Special Construction Techniques: The materials, quality control procedures, and any special con-struction techniques should be described. The sequence of installation of the new fuel racks, and a description of the precautions to be taken to prevent damage to the stored fuel during IV-5 TABLE Load Combination Elastic Analysis Acceptance Limit D+L Normal limits of NT 3231.la D+L+E Normal limits of NF 3231.la D + L + To Lesser of 2Sy or Su stress range D + L + To + E Lesser of 2Sy or Su stress range . l D + L + Ta + E Lesser of 2Sy or Su stress range ! D + L + Ta + El Faulted condition limits of NF 3231.lc l Limit Analysis 1.7 (D + L) Limits of XVII-4000 of Appendix XVII of ASME Code Section III 1.7 (D + L + E) 1.3 (D + L + To) 1.3 (D + L + E + To) 1.1 (D + L + Ta + E) Notes: 1. The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan where each term is defined except for Ta which is defined as the highest temperature associated with the postulated abnormal design conditions.

2. Deformation limits specified by the Design Specification limits shall be satisfied, and such deformation limits should preclude damage to the fuel assemblies.

2(f3 , 10/5/9912:36 PM

 !;                                                   WAAx/ws{puGIEMLELISSEML9%nNJTDNMEJ1tm        i
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l.g g 3. The provisions of NF 3231.1 shall be amended.by the

  '*                requirements:of the paragraphs c.2, 3,-and 4 of the Regulatory Guide 1.124 entitled " Design Limits and Load Combinations lior Class'l Linear-Type Component Supports."                     l L                                         IV-6 1-i
                            .                                                                      9 e

3,r3 10/5/9912:36 PM

9 ,. . i p afcg UNITED STATES Iog NUCLEAR REGULATORY COMMISs!ON [ .,3.( g WASHINGTON, D. C. 20555 s" !

              *% ..s ...
                          ]                                   SEP 2 31981 e

MEMORANDUM FOR: Raymond F. Fraley, Executive Director Advisory Committee on Reactor Safety FROM: Guy A. Arlotto, Director Division of Engineering Technology Office of Nuclear Regulatory Research

SUBJECT:

DRAFT 1, REGULATORY GUIDE 1.13 REVISION 2,

                                         " SPENT FUEL STORAGE FACILITY DESIGN BASIS" Enclosed for initial review of the ACRS Regulatory Activities Subcomittee are 20 copies of Revision 2 to Regulatory Guide 1.13 (Enclosure 1) and 20 copies of the Draft Value/ Impact Assessment (Enclosure 2).

The draft regulatory guide is a proposed revision to Regulatory Guide 1.13

                       " Spent Fuel Storage Facility Design Basis," which is being revised to endorse ANSI N210-1976/ANS 57.2, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations."

The draft regulatory guide, which was originally scheduled for review at the September 9th' meeting, was withdrawn to insure the incorporation of all necessary input from Division Offices. Since this draft is preliminary, additional staff efforts, including review and resolution of public comments, will be necessary prior to implementation of a regulatory position. ACRS Regulatory Ac vities Subcomittee coments and recommendations are requested on the pro sed regu Sry position. l Guy k.Arlotto, Director Divi sion of Engineering Technology Office of Nuclear Regulatory Research cc: Public Document Room

Enclosures:

as stated

    - . .       . _ _ =
              ~

Draft - 9/22/81-cif - CSchulten- Job A#1 Enclosure 1 l

                              .                                                                                     1
                   -1                     DRAFT 10F REVISION 2 TO REGULATORY GUIDE 1.13                          .

2 SPENT FUEL STORAGE FACILITY DESIGN BASIS , ( 3 A. INTRnnilCTJDN- l

                                                                                    <                               l 4          General Design Criterion 61, " Fuel Storage and Handling and Radioactivity          i

! 5 Control," of Appendix A, " General Design Criteria for Nuclear Power Plants," 6 to 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," 7 requires that fuel storage and handling systems be designed to assure adequate l 8 safety under normal and postulated accident conditions. It also requires that 9 these systems be designed (1) with a capability to permit appropriate periodic 10 inspection and testing of components important to safety, (2) with suitable 11 shielding for radiation protection, (3) with appropriate containment, confine- l 1

                  ,12    sent, and filtering systems, (4) with a residual heat removal capability having            i 13    reliability and testability that reflects the importance to safety of decay 14-  heat and other residual heat removal, and (5) to prevent significant reduction             i 15- in fuel storage coolant inventory under accident conditions. This guide 16   describes a method acceptable to the NRC staff for implementing this criterion.            l g7                                   B. DISCUSSION 18          Working Group ANS-57.2 of the American Nuclear Society Subcommittee ANS-50 19    has developed a standard which details minimum design requirements for 10 CFR             j 20    Part 50 light water reactor spent fuel storage facilities at nuclear power

! 21 stations. This standard was approved by the American National Standards 22 Committee N18, Nuclear Design Critaria. It was subsequently approved and 23 designated ANSI N210-1976/ANS-57.2, " Design Objectives for Light Water Reactor . 1.13-1 L

v.

      ~
        ~

Draft - 9/22/81-cif

CSchulten- Job A#1 1 Spent Fuel Storage Facilities at Nuclear Power Stations" by the American National ,

2 Standards Institute on April 12, 1976. 3 These facilities must be designed to: { l 4- a. Prevent loss of water from the fuel pool that would uncover fuel. ! 5 b. Protect the spent fuel from mechanical damage. 6 c.' Provide the capability for limiting the potential'offsite exposures 7- in the event of significant release of radioactivity from the fuel. ; l 8 If spent fuel storage facilities are not provided with adequate protective i 9' features, radioactive materials could be released to the environment as a result

          '10-     of either loss of water from the storage pool or mechanical damage to fuel within 11     the pool.

12 1. Loss of Water from Storage Pool 13 Unless protective measures are taken, loss of water from a fuel storage 14 pool could cause overheating of the spent fuel, resultant damage to fuel clad-

          =15      ding integrity, and could result in a release of radioactive materials to the 16    environment. Naturat events, such as earthquakes or high winds, could damage 17    the fuel pool either directly or by the generation of missiles.       Earthquakes or    l 18    high winds could also cause structures or cranes to fall into the pool. Design-19     ing the facility to withstand these occurrences without significant loss of 20    watertight integrity would alleviate these concerns.

21 Dropping of heavy loads, such as a 100-ton fuel cask, although of low 22 probability, should be considered in plant arrangements where such loads are ; 23 -positioned or moved in or over the spent fuel pool. Cranes which are capable 24 of carrying heavy loads should be prevented, preferably by design rather than

           '25      interlocks, from moving into the vicinity of the pool.

1.13-2

Draft - 9/22/81-cif ,. CSchulten- Job A#1 1 ,The negative pressure in the fuel handling building during movement of spent fuel should be at least m s 3.2 mm (-0.125 inches) water gauge to pre- " 2 3 vent exfiltration and to assure that any activity released to the fuel handling 4 building will be treated by an engineered safety feature (ESF) grade filtration 5 syctem before release to the environment. 6 Even if the measures described above which are used to maintain the desired 7 negative pressure are followed, small leaks from the building may still occur as 8 a result of structural failure or other unforeseen events. For example, equip-9 ment failures in systems connected to the pool could result in loss of water l l 10 from the pool if this loss is not prevented by design. A permanent fuel pool- ' 11 coolant makeup system with a moderate capability, and with suitable redundancy 12 or backup, could prevent the fuel from being uncovered if these leaks should 13 occur. Early detection of pool leakage and fuel damage could be provided by ) 14 both pool-water-level monitors and radiation monitors. Both types of monitors

       .15   should be designed to alarm both locally and in a continuously manned location.

16 Timely operation of building filtration systems can be assured if these systems 17 are actuated by a signal from local radiation monitors. 18 2. Mechanical Damage to Fuel 19 The release of radioactive material from fuel may occur during the refueling 20 process, and at other times, as a result of fuel-cladding failures or mechanical 21 damage caused by the dropping of fuel elements or the dropping of objects onto 22 fuel elements. 23 Missfies generated by high winds are also a potential cause of mechanical 24 damage to fuel. This concern could be eliminated by designing the fuel storage 1.13-3 ;q ,,

l N ,* , Draft - 9/22/81-cif l ,4 , CSchulten- Job A#1 I' 1 facility to preclude the possibility of the fuel being struck by missiles ,

              .2    . generated by'high winds.

3- 3. Limiting Potential Offsite Exposures

              .4             A relatively small amount of mechanical damage to the fuel or fuel over-5 ' heating might cause significant offsite doses of radiation if no dose reduction
              .6       features are provided. Use of a controlled leakage building surrounding the l

l' 7 fuel storage pool, with associated capability to limit releases of radioactive 8 material resulting from a refueling accident, would appear feasible and do much 9 to eliminate this concern. 10 For the spent fuel pool cooling, makeup and cleanup systems, the staff 11 will consider the design acceptable if it includes seismic Category 1 and 12 tornado protection for the water makeup source and its delivery system, the 13 ' pool structure, the building housing the pool, and the storage building's

             ,14       filtration-ventilation systems. The pool building's filtration-ventilation 15        systems should be designed to meet the guidelines of Regulatory Guide 1.52, 16       " Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-17        Feature Atmosphere Cleanup System Air Filtration and Absorption Units of Light-18       Water-Cooled Nuclear Power Plants."                            ,

19 In all activities involving personnel exposure to radiation, attention 20 should be directed toward keeping occupational radiation as low as reasonably 1-21 achievable.(ALARA). Efforts toward maintaining exposures ALARA should be 22 included in the design, construction, and operational phases. Guidance on 23 ' maintaining exposures ALARA is provided in Regulatory Guide 8.8, "Information 24 Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power 25 Stations Will Be As Low As Is Reasonably Achievable." 1.13-4

 .                                                                                                         i
                                                                                                           )
             -                                                                      Draft - 9/22/81-cif C5chulten- Job A#1     I l                                                C.      REGULATORY POSITION 3      '

l 2 The requirements that are included in ANSI N21'0-1976/ANS-57.2, " Design ! l 3 Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear 4 Power Stations"1 are generally acceptable to the NRC staff. The staff has 5 determined that this standard provides an adequate basis for complying with 6 the requirements of General Design Criterion 61 " Fuel Storage and Handling and 7 Radioactivity Control" of Appendix A " General Design Crikeria for Nuclear Power 8 Plants" to 10 CFR Part 50 as related to light water reactors and subject to 9 the following clarifications and modifications: 10 1. The example in Section 4.2.4.3(1) should be modified. The inventory -\ 11 of radioactive materials that could possibly leak from the spent fuel building 12 shouldcorrespondtotgamountpredictedtoleakunderthepostulatedcaximum 13 damage conditions resulting from the dropping of a spent fuel assembly in the 14 spent fuel building. However, in any event, the inventory should not be less 15 than the amount available due to rupture of all fuel rods of a spent fuel assembly. 16 Other assumptions in the analysis should be consistent with those given in 17 Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radio-18 logical Consequences of a Fuel Handling Accident in the Fuel Handling and Storage 19 Facility for Boiling and Pressurized Water Reactors.n2 80 2. In addition to meeting the requirements of Section 5.1.12 the maximum 21 potential kinetic energy capable of being developed by those objects handled' 22 2: 2 Copies may be obtained from the American Nuclear Society, 555 North Kensington 24 Avenue, La Grange Park, Illinois 60525 25 2 Copies of Regulatory Guicies may be obtained from the U.S. Ntclear Regulatory 26 Commisson, Washington, D.C. 20555. l 1.13-5 L

Draft - 9/22/81-cif CSchulten- Job A#1 1 above stored spent fuel, if dropped, is not to exceed the kinetic energy of ,j 2 one fuel assembly innd its associated handling tool when dropped from the height 3 at which it is normally handled above the spent fuel pool storage racks. 4 3. In addition to meeting the requirements of Section 5.1.3, boiling of 5 the pool water may be permitted only when the resulting,thermal loads are 6 properly accounted for in the design of the pool structure, the storage racks, 7 and other safety-related structures, equipment, and systems. 8 4. In addition to meeting the requirements of Section 5.1.3, the fuel 9 storage pool should be designed (a) to keep tornado winds and missiles generated 10 by these winds from causing significant loss of watertight integrity of the 11 fuel storage pool and (b) to keep missiles generated by tornado winds from j 12 striking the fuel. These requirements are discusu ' in Regulatory Guide 1.117,

        ,13  " Tornado Design Classification." The fuel storage building, including walls 14  and roof, should be designed to prevent penetration by tornado missiles or from 15  seismic damage to assure that nothing bypasses the ESF grade filtration system in the containment building. Inhepnt an earthqtrate-trr e L5rhn' ado-Mesile 16 17  danag6tament-and4tAfuepgql.4eoling7yrter, no 18   6 i

19 'aittsme_.r:dWAvirty. , 20 5. In addition to meeting the requirements of Section 5.1.5.3, provisions I 21 should be made for handling highly radioactive non-fuel leeM'-t:d components l 22 in fuel pools. Either the design of the retrieval system or administrative 23 controls should be included which would prohibit unknowing retrieval of 24 irradiated components. 1.13-6

u l

Draft - 9/22/81-cif

 .-.                                                                                CSchulten- Job A#1      j 1          ,6. In addition to meeting the requirements of Section 5.2.3.1, an interface    ,

2 between th'e cask venting system and the anxinMed building ventilation system 3 should b'e provided. This interface would provide for the proper handling of l 4 the " vent gas" generated from filling a dry, loaded cask with water and thereby 5 minimizing personnel exposure from the untreated off gas.

           '6           7.   .In order to limit the potential offsite release of radioactivity during 7     a Condition _IV-fuel handling accident, Section 5.3.3 should include the require-8     ment that the released radioactivity be either contain~d  e    or removed by filtration
         ~9       so that the dose to an individual is less than 10 CFR Part 100 guidelines.

l 10 The calculated offsite dose to an individual from such an event should be wel1 11 within (approximately 25% of) the exposure guidelines of 10 CFR Part 100 using i 12- appropriately conservative analytical methods and assumptions. In order to 13 assure that released activity does not bypass the filtration system, the 14 engineered safety feature fuel storage building ventilation should provide and 15 maintain a negative pressure of at least algos 3.2mm (20.125 inches), water 1 16 gauge within the fuel storage building. l 17 8. In addition to the requirements of Section 6.3.1, overhead handling 18 systems used to handle the spent fuel cask should be designed such that travel 19 directly over the spent fuel storage pool or safety-related equipment is not j l l 20 possible. This should be verified by analysis to show that the physical structure l 21 under all cask handling pathways will be adequately designed so that unacceptable i 22 damage to the spent fuel storage facility or safety-related equipment will not i 23 occur in the event of a load drop. 1.13-7 u

FI . l Draft - 9/22/81-cif ! . CSchulten- Job A#1 ! I^ l 1 9. In addition to the references listed in Section 6.4.4, Safety Class 2- 3, Seismic Category I and safety-related structures and equipment should be ! 3 subject to a quality assurance program which meets the applicable provisions j 4 of Appendix B to 10 CFR Part 50. Further, those programs should obtain guidance ! 5 from Regulatory Guide 1.28 endorsing ANSI N45.2 " Quality Assurance Program 6 Requirements for Nuclear Facilities" and the applicable provisions of ANSI N45.2 7 daughter standards endorsed by Regulatory Guides. , 1 8 The Regulatory Guides endorsing the applicable ANSI N45.2 daughter stan-9 dards are as follows: 10 1.30 Quality Assurance Requirements for the Installation, Inspection, 11 and Testing.of Instrumentation and Electric Equipment (N45.2.4). 12 Quality Assurance Requirements for Packaging, Shipping, Receiving, 1.38 13 Storage, and Handling of Items for Water-Cooled Nuclear Power

           '14                   Plants (N45.2.2).

15 1.58 Qualification of Nuclear Power Plant Inspection, Examination, 16 and Testing Personnel (N45.2.6). 17 1.64 Quality Assurance Requirements for the Design of Nuclear Power 18 Plants (N45.2.11). 19 Quality Assurance Terms and Definitions (N45.2.10). 1.74 20 Collection, Storage, and Maintenance of Nuclear Power Plant 1.88 21 Quality Assurance Records (N45.2.9). 22 Quality Assurance Requirements for Installation, Inspection, 1.94 23 and Testing of Structural Concrete and Structural Steel Dunng 24 the Construction Phase of Nuclear Power Plants (N45.2.5). 1.13-8

  ....;         =               ... . . =
       -                                                                          Draft - 9/22/81-cif

.. CSchul+.en- Job A#1 1 1.116 Quality Assurance Requirements for Installation, Inspection, 2 and Testing of Mechanical Equipment and Systems (N45.2.8). 3 1.123 Quality Assurance Requirements for Control of Procurement of 4 Items and Services for Nuclear Power Plants (N45.2.13). 5 10. The spent fuel pool water temperature of 65.6*C (150'F) stated in Sec- ) 6 tion 6.6.1(2)(a) exceeds the NRC staff recommended limit. With the normal 7 cooling system in operation, the pool water temperature should be kept at 8 or below 60*C (140*F) with full core offload except when the pool water 9 temperature is based on comparative analyses of the pool conditions that i 10 have been founo acceptable previously. The spent fuel pool water tempera-11 ture recommended ifnits for normal and abnormal cases are indicated in the 12 table below. 13 NORMAL OPERATION ; 1 14 Case I Case II s ! 15 . both trains operational . both trains operational l 16 . normal refueling . full core offload 17 . pool full of spent fuel . pool full of spent fuel 18 . Maximum operating temperature Maximum operating temperaturd, 19 < 48.9'C (120 'F) < 60'C (140' F) 20 to protect the ion exchange based on fogging criteria and 21 personnel comfort resin from degradation O e 1.13-9

Draft - 9/22/81-cif CSchulten- Job A#1 1 - ABNORMAL OPERATION 2 Case III Case IV

                                                                                                                                                                                         .t 3                   . one train operational                    . no cooling loops operational.

4 . normal refueling . full core offload 5 . pool full of spent fuel . pool full,,of spent fuel 6 Maximum operating temperature Pool boiling permitted 7 <60*C (140*F) e v 8 11. A nuclear criticality safety analysis should be performed in accordance 9 with Annex A for each light water reactor spent fuel storage facility that 10 involves the handling, transfer, or storage of spent fuel assemblies. Il 12. Sections 6.4 and 9 of ANS 57.2 lists codes and standards that are reperenced 12 in this standard. Endorsement of ANS 57.2 by this regulatory guide does 13 not constitute an endorsement of the referenced codes and standards. 14 D. IMPLEMENTATION

            .15               The purpose of this section is to provide information to applicants regard-16 ing the NRC staff's plans for using this regulatory guide.

17 This guide reflects current NRC staff practice for construction permit 18 review. Therefore, except in those cases in which the applicant proposes an 19 acceptab,le alternative method for complying with specified portions of the 20 Commissi5n regulations, the methods described herein will be used in the 21 evaluation of license applications docketed after . 1.13-10

      .                                                                     Draft - 9/22/81-cif CSchulten- Job A#1 1

el#6EK A & 2 Nuclear Criticality Safety 3 11 Scope of Nuclear Criticality Safety Assessment 4 1.1 A nuclear criticality safety analysis shall be performed for each 5 lightwaterreactorspentfuelstoragefacilih,ysystemthatinvolves 6 the handling, transfer, or storage of spent fuel assemblies. 7 1.2 The nuclear criticality safety analysis shall demonstrate that 8 each reactor spent fuel storage facility system is subcritical 9 (k,ff shall not exceed 0.95). 10 1.3 The nuclear criticality safety analysis shall include consideration 11 of all credible normal and abnormal operating occurrences, including: 12 a) Accidental tipping or falling of a spent fuel assembly 13 b) Accidental tipping or falling of a storage rack during transfer 14 c) Misplacement of a spent fuel assembly 15 d) Accumulation of solids conthining fissile materials on the 16 pool floor or at locations in the cooling water system. 17 e) Fuel drop accidents 18 f) Stuck fuel assembly / crane uplifting forces 19 g) Horizontal motion of fuel before complete removal from rack 20 h) Placing a fuel assembly along the outside of rack 21 i) Objects that may fall onto the stored spent fuel assemblies 1.13-11

Draft - 9/22/81-cif ;

   .                                                                          C5chulten- Job A#1 1      ,1.4 At all locations in the reactor spent fuel storage facility where 2             spent fuel is handled or stored, the nuclear criticality safety 3             analysis shall demonstrate that criticality could g occur without 4             at least two unlikely, independent, and concurrent failures or 5             operating limit violations.

6 1.5 The nuclear criticality safety analysis shall explicitly identify 7 spent fuel assembly characteristics upon which subtriticality in the 8 reactor spent fuel storage facility depends. 1 l 9 1.6 The nuclear criticality safety analysis shall explicitly identify I 10 design limits upon which subcriticality depends that require physical 11 verfication at the completion of fabrication or construction. l l l 12 1.7 The nuclear criticality safety analysis shall explicitly identify 13 operating limits upon which suberiticality depends that require 14 implementation in operating procedures. i 15 2. Calculational Methods and Codes 16 Methods used to calculate subcriticality shall be validated in accordance 17 with Regulatory Guide 3.41, " Validation of Calculational Methods for Nuclear 18 Criticality Safety." (Endorses ANSI N16.9-1975) i

l 1.13-12 Q ,,

       -                                                                            Draft - 9/22/81-cif
   ,                                                                                CSchulten- Job A#1 1    3. , Method to Establish Subtriticality
           -2        3.1 T       valuated multiplication factor of fuel in the spent fuel 3              stora e racks under normal and credible abnormal conditions shall 4              be equa to or less than an established maximum allowable multi-5              plication actor k,; i.e.,
                                                                                      /

6 k, < sk, (Eq.,I)

7 where /

8 k, = the evaluatediaaximum multiplicaton factor of fuel in the 9 ' spent fuel stor\ age racks, including any necessary allowance 10 for statistical u ertainties in the calculational technique t \ 11 such as in Monte Carlo calculations. l 12 The maximum allowable multiplication factor shall be calculated 13 from the expression: \

14

k, = kc - Ak u - Ak, \ (Eq. 2) I 15 where 16 d = k,7f c computed for the most reactive fuel assembly at the most . l / s , p reactive point by the same calculatiorial method which was used 17 . i

                                                                             ~                         c i                                                           .

18 r-the-benchmark uperiment

                                              /
                                                                              \                      -

19 Note: kg Ms the value of k,ff that results\fromthecalcu- l 20 Ntion of the benchmark experiments usi'ng a particular

21 / calculational method. The value representsga combina-

22 tion of theoretical technique and numerical da . (For 23 more detail, see Regulatory Guide 3.41, "Validati n of 24 Calculational Methods for Nuclear Criticality Safet ") . 1.13-13

    -                                                                             Draft - 9/22/81-cif                 i
 .;                                                                               CSchulten- Job A#1 1     ,

Aku = The un y in t nchmark experiments. 2 Ak, = The vaba-tequired to u ure r an accepted margin of subcriticality.

3 3.2 Ak ushall include both uncertainties in the benchmark experiments as 4 well as uncertainties in the bias which result from extrapolation of the ) 5 benchmark experiments into the range of parameters encountered in the spent

                                                                             ,.                                       l 6       fuel storage rack design.                                                                           )

s l 7 3.3 Ak, shall provide an adequate margin of subtriticality under the 8 operating limitations and Design Events I through IV, and shall be no l 9 less than 0.02 (new fuel when stored dry).* s l 10 3.4 Inc.the absence of information that justifies a smaller margin of 11 subcriticality, value of 0.05 shall be assumed for Ak, for the design i 12 of spent fuel storage racks (spent fuel). 13 4. Storage Rack Analysis Assumptions l 14 4.1 [The-f uei-ns s embly- as sumed- f or-sterage-f acili ty- de si gn-s haii-be-ene 15 ef-the-foliewing-] The spent fuel storage rack module design shall be 16 based on one of the following assumptions for the fuel: 17 a) the most reactive fuel assembly to be stored at the most 18 reactive point in the assembly life [with-ne-aliewance-for 19 fission prodnet-centent-doe-te-born-ap); or 20 21 " Additions shown by underline and a vertical line in each margin. Deletions 22 shown by brackets and crossouts. 1.13-14 I

f - Draft - 9/22/81-cif

   .,-                                                                       CSchulten- Job A#1 1 ,

b) the most reactive fuel assembly to be stored based on a minimum 2 - confirmed burn up. [if-credit-is-taken-for-bornap--an-aliewabie 3 foei-assembiy-reactivity-shali-be-estabiished and-it-shali-be 4 shown-by-actuai-measurement-that-each-feei-assembiy-meets-this 5 eriterien-before-it-is-afiowed-te-be pieced-in-sterage-] (See 6 Annex B.) 7 Both types of rack modules may be present in the same storage 8 pool. 9 4.2 Determination of the most reactive spent fuel assembly shall include 10 consideration of the following parameters: 11 . maximum fissile fuel loading, 12 . fuel rod diameter, 13 . fuel rod cladding material and thickness, 14 . fuel pellet density, 15 . fuel rod pitch and total number of fuel rods within assembly, 16 . absence of fuel rods in certain locations, and 17 . burnable poison content. 18 4.3 The fuel assembly arrangement assumed in storage rack design shall l 19 be the arrangement that results in the highest value of ks considering: 20 a) spacing between assemblies, 21 b) moderation between assemblies, and 22 c) fixed neutron absorbers 'oetween assemblies. 1.13-15

          .                                                                    Draft - 9/22/81-cif CSchulten- Job A#1 1  4
                   ,.4   Determination of the spent fuel assembly arrangement with the highest        ,

2 value of ksshall include consideration of the following: 3 a) eccentricity of fuel bundle location within the racks and 4 variationsinspacingamongadjacentbundles, 5 b) dimensional tolerances, 6 c) construction materials, 7 d) fuel and moderator density (allowance for void formations and 8 temperature of water between and within' assemblies), 9 e) presence of the remaining amount of fixed neutron absorbers in ! 10 fuel assembly, and I 11 f) presence of structural material and fixed neutron absorber in i 12 cell walls between assemblies. l l l 13 4.5 Determination of burn up for storage shall be made in racks for which 14 cred'it is taken for burn up. The following methods are acceptable: l 15 a) a minimum allowed fuel assembly reactivity shall be established and 16 a reactivity measurement shall be performed to assure that each assembly 17 meets this criterion; or 1 18 b) a minimum fuel assembly burn up value shall be established as deter-19 mined by initial fuel assembly enrichment or other correlative param-20 eters and a measurement shall be ryrirped to assure each fuel assembly 21 meets the established criterion; o,- 1.13-16

Draft - 9/22/81-cif

   ..                                                                      CSchulten- Job A#1
                                                                  ~

1 ,c) a minimum fuel assembly burn up value sha11 be established as deter-l

            ~2         mined by initial fuel assembly enrichment or other correlative param-l 3        eters and an analysis of each fuel assembly's exposure history shall 4        be performed to determine its burn up. The analyses shall be performed

5. under strict administrative control usina approved written procedures.

6 The procedures shall provide for independent checks of each step of 7 the analysis by a second cualified person usi%a nuclear criticality 8

safety assessment criteria described in Section 1.4. 9 The uncertainties in detemininc fuel assembly storage acceptance criteria 10 shall be considered in establishing storage rack reactivity, and auditable 11 records shall be kept of the method used to determine fuel assembly storace 12 acceptance criterion for as lona as the fuel assemblies are stored in the 13 racks.

         ' 14     Consideration shall be given to the axial distribution of burn up in the 15    fuel assembly and a limit shall be set on the length of the fuel assembly 16    which is permitted to have a lower average burn up than the fuel assembly 17    avernoe.

18 5. Use of Neutron Absorbers in Storage Rack Design 19 5.1 Fixed neutron absorbers inay be used for criticality control under 20 thefollowingconditions: s

21. a) The effect of neutron-absorbing materials of construction or 22 added fixed neutron-absorbers may be included in the evaluation 1.13-17

Draft - 9/22/81-cif J' CSchulten- Job A#1 1 . if they are designed and fabricated so as to preclude inadver-2- tent removal by mechanical or chemical action.

3 b) Fixed neutron' absorbers shall be an integral, non-removable part 4 of the storage rack. 5 c)' When a fixed neutron absorber is used as the primary nuclear 6 criticality. safety control, there shall be provision to: 7 1) initially confirm absorber presence in the storage rack, 8 and j J t l 9 2) periodically verify continued presence of absorber. ! i

             .10      5.2 The presence of a soluble neutron absorber in the pool water        !

l

11 . shall not normally be used in the evaluation of k,. However, when l 12 calculating the effects of Condition IV faults, realistic initial 13 conditions (e.g., the presence of soluble boron) may be assumed for l l 14 the fuel pool and fuel assemblies. i 1.13-18 l

c .: .

            .                                                                           Draft -'9/22/81-cif
    . .  .                                                                            CSchulten- Job A#1 l

l . l 'l ANNEX B l l l 2 Most Reactive Fuel Assembly to be Stored 3 Based on a Minimum Confirmed Burnup e 4 If credit is to be taken for fuel burnup in the design of spent fuel storage l 5 racks, an acceptable basis for setting and meeting the' limit must be established. 6 The rationale for this basis will evolve from many rather complex considerations. l 7 Consideration should be given to the fact that the reactivity of any given spent fuel assembly will depend on initial enrichment, 2350 depletion, amount of 8 9 burnable poison, plutonium buildin and fission product burnable poison depletion, l- 10 and the fact that the rates of depletion and plutonium and fission product

              '11  buildits are not necessarily the same.

l 12 Consideration should be given to how burnup limits are selected and 13 specified for a particular fuel type: 14 The allowable 2ssU depletion in the spent fuels without burnable poison 15 must not be set too high. If too much depletion is credited in the analysis L 16 compared to the range of 885U depletion in spent fuel assemblies to be 17 stored, the design could be nonconservative from the standpoint of 18 criticality safety. On the other hand, if too little depletion is credited 19 in the analysis compared to the spent fuel to be stored, then the design 20 will be conservative. Thus a maxima depletion to be allowed in design ; l l 1.13-19 ! l < - - - - - 1

7 ] L .- . J 6- Draft - 9/22/81-cif CSchulten- Job A#1 l l 1 ,can be established consistent with the range of assU depletions expected L l 2 in the spent fuel assemblies to be stored. (This limit would then l 1 l 3 correspond to the minimum depletion that would be allowed in a particular 4 fuel assembly type destined to be stored in the racks.) l 5 The allowable plutonium content in the spent fuel upon which design would 6 be based must not be set too low. If design is based on too little pluto-7 nium compared to the range of plutonium concentrations that may be in the 8 spent fuel assemblies to be stored in the racks, the design could be non-9 conservative from the standpoint of nuclear criticality safety. On the 10 other hand, if too much plutonium is credited in the analysis of the 11 storage racks compared to the spent fuel assemblies to be stored, then 12 the design would be conservative. Thus, a minimum plutonium content to

13 be allowed in desian can be established consistent with the range of 14 plutonium concentrations expected in the spent fuel assemblies to be stored. 15 (This limit would then correspond to the maximum plutonium content that 16 would be allowed in a particular fuel assembly type destined to be stored 17 in the racks.) 18 Credit for fission product content presents special problems, such as the 19 identities and quantities of the various fission products present and how L 20 to evaluate the effect of decay rates on the credit taken. The allowable 21 fission product content in the spent fuel upon which design would be based

      - 22    must not be set too high. If design is based on too high of a fission 23   product content compared to the range of fission product concentrations 24    that may be in the spent fuel assemblies to be stored in the racks, the 1.13-20

F... . I l . Draft - 9/22/81-cif ( CSchulten- Job A#1 l 1 . design could be non-conservative from the standpoint of criticality safety. - t , 2 On the other hand, if too few fission products are credited in the analysis l 3 cf the racks compared to the spent fuel assemblies to be stored, then the 4 design would be conservative. Thus, with proper consideration a maximum l 5 fission product content to be allowed in design could be established consis-6 tent with the range of fission product concentrations expected in the spent i

                                                                                                        ,-                                  j 7         fuel to be stored,                                                                                             l l

l 8 (This limit would then correspond to the minimum ' fission product content 9 that would be allowed in a particular fuel assembly type to be stored in j 10 the racks.) 1 11 Finally, consideration should be given to the practical implementation of

                 -12 the spent fuel screening process.                     Factors to be considered in choosing the 13 screening method should include:                     [Bepietion-of.224l-and picteniem-and-fission 14 product-beiidin-cannet-be-easily-et practicaily-determined-anaiyticaliy:--An 15 ebvices-approach-wenid-be-to-transiste-the-aiiewabie-bornep-to-a-net-ailewable

! 16 f e ei- a s s e mbly- re a c ti v i ty- and- th e n-me ss ere- e v e ry- f e ei- a s s e mbiy- te- c e n fi rm-th at 17 the-minimum-etiterien-is-met;] i l 18 - accuracy of the method in determining the storage rack reactivity

19 - reproducibility of the result, i.e. , what is the confidence 'in the I

20 result? l 21 - simplicity of the procedure; i.e. , how much disturbance to other opera- ] 1 tions is involved?: 22 l 23 - accountability, i.e., ease and completeness of recordkeeping; and l 24 - auditability, i i 1.13-21 ,, l l l L. i

0 Draft - 9/22/81-cif j r CSchulten- Job A#1 l . Enclosure 2 1 1 VALUE/ IMPACT ASSESSMENT ON NUCLEAR POWER PLANT 2 SPENT FUEL STORAGE FACILITY DESIGN i 3 1. PROPOSED ACTION 4 1.1 Description 4 5 Each nuclear power plant has a spent fuel storage facility. General Design 6 Crit $ria61,"FuelStorageandHandlingandRadioactivityControl"ofAppendixA, ! 7 " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic 8 Licensing of Production and Utilization Facilities," requires that fuel storage 9 and handling systems be designed to assure adequate safety under normal and 10 postulated accident conditions. The proposed action would provide an acceptable , 11 method for implementing this criterion. This . action would be an update of. 12 Regulatory Guide 1.13, " Spent Fuel Storage Facility Design Basis." 13 1.2 Need for Proposed Action 14 Since Regulatory Guide 1.13 was last published in December of 1975, addi-15 tional guidance has been provided in the form of ANSI standards and NUREG reports. 16 The Office of Nuclear Reactor Regulation has requested this guide be updated. 1.13-22 I l

l

        .                                                                     Draft - 9/22/81-cif  i
   .-                                                                         CSchulten- Job A#1   l 1 1.3 .Value/ Impact of Proposed Action 2       1.3.1 NRC 3       The applicants' basis for the design of the spent fuel storage facility 4 will be the same as that used by the staff in its review of a construction        '

5 permit application. Therefore, there should be a minimum of cases where the 6 applicantandthestaffradicallydisagreeonthedesighcriteria. 4 7 1.3.2 Government Agencies 1 8 Applicable only if the agency, such as TVA, is an applicant. j j 9 1.3.3 Industry 10 '. The value/ impact on the applicant will be the same as for the NRC staff. 1 l 11 1.3.4 Public ) 12 No major impact on the public can be foreseen. ! 4

13 1.4 Decision on Proposed Action 14 The guidance furnished on the design basis for the spent fuel storage 15 facility should be updated. l _16 2. TECHNICAL APPROACH 17 The American Nuclear Society published ANS-57.2 (ANSI N210), " Design ,

          , 18  Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear 19 Power Stations." Part of the update of Regulatory Guide 1.13 would be an 1.13-23                                          )

i i

L - . Draft - 9/22/81-cif

       ?                                                                             CSchulten- Job A#1 1  evaluation of this standard and possible endorsement by the NRC. Also recommenda-2' tions made by Task A-36 which were published in NUREG-0612, " Control of Heavy 3  Loads at Nuclear Power Plants" would also be included.

4 3. PROCEDURAL APPROACH e 5 Since Regulatory Guide 1.13 already deals with the proposed action, logic 6 dictates that this guide be updated. 7 4. STATUTORY CONSIDERATIONS 8 4.1 NRC AUTHORITY 9 This guide would fall under the authority and safety requirements of the 10 Atomic Energy Act of 1954, as amended. In particular under General Design 11 Criterion 61, Appendix A, 10 CFR Part 50 of the NRC's implementing regulations. 12 4.2 Need for NEPA Assessment 9 13 The proposed action is not a major' action as defined by 10 CFR Part 51.5(a)(10) 14 and does not require an environmental impact sta'tement. 15 5. CONCLUSION 16- Regulatory Guide 1.13 should be updated. 1.13-24

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.X-MET XRF analyzers INTRODUCTION Unlike other systems which are O Stainless and High Temperature I limited to 21 or fewer elements and Steels ,

Fast and accurate material identifica- fewer types of alloys, the X MET is ex- O Chrome Moly Steels , tion is required in many areas of the tremely versatile. This advanced de- O 'Ibol Steels sign truly represents a significant O Alloy Steels (with greater than ' metals industry, including production, fabrication, inventory control and benefit of the new generation portable 1 % of either C ; Ni, Cu, Mn, scrap sorting. X-ray fluorescence spec- alloy analysis systems. Mo) trometry has, over the last twenty The state of-the-art X-MET portable O Nickel Based Alloys (inconels, gained the recognition of metal- alloy analysis system is powerful yet Hastelloys, Monels, etc.) sts as a significant tool in materi- simple to use it can be factory O Cobalt Based Alloys (Stellites, al entification. The speed, reliability, calibrated to pmvide a direct readout Haynes, etc.) and non<!estmetiveness of x ray of alloy name in five seconds or less. O Copper Based Alloys (Brasses, 4 fluorescence spectrometry make it Many of the 30,000 alloy types in use Bronzes, Cupro-nickels, etc.) l suitable not only for laboratory appli- today are applicable, including: O Titanium Based Alloys i cations, but also for field and plant O Aluminum Based Alloys use. . O Magnesium Based Alloys 1 The successful expansion of x-ray O Zine and LeadlZine Allo >s fluorescence analysis from laboratory 0 Exotics (Zirconium Alloys, to plant environments was prompted Molybdenum Alloys, etc.) by the development of portable analyzers made possible by: 1 (i.) the use of small, sealed radio-isotope sources used to excite the characteristic x rays of the sample; (ii.) the availability of powerful microprocessors; and ._ J (iii.) the use of rechargeable bat- 'KpK 5 in ,I, l teries to make the instrument independent of AC powet y .q g 'f. . 7.' . dyg 4a Y g - l Metorex's X-MET x-ray fluores o w. . .

                                                                                                                      -                                     !i cence analyzer makes use of these de-                                     .                            o                                                       ,

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x-ray analyzer. The X MET also allows for data storage and processing, a .. u V: w -*e. O Ej ; task previously assigned to an off-line . .. . _ , I computer. s

A careful examination of the specifi- "

I cations of thousands of alloys cunent- ., , ly used reveals that there are 40 to 50 .. ' f.9. , elements involved in the alloying e/ s process, with 10 to 20 typically %g , present in any given alloy. In compari- , Z son, the X MET is capable of measur- , n, i 'i J ing odic all 80from of atomic the elements number in the[ri-Y.~ N4 M l table - F -- - 5 aluminum, thmugh atomic numben l 92, uranium. . ., thlYh 3;[iys 6MM i . 3.v g pg . ~_ j 2

ALLOY IDENTIFICATION In addition to a rapid and positive The preferred source for a hand-identification, the X MET is capable portable instrument is a sealed radio. Alloy identification can be defined as of providing alloy elemental composi- isotope which emits x-rays or low a process of ascertaining those tion i.e., concentration of mWor alloy. energy gamma rays. Such sources are characteristics of a given material by ing e(lements displayed as percentages rugged, free from drift problems and which it is definitively recognizable or with their respective element sym- very compact. Typical sources are only known. bol(s)] in about 20 seconds, depend- 8 mm in diameter by 5 mm thick, Various techniques have been used ing on precision and accuracy needed. weighing about 2 grams. Their output over the years for alloy sorting or With mislabeling of delivered alloy is about 6 orders of magnitude less identification. The traditional ones in- materials occurring from a few per- than that of an x-ray tube, which clude color recognition, magnetism, cent up to 10 % or more, and with results in only minimal potential radi-spark testing, differences in apparent the chance of mixups on production ation hazard (for the same reason, the density, and chemical spot tests. More floors, in salvage operations, on job. 'use of classic wavelength dispersive sophisticated methods are based on sites, etc., the economic and product crystal spectrometers with high x-ray thermoelectricity and optical emission liability concerns increasingly justify tube intensities and high power re-spectroscopy. In general, these investment in such rapid, positive quirements is impractical). methods require an experienced op- identification devices. For optimum performance, x-rays erator to complete the identification, must be measured with high geo. based on the results of the measure. metrical efficiency, and the detector ment. The same applies to a conven. PORTABLE X RAY must be capable of discriminating be-tional full scale chemical analysis of FLUORESCENCE ANALYSIS tween x-rays from neighboring ele-an alloy, which must be followed by a ments without further significant loss search through composition tables to X-ray fluorescence spectrometry is a of x-ray photons. Gas 4111ed propor-find the matching alloy name or grade comparative analytical technique tional counters have proven them-designation. which utilizes the physical principles selves over the years as the most relia. The portable, microprocessor-based of the interaction of x-rays or gamma ble detectors used in portable x-ray X-MET offers a real breakthrough by rays with matter. When a sample is analyzers. relieving the operator from decision exposed to a beam oflow energy (1 to Until recently, the n: solution of making. All that is necessary for anal- about 100 kev) x-rays or gamma rays, proportional counters has not been ysis is to expose the sample to the in- the main result is excitation in the good enough to avoid the need for strument for a few seconds, and then sample of the characteristic x-rays of balanced filters. However, new de-read the final identification from the its elements. It is therefore possible to velopments in proportional counter display or printout. Search-match analyze the sample both aualitatively technology (as used in the X-MET) technology is employed which (recognition of elements by their have yielded detectors which signifi-eliminates the need for analysis and unique x-ray patterns) and quantita- cantly improve the resolution (1214 % judgement procedures, tively (amount of element in the sam- for the MnK line). This, coupled with The X-MET provides direct storage pie is proportional to the intensity of a superior microprocessor (Motorola of up to 400 precalibrated alloy signa- its characteristic x rays). 68000) for spectral processing, has tures, and easy replacement by the The X-MET portable x ray analyzer, resulted in the availability of the user can be done on the spot as new configured for alloy analysis and iden. hand-portable X MET x-ray fluores-alloy identification needs arise. Refer- tification, consists of a hand-held cence analyzer capable of simultane-ence signatures may be custom probe and an electronic unit. The ous multielement analysis, named for maximum user con- probe contains an x ray source to ex-venience. Labels such as bin number, cite the sample, and a detector which serial number, melt number, etc. may resolves the x-rays and measures their be used in place of, or in addition to, intensities. The electronic unit accepts alloy common name and/or alloy the signal from the probe, performs proprietary name. all necessary data processing and dis-plays the result. It also contains the power supplies, operator interface and an I/O port for peripherals such as a printer, data logger or personal com-puter (PC). J 3 i i

                                                                                   .                                                              1 X-MET 880                                                                                                                                    l THE X-MET SYSTEM                             searchimatch pattern recogni-tion techniques for alloy identifi-With the X-MET portable alloy analy-                  cation. In addition, the sis system, measurements are totally                  microprocessor provides the                                                    '

non destnictive and can be made un- capability for on-line computa-der extremes of environmental condi- tion of elemental alloy composi-tions ranging from high dust indoor tion. l emironments to very cold, hot or wet , l outdoor emironments. weh*t we e tures a 9(ed e t-her. FIELD PROGRAMMING

                                                                                                                    %w metically seale electronic unit totall-       With a fully calibrated unit (as sup.

ing enly 8.5 kg. It is designed to fit in j a small water repellant backpack for Plied from the factory) operator tra,in-ng is mmimal, and mutine use is ex-y j user convenience in field transport. tremely simple. Howevec unlike other i Pbwer fpr field operations is received p rtable alloy analyzers of earlier de- - lug-in (mm long lasting lead gel-cell, fysign, lastthe X-MET allows for ea.sy and rechargeable batteries that easi over 10 hours of continuous use, rapid field reprogrammmg. A special c de provides operator access to g without the disadvantage of" memory,, effect such as is typical of NiCad type programmmg func,tions which allow n-the-job customized calibrations for batteYX E$ system has reven its either the identification mode (alloy Th worth in hundreds ofinstaflations type) or the assay mode (composition readout). Addition of references in the

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throughout the world. It has prevent- mode is extremely simple. ('Iype ed thousands of costly mix-ups and [D;, ADD", then measure a reference in tialliabill bl b d standard for 100,to 200 seconds).

                                                                                                                                                ,d mg decision-ma           datbor me ur-        Calibration traimng, to allow the user                                                    e gical specification analysis.                 to create special or custom calibra-       '~ 7. -

tions in the assay mode, is provided y by the X-MET representatives world-KEY FEATURES . wide. The keys to such a wide range of capabilities and such high perfor- OPTIONAL ITEMS mance in a portable system are: 1.) The high resolution proportion. To facilitate field programming,

al detector which gives good Metorex offers a set of100 reference To dokument the data from X MET, performance and high reliability alloys, all of which include certificates Metorex offers a lightweight, at ambient temperatures without of analysis showing the participating portable, batte operated terminall the need for x-ray filters or s printer, which directly into the -

labs and verifying each laboratory'in-method b.'. sed on NIST (National X-MET. For data handling liquid nitrogen dewars. 2.) The microprocessorisoftwarel stitute of Standards and 7bchnology) needs, the X-MET results may be col-electronics package which takes traceability. The set covers the 100 lected on a plug-in data logget for most commonly used alloys, that also later readout, or may be interfaced full advantage of the fundamen-tal improvements in detection represent the optimum calibration directly to a PC system. Metorex capability, while allowing simple suite for each alloy type. (Request can supply standard software to facili-Brochure No. BNRM-1 for further tate use and enhance PC data cap-straightforward manlmachine in-terfacing and providing powerful alloy standards information.) ture, as well as, data reduction. i (

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p i , SOURCE OPTIONS r :! N Probe All ying Elements f a type Sources Measured EAPS Cd-109 Ti. V (moderate

 /                                       Q                                                                                                         sensitivity), Cr, Mn, Fe, Co, Ni, Cu,2n, Zr, Nb, Mo, Hf, Ta,
                  -                                                                                                                                W,Pb DOPS Cd-109,           Same as a single Fe 55         Cd-109 source, plus EXCITATION                         Another example of the benefit of a                                          excellent perfor-SOURCE OPTIONS                             dual-source probe is the extended ele.                                       mance for Ti, V ment analysis range offered by the                    DOPS Cd-103,           Same as a single The system is routinely configured                     combination of Am-241 and Cd-109.                              Am 241        Cd 109 source, plus with a Cd-109 excitation source which                  This source combination extends the                                          Rh, Pd, Ag, In, Cd, prosides excel!ent performance for                     elemental analysis range so that it in-                                      Sn,Sb common alloying elements such as                       cludes tin (Sn) in, for example, copper               DOPS Cm-244,           same as chromium (Cr), manganese (Mn), iron                    and titanium based alloys, The combi-                          Am-241        Cd-109/Am-241 dual (Fe), cobalt (Co), nickel (Ni), copper                 nation of Am-241 and Cm-244                                                  source w/slightly (Cu), zine (Zn), niobium (Nb),                          sources allows analysis for elements in                                     reduced sensaivity molybdenum (Mo), zirconium (Zr),                        the same spread as the Am 2411                                              for Nb and Mo, in-tungsten (W) and lead (Pb). A dual-                     Cd-109 range, but with slightly                                             creased sensitivity source probe is also available which                    reduced sensitivity for Nb and Mo and                                       for Ti, V, Cr, Mn, Fe' can contain two excitation sources,                     somewhat incmased sensitivity for Ti,                                       Co, Ni, Cu,2n, Zr, such as Cd-109 and Fe-55. With this                     V, C:; Mn, Fe, Co, Ni, Cu, Zn, W and                                        Hf, Ta W, Pb combination, the performance of the                     Pb. In addition to the important capa-                                      Al Si, P, S' Ti', V

SLPS Fe 55 Cd-109 excitation source, for the ele- bility of Sn analysis, combining the Cr ments listed above, is maintained; Am-241 source with either Cd-109 or Cm-244 allows analysis for other so. SAPS = SURFACE ANALYSIS PROBE l while the elements titanium (Ti) and vanadium (V), which are more effi- , called " heavier"" elements, which SYP? "S R C T EL E PROBE ciently measured usmg the Fe-55 exci- may be important for some alloys. *SLPS with Fe-55 prcudes the hghest poss@e tatien source, are brought into the sensaway (see taue 6)._ i

               " excellent performance" category.

i

                 " K shell ray ene gies above 18 kW are considered to arise from " heavy" elements For example, atomic numbers 44 thru 56 (Ru, Rh, N Ag, Cd,In, Sn, Sb,1h, I, Xc, Cs and Ba).

5

i { ' Source Se.ection for DiTerent E ements l . l Cl Ar R i. Ca Se T4 Y Cr K l At 86 P S Rb Sr Y Er Nb Mb Te Ru Rh Pd AgCd In Sn Sb Te i KeCe Bala ce L Re ir Au Tl Bi At Fr Ae Pa pg55 :::::::: : W Os Pt Hg Pb:Po: Rn : : Re

: :Tb:::U M

[ t f f i a f i i C. se 6 v C, a'. e. e. Ai eu !. e. Le L Pm Eu Tb Ho Tm Lu To Re It Au Tl an I y la Pr E -$ { IIi 4 a i i e i i i i s iiiii a i Nd $m Gd Dy Er Yb Hf W Os Pt Hg Pb iisi Be Ce i de Mn Fe Ce Ni de Zn be be As la Pr Pm Em Tb Ho Tm La To Re ir Au TI I de d a Dy kr hbHf Y be Pt Hg Pb K:re : : C. Hi Cu z. i

c. ce e Hf W Os Pt Hg Pb L ;:::::: :::?

To Re ir Au Tl h . Fe't

                                                                     '             a l

3l u l l l 5 l c g I l~ j l 9 3 4 5 6 7 8 10 1,5 2 The elements shown on the empha- The basic rule for selecting the CHOICE OF SOURCE sized line comprise the elements for source with the help of the table is To span the whole element range which the source is best suited. The that the K-lines in the emphasized three different isotope sources are elements on the unemphasized line areas are to be preferred. needed. Four sources with different Can be analyzed, though normally activities are available, with reduced accuracy. 6

i~ , 2 1 PERIODIC TA LE OF ELEMENTS He H 40 10 6 8 9 10 ' 5 3 4 Asome 8 C l7N O F Ne Na* 19 0 20 2 Li Be 10 8 12 0 14 0 16 0 I 69 90 SYMBOL 13 14 15 16 17 18 Amme Al Si P S Cl Ar 11 12 Na Mg 27 0 28 1 31 0 32 1 35 5 39 9 23 0 24 3 31 32 33 34 35 36 28 29 30 19 20 21 22 23 24 25 26 27 Cu Zn Ga Go 74 As Se Br Kr K Ca Sc Tl V Cr Mn Fe Co Ni 63 5 65 4 69 7 72 6 9 79 0 79 9 83 8 52 0 54 9 55 8 58 9 58 7 39 1 40 1 45 0 47 9 50 9 49 50 51 52 53 54 44 45 46 47 48 38 39 40 41 42 43 in Sn Sb To I Xe 37 Zr Nb Mo Tc Ru 102 Rh Pd Ap9 Cd 114 8 118 7 1218 127 6 126 9 131 3 Rb Sr Y 98 101 1 9 1064 10 112 4 J 85 5 87 6 88 9 91 2 92 9 95 9 82 83 84 85 86 77 78 79 80 61 56 57* 72 73 74 75 76 Pb B1 Po At Rn I 55 Cs Ba La Hf Ta W Re Os ir Pt Au Hg' 'Tl

                                                                                                                                          "  ' ' ' 2'                   '     "'

J '' "'5'''''''''''''''5' K ',', ',, 69 70 71

    ** *'                               Fr Ra Ac                                                      62       63      64     65       66      67      68 Yb       Lu 59    60      61 Tb Dy Ho                 Er     Tm Pr Nd Pm Sm Eu Gd 158 223 226        227 173 0 175 0 m               v                                                                        145   150 4 152 0 1573             9 is2 5 164 9 te7 s 168 9 140 1 140 9 144 2                                                                       101     102      103 H                  Le                                                                                               96    97        98      99     100
                                                                   "    90        91    92     93      94       95 Am      Cm 8k            Cf      Es Fm Md No                     Lr 4

Th Pa U Np Pu 251 254 257 256 254 257 238 0 237 244 243 247 247 232 0 231 As he Br Kr Rb Sr k dr bb ko e u l Bi At Fr Ac Pa

     >. . . .......                                Cd'09 Pb Po Rn RsTh U Tr      Rh         Ag       in Eb I              Ce la Pr Pm Eu Tb Ho, To Nb Ce Nd          Sam bd Dy Er a Se     hr kr b Sr   hr           le      Ru       Pd        Cd         Sn Te        e Pb Po Rn Re Th U g 241

a. . . . . iii.

E ' L ' ir' E P. L Arr 2" Cm 2 Cd* li

            .       . i  di 15 i . . .             i 20 i    i i i ii...i 30 e

40 i bi 60 kev i . SOURCE SPECIFICATION Source activities are in the range Preferred 00 M th p outputs Element Ranges Detector h 5O , Half life Emission ptope K linc3 ' Llines Mn K X rays SIV Nb Ce No ! Fe-55 2.7 years La-Pb Ar 17.8 years Pu L X-rays TI-Se Cm-244 Cr Mo Tb-U Ar Cd-109 Am-241 1.3 years 433 years Ag K X-rays Gamma rays Zn-Nd Hf-U Ar etOrGX l at 59.6 kev l 7 t e

                                                                                                               ~                                                               '

l l l i l l . APPLICATIONS (or sources with dual source probe) source. With this configuration, the I . and a 7 kg microprocessor-based elec- X MET can positively identify 303 Examples of Applications are: tronic unit which processes and stores from 304 with 100 % success. A simi-O Incoming Inspection all data. The probe can measure sizes .lar example is the palt 410 and 416 O On-Site Alloy Venfication both smaller and larger than the stainless steel, which also differs by 0 Quality Control measuring aperture while maintaining 0.3 % sulphur and can be handled in O Stock Control correct identification (l.D.) results. the very same mannec It should be O Scrap Upgrade and Classification Although the probe must be in con- noted that sulphur and the other ele-O Melt Analysis tact with the sample and must cover ments of atomic number greater than O Weld Analysis the measuring aperture for quantita- 12 can be measured with this probe O Maintenance Assessment tive results (elemental composition), using an air path. O Construction Site PMI(Positive the alloy soit (l.D.) mode results are in one case, this light element Material Identification) automatically corrected for undersize probe proved its excellent sensitivity samples (below 6x21 mm) and non- for titanium in steels by analyzing the Products analyzed include all sizes, contacted samples (up to 19 mm away residual metal particles in sandpaper shapes, and finishes. fmm the probe). One extreme exam- and consistently identif>ing 304-vs-321 ple of an undersize sample that identi- correctly in 5 seconds based on the Examples of Products Analyzed are: fies correctly in 5 seconds with no Ti signal from the alloy particles re-O Sheets O Blades special procedure (such as taping a tained on the sandpaper (request the O Ingots O Fasteners bundle of rods together) is the separa- detailed report from your sales agent). , O Billets O Valves tion of 0.8 mm diameter Inconel weld. With such high sensitisity for tita- 1 O Castings O 'Ihnks ing rods. nium, the task of separating 304 from i O Plates O Sludges The standard alloy system can be 321 can be accomplished in just one O Rods 0 Turnings upgraded any time by the addition of (1) second with 100 % confidence. O 'Ibbes O Powders a Surface Light Element Probe. The Using a standard surface anal > sis O Bars O LiquidDigestions unique capability to analyze probe with Cd-109 source, the sorting O Bolts O Cutting Oils " light""* elements with a portable routine has proved itself also to work , analyzer with no "special" considera- with a 100 % success rate on separa-For most products, little or no sample tion (such as helium purge or vacuum tion of stainless steel 303 and 303Se, preparation is required: simply place path) allows such presiously unattain- where the only difference is 0.3 % the pmbe on the material, pull the able field analyses as sulphur (S)in selenium. trigger for a few seconds, then read steel, silicon (Si) in aluminum, Si and out the results. Ti in nickel based alloys, etc. It is well known that the stainless PERFORMANCE Examples of Applicable Industries in- steels 303 and 304 differ only by clude: 0.3 % sulphur, which makes separa. Examples of performance for a wide O Fossil and Nuclear Power tion of these two grades extremely range of applications are given in the O Metallurgical Manufacturing in- difficult and challenging. However, the following tables. Note that the data dustry identification mode handles this task given represent typical precision and very well with the use of a surface RMS values. In many cases, even O Metal Service and Distribution Centers light element probe and a Fe-55 better values can be achieved. O Metal Scrap Recycling Opera- tions .: '. 'Tg ,9'~' O Foundries A !

                                                                                                      ....,m          -
                                                                                                                             ~% - ad*:s                    -

7 ~ - w .aL O Analysis Senice Labs , c i i- .

1 O Chemical Process Industries O Construction Engineering O Refining and Petrochemical I f .

a g,y . O Pulp and Paper . i f - *y; P 4

                                                                                                                                                             ~'          ~

O Metals Fabrication ,i O Military Hardware m g Mjg.j UNIQUE CAPABILITIES ~ h

                                                                                                                   'I
                                                                                                                                                           .1 '            .

The standard alloy system consists of a 1.5 kg trigger-actuated probe with

                                                                                                                                                          ;t~.  -

6x21 mm measurement area (other - J sizes optional), an excitation source *1 1

K shell x-ray enerspes below 5 kev are considered to be thow of the light elements, for example. atomic numbers 13-23 (A1. St. P. S. Ct. Ar. K. Ca. Sc. Ti and VL 8

l TABLE 1.

SUMMARY

OF THE RESULTS OF IDENTIFICATION PROCEDURE identification Results for Alloy Group In Model Measured Elements Typical Useage

                                                                                                      % Feasible Ti, Cr, Fe, Co, Nickel Alloys (21 ref.)                Ni, Cu, Nb,                                          100.0           -

Mo, W Mn, Fe, Ni, Copper Allop (15 ref.) Cu,2n,Pb, 90 to 100 Sn High Temp. k ;,MCu, 90 to 100 hanr Nb, Mo Cr/Mo Steels (6 ref.) Cr, Fe, Ni, Mo 95 to 100 Carbon and Low Alloy Cr, Fe, Ni, Mo 65 Steels (9 ref.) Ti, V, Cr, Titanium Alloys (16 ref.) Mn, Cu, Zr, 95 to 100 Mo Aluminum Alloys (8 ref.) Cu Within thegoue dentifcation Several indivdual models were set up for varcus alloy groups using a probe equipped with a 5 mci Cd 100 souroa in each model first a library of reference alloys was created by measunng each reference for 200 asc. Then each role-rence sample was measured. as an unknown, at least ten times for 5 sec and the percentage of correct or incorrect dentrfcatons was recorded. The percent correct or incorrect contifcaton within each alloy group was calculated for all alkhs tested within a given alloy group The results are hated in the abcMi Tabla X-MET SYSTEM PERFORMANCE DATA FOR CARBON, LOW ALLOY

TABLE 2. s AND Cr/Mo STEELS j 1 l MEASUREMENT CONDITIONS: Probe: DOPS: Slot Aperture, Dual Source Meas. Time: 300 sec. Source Cm-244 Cd-109 1 Element Cr Mn Fe Ni Cu Mo Concen'Jadon 0-35 % 0-1 % O-1 % Range 0-9 % 0-1 % 90 -100 % RMS Error a) D7 % .13 % 45 % .2 % D6 % D15 % Precision of b) D6 % .13 % .24 % .1 % D15 % D10 % Measurement J l a) Root Mean Square Error a around the cafibraton nne feed (t.SO Method) to the expenmental data poents. b) One standard doveton due to counting statstos; value reponed is vald for the measurement trne gwn in bp of l l the table. 9

AND STAINLESS STEELS . MEASUREMENT CONDITIONS: Probe: DOPS; Slot Aperture. Dual Source Meas. Time: 100 sec. Source Fe-55 Cd-109 j Elernent Ti Cr Mn Ni Cu Nb Mo centration 0-2 % 0-25 % 0-15 % 0-35 % 0-35 % 0-1 % O-3.5 % RMS Error a) D9 % 30 % .40 % 35 % .14 % D1 % D2 % i sen ) e ent DS % .15 % . .20 % 35 % .14 % D1 % D1 % a) Root Mean Square Error a around the cahbraton 6ne feed (LSO Method) to the experdtal data points. o b) One standard devaton due to counting statstes; wius reponed a veld for the rnessurement tune given n top of the table. TABLE 4. X-MET SYSTEM PERFORMANCE DATA FOR NICKEL, COBALT ALLOYS MEASUREMENT CONDITIONS: Probe: SAPS; Slot Aperture Source: Cd-109 Meas. Tirne: 100 sec. Element Cr Fe Co Ni Cu W Nb Mo knge" 0-30 % 0-67 % 0-60 % 0-100 % 0-32 % 0-15 % 0-5 % 0-28 % RMS Error a) .65 % 315 % .80 % 1D % .40 % 38 % .04 % .18 % s ) hr9 .15 % .20 % .40 % .5 % 30 % .11 % D2 % .18 % nt I a) Root Mean Square Error around the cahbraton kne fmed (LSO Method) b the expenmental data ponts. b) One standard devoton due e counting statstes; value reported a vahd for the measurement time given l n top of the table. j TABLE 5. X MET SYSTEM PERFORMANCE DATA FOR COPPER ALLOYS (BRONZES & BRASSES) MEASUREMENT CONDITIONS: Probe: DOPS; Slot Aperture, Dual Source Meas. Tirne: 100 sec. Source Cm-244 Am 241 Element Fe Ni Cu Zn Pb Cd Sn oncentration 0-5 % 0-30 % 60-100 % 0-40 % 0-8 % 0-0.1 % 0-1 % RMS Error a) .15 % 35 % .45 % 31 % 3% D20 % DOS % 0,gs;;f,,1e) .,* .25 % .45 % .25 % .i % o08 % o05 % a) Root Mean Square Error around the cahbraton kne feed (LSO Method) to the experrnental data pointt b) One s'andard deviaton due e countno statstes; value reponed a veld for the rnessurement trne gren 10 m ion or tne teei.

4 l j i TABLE 6. X MET SYSTEM PERFORMANCE DATA FOR ALUMINUM ALLOYS I Probe Source SLPS (No)/Fe-55 DOPS (Hp Ar) dual sourceCM-244' + Arn-2418 Time 240 seconds f 60 seconds I Elements Si Ti Cr Mn' Fe' Ni' Cu' Zn' Pb' Sn' 0-12 % 0 .16 % 0 .25 % 0-A % 0-1.2 % 0-5 % 0-5 % 0-35 % 0-35 % 0 .25 % noe RMS Error a) .22 % D015 % D2 % D4 % D4 % m% m% De % D6 % D2 % eson b) g .20 % D01 % DOS % D3 % D1 % D04 % DOS % DOS % DOS % 403 % a) Root Mean Souare Error 2 around the cahbraton hne feed (LSO Method) to the experrnental data porna. b) One standard devaton due to counting stats +cs; wiue reported a vand for the measurement sme gran m top of the tabe TABLE 7. X MET SYSTEM PERFORMANCE DATA FOR CARBON, LOW ALLOY ANrsCr/Mo STEELS MEASUREMENT CONDmONS: Probe: SAPS; Slot Aperture Source: Cd-109 Meas Time: 300 see Element Cr Mn Fe Ni Mo tration Ra ge 0-9 % 0-1 % 90 -100 % 0- 35 % 0 .1 %

RMS Error a) .2 % 35 % .45 % .15 % D15 % e e nt .2 % .% 3% .1 % 10 % a) Root Mean Square Error e around the cahbraton kne fitted (LSO Method) to the experrrental data poens. I b) One stancard eevaten due to counting statstes; wiue reported a vehd for the measurement time given I in top of the tabh TABLE 8. X MET SYSTEM PERFORMANCE DATA FOR TITANlbM ALLOYS MEASUREMENT CONDmONS: Probe: SAPS Source: Cd 109 Meas Time: 100 sec . I Element V Cr Ni Zr Nb Mo Concentration Range 0-8 % O-6 % O-1 % 0-5 % 0-2 % 0 -15 % ea e nt a) .2 % .2 % 3% D2 % D01 % M I l a) One standard devoton due to counting statstos; value reported is.vehd for the messarement tune giwn n top of the tabla 11

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  '                           Request for Additional Information
      -                          Shearon Harris Power Plant, Docket No. 50-400 Request for License Amendment, " Spent Fuel Storage," dated December 23,1998 The following request for additional information (RAI) is for the purpose of developing an inspection plan and performing safety evaluations for the spent fuel pool (SFP) 'C' and 'D' piping, as described in the licensee's submittal, by letter dated December 23,1998.

The term " original construction", as used herein, applies to the construction performed under the licensee's N certificate. The term " weld" applies to welders, weld joint, and all material associated with the weld. ,. j

l. Existing Piping System !

A. Detailed descriotion of the oroDosed chanae: . y

1. Provide isometric drawings (isometrics) showing all piping and piping systems withi e S

the scope of the proposed alternatives; i.e., for fuel poo! cooling and cleanup syste (FPCCS) and component cooling water system (CCWS) piping and for continuance of design and construction without an N stamp.

2. Provide weld matrixes that list all the welds (each weld should be uniquely identified and traceable to I.A.1. above) within the scope of the attematives.

3. In the matrixes, or on the isometrics, identify the piping material (ASME/ ASTM Specification), weld material (ASME/ ASTM Specification), the existence of all required material documentation, and any specific missing documentation. Identify each missing document for each weld. Identify the method (s) used for reconciliation of each type of missing document (e.g., missing Certified Material Test Report reconstructed with complete chemical analysis run on shavings taken from the material). For the sampling and testing methods used for reconciliation, identify references used for guidance (i.e., NRC DG-1070, ASME, or EPRI). Explain any differences between the sampling / testing methods and the selected referenced guidance.

4. In the matrixes or on the isometrics, identify inaccessible non-embedded welds and embedded welds (all other welds should be accessible).

1

5. On the isometrics, indicate the specific location of each weld listed in 1.A.2. and identify the boundaries of the systems that are considered safety related. Identify all non safety related items that appear on the isometrics.

6. Identify in the matrixes, or on the isometrics, the welds that will be or have been ' :

inspected or reinspected that have Code documentation, welds that have been l OPTONAL FORM 99 (7 90) T f 2.N 1 F AX TR ANSMITTAL . e n,.

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, NSN 7540-01-317-7308 5099-101

I Request for Additional Information Shearon Harris Power Plant, Docket No. 50-400 Request for License Amendment, " Spent Fuel Storage," dated December 23,1998 inspected that do not have Code documentation, and welds that will be or have been inspected or reinspected not to Code. For the welds that will be or have been inspected or reinspected but not to Code, describe the inspection technique, acceptance criteria, and documentation. Identify the edition and addenda of ASME Code that will be or has been used for the above inspections and reinspections.

7. Identify any non safety related items installed during the original construction that will be upgraded to safety related status by 'his amendment; e.g., will any of the non g,4 safety related ANSI B31.1 piping (Enclosure 8, page 7 of the submittal) be upgraded?

8. Identify any commercial grade items installed during original construction. If dedication was used during original construction, is documentation of the dedication N,4 program available for review? Are the dedication packages for items available for review?

9. Identify any commercial grade items requiring dedication that will be used to complete construction.

10. Was the piping system constructed in accordance with a 10 CFR Part 50, Appendix B program? Is the construction Appendix B program documentation available for review? If construction was performed under a different program, identify the program. Is the program documentation available for review?

11. Are the work control procedures and hold point sign-off documents from the original construction available for review? If these documents are required by Code, which ,

documents are missing?

                                                                                                                . W)

12. Provide a list of the qualified welders who worked on the original construction and identify the ones qualified to weld stainless steel piping. Are historical qualificatio records for these welders available for review? If not, provide an explanation to

                                                                                                              )[,
                                                                                                            - g support acceptability. For welds missing welder identification, how will weld integrity be established?

B. Aeolicable reaulations for welds and oicina systems within the scope of the oroDosed alternatives

1. Identify the edition and addenda of Code and any code cases that were used for original construction of the welds and piping systems.

2 l

e

  /                                 Request for Additional Information Shearon Harris Power Plant, Docket No. 50-400 Request for License Amendment, " Spent Fuel Storage," dated December 23,1998

2. Identify the edition and addenda of Code and code cases that will be used to complete construction of the piping systems. Identify any exceptions to Code requirements and justifications for these exceptions.

3. Identify the edition and addenda of Code and code cases that were or will be used for repair and replacement of welds and piping?

4. Provide a matrix (See I.A.2.) that identifies the specific pEragraphs in Code applicable to each weld. Identify documentation deficiencies for each weld. Identify any exceptions to Code requirements? Provide attematives and justifications for (t these exceptions.

5. Identify the ASME requirements, including administrative requirements, that were completed prior to stoppage of the original construction of the piping systems. Is documentation of these completed requirements available for review? What ASME data reports were filed and their filing dates?

6. Identify ASME survey inspections conducted prior to stoppage of the original construction of the piping systems. Are documented results available for review? [VP 7.

V4 f ['i , /~. Identify third party inspections (e.g., Hartford, ANI) conducted prior to stoppage of the original construction of the piping systems? Are these reports available for /

      .             review?

8. With regard to piping system components / services performed by others, are documented validations of these vendors services available for review?

ll Completion of Piping System (General) ,e

1. Does CP&L have an active construction permit for Shearon Harris? /

2. Identify the differences between HNP's proposed construction program to complete the SFP C and D and the original construction program under HNP's N certificate.

How will these differences be reconciled?

3. Will data packages be prepared?

4. What third party verification is planned?

3

~..

 #                               Request for Additional Information Shearon Harris Power Plant, Docket No. 50-400 Request for License Amendment, " Spent Fuel Storage," dated December 23,1998 Ill. _ Specific Comments on Submitted Information (Enclosure 6, December                   ubmittal)            i

1. What was the basis for selecting the four extemally accessible field welds for intemal examination (p6/7)? Identify these welds in the matrix provided in response I.A.2 above.

                 /2 s With reference to the " substantial portion of the embedded piping and field welds"                l v'    (p7), identify these welds in the matrix provided in response 1.A.2 above.

3. Provide the inspection procedure used for remote inspection of embedded welds.

4. With reference to the remote inspection of the embedded welds, identify the critical characteristics that will be verified and the acceptance criteria to be used. 'vg[.f g

5. Provide the results of the remote inspection with any ide sied discrepancies. 8h v

6. Provide a completed weld data report, representative of those that were discarded (analogous records exist for the licensed unit). Identify the critical characteristics

and explain how, in lieu of records, each will be validated (see I.A.3. and I.A.11. above).

7. With reference to the procurement specification (SS-021, Purchasing Welding ,

Materials for Permanent Plant Construction) (pg), did other specifications for other l filler materials exist? What assurances are provided that these other filler materials j were not used for the embedded piping? !

8. Provide any updates / supplements to the Alternative Plan (p 10) as they become available. ;

9 With referenced to the "large percentage of embedded field welds" that will be ) inspected (p 10), identify these welds on the matrix provided (see 1.A.4. above). l Provide technical justification for not examining the remaining welds.

10. Explain what is meant by the statement that internal examination of the embedded weids provides a mecsure of quality assurance beyond Code requirements (p11). ,

What additional physical or material attributes will be verified? :

                 ' 11. The submittal refers to opinions by Bechtel and Hartford Steam Boiler concerning the benefits in accordance with an N certificate program (p.12). Are these opinions documented and available for review?

4

l

1 Request for Additional Information Shearon Harris Power Plant, Docket No. 50-400 Request for Lic'ense Amendment, Spent Fuel Storage," dated December 23,1998 i

12. Provide a matrix comparing the specific ASME Section 111 requirements with the Corporate QA Program.

8

         'b\ 13. Provide a matrix comparing the specific ASME Section lli requirements with the corresponding Section XI requirements.

J

14. Provide documentation of the referenced comparison (p12) of approved ASME 1 Section 111 Construction QA Program Manual with the effective Corporate 10CFR50, Appendix B QA Program. p w be# 5't

15. Provide documentation of the supplemental quality assurance requirements that have been developed (p13) specifically for the purpose of addressing differences between ASME Section lli quality assurance requirements and the Corporate 10CFR Appendix B QA Program.

i 5

p 8

                /          \                       UNITED STATES g          ,g           NUCLEAR REGULATORY COMMISSION 2                  WASHINGTON, D.C. 20565-0001
                \, *..../

FACSIMILE TRANSMISSION DATE: d/ ) ' I TO: g - FAX NO: 911 - 362. - Z~7 o / TEL NO: %f.In.-Zr]o FROM: klCk Mr.4F1C 4-U.S. NUCLEAR REGULATORY COMMISSION l OFFICE OF NUCLEAR REACTOR REGULATION FAX NO.: (301) 415-2102 TEL NO: 16/ y/5- /373 I PAGE 1 OF 2 -PAGES l i REMARKS: Kevin -

                          #ert. qfe. que ob3 h dsustwe dun 9 A               ,

Co.,ferew. Ca 0 ts h hari 3'-

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D' i REQUEST FOR ADDITIONAL INFORMATION SHEARON HARRIS SPENT FUEL STORAGE CAPACITY INCREASE Reactor Systems Branch l^ l 1) Although the bumup criteria for storage in Pools C or D will be implemented by administrative procedures to ensure verified bumup prior to fuel transfer into these pools, an administrative failure should be assumed and evaluation of a fuel assembly l misloading event (i.e., a fresh PWR assembly inadvertently placed in a location I

                   .estricted to a bumed assembly as per TS Fig. 5.6.1), should be analyzed.

2) How will the bumup requirements needed to meet TS Fig. 5.'6.1 be ascertained for fuel assemblies shipped from other PWR plants (Robinson)?

3) The fuel enrichment tolerance is specified in Section 4.5.2.5 as +0.0/-0.05. Why isn't a positive tolerance of +0.05 assumed (i.e.,5.0+0.05 weight percent U-235)?

4) Justify that the allowance that was assumed for possible differences between the fuel vendor and the Holtec calculations is sufficient to also encompass bumup calculational uncertainties.

5) The summary of criticality safety calculations shown in Tables 4.2.1 and 4.2.2 indicate

(' . that the total uncertainty is a statistical combination of the manufacturing tolerances but do not indicate methodology biases and uncertainties. Were these included? I 1 i i

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   . po            h                              UNITED STATES c

g j j NUCLEAR RESULATORY COMMISSION WASHINGTON, D.C. 20665-0001 o

        %, . . . . . p#                                  March 24, 1999 Mr. James Scarola, Vice President Shearon Harris Nuclear Power Plant Carolina Power & Light Company Post Office Box 165, Mail Code: Zone 1 New Hill, North Carolina 27562-0165

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE ALTERNATIVE PLAN FOR SPENT FUEL POOL COOLING AND CLEANUP SYSTEM PIPING - SHEARCN HARRIS NUCLEAR POWER PLANT (TAC NO. MA4432)

Dear Mr. Scarola:

By letter dated December 23,1998, you requested a license amendment to revise Shearon Harris Nuclear Power Plant Technical Specification (TS) 5.6, " Fuel Storage," to increase the spent fuel storage capacity by adding rack modules to pools 'C' and 'D.' Enclosure 8 of your submittal provided a detailed description of the propossd altsmatives to demonstrate compliance with ASME B&PV Code requirements for the cooling and cleanup system piping in accordance with 10 CFR 50.55a(a)(3)(i). During the course of its review, the NRC staff has determined that additionalinformation is necessary to complete its review. The enclosed request for additionalinformation regarding

          . your proposed alternative plan was discussed with your Licensing staff on March 9,1999. A mutuattj agreeable target date of April 30,1999, for your response was established. If circumstances result in the need to revise the target date, please call me at the earliest opportunity.

Sincerely, Richard J. Laufer, Project Manager Project Directorate 113 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-400 l

Enclosure:

As stated cc w/ encl: See next page

p.

   '       ?                                     Shearor. Harris Nuclear Power Plant I'    Carolina Power & Light Company          Unit i
         .cc:

f,t. William D.' Johnson . l Vice President and Corporate Secretary Director of Site O , Carolina Power & Light Company Carolina Power &perations Light Company , i- Post Office Box 1551 Shearon Harris Nuclear Power Plant j - Raleigh, North Carolina 27602 Post Office Box 165, MC: Zone 1 L New Hill, North Carolina 27562-0165 l Resident inspector / Harris NPS c/o U.S. Nuclear Regulatory Commision Mr. Rot, art P. Gruber 5421 Shearon Harris Road Executive Director New Hill, Wrth Carolina 27562-9998 Public Staff NCUC Post Office Box 29520

        - Ms. Karen E. Long                       Raleigh, North Carolina 27626 l          Assistant Attorney General ,

Stats of North Carolina Chairman of the North Carolina

Post Office Box 629 Utilities Commission Raleigh, North Carolina 27602 Post Office Box 29510 Raleigh, North Carolina 27626-0510 Public Service Commission State of South Carolina Post Office Drawer .

Mr. James Scarola - Columbia, South Carolina 29211 Vice President-Harris Plant l Carolina Power & Light Mr. John H. O'Neill, Jr. . Post Office Box 165, MC: Zone 1 Shaw Pittman, Potts & Trowbridge New Hill, NC 27562-0165 2300 h Street, NW. Washington, DC 20037-1128 Mr. Vernon Malone, Chairman l Board of County Commissioners of Wake County P. O. Box 550 Mr. Mel Fry, Director Raleigh, North Carolina 27602 ' Division of Radiation Protection l N.C. Depc-tment of Environment and Natural Resources 3825 Barrett Dr.- . Mr. Richard H. Givens, Chairman Raleigh, North Carolina 27609-7721 - Board of County Commissioners , of Chatham County P. O. Box 87 Mr. Terry C. Morton - Pittsboro, North Carolina 27312 Manager 4 Performance Evaluation and Ms. Donna B. Alexander, Manager '

          - Regulatory Affairs CPB 9              Regulatory Affairs Carolina Power & Light Company          Carolina Power & Light Company Post Office Box 1551                    Shearon Harris Nuclear Power Plant Raleigh, North Carolina 27602-1551      P.O. Box 165, Mail Zone 1 New Hill, NC 27562-0165 Mr. Bo Clark Plant General Mana er- Harris Plant     Mr. Johnn H. Eads, Supervisor Carolina Power & Li ht Company          Licensin     egulato Programs Shesron Harris Nuc ar Power Plant       Carolina o var & Li ht Company P.O. Box 165                            Shearon Harris Nuc ar Power Plant New Hill, North Carolina 27562-0165     P. O. Box 165, Mail Zone 1 New Hill, NC 27562-0165

 ,                            Request for ' Additional Information Shearon Harris Nuclear Power Plant alternative plan for spent fuel pool cooling and cleanup system piping The term " original construction," as used herein, applies to the construction performed under the licensee's N certificate. The term " weld" applies to welders, weld joint, and all material associated with the weld.

I. Existing Piping System A. Detailed descriotion of the orocosed chanae:

1. Provide isometric drawings .(isometrics) showing all Code-related piping and piping systems within the scope of the proposed alternatives; i.e., for fuel pool cooling and cleanup system (FPCCS) and component cooling water system (CCWS) piping.

Provide isometric drawings to be used for continuance of design and construction without an N stamp.

2. Provide weld matrixes that list all the welds (each weld should be uniquely identified and traceable to 1.A.1. above) within the scope of the alternatives.

3. In the matrixes, or on the isometrics, identify the piping material (ASME/ ASTM Specification), weld material (ASME/ ASTM Specification), the existence of all required material documentation, and t.ny specific missing documentation. Identify each missing document for each weld. Identify the method (s) used for reconciliation of each type of missing document (e.g., missing Certified Material Test Report reconstructed with complete chemical analysis run on shavings taken from the material). For the sampling and testing methods used for reconciliation, identify references used for guidance (i.e., NRC DG-1070, ASME, or EPRI). Explain any differences between the sampling / testing methods :and the selected referenced guidance. For chemical analysis, identify sample size and chemical analysis (mean and standard deviation for each element) for each analyzing technique.

4. In the matrixes or on the isometrics, identify inaccessible non-embedded welds and embedded welds (all other welds should be accessible).

5. On the isometrics, indicate the specific location of each weld listed in l.A.2. and identify the boundaries of the systems that are considered safety-related. Identify all non-safety-related items that appear on the isometrics.

l

6. Identify in the matrixes, or on the isornetrics, the welds that will be or have been inspected or reinspected that have Code documentation, welds that have been inspected that do not have Code documentation, and welds that will be or have been inspected or reinspected not to Code. For the welds that will be or have been inspected or reinspected but not to Code, describe the inspection technique, acceptance criteria, and documentation. Identify the edition and addenda of ASME Code that will be or has been used for the above inspections and reinspections.

1

l 1 1 l - i Request for Additional Information Shearon Harris Nuclear Power Plant alternative plan for spent fuel pool cooling and cleanup system piping l Identify any non-safety-related items installed during the original construction that will be upgraded to safety-related status by this amendment; e.g., will any of the non-safety-related ANSI B31.1 piping (Enclosure 8, page 7 of the submittal) be upgraded?

8. Identify any commercial grade items requiring dedication that were installed during original construction. For these items, is documentation of the dedication program available for review? Are the dedication packages foriterhs available for review? 1

9. Identify any commercial grade items requiring dedication that will be used to complete construction.

10. Was the piping system constructed in accordance with a 10 CFR Part 50, ;

Appendix B program? Is the con,iruction Appendix B program documentation ! available for review? If construction was performed under a different program, identify the program. Is the program documentation available for review?

11. Are the work control procedures and hold point sign-off documents from the original construction available for review? If these documents are required by Code, what documents are missing?

12. Provide a list of the weld procedure specifications (WPS) used anc ; heir procedure

       .              qualification records (PORs). For welds missing welder identification, how will weld integrity be established?

B. Acolicable reaulations for welds and oicina systems within the scope of the proposed alternatives

1. Identify the edition and addenda of Code and any Code cases that were used for original construction of the welds and piping systems. If not the same for all the welds, identify the Code requirements for each weld or groups of welds.

2. Identify the edition and addenda of Code and Code cases that will be used to complete construction of the piping systems. Identify any exceptions to Code requirements and justifications for these exceptions.

3. Identify the edition and addenda of Code and Code cases that were or will be used for repair and replacement of welds and piping.

4 Provide a matrix (See I.A.2.) that identifies the specific paragraph in Code that is applicable to missing weld documents. Identify documentation deficiencies for each weld. Identify any exceptions to Code requirements. Provide alternatives and justifications for these exceptions. l 2

r' y Request for Additional Information Shearon Harris Nuclear Power Plant alternative plan for spent fuel pool cooling and cleanup system piping

5. Identify the ASME requirements, including administrative requirements, that were completed prior to stoppage of the original construction of the piping systems, is 4 documentation of these completed requirements available for review? What ASME data reports were filed and their filing dates?

6. Ider.tify ASME survey inspections conducted prior to stoppage of the original l wnstruction of the piping systems. Provide documentation for representative internal / external audits conducted during the peak construction periods for the welds in question (1978-1979), particularly in the areas of work control, welding, material traceability, and records.

7. Identify third party inspections (e.g., Hartford, ANI) conducted prior to stoppage of the original construction of the piping systems. Provide a representative sample of documentation for these inspections.

8. With regard to piping system components / services performed by others, provide documented validations of these vendor services. Provide the documentation of audits of the supplier of prefabricated piping.

Il Completion of Piping System (General)

1. Identify the differences between HNP's proposed construction program to complete the SFP C and D and the original construction program under HNP's N certificate.

How will these differences be reconciled?

2. Will data packages be prepared?

3. What third party verification is planned?

i 111. Specific Comments on Submitted Information (Enclosure 6, Decem' car 28 Submittal)

1. What was the basis for selecting the four externally accessible field welds for internal examination (p6/7)? Identify these welds in the matrix provided in response I.A.2 above.

2. With refeience to the " substantial portion of the embedded piping and field welds" (p7), identify these welds in the matrix provided in response I.A.2 above.

3. Provide a summary of the inspection procedure used for remote inspection of embedded welds.

3 I

C .

a Request for AdditionalInformation Shearon Harris Nuclear Power Plant

          ' alternative plan for spent fuel pool cooling and cleanup system piping

4. With reference to the remote inspection of the ernbedded we!ds, identify the critical 4 characteristics that will be verified and the acceptance criteria to be used.

5. Provide the results of the remote inspection with any identified discrepancies.

l

6. Provide a completed weld data report, representative of those that were discarded l (analogous records exist for the licensed unit). Identify the critica1 characteristics l and explain how, in lieu of records, each will be validated (see I.A.3. and I.A.11.

above).

7. With reference to the procurement specification (SS-021, Purchasing Welding Materials for Permanent Plant Construction) (p9), did other specifications for other filler materials exist? What assurances are provided that these other filler materials were not used for the embedded piping?

l

8. Provide any updates / supplements to the Altemative Plan (p 10) as they become

! available. 9 With referenced to the "large percentage of embedded field welds" that will be

inspected (p 10), identify these welds on the matrix provided (see f.A.4. above).

Provide technical justification for not examining the remaining welds. l

10. Explain what is meant by the statement that internal examination of the embedded welds provides a measure of quality assurance beyond Code requirements (p11).

What additional physical or material attributes will be verified?

11. The submittal refers to opinions by Bechtel and Hartford Steam Boiler concerning l the benefits in accordance with an N certificate program (p.12). Please provide l documented endorsements.

i

12. Provide a copy of the site ASME Section 111 QA program used during original construction.

l

13. Provide a copy of the Corporate QA program that will be used to complete construction. Provide a list of implementing quality control procedures for welder !

qualification, weld procedures, inspections, documentation, etc.

14. Provide a copy of the sup'plemental quality assurance requirements developed to augment the Corporate QA program, which was based on review of the approved l Construction OA Program at the time of construction versus the existing Corporate l QA Program.

t 4

Request for AdditionalInformation Shearon Harris Nuclear Power Plant alternative plan for spent fuel pool cooling and cleanup system piping

15. Provide documentation of the referenced comparison (p12) of approved ASME Section lli Construction OA Program Manual with the effective Corporate 10CFR50, Appendix B QA Program.

16. Provide documentation of the supplemental quality assurance requirements that have been developed (p13) specifically for the purpose of addressing differences between ASME Section lil quality assurance requirements and the Corporate 10CFR Appendix B QA Program.

\ l 5 I

[ Richard _l,suLr - TS_ hah 18.wpd pgg h' '* ygvu) $gpu) % cum 0-i 4 Cover Letter Draft: TQ % (b 4- "'3u -wi- /3M [ 7 gg g

We are currently reviewing your request, submitted by letter date'$ 6ecembe723,3998, for a license amendment to place the Harris Nuclear Plant spent fuel pools *C" and *D" in service. By letter dated April 30,1999, you provided a response to our request for additional information required to complete the review of the proposed alternative piping plan. In conjunction with review of this information, a telecon was held on August 19,1999 to further clarify certain, attached items, for which your staff has agreed to docket its response. l

Weld Material

1. In Enclosure 4, " Metallurgy Unit Report for Spent Fuel Pool Wel Metal Composition analy. sis" of our request for additional information (RA1), explain how the Metorex X-Met 880 Alloy Analyzer discriminates between the different standards that you used in your analysis.

What are the chemical element ranges associated with the different standards that you used? What determines a match on a particular standard. What chemical elements are not included in the " Match" determination and how are these elements reconciled? j l

2. Provide assurance that the ferrite numbers are acceptable for A-No. 8 weld wire (ND-2433) l used in welds with missing weid wire documentation. l 1

3. In Enclosure 6, " Lab Test Reports," of your response to our RAI, explain the chemical

(' analysis in the Table associated with POR 6(c), dated 11/15/84, page 2 of 2, labortatory test No. 9-2149. What roll (s) are associated with the base material, weld, and standard (s)? j What criteria was used to determine acceptability.

4. For the piping and welds examined internally, provide a discussion of the examination results. What inspection criteria is used for evaluating the piping and welds for corrosion and fouling? Describe the corrosion and fouling inspection procedure and inspection personnel qualification process. For the embedded welds not examined internally, describe what is preventing their examination.

5. What are the chemical analysis for steel welds 2-CC 3-FW-207,2-CC-3-FW-208, and 2-CC-3 FW-2097

6. Provide the paper trail that identifies a specific weld material to a specific weld on the isometric drawings, i.e. show that the weld material being verified with the Metorex X-Met 880 was specified for that location. Identify missing documentation that breaks the paper trail,if any.

7. Disc chemical analysis and any other analysis performed on the water in the FPCCS and[CCW of the SFP C and D? Where did the water come from? Discuss any diffeTances between the chemical analysis and any other analysis and the original water source. Provide the staff with representative analysis of the water.

8. In Enclosure 8 "Hydrotest Records for Embedded Spent Fuel Pool Cooling Piping and Field

(, Welds," of your response to our RAI, you provided signed hydrostatic test reports for 13

  'tRigErd Lauf r - TSJROrrt8.wpd                                                                               tregy, E f

D embedded welds. Starting with the signed hydrostatic test report, back track through procedures and program requirements to the point where the missing document (s) were verified as being complete. In another words, identify the specific procedural and program controls requiring verification of completion of the missing documentation (manufacturing / fabrication records, weld data records,' updated isometric drawings, and inspections) starting backward from the hydrostatic test report.

9. Identify the concrete pouring procedure that requires checking for the welder symbol and a successful hydrostatic test before pouring.

10. Describe how the liner leak tests support weld integrity for welds 2-SF-8-FW-65 and2-SF-8 FW-66 (Enclosure 3 of your response to NRC's RAI). For these two welds, back track through procedures and program requirements to the point where the missing documents were verified as bein0 completed.

Condition of Equipment

11. What was the condition of layup for the partially completed piping.

12. Describe precautions that were taken to protect system components (e.g., pumps, valves, heat exchangers, piping) from deleterious environmental effects during layup.

13. Summarize the activities being taken to ensure the acceptable quality of equipment before being retumed to service.

e Embedded Welds

14. Only 6 of the 15 embedded welds will be inspected. Discuss the physical limitations of the inspection equipment that limits inspectability.

15. Why was visual inspection rather than, say, ultrasonic inspection chosen to examine the integrity of the embedded welds?

16. Discuss why the decision not to inspect all of the embedded welds will result in an acceptable level of quality and safety.

Post-Modification Testing

17. Describe the post modification testing to be performed to ensure that the system (s) will satisfy all design regr'rements, include description of hydrotests to verify the

( integrity of the system pressure boundaries, flushing to ensure unobstructed flow l f Ts_hmisDRAFT, September 10,1999 f\

pag, y mingnumwmernsamw _

q- through system components, and preoperational functional testing under design flow / heat loads. I e { ( \. .. - . m .s.,,. ,40.,,,, i

.I

p / M' j tf - l4 DOCKETED l USFOctober 5,1999 UNITED STATES OF AMERICA l , NUCLEAR REGULATORY COMMISSION 9 U -6 All :07 BEFORE THE ATOMIC SAFETY AND LICENS54G BOARD > AbE F In the Matter of )

CAROLINA POWER & LIGHT COMPANY ) Docket No. 50-400-LA

                                                            )       ASLBP No. 99-762-02-LA (Shearon Harris Nuclear Power Plant)                )               '! < -
                                                            )

CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S RESPONSE TO ORANGE COUNTY'S FIRST SET OF DISCOVERY REQUESTS TO NRC STAFF"in the above captioned proceeding have been served on the following through deposit in the Nuclear Regulatory Commission's internal mail system or as indicated by an asterisk, by first-class mail and by electronic mail (e-Mail) transmission where indicated this 5th day of October,1999: G. Paul Bollwerk,III, Chairman Frederick J. Shon Administrative Judge Administrative Judge

      - Atomic Safety and Licensing Board                   Atomic Safety and Licensing Board Mail Stop:T 3F-23                                  Mail Stop: T-3F-23 U.S. Nuclear Regulatory Commission                  U.S. Nuclear Regulatory Commission Washington, DC 20555-0001                           Washington, DC 20555-0001 (E-mail: GPB@NRC. GOV)                              (E-mail: FJS@NRC. GOV)

Dr. Peter S. Lam Office of the Secretary Administrative Judge ATTN: Rulemaking and Adjudications Atomic Safety and Licensing Board Staff Mail Stop: T 3F-23 Mail Stop: O 16-C-1 U.S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Washington, DC 20555-0001 (E-mail: PSL@NRC. GOV) (E-mail: HEARINGDOCKET

                                                              @NRC. GOV) l l

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Office of the Commission Appellate James M. Cutchin, V l _ Adjudication Atomic Safety and Licensing Board Mail Stop: O 16-C-1 - Mail Stop: T 3F-23 .l U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission l Washington, DC 20555-0001 Washington, DC 20555 (E-mail: JMC3@NRC. GOV) Diane Curran, Esq.* John H. O'Neill, Jr.* , Harmon, Curran, Spielberg William R. Hollaway* i

               & Eisenberg, L.L.P.                      Counsel for Lic'eYisee 1726 M Street, N.W., Suite 600            Shaw Pittman Potts & Trowbridge            I Washington, DC 20025                       2300 "N" Street, N.W.

(E-mail: deurran@harmoncurran.com). Washington, DC 20037-1128 (E-mail: john _o'neill@shawpittman.com, I william.hollaway@shawpittman.com) j l i Steven Carr* l Legal Department Carolina Power & Light Co. 411 Fayetteville Street Mall P.O. Box 1551- CPB 13A2 - l Raleigh, North Carolina 27602 (E-mail: steven.carr@cplc.com) Susan L. Uttal Counsel for NRC Staff 1 l L}}

ML20212L144

Contents

Text

SUMMARY

Enclosures:

Enclosure:

SUBJECT:

Enclosures:

SUMMARY

SUBJECT:

Dear Mr. Scarola:

Enclosure:

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