ML20033B513

Index & Certification to Amend 30 to Application for License to Mfg Floating Nuclear Plants

ML20033B513
Person / Time
Site:	Atlantic Nuclear Power Plant
Issue date:	11/23/1981
From:	Cowan B, Keuride OFFSHORE POWER SYSTEMS (SUBS. OF WESTINGHOUSE ELECTRI
To:
References
NUDOCS 8112010483
Download: ML20033B513 (3)

Text

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COLKETED UNITED STATES OF AMERICA INNEC NUCLEAR REGULATORY COMMISSION

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BEFORE THE ATOMIC SAFETY AND LICEN N

R r M u^5 EECRETARY U I'M gjffC b

In the Matter of OFFSHORE POWER SYSTEMS Docket No. STN 50.

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(Manufacturing License for 0 ggy Q Floating Nuclear Power Plants)

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g INDEX AND CERTIFICATION q

In response to the requirements set forth in f0? 1CJ S 50.30, subsection (c) (2), the application in the above-captioned proceeding is hereby amended:

I.

Letter and Oath Letter of A.

R.

Collier dated November 17, 1981 forwarding to the Director of Nuclear Reactor Regulation Amendment No. 30.

II.

Change Pages and Instructions Amendment No. 30, dated November 17, 1981, consists of an un-bound series of change pages (three-hole punched) for inser-tion in the Plant Design Report.

Each set of change pages is accompanied by an outline entitled " Instructions for Entering Amendment No. 30 in the Plant Design Report."

I hereby certify that the above-identified documents are Amendment No. 30 to the Plant Design Report, dated November 17, 1981, and are the most recent update of the application material.

Amendment No. 30, along with the information previously served, constitutes the complete contents of the Plant Design Report of Offshore Power Systems Application for a License to Manufacture Floating Nuclear Plants as of the date of this "Index and Certifi-cation."

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maeaaA,

ih Counsel for Applicant, Offshore Power Systems Dated:

November 23, 1981 6.

8112010483 811123

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DR ADOCK 05000437 PDR k

e UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of OFFSHORE POWER SYSTEMS Docket No. STN 50-437 (Manufacturing License for Floating Nuclear Power Plants)

CERTIFICATE OF SERVICE I hereby certify that copies of Amendment No. 30, dated November 17, 1981, to the Plant Design Report of Offshore Power Systems Application for a License to Manufacture Floating Nuclear Plants was forwarded via U.S. Mail (First Class), post-age prepaid, with contents of each package further identified by a companion document entitled "Index and Certification," this 23rd day of November, 1981, in accordance with the service list identified as Attachment No. 2 and incorporated herein by reference.

7 vic17 y

Coun el for Applicant O

shore Power Systems 8000 Arlington Expressway P.

O.

Box 8000 Jacksonville, Florida 32211 (904) 724-7730 1

e' ATTACHMENT 2 Sheldon J. Wolfe, Esq., Chairman Atomic Safety and Licensing Board U.S.

Nuclear Regulatory Commission Washington, D.C.

20555 Dr. David R.

Schink, Member Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washignton, D.C.

20555 George A.

Ferguson, Member Atomic Safety and Licensing Board School of Engineering Howard University 2300 5th Street, N.W.

Washington, D.C.

20059

.Dr.

David L.

Hetrick, Alternate Member Atomic Safety and Licensing Board Professor of Nuclear Engineering The University of Arizona Tucson, Arizona 85721 Alan S.

Rosenthal, Esq., Chairman Atomic Safety and Licensing Appeal Board Panel U.S.

Nuclear Regulatory Commission Washington, D.C.

20555 Docketing & Service Section (original & two copies)

Office of the Secretary U.S.

Nuclear Regulatory Commission Washington, D.C.

20555 Barton Z.

Cowan, Esq.

John R.

Kenrick, Esq.

Eckert, Seamans, Cherin & Mellott i

42nd Floor,.600 Grant Street Pittsburgh, Pennsylvania.15219 Director-(2 copies)

Division of Nuclear Reactor Regulation U.S.

Nuclear Ragulatory Commission Washington, D.C.

20555 L

I ARC-81-1016 00tKETED eWC Offshore Power Systems 8000 Arlington Expressway 904 -

4-Box 8000, Jacksonville, Florida 32211 Tele x : 568406

? SECitETARY

,utit;G & SERVlCE November 17, 1981 BRANCH Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, DC 20555 Re

NRC Docket No. STN-50-437; Offshore Power Systems' Application for License to Manufacture Floating Nuclear Plants, Amendment No. 30.

A. R. Cotiler President

Dear Sir:

Offshore Power Systems hereby amends its ;pplication for License to Manufacture Floating Nuclear Plants by filing Amendment No. 30, dated November 17, 1981, to the Plant Design Report.

Amendment No. 30 deals solely with the subject matter of the Plant Design Report previously submitted and incorporates the site envelope change agreed upon at the October 16,1981 m5eting of the Advisory Comittee on Reactor Safeguards. Amendment No. 30 also incorporates the containment design comitment contained in Mr. P.B.

Haga's letter FNP-PST-078, dated September 17, 1981.

Amendment No. 30 also includes other changes and errata to the text of the Plant Design Report. Instructions for the use of replacement pages are provided.

Amendment No. 30 consists of:

1.

Three (3) originals of this letter.

2.

Sixty-seven (67) conformed copies of this letter.

3.

Seventy (70) sets of change pages, described above.

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Page #2 -_ Director of Nuclear. Reactor Regulation Consonant with 10 CFR 2.101, this amendment to the Offshore Power Systems' Application.for License to Manufacture Floating Nuclear Plants is being served on those persons identified in Enclosure 2 to the Nuclear Regulatory Commission letter signed by Roger E. Boyd and dated August 6,1976, including the Honorable Jake M.

Godbold, Mayor of the City of Jacksonville, Florida, s/ A. R. Collier A. R. Collier President Attest:

s/ Y. W. Campbell V. W. Campbell Secretary Sworn to and subscribed before me, this 17th day of November, 1981.

s/ Joyce Faye Smith Joyce Faye Smith Notary Public, State of Florida at Large My commission expires: 10/05/82 l

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00ChETED 0FFSHORE POWER SYSTEMS USNRC O

DOCKET NO. STN 50-437 INSTRUCTIONS FOR ENTERING AMENDMENT NO. 30W NOV 25 P5:i2 IN THE PLANT DESIGN REPORT "T r SECRET /SY

-i::c & service i

1.

Remove and insert pages in the text of the Plani. Design # Ep@t in R

accordance with the following tabulation.

Remove Page(s)

Insert Page(s)

Chapter 1 (Vol. 1) 1.5-13, 14 1.5-13, 14 Chapter 2 (Vol. 1) 2.1-7, 8 2.1-7, 8 2.1-9, 9a 2.1-9, 9a 2.1-15 2.1-15 O

Appendix C (Vol. 8)

C-124, 125 C-124, 125 C-139, 140 C-139, 140 O

Amendment 30 November 17, 1981

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Supporting Site Information (Hydrology, Geology and Seismology)

The development of the site envelope parameters for the Floating Nuclear Plant has been accompanied in many cases by an investigation of site data and site evaluation techniques over a wide range of potential sites.

Included has been data on maximum water levels under various conditions of storm and tide and generic reviews of geologic and seismic requirements for sites on the Atlantic and Gulf coasts.

In Amendment 4 further information is provided on such subjects as the state-of-the-art in geophysical exploration and geophy-sical techniques applied to definition of stability and freedom from subsidence; 4

a the seismic relationship between the Continental Shelf and the Coastal Plain, 1

j and the relationship between seismic events and tsunamis.

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Development work of this nature will continue as considered necessary to I

support the envelope of postulated site parameters for the Floating Nuclear j

Plant.

f Core Ladle Information e

Information required to finalize the design of the core ladle is described in 30 i

the core ladle design report which is referenced in Section 1.2.12 of the j

Plant Design Report.

l 1.5-13 Amendment 30 November 17, 1981 i

REFERENCE 1.

" Topical Report - Safety Related Research and Development for West-inghouse Pressurized water reactors - Program Summaries - Fall 1972",

WCAP-8004, Decenter 1972.

2.

Advisory Comittee on Reactor Safeguards Repsrt on the Atlantic Gen-erating Station, October 18, 1973.

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Anendment 13, 1975 1.5-14 January 21, 1975

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SITE ENVELOPE,

.., 1 Requiroent for

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PDR

$1te Envelope Envelope Envelope Section Parameters Parameter Parameter Limit Reference Vital areas must not flood during Maximum MLW depth, Basin water depth at MLW must satisfy all of the 2.3 the postulated sinking emergency (Note 1) following conditions:

a) MLW < 76' minus 10 percent exceedance high spring 22 tide minusT/TUO year storm surge minus allowance for wave crest adjacent to vital structures b) MLW i 76' minus 10 percent exceedance high spring l22 tide minus tsunamt minus allowance for wave crest adjacent to vital structures (Note 2)

."le.t must..:,; ground under the Minimum MLW depth Basin water depth at MLW must satisfy all of the following 2.3 influence of environmental loads (Note 1) condition *:

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a) MLW > Plant Draft plus Maximum downward corner displace-O ment (Note 3) produced by the design basis tornado b) MLW 1 Plant Draft plus 10 percent exceedance low spring l 22 tide lus drawdown from stillwater level produced by the p us Maximum downward corner displacement (Note 3) produced by the PMH at conditions of maximum stonn drawdown c) MLW 1 Plant Draft minus 10 percent exceedance high spring l22 tide minus storm surge produced by the PMH plus Maximum downward corner displacement (Note 3) produced by the PMH at conditions of storm surge d) MLW > Plant Draft plus 10 percent exceedance low spring l 22 tide plus drawdown produced by tsunami (Note 3).

Note (1}: The equations in the " Envelope Parameter Limit" column define limits of acceptable MLW depth which must be satisfied throughout the life of the plant. Deviations from the nominal elevation of the basin floor at each specific site must be taken into account in order to determine the range of water depths at MLW which might.be encountered during the life of the plant; expected maximum and and minimum MLW depths are then compared to the limits established'by the above equations.

N ote (2):

N For river sites, the site characteristics that need to be combined and compared to the 76' maximum water depth are:

g3 Operating Basis Flood level in basin

+OperatingBasisStormSurgeinbasin(StandardProject)

L1 in 100 yr stom)

g

+ Allowance for wave adjacent to vital structures 5

Note (3): Including static trim in addition to motion produced by environmental loading,

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TABLE-2.1-1 (CONT)

Requirement for PDR Site Envelope Envelope Envelope Section Parameters Parameter Parameter Limit Reference Plant design basis motion a) Plant response spectra at four a) Horizontal Component:

must not be exceeded specified locations (expressed

1) PMH,.10g

. 3.7.1 in terms of equivalent static

2) Tornado with continu3us accelerations) basis motion,.109

3) SSE with continuous basis motion,.20g Vertical Component:

1) PMH,.109

2) Tornado with continuous basis motion,.10g b) Fr;ee-field ground response spectra b)

Horizontal Conoonent SSE, 0.30g ro 3')

5 Vertical Component

SSE, 0.20g c) Maximum design basis angular c) 3 displacen;ent about any axis in the horizontal plane due to combined pitch and roll (Note 4) gy Note (4):

It is not an implied requirement that the minimum MLW depth at all sites accommodate B. 'd the platform corner displacen'ent associated Q@

with 3 motion (approximately 14.4 feet).

Rather, the minimum MLW depth for each site

." 8 is a function of the motion determined for 0

the specific site; this motion must not exceed 3.

TABLE 2.1-1 (CONT)

Requirement for PDR Site Envelope Envelope Envelope Section Par meters Parameter Parameter Limit Reference Plant operating basis motion a) Plant response spectra at four a) Horizontal Component:

3.7.1 must not be exceeded during specified locations (equivalent

1) OBE with continuous operating basis events static accelerations) basis motion 10g i

2)Operatingbasiswind and wave 05g Vertical component:

1) Operating basis wind and wave,.05g b)

Free-field ground response b) Horizontal Component:

i spectra OBE,.159

.N Vertical Component:

)

7 OBE,.10g e

0 c) Maximum operating basis angular c) 2 displacement about any axis in the horizontal plane due to combined pitch and roll Plant continuous basis a) Plant response spectra at four a) Horizontal Component:

3.7.1 motion must not be specified locations (expressed Continuous basis wind and exceeded during continuous in terms of equivalent static wave. 0.015g basis wind and wave accelerations)

Vertical Component:

gg Continuous basis wind and 2

gg wave, 0.015g 30 cr -a b) Maximum continuous bases angular b) 0.5 displacement about any axis in

." 8 the horizontal plane due to com-bined pitch and roll

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TABLE 2.1-1 (Continued)

Requirement for PDR Site Envelope Envelope Envelope Section Parameters Parameter Parameter Limit Reference Pressure loads on the plant a) Tornado a) Rotational speed: 290 mph 3.3 &

superstructures must not Translation speed: 70 mph 3.8 exceed the design value (max.) 5 mph (min.)

Pressure drop:

3.0 psi b) Design basis wind b) Fastest mile wind speed, (PMH) 204 mph c) Operating basis c) Fastest mile wind speed, wind 160 mph Basin water must not exper-Basin Ice Continuous sheet of basin ice must 2.7,3 y

L ience a "hard freeze" not occur or must be prevented by g

Owner action Maximum basin water tempera-Maximum basin water 950 2.7.3 ture must not exceed the temperature design basis of safety re-lated cooling water system Minimum air temperature at Air temperature

-15 F (Note 1) 2.7.2 0

the water surface (0-5 meters) 24 must not be less than the de-sign service temperature of Py the hull steel em kh Minimum basin water tempera-Minimum basin water 28.6 F 2.7.3 w@

ture must not be less than temperature the design service tempera-mW ture of the hull steel S

u Note 1: Lower temperatures may be accommodated by hull material changes as describert in Section 3.12 24

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i TAGLE 2.1-1 (Continued)

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SITE EllVELOPE 1

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Requirement. for PDR Site Envelope Envelope Envelope Section Parameters Parameter Parameter Limit Reference f

Site design features Site configuration, Site features at riverine 2.10.5 j

shall be incorporated natural and raan-made and estuarine sites shall i

at riverine and estuarine site features permit a degree of liquid sites, as necessary, to pathway source interdiction mitigate the environ-which is equivalent to that j

mental consequences of a which was found in NUREG-0502 l

postulated core melt to be generally achievable in accident a land-based plant or that 30 n3 1.

specified by NRC in the 1

2.

future as appropriate for

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land-based plants.

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REGULATION 10CFR50.34(e)(3)(ii)

Subject:

Quality Assurance List To satisfy the following requirements, the application shall provide sufficient infonnation to demonstrate that the requirement has been met.

This information is of the type customarily required to satisfy 10 CFR 50.34(6)(1) or to address the applicant's technical qualifications and management structure and competence.

Ensure that the quality assurance (QA) list required by Criterion II, App.

B.,

10 CFR Part 50 includes all structures, systems, and components important to safety.

(I.F.1) 0FFSHORE POWER SYSTEMS RESPONSE Implementation of NRC requirements assure that QA measures are applied to a comprehensive set of structures systems and components important to safety.

In addition to NRC requiements, Offshore Power Systems utilizes an internal classification system which adds many structures, systems and components to the Quality Assurance list which might not otherwise receive more than standard commercial Quality Assurance measures. The application of both NRC and OPS Quality Assurance requirements to the FNP are outlined below.

O The general design criteria for nuclear power plants are contained in 10 CFR 50, Appendix A. These criteria provide a broad definition of plant structures, systems and components important to safety. The basic response of Offshore Power Systems to each of the General Design Criteria is contained in Section 3.1 of the Plant Design Report. NRC Quality Assurance regulations are contained in 10 CFR 50, Appendix B. Throgh the mechanism of regulatory guides the NRC has imposed Appendix B Quality Assurance requirements to various structures, systems and components, based on the characteristics of the particular structure, system or component concerned.

The QA program of OPS will apply to those structures, systems, and compo-nents derived from the following documents:

o Appendix A of 10CFR50 - " General Design Criteria for Nuclear Power Plants." All structures, systems, and components important to safety are those that provide reasonable assurance that the FNP can be operated without undue risk to the health and safety of the public.

Amendment 28 C-124 July 15,1981 l

These structures, systems, and components can be derived from the General Design Criteria.

o Regulatory Guide 1.26 identifies the nuclear plant fluid systems which fall into quality classifications A, B, C and D. Offshore Power Systems complies with this Regulatory Guide; however, industry safety classi-fications (1, 2, 3 and Non-Nuclear Safety or NNS) are used in place of quality groups A, B, C and D. Offshore Power Systems procedures require appropriate Appendix B Quality Assurance measures for all systems and components classified as Safety Class 1, Safety Class 2, " fety Class 3 or NNS.

o Regulatory Guide 1.29 requires that the Quality Assurance Program of 10 CFR 50, Appendix B be applied to each of the structures, systems and components listed in Regulatory Positions 1,

2 and 3.

The Quality Assurance measures invoked by Offshore Power Systems for Floating Nuclear Plant structures, systems and components complies with Regula-tory Guide 1.29.

o Regulatory Guide 1.120 establishes the QA requirements for the Fire Protection System. These requirements are unchanged from those of Branch Technical Position APCSB 9.5-1 ( Appendix A) which were committed to in Offshore Power Systems Report RP06A30, " Floating Nuclear Plant Fire Protection Evaluation", September,1977.

o Regulatory Guide 1.143 supplements Regulation Guides 1.26 and 1.29 for Radwaste Systems. Regulatory Position 6 of this guide detail s an acceptable Quality Assurance Program for Radwaste Systems. Offshore Power Systems will meet or exceed the requirements of Position 6 in future design and manufacturing activities.

Offshore Power Systems engineering procedures require the responsible engi-neers to classify each Floating Nuclear Plant structure, system and compo-30 nent using a pre-defined set of classifications. This involves performing an engineering analysis eva' fating the functional use of each item derived from the above documents to detennine it's importance to safety and the Amendment 30 C-125 November 17, 1981

Containment Functional Capability The Floating Nuclear Plant steel containment vessel consists of the containment shell and the containment base plate as described in the Plant Design Report (PDR) Sections 3.8.2.6 & 3.8.2.8, respectively.

The current design of the containment vessel is based on the uniform internal design pressure of 15 psig given in PDR Section 6.2.1.2 and the non-uniform transient pressures given in PDR Chapter 15. Analyses of the containment functional capability have been performed for the current containment design. The results are sumarized in Figure C-15.

In the analyses, the following calculation methods and design param-eters were considered:

1.

Shell capability was determined 3s the pressure producing gross yield behavior. Yield was based on Von Mises criterion.

2.

Actual yield stress used in the calculation was assumed to be equal to 120% of the specified minimum yield stress.

3.

A hand calculation was performed on the shell with smeared out hoop stiffeners. This approach was verified by finite element s

elasto-plastic analyses of panels with discrete longitudinal and hoop stiffeners. The latter analyses were derived from work done by OPS for the Sequoyah and McGuire / ice condenser containment.

l 4.

Platfonn capability was calculated using plastic analysis methods.

5.

Evaluation of the shell/ platform interface was based on the area of the platform structure backing up the containment shell as shown in Figures C-16 and C-17, and Table C-10.

6.

Buckling analyses of the torispherical dome of the containment shell and the spherical cap of the equipment access hatch were based on realistic buckling criteria.

M "An Analysis of Hydrogen Control Measures at McGuire Nuclear Station,"

O Volume 2, Section 4.2.5, Duke Power Company, November 17, 1980.

Amendment 28 C-139 July 15,1981

Modifications to the containment were investigated and the results indicated that the containment functional capability can be increased from 55 psig to a pressure of 80 psig within the existing design concept and without excessive impact on the plant design (see Table C-11.

The containment vessel will be upg"aded to meet the requirements of the ASME Code Service Level C Liraits, excluding evaluation of in-stability, considering pressure and dead load alone, during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction accompanied by hydrogen burning. Resul ts to date indicated that the hydrogen burning pressure load is considerably less than 45 psig. Therefore, a minimum internal pressure of 45 psig is specified for the above design consideration. In addition, the con-tainment vessel will be designed to meet ASME Service Level A stress limits for a design internal pressure of 25 psig at ambient tempera-30 ture, including the effects of dead load. This pressure load will not be included in other load combinations such as those including the operating basis earthquake.

Internal Containment Structures Based on our review of analyses performed on ice condenser contain-ments, it is our judgment that differential pressures across internal structures during hydrogen burning will not challenge the integrity of I

those structures. We therefore conclude that there is reasonable i

assurance that containment internal structures can withstand the effects of hydrogen burning. Analysis will be perfomed to define the environmental conditions inside containment within two years after issuance of the Floating Nuclear Plant Manufacturing License. These analyses will be utilized to confirm that the containment internal structures accomodate hydrogen burning.

Equipment Survivability The systems necessary to maintain containment integrity will be j

designed to perfom their function under the conditions calculated to occur during the operation of the distributed ignition system.

The identification, location, evaluation, and protection (if necessary) of I

equipment associated with such systems will be established during the FNP final design.

Amendment 30 C-140 November 17, 1981