ML19242D918

SER Re Operation of Facility.Suppl 1 to NUREG-75/100

ML19242D918
Person / Time
Site:	Crane, Atlantic Nuclear Power Plant
Issue date:	03/16/1976
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0054, NUREG-54, NUDOCS 7909040257
Download: ML19242D918 (50)

Text

NUREG-0054 y

(Supplement No.1 to NUREG-75/100) p3 1

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U. S. N uclear 4

Regulatory ^ommission related to operation of Office of Nuclear Reactor Regulation Offshore Power Systems Docket No. STN 50-437 Floating Nuclear Plants (1-8)

March 16,1976

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NUREG 0054 Supplement No. 1 to MUREG-75/100 SUPPLEMENT NO. 1 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION IN THE MATTE 9 (?

0FFSHORE POWER SYSTEMS FLOATING NUCLEAR PLANTS (1-8)

DOCKET NO. STN 50-437 MARCH 16, 1976

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TABLE OF CONTENTS la30

1.0 INTRODUCTION

AND GENERAL DISCUS 5 ION.

1 1.1 Genera! Rackground.

1 1.6 Site-Related Design Envelope.

1 1.10 Outstanding Issues.

1 2.0 FLANT-SITE INTERFACES.

9 2.6 Mooring Systen.

9 2.8 Site Environment.

9 2.8.1 Meteoroloqy.

9 2.8.1.1 aegional Clinatalogy.

9 2.3.2 Wind Convergence Over A Breakwater.

9 2.10 Site Accidents.

9 2.10.2 Shippino Acciderts.

9 3.0 DESIGN CRITERIA FOR STPUCTLF.ES, COMPONENTS, EQUIFMENT AND SYSTEMS.

11

3. 5 Missi)c Protection Cr iteria, 11 3.5.1 Tornado Missiles.

11 3.8 Design of Seismic Category I St actures.

11 3.8.1 Containment (Steel).

11

-.8.4 Air Blast Procedures.

Il 3.11 Platfo.. Structure.

11 3.11.1 Hull Material.

12 3.11.3 Corrosicn Control.

12 6.0 ENGINEERED SAFETY F[ATURES.

13 6.2 Containment Systems.

13 6.2.1

-ontainment Functional Design, il L.2.4 Containment Isolation Systen.

13 6.2.8 Main Stean Line Break Inside Containment.

13 6.3 Erergency Core Cooling System.

14 6.3.3 Perforrance Evaluation.

14 7.0 INSTRU"ENTATION AND CONTROL.

15 7.3 Engineered Safety Features Actuation System.

15 7.3.5 Application of the Single Failure Criterion to Manually-Controlled, Electrically-Operated Valves.

15 7.5 Safety-Related Display Instrumentation.

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TABLEOFCONTENTS(Continu_edl Pale 8.0 ELECTRIC POWER SYSTEMS.

16 8.2 Offsite Power Systems.

16 16 8.4 Physical Independence of Electric Systems.

18 9.0 AUXILIARY SYSTEMS.

18 9.5 Other Aux'liary Systems.

18 9.5.1 Fire Protection System.

10.0 STEAM AND POWER CONVERSION SYSTEM.

19 10.2 Turbine Generator.

19 19 10.2.1 Design Considerations for Flcatina Nuclear Plant.

11.0 RADI0 ACTIVE WASTE MANAGEMENT.

20 12.0 RADIATION PROTECTION.

21 22 15.0 ACCIDENT ANALYSES.

22 15.4 Radiological Consequences 15.4.1 Loss-Of-Coolant Accidert Lose Mode.

22 1

15.4.2 Fuel Handling Accident Dose FLdel.

42 16.0 MANUFACTURING CONDITIONS.

25 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.

26 20.0 FINANCIAL QUALIFICATIONS.

23 20.1 Introductian.

20 20.2 Pricing Policy and Manufccturing Cost Estimates.

23 20.2.1 Pricing Policy.

28 20.2.2 Manufacturing Cost Estimates ;nd Sources of Funds.

29 20.3 Conclusions.

30

21.0 CONCLUSION

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LIST OF TABLES Table e

1.2 (Pevised) floating Nuclear Plant Site Design Envelope.

2 15.2 (Pevised) Assunptions Used In The Calculation Of Less-Of-Coolant Accident Doses.

23 15.3 Assumption Used In The laiculation Of fuel Handlinq Accident Doses.

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24 15.4 (Revised) Radiological Accident Consequences.

24 APPENDICES Appendix A - Contir.uation of The Chronology of PegJlatory Radiological Peview of Floating Nuclear Plants I-8 A1 Appendix B - Interin Peport of the Advisory Connittee on Eeactor Safe-cuards, dated Decerber 10, 1975 g_1 Appendix C - Corrunica tion f rer, U.S. Departrent of Correrce, National Oceanic and Atrospheric Administration, Environrental f a ta Service, Noverber 5,1975 C-1 7 j r

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i.0 INTRODUCTION AND GENFRAL D ECUSSION 1.1 General Packgrnun_d_

The Nuclear Regulatory Comission's (Cuarnission) Safety Evaluation Report in the ratter of the application of Offshore Power Systers (herein;f ter referred to as the applicant) for a license to nanufacture eight standaraized floatina nuclear plants was issued on September 30, 1975. In that Safety Evaluation Report we identified (1) two matters requiring additional inforration from the applicant, (2) three ratters where our review of infornation submitted by the applicant was not yet complete and (3) six natters wherein the applicant's prorsed design diff ered fron staf f requirerents.

The purpose of this SupDienent is to update the 53fety Evaluation Report by pro-viding (1) our evaluation of additional inferration submitted by the applicant since the Safety Evalmtien Report was issued, (2) our evaluation of the matters where we htd not comp!eted ou review of information subritted by the aWiicant when the r

Safety Evaluation Peport was issued and (3) cur responses to the coments nace t;y the Advisory Connittee on Peactor Safeguards in its rercrt dated Cecerber 10, 1975.

Tte reas of primary U.S. Coast Suard eview and inspection responsibilities are included in the 54fety Evaluation Peport anJ this Supplement. These areas are delineated in the 7e crandr: of Understarding Between the U.S. Coast Ga rd and Tre U.S. Atmic Erergy Ccvission for Peplation of Floating Nuclear Pcwer f lants,'

Manuary 4, 1974 The i;.S. Coast Guard will issue a letter of acceptance indicatira its sitisfaction with tre preliminary design information relatiag to its review of the application.

Except f or the ap;endices, each of the followin; sections o' this Supplement is ru-tered the same a s the sec t h " of the Safety Evaluation Peport that is teing updated, and the discussicns are supple entary to and act in lieu of the discussion in the Safety Evalu3tien Percrt. Appendix A is a continuation of the chronology of th staff's principal actions related to the processing of the applicatien. Arrendix B is tha Interi~ Report of the Advisrry Com ittee on Peactor Safeguards on the Floatinq Nuclear plant.

unication from the National Oceanic ani Atmospheric n conr Ad inistration is included as Appendix C.

l.6 Site lelated Cesian Envelo:e The site-related 6 sign envelge parameters are surrarized in Table 1.2 of the Sa fety Evaluation Report. This taole has been revised to reflect our present evala-ation. For convenience, the revised table has teen reproduced in its entirety in this Supple-ent. Except for iters (12), (13) ard (14) the table is essentially the 53 e as the table in tne Safety Evalu3 tion Pepcrt. Additions or changes are identi-fied by a vertical nargin bar.

1.10 Outstandinc Issues In Section !.10 cf the Safety Evaluation Pe ort, we listed a number of outstard-s in issues. All of tne outstanding issues have been resolved with the single excep-tion that evaluation of emergency core cooling syste~ design in acccedance with 4:en dix K to 10 CFR Fart 50 is not yet complete (Section 6.3.3).

The resolutien of this ra tter will 9 r enrted in a future supplenent to the Safety Evalu3 tion Peport.

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TACLE 1.2 (REVISED)

FLOATING NUCLEAR PLANT SITE DESIGN ENVELOPE Plant Desigr.

Report Requirement for Section Site Envelope Envelope Envelope Parameters Parameter Para eter Limit Reference (1) Tital areas nust not flood during the Maximum nean low water depth Basin water depth at mean low-water must satisfy 2.3 postulated sinking emergency (Note 1) all of the following conditions (Note 2):

(a) Mean low wa*.er 76 ft minus 10 percent exceedance high spring tide minus 1/100 year storn surge minus allowance for w3ve crest adjacent to vital structures.

(b) Mean low water 176 ft min _us 10 percent l

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exceedance high scring tide minus maximum tsunami minus allowance for wave crest hl adjacent to vital structures.

C (2) Plant must not ground under the Minimum nean low water depth Basin w3ter depth at rean low water must satisfy 2.3 influence of enviro nental loads (Note 1) all of the fellowing conditions (Note 3):

o (a) Mean low water Plant Craft plus maximum downward displacement produced'~by the

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design basis tornado.

(b) Mean low water _ Plant Draft p_lus 10 per-cent exceedance low spring tide plus draw-down from stiliwater level produced by the probable maximum hurricane plus maximum downward corner displacement produced by the probable naximum hurricane at condi-tions of maximum storm drawdown.

'd (c) Mean low water _ Plant Draf t mirus 10 percent high spring tide minus storm surge produced by the probable maximum hurricane plus maximum downward corner displacement s.

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produced by the probable maxinum hurricane J

at conditions of storn surge.

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produced by tsunri.

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Plant Design Pequirement for Report Site Envelope Envelope Envelope Section Pa rame ters Pa rame t er Paraneter Limit Reference (3) Plant design basis motion nust not be (a) Plant response spectra at (a) Harizontal Component:

3.7.1 exceeded four specified 'ocations (1) Probable naximum hurricane, 0.109 (expre ssed in terms of (2) Tornado with continuous basis motion, equivalent static 0.10g accelerations)

(3) Safe shutdown earthquake with ccntinu-vus basis motion, 0.209 sj iM Vertical Component:

c, (1) Probable maximum hurricane, 0.109 (2) Tornido with continuous basis motion, 0.109 (3) Vertical corponent due to horizontal safe shutdown earthquake with continu-ous basis motion, 0.05g CD (b) Ground response sprectra (b) Vertical component only, safe shutdown earthquake, 0.20q (c) Maximum design basis (c) 3 degrees

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angular displacement about any axis in the horizontal p'ane due to combined pitch N

and roll (Note 4)

(d) Ground response spectra (d) Horizontal Component: Operating basis with plant in sunken earthquake, 0.15g p' 1-condition f

Vertical Component: operating basis L-ea rthqua ke, 0.109 (4) Plant operating bais motion nust not (a) Plant response spectra at (a) Horizontal Comper.ent:

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be exceeded dJring operating basis four specified locations (1) Operating basis earthquake with events (equivalent static continuous basis motion, 0.10g accelerations)

(2) operating basis wind and wave, 0.05g CN

TAELE 1.2 !CEVISED) (Continued)

Plant Design Requirenent for Feport Site Envelope Envelope Envelo;e Section Prameters Parceter Ptrareter Linit peforence Vertical Corponent:

(1) iertical corpcnent due to horizontal cperating basis earthquake with contin-NJ ucus basis notion, 0.025q IN)

(2) npera tinq basis wind and wave, 0.05g CD (b) Maximum operating basis (b) 2 degrees angular displacement about any au s in the horizontal

,q M-plane due to conbined pitch and roll (Note 4)

-2 (5) Plant continuous basis 4 ; tion nust

(?) Plant response spectra at (a) Horizontal Corponent: Continuous basis 3.7.1 not be exceeded during continuous (wr specifird locations wind and wave. 0.015g basis wind and wave (expres.M in terms of equivalent st W e Vertical Component: Continuous basis wind accelerations) and wave, 0.015q

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(b) Maximum continuous basis (b) 0.5 degrees N

angular displacement about any axis in the hor zontal

-- Q plane due to combined pitch and roll (b) Pressure loads on the plant super-(a) Tornado (a) Potational s; eed. 290 niles per hour 3.3 5 3.8 mM.

structures rust not exceed tho design Translational speed: 70 miles per hour value (maxirun), 5 miles per hour (ninimum);

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3.0 pounds per square inch.

(b) Desio basis wind (probable (b) fastest nile wind speed, 204 miles per hour 1

naxi u i hurricane)

(c) Cperating basis wind (c) Fastest nile wind speed,160 niles per hour (7) Basin water must not experience a Basin Ice Continuous sheet of basin ice must not occur 2.7.3 "hard freeze" or nust te prevented by utility-owner action.

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(C Maxi un basin water tenperature rust Ma xi~ n basin water ter perature 85 degrees Fahrenheit 2.7.3 J

not exceed the design tesis of safety-related cooling water system.

TAELE l.2 (REVISED) (Ccntirued)

Plant Design Requirement for Report Site Envelope Envelope Envelope Section Parareters ParaTeter Ptrareter Linit Peference (9) Minimum air temperature at the sea Air terperature

-5 degrees Fahr enhei t 2.7.2 surface (0-5 reters) must not be less than the design service tenperature of the hull steel (10) Minimum basin water temperature rust Minimum basin witer terperature 2b.6 degrees Fahrenheit 2.7.3 not be less thar. the design service temperature of the hull steel q

('Y (11) Precipation must not overload plant Frecipitation rate (rainfall or 13 inchee per hour 2.7.6 L'

roof structures waterspout)

(12) A class of accidents, the conse-a) P (aircraf t crash)

(a) through (c): P _19' / yea r 2.9 quences of which could exceed the b) P (flarrable vapur cloud) plant design basis, must have a c) P (toxic chenical spill) low probability of occurrence d) P (explosion 2 pounds per (d) P 10 / year, or doronstrate site features p

square foot reflected prevent explosion f rom occurring near enough w

overpressure) to the plant to produce 2 pounds per square e) P (toxic vapor cloud) foot reflected overpressure (e) F

?O' / year or dencnstrate that concentra-tinn of toxic vapor at control room and ervrger c y relocation area intakes does not exceed i mi ts given in Table 2.9.1 (13) Accident dose offsite must not While body dose, thyroid dose The cnr:bination of plant accident releases, 2.3.2 exceed 10 CFR 100 atmospheric diffusion, exclusion boundc y radius, and low population zone rodius must result in doses less than or equal to 10 CFR 100 licits. For deternining exclusion boundary, the two-Pfur,/Q Value at the boundary should be 1.97 10 sec/m' or less (14) Norral operating doses rust not Whole bedy dose and thyroid The cor bina tinn of normal plar,t opera ting 2.R.1

~ '1 N-J etcred 10 CFR 50, Appendix I dose from qaseous e'tluents, releases, atmospheric diffusicn, and site dose frn' liquid effluent' boundary r ust result in doses less than or ecul to 10 Cf; E3, fppendix I linits for caseous effluents; doses from liquid effluents m

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Tmpendix I limit.

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T* ELE 1.2 (PEVISED) (Cont; ued)

Plant Design Pemrt Requirenent for Site Envelere Envelope Envele,e Section Pa ra-e t e rs Paraneter Fa rameter I inn t Re f erer.c e (15) Basin flcor nust be adequate to (a) Flatness deviations (a) - 2 foot fro i nean plane and 10 f oot 2.5.2.1 support the plant in the sunken in-plane extent ccndition (b) Bearing strength (b) 1600 pounds per square foot (16) The rcoring systen riust:

(a) Transnit loads at the plant (a) location of plant / mooring (a) 'ive feet above plant botton near the

-.6 nooring foundations system corners of the plant (b) not overload the plarit nooring (b) transrii tted nuoring systen (t) To te specified during detailed design foundations loads (c) allow level and non-level sink-(c ) nuoring spten (c) O to 6 degrees sinking

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(17) Plant r,ust be prevented f rom Site configuration, nooring sys-Site degendent 2.6 and 2.10.2 colliding with site structures ten and other site structures (lP) A reliable source of offsite power (a) Separation and availability (a) Gereral Design Criterion 17 2.10.1 O

r'ust be provided of circuits (b) Nr ber of circuits (b) General Design Criterion 17 or as required

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DJ for continuity of alternating current power, whittever is greater (c) Integrity of the PCwer co'-

(c) Must renain functional during operating nection with the plant basis events evperienced at the specific site

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per year (19) Either the ensite or offsite alter-The combined probability of (1)

P nating current power systen nust be a loss of offsite power for a continuously available period in excess of seven days and (2) inability to replenish diesel fuel during a continuous seven-day period conir.cident with the loss of offsite power q

_, (20) A fuel oil spill occurring outside Site protective structure 100 fect f ron plant 2.9.4.1

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Plant Design Requirecent for Report Site Envelope Envelope Envelo,m Section Pa ra cy t e r s __

N rame_ter Pa raye t er __L im t Pef erenc e (21) Site design basis accidents and Site nissiles Irpact or penetration equal to or less than:

2.9 and environnental conditions r'ust not Tame (a ) 25 ton boa t, 50,000 pounds oroduce missiles which prevent (b) Wood plank, + inches x 12 inches x 12 feet, achieving safe shutdown 200 pounds (c) Steel pioe, 3 inches diameter, schedule 40, 10 feet Icng, 73 pounds (d) Steel rod, 1 inch diameter, 3 fret Icnq, 8 pounds b

(e) Utility pole, 13-1/2 inches dia eter, CD 35 feet long, 1,490 pounds (22) Vessels which can penetrate the first Site structure Irpact on the plant equivalent to a ship of 2A.8 inboard bulkhead or t; reach more than 3,500 tons at 13 knots c--

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two watertight compartr ents must be prevented from striking the plant with a velocity that would cause this

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damage (23) Operating basis wave in the basin must Waves in basin The mean wave height between crest and trougn 3.12.2.2.1 not exceed the operating basis valua associated with a wave length between 350 and for the platform hull 550 feet nust not esceed 6 feet (24) Design basis wave in the basin nust Waves in basin The rean wave height between crest and trough 3.12.2.2.1 not exceed the design basis value associated with a wave length between 350 and for the platform hull 550 feet must not exceed 10 feet (25) Corrosion of the iv.ersed surf aces (a) Mirinum post-polarization, (a) -0.85 volts (Versus copper-copper sulfate 9.0.J q

of the platforn hull rust be con-current-off negative hull reference electrode) trolled by a suitable cathodic pro-

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tection system (b) Polarization capacity (b) Achieve polarization within f;0 days at 90 percent current capacity taking into account stray currents Maintain polarization at 75 percent current y,

N capacity taking into account stray circuits C

(c) Redundancy / reserve capacity (c) Maintain polarization with single corponent g

failure taking into account stray currents

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TABLE 1.2 (REVISED) (Continued)

Plant Design Report Requirement for Site Envelope Envelope Envelope Section Pa rame t ers Para eter Paranet r Limit Referenc e (d) Nurter of rectifiers / anode (d) 8 mininun groups (e) Rectifier control (e) Automatic by hull-mounted reference electrodes (f) Interferente from other (f) Elirinate by bonding together electrically

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structures all submerged steel structures C:

(q) Performance nonitoring (q) Progran to be imnlemented by cwner

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Note (1): The equations in.ne " Envelope Parameter Limit" column define limits of acceptable me3n low witer (MLW) depth which nust be satisfied throughout the life of the plant. Deviations from the nuninal elev3 tion of the basin floo at each specific site must be taken into account in order to deternine the range of water depths at MLW which might be encountered during l'_.'

the life of the plant; expected naximum and minimun MLW depths are then compared to the limita established by tne above equations, co hate (2); for river sites, the site charac teristics that need to be conbined a. d compared to the 76 f eet raximun wa t?r depth are:

Operating Basis Flood level in basin (Standard Project flood)

+0perating Basis Stena Surge in basin (1 in 100 year storn)

+ Allowance for wave adjacent to vital structures Note (3): Including static trin in addition to notion produced by environmental loading.

Note (4): It is not an implied renuirenent that the mininun MLW depth at all sites acconnodate the platf erm corner displacenent associated with 3 degrees.

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2. a PLANT SITE INTERFACE 3 2.6 foco ri n1 y_s tem S

We stated in the Safety Evaluation Report that is is technically feasible to design and Duild a satisfactory mooring systen and that we would evaluate the adequacy of each mooring sy; tem as proposed by the utility-cwner of the plant. A total of twelve nooring system anchor points will be provided cn three sid:.s of the hull struc-ture. The specific anchc point load corponents are specified in the Fiant Design Report. The mooring system anchor points are discussed in Section 3.11.5 of the Safety Evaluation Report. Each utility-owner ruay utilize as rany of the anchor points speci-fica on the hull as requiru by its particular trooring schene. Each plant ray be noored dif ferently and ray include dif ferent degrees of redundancy depending upon the m6rgins of safety used in the design. We will evaluate the adequacy of each nooring system, including its degree of recundacy, as proposed by the utility-owner of each plant.

2.8 Site Environnent

2. <.1 f*eteorology 2.6.1.1 Reaional Climatolosj

• n the Safety Eval etion Report we stated that our design requirer'ent for the crerating t asis sustainod wird speed at a heir of 30 feet above sec level with a return period of 100 years 15 160 miles per hcur.

This requirenen; is based on data provided by the ?,ational Oceanic and Atrospheric A kinistration (see Appendix C).

The applicant in Arendacnt 21 to tre Plant Design Report has stated that the plant will be designed fcr an operating basis wind speed of 100 miles per hour. We consider this ratter resolsed.

2.8.2 Wird Conyergence Over A Cre3kwater In the Safety Evaluation Report it was noted that the applicant proposed a series of wind tunnel tests to determine the ef fects af convergence over the breakwater on wind loads on the plant. Also, the applicant corriitted to designing the plants to accorr.edate any increase in loadings indicate, by these tests. The results of the tests are presented in the applicant's Repsrt No. TR-lf ?, " Wind Tunnel Study of Wind Forces on a Floating Oclear power Plant,' whitn wir

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extremely reliable structure. On the basis of the above, we have not felt that it is warranted to require further verification of the design by neans of platform strain and defomation measurements.

The plant will be instrumented to monitor plant motions during tow to provide a means of determining whether plant systems and components are overstressed. Instru-mentation will be provided to supply data on the motion of the plant due to wind, waves and earthquakes during operation of the plant. In its review of the Atlantic Generating Station application, the staf f has also discussed provisions for monitoring the forces in the mooring system and the ability to correlate these forces with the plant rotions.

We intend to evaluate the 1eed for such instrumentation programs with the utility-owner of each plant.

The plans for visual inspection and nondestructive testing of the platfom structure are described in Chapter 3 of the Plant Design Report and in particular Section 3.12.6 (Corrosion Control) and 3.12.7 (Inspection and fdaintenance After Con-s truc ti on). They are further amplified in the applicant's Report No. AD-7100-14A85.

"fNP Platform Hull Dry Docking Equivalency." The staff and the U.S. Coast Guard evalua-tion of these plans is discussed in Section 3.11 of the Safety Evaluation Report.

We cc.nclude that the design criteria and design controls discussed above and presented in the Plant Design Report provide adequate verification of the structural design. The staff will further verify the adequacy of the design in its review of the applicant's Final Design Report and will determire then if there need be any additional requirements.

3.11.1 Hull Material We stated in the Safety Evaluation Report that we require that the design criteria f or the hull material include Charpy-V-Notch procedure qualification testing. The applicant in Amendment 21 to the Plant Design Report revised the design critt 'a for the hull material to treet our requirements. The applicant will also undertake a test program to establish the suitability of the Dynamic Tear test in the heat af fected zone. The test program and results will be submitted to the Nuclear Regulatory Connission and the U.S. Coast Guard for review and approval. If the test progran and results prove conclusively that the Dynamic Tear test can be used in the heat affected zone, the applicant proposes to substitute it for the Charpy-V-Notch testing for qualification and production testing. We find this acceptable. We consider this matter resolved.

3.11.3 Corrosion Control We have examined the platform hull splash zone corrosion protection system with regard to the practicality of repair or renewal. The splash zone has severe protection requirements because continual wetting of the surface by aerated sea water is alternated with exposure to the at osphere. It is recognized that the hull coating will not have unlimited life and that maintenance will be necessary. This has been anticip6ted and provision has been nade for this eventuality.

The applicant proposes a silica-filled catalized epoxy coating that has a high tolerance of wet conditions during coating application. In addition, the nature of the coating is such that local repairs can be made underwater. However, if extensive repairs are necessary or if recoating is indicated, this can be facilitated by triming the platform. The platform trim system is designed to provide controlled ballasting of

+ 1 degree for maintenance condition. Alternately, cofferdam techniques could be employed without the need for tilting the platform.

We therefore conclude that a splash zone corrosion control system that may have a life of less than 40 years may be used, since adequate reans are available to perform repairs or recoating without causing deviation fram the floating nuclear plant design limits.

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6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systens_

6.2.1 Containment Functional Design _

In AmendNnt 19 to the Plant Design Report the applicant increased the naxirum tecperature of the ultimate heat sink from 85 degrees Fahrenheit to 95 degrees Fahren-heit to allow expanded / riverine sitings of a floating nuclear plant. This change required the design capacity of the containment spray pumps to be increased from 2400 gallons per minute to 2800 gallons per minute. The containment spray heat exchangers and systen piping have also been upgraded, and the flow area of the containment su~p screen assemblies has been increased. All containment analyses sensitive to the changes in the containnent spray systen design have been repeated. The results indicate that:

(1) the margin between the available net positive suction head at the spray pump inlet and the required net positive suction head has increased from 72 percent to 80 percent of the required net positive suction heat; (2) the containrent capability with regard to bypass stean flow from the containment lower compartment to the upper compartment is essentially unchanged; and, (3) the maximun calculated containment pressure has decreased fron about 12.5 pounds per square inch, gauge to about 12.2 pounds per square inch, gauge.

Our previous conclusion that the designs of the prinary containment and the con-tainment heat renoval systems are in accordance with the appropriate General Design Criteria renains unchanged.

6.2.4 Contaiwent Isolat:on Systems In the Safety Evaluation Report for the floating nuclear plant, we reported that the applicant proposed the use of simple check valves outside containrent as isolation valves for the nain and auxiliary feedwater lines. The use of simple check valves outside containment for this purpose is expressly prohibited by General Design Criterion 57.

In discussions with the applicant regarding this design feature, the applicant connitted to upgrade the design of the main feedwater line to seinic Categvy I and ANS Safety Class 2 from the containment penetration up to and including the feedwater regulator valve, and the design of the auxiliary feeduater line to seismic Category I and ANS Safety Class 2 from the containnent up to and including the auxiliary feedwater stop valves.

The applicant h3s further comitted to reclassify the feedwater regulator valve and the auxiliary feedwater stop valves as containrent isolation valves.

These connitments were reported in the Novecber 7, 1975, meeting of the ACRS and have been documented in Arendment 21 to the Plant Design Report for the floating nuclear plant.

We therefore conclude that containaent isolation provisions for the main and auxiliary feedwater lines are in corpliance with the requirerents of General Design Crite -ion 57 and are acceptable.

6.2.8 Main Steam Line Break Inside Containment We have reevalu3ted the floating nuclear plant with regard to the containrent pressure and temperature response to a main steam line break inside containrent.

Car recent review of the Westinghouse Electric Corporation LOTIC-1 and LOTIC-2 codes revealed the method of calculating heat transfer from a superheated staan environment to p;ssive heat sinks in the containment lower compartment to be not conservative.

The LOTIC-1 code as used by the applicant to analyze the containment pressure and terperature response to a main stean line tweak.

In a recent connunication the appli-cant indicated recognition that the LOTIC-1 code is not capable of accurately calcu-lating the containment terperature and pressure in the superheated steam region.

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February 27, 1976, the applicant has committed to reanalyze the steam line break accident and determine the resulting containment temperatures using an appropriate code which we have found to be acceotable.

With regard tu containment pressure resulting from a postulated steam line areak inside containment, it is our judgment that the containment design pressure of 15 pounds per square inch, gauge will not be exceeded. The naximum containment pressure calculated for the containment prior to complete meltout of the ice conderser is 9.9 pounds per square inch for the design basis loss-of-coolant accident. The contain-c'ent peak pressure prior to complete ice melt is a function of the mass and energy release rates into the containment. Since the loss-of-coolant accident nass and energy release rates are more than double the mass and energy release rates for a steam line break, the containment pressure will not exceed 9.9 poands per square inch as long as there is ice in the ice condenser. The applicant has provided sufficient information to show that, considering single failure in the feedwater system and manual isolation in the auxiliary feedwater system, flow from the steam line will be terminated prior to complete ice melt in the ice condenser, and as a result, the containment design pressure would not be violated for a main steam line break inside containment.

We therefore conclude that the applicant's comnitrent to reanalyze the containment response when an acceptable code is available is acceptable at this stage (preliminary design) of the licensing process.

6.3 Emergency Core Cooling System 6.3.3 Performance Evaluation _

The evaluation of emergency core cooling system design in accordance with Appendix K to 10 CFR Part 50 is not complete. Our evaluatien and conclusions will be included in a supplenent to the Safety Evaluation Report.

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7.0 INSTRUMENTATION AND CONTROL 7.3 Engineered Safety Features Actuation System 7.3.5 Applicati,on of the Single Failure Criterion To Manually-Controlled, Electrically-Operated Valves We are currently perfoming a generic review of a Westinghouse proposal to elimi-nate from the design basis those single failures in emergency core cooling system valve control circuitry that result in spurious valve actuation. The resolution of this issue is directly applicable to the hot leg injection valves in the floating nuclear olant design. In Amendment 21 the applicant committed to confom to the generic Westing-house Electric Corporation resolution of this matter. On the basis of the applicant's conmitnent, we conclude that the design is acceptable.

7.5 Saf ety-Related Display Instrumentation We stated in the Safety Evaluation Report that the design criteria for the safety-related disolay instruner'ation are presented in RESAR-3 Section 7.5, including instrumentation for post-accident monitoring and safe shutdown. Regarding this aspect we consider that the range of safety-related display information is adequate to enable the plant operator to take correct action during and af ter an accident. The indicators and recorders referenced in these secions will be mounted in the main control room in a ranner consistent with the functional requirements of plant operation. All information and ccntrol facilitics required during the course of an accident and post accident recovery will be located in an area within the control room that will be utilized exclusively for accident mode operation. In those cases when the information displayed for accident monitoring is also required for normal operation, the same inforr"ation channel will be employed. The information displays required for normal operation will be identical in range and fomat to tnose used for accident monitoring and will be located in the "nornal operation area" c6 the control roon. We therefore conclude that the proposed design of the safety-related display instrumentation reets our require-rents and is acceptable.

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8.0 ELECTRIC F0WER SYSTEMS 8.2 Offsite Power Systen The design of the offsite power system includes two physicaily independent 100 percent capacity n ' iary transformers to be used with two physically independent sion circuits arranged in such a nanner as to ninimize the plant-to-shore tre.

t i likelihood of their ;imultaneous failure under operating and postulated accident and environmental conditions and so that both are inrediately available in the event of loss of one offsite circuit. In addition, the design provides for nonual switching capability to connect a third transmission circuit (cable) in place of either of the normal operating circuits to be used in the event of extended loss of one of fsite cable. The design of the transmission circuits, including the flexible connection, however, is outside the scope of the Plant Design Report and will be evaluated pursuant to a specific utility-owner site.

Therefore we conclude that the design of the offsite power terminations and associated circuitry provided in the design of the floating nuclear plant satisfies General Design Criteria 17 anj is acceptable.

8.4 Physical Independence of Electric Systens Additional infornation has been provided by the applicant regarding fire pre-vention and control. The applicant has indicated that the cable design for the floating nuclear plant was selected to provide an optinum balance of electrical, physical, aging, and water absorption characteristics, and flame retardant and mechanical proper-ties. The cable flame retardancy is considered capable of providing acceptable nargins in excess of its postulated fire e g osure. Power and control cabit insulation will be ethylene-propylene rubt;er-base with an overall jacket of neoprene or hypalon or will be an insulation having prop?rties equal to or better than the above insulation.

Instrumentation cables will vary with the type of signal conveyed 5ut will neet the insulation properties of power and control cables. Power and control cables have been subjected to Underwriter's Laboratories (UL), National Electrical Manufacturers Asso-ciation, Institute for Power and Control Engineers Association, and other flame tests to prove cable reliability.

Fire retar$nt wiring will be utilized throughout the control boards for both redundant and non-redundant circuits as an additional safety factor Cable penetrations through fire rated walls and floors will be designed and con-structed such as to raintain the barrier integrity without transnitting flane for the rating duratico. Design criteria which are presently being developed by industry (for example, the Institute of Electrical & Electronic Engineers Power Generation Committee) will be reviewed and evaluated for adoption as appropriate to all stop points. Design consideration will address but not be linited to:

(1) Stop naterial and its rating characteristics.

(2) Test rethods and qualifying procedures.

(3) Installation quality assurance procedures.

(4) Modification procedures (adding cables af ter stop installation).

(5) Suggested periodic inspection procedure.

The floating nuclear plant design includes a fin e protection system designed.to prevent, detect, extinguish, limit or control fire and its hazards and damaging ef fects, both inside the floating nuclear plant and inside a breakwater basin (also see Section 9.5.1 of this Supplement). All areas within the floating nuclear plant which contain hazard]us materials, vital equipnent, or equipment important to safety will be protectod from fire exposure by eitaer, or a combination of, fire resistive barriers, spatial separation, or fire detection and autonatic and nanual extinguishing systems. Au tona ti c wet pipe sprinkler systens will be provided in areas of high cable density such as the 16

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cable pull area and the containment electrical penetr? tion areas. Manually actuated carbon dioxide fire protection systems will be provided for the rooms housing the safety related control and instrumentation racks and the diesel generators.

A fire and smoke det 'ction alarm systec provided in the design will give immediate audible and visual alarms. This system will also monitor the status of the automatic fire extinguishing systems The design philosophy used seeks rapid identification of the location of a fire so that corrective measures may be taken to limit damags The monitored regions of the plant are divided into functional areas. The detection systen for each area will be independent of every other area, except for a connon alarm panel in the control room. Our review of the design of electrical control and instrumenta-tion systems important to safety included consideration of potential fire propagation to redunde.nt safety systems. We conclude that the prcposed design criteria and commit-ments in this regard meet present staff requirements and are acceptable.

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9.0 AUXILIARY SYSTEfiS 9.5 Other Auxiliary Systems.

9.5.1 Fire Protection System Subsequent to the issuance of the Safety Evaluatic. Report the applicant provided additional information regarding its fire protection and detection system design.

Our review of this information is sumarized in the following paragraphs.

The applicant has initiated a test program to verify suitability of the external wall and weir material to withstand external fires, resulting from a service barge accident within 100 feet of the plant, and from a petroleum tanker spill and ensuing fire occurring beyond 100 feet from the plant.

Tne system design to be tested consists of a weir designed to distribute cooling water on the exterior netal wall surfaces. The test result will be used to establish curves of heat flux versus time, ari wall temperature versus time for a spectrum of fires using various materials of construction. The applicant hrs statea that the staf f and U. S. Coast Guard would be kept informed of tne progress of the tests pro-gram and of design developments.

An external fire detection syster to alarm in the control room has been incor-parated in the design of the plant. The applicant has comitted to a test program to detertiine heat rate ari wall temperature curves for sustained heat fluxes which will be used to establish the location and type of detection equipment necessary to pro-tect the plant from such external exposure fires.

The design of the internal fire detection and alarm systems is based on the use of monitored zones. The detectors will be primarily located in unnanned areas not protected by automatic fire extinguishing systems. Standpipe hose stations will be locatcd at all elevations so that all parts of the plant are within reach of two hose streams from dif fer ent hydrants. The applicant has stated that the final design of the fire protection and detection system will reflect considerations of the recommendations of the staf f report, "Recorrendations Related to Browns Ferry Fire,' NUREG-0050, February, 1976.

Based on our review of the systems for detecting and protecting against fires, both internal and external to the planc, and conformance to the requirements of General Design Criterion 3, we conclude that the design criteria and proposed test programs are acceptable.

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10.0 STEAM AND POWF R CONVERSIGH SYSTEM 10.2 Turbine Generator 10.2.1 Design Considerations for Floating Nuclear Plant The platform hull does not provide the rigid base for the turbine foundation found in a conventionally constructed power plant. As a result, the applicant has evaluated the turbine generator design in light of the extremes in deflection that it will experience due to platform hogging and sagging. The analysis also included the inertial and gyroscopic forces associated with the platform motions. Although not cor,pletec', 'he preliminary results have indicated the ability of the turbine generator /

turbine foundation system to accon odate the anticipated platform deflections and motions. To minimize the effects of platform deflections transmitted to the turbine-generator machinery, selective alignment of the turbine-generator will be done. This procedure is similar to '. hat used on land based plants, where the effects of operating conditions such as vacuum loads, are compensated for in alignment of the machinery during erectice. In addition, as stated in the Safety Evaluation Report, the applicant has committed to analyze and test the turbine generator / turbine foundatisn/ platform system to verify that the turbine foundation adequately decouples the turbine generator f rom the platform so as to minimize the turbine-induced vibrations in other corponents.

We conclude that it is feasible to design and install a turbine generator / turbine foundation system which will function properly on the platform. We will evaluate the final design and verification analysis during our review of the final plant design report.

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11.0 RAJI 0 ACTIVE WASTE MANAGQtENT be stated in the bafety Evaluation Report that our evaluation of the capability of the liquid and gaseous radioactive waste treatment systems to meet the dose design objectives of Appendix I to 10 CFR Part 50 was not complete. The applicant, in Amend-ment 21. stated that the design objectise of the plant is to caeet the guidelines for quantitics of radioactivity released set forth in the Annex to Appendix I (Septer:ber 4, 1975). AoLiticaally, the applicant in its letter of March 4, 1976, indicated that for the broad siting spectrum, the annual cverage doses from liquid and airborne activity would also meet the dose guidelines specified in the Annex. The annual average dose estimates for liquid discharges and for discharged airborne activity were evaluated using conservative meteorology. Based on our evaluation of the design capability and design objectives of the radioactive waste management systen we conclude that these systrms will meet the dose objectives of Appendix ! to 10 CFR Part 50 for a broad siting spectrum, Each utility-owner however, will be required to verif y and demonstrate that a specific site is in conformance with the plant site design envelope parameter limit.

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12.0 RADIATION PROTECTION During the radiation protection review of the floating raclear plant application, careful attention was given to the evaluation of whether occupational radiation expo-sures would be as low as reasonably achievable during operation of the plant. The applicant's preliminary dose assessrce1t gave an e.tiriate of exposure that we found r

dcCeptabie, when considered at tr.e present stage of the preliminary design. The addi-tional infonr.ation provided by the applicant in response to coments made by the Advisory Comt ittee on Reactor Safeguards in its Cecenber 10, 1975 report indicates that unlimited access areas will be designed to nave exposure levels below 0.1 millirem per hour.

In view of the applicant's acceptable proposed implementation of as low as reasonably achievable design criteria, and the additinnal indication that unlimited access areas will be designed for especially low dose rates, we believe that we can continue to expect that occupational radiation exposures will be as low as reasonably achie.50le. We will review the calculations and design estimates of specific area dose rates docing our review of the final plant design report.

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15.0 ACCIDENT ANALYSES 15.4 Radiological Casequences 15.4.1 Loss-of-Coolant Accident Dose Model In Amendments 19, 20 and 21, the applicant submitted additional information on the secondary containment volumes treated by the annulus filtration system and the txhaust and recirculation flow rates of the annulus filtration system following the occurrence of a postulated loss-of-coolant accident. We have incorporated this information into our loss-of-coolant accident dose rodel in order to evaluate the radiological con-sequences of the accident. In addition, the absorption of the low energy beta radia-tion in the surface tissues of the body was not included in the calculation of the whole body doses. The assumptions and parameters used in our analysis are listed in Table 15.2 (Revised).

Since the floating nuclear plant is a standard plant with no specific site boundary distrces or meteorology, we dete.uined the limiting atmospheric dispersion factor (X/Q value) required by a site in order to ceet the guideline doses of Regulatory Guide 1.4

" Assumption used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactor" We calculated that a site with a two-hour atmospheric dispersion value o.' about 1.9 x 10-3 seconds per cubic meter or less at the exclusion area boundary is required to meet both the thyroid and whole body guideline dases of 150 rem and 20 rem, respectively, for the oesign basis loss-of-coolant accident;

• The limiting long term (30-day) atmospheric dispersion values required to neet the guideline doses at the low population zone distance were not determined in this analysis and the suitability of a site with regard to the low popula-tion zone deses will be evaluated for each individual floating nuclear plant site.

However, based on previous analysis of nuclear power plants of similar size and with similar engineered safety features, and the analysis of the two-hour doses for the floating nuclear plant, there is reasonable assurance that the floating nuclear plant will meet the guideline doses of Regulatory Guide 1.4 at low-population zone distances comparable to those of recently approved sites.

15.4.2 Fuel Handling Accident Dose Model The radiological consequences of a fuel handling accident have been evaluated based on the limiting two-hour atmospheric dispersion value of 1.9 x 10-3 seconds per cubic meter deternined for the loss-of-coo? int accident. The assumptions used in the analysis of the fuel handling accident are given, for convenience, in Table 15.3 of this Supplement (identical to that appearing in the Safety Evaluation Report) and the calcu-lated doses (27 rem to the thyroid and 2 rem to the whole body) are shown in Table 15.4 (Revised) of this Supplement.

• Of the sites we have previously evaluated, all of which were on lanJ approximately 90 percent had two-hour dispersion values equal to or less than 1.9 x 10-3 seconds per cubic reter at their exclusion arca boundaries.

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TABLE 15.2 (Revised)

ASSUMPTIONS USED IN THE CALCULATION OF LOSS-OF-COOLANT ACCIDENT DOSES Power Level, megawatts thermal 3579 Operating Time, years 3.0 Containment Volumes, cubic feet 5

Lower Compartment 4.08 x 10 5 Ic_ Condenser 1.25 x 10 5

Upper Compartment 7.37 x 10 Primary Containment Leak Rate, percent per day 0-24 hours 0.5 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25 Bypass Leakage Fraction 2 percent of primary containrent leakage Ice Condenser Elemental Iodine Raroval Efficiency 30 percent per pass Period of effectiveness 0.17 Hour to 0.47 Hour Containment Spray System Effective Volume, cubic feet 7.37 x 10 Renoval Rates, inverse hours Elemental Iodine 4.5 Farticulate Iodine 4.0 Organic Iodine O

Secondary Containrent Volume Treated 5

Sy Annulus Filtration System, cubic feet 6.28 x 10 Mixing Fraction, percent 50 Filter Efficiencies, cercent Elemental Iodine 95 Organic 'odine 95 Farticulate Iodine 99 Annulus Filtration System Flow Distribution ExhaJst Flow Recirculation Flow, Time Step _

Cubic feet per minute Cubic feet _per minute 0-10 seconds 0

0 10-300 seconds 6000 2000 300-600 seconds 4500 3500 600-1100 seconds 2600 5400 1100-1700 seconds 500 7500 1700-2700 seconds 2000 6000 2700 seconds - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1300 6700 2 - 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 350 7650 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 250 7750 23 u

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TABLE 15.3 ASSUMPTIONS USED IN THE CALCULATION OF FUEL HANDLING ACCIDENT DOSES 3579 Power, megawatts thermal Number of Fuel Rods Damaged 264 Total Number of Fuel Rods in Core 50,952 Peaking Factor of Damaged Rods 1.65 Shutdown Tine, hours 100 Inventory Released from Damaged Rods, percent (Noble Gases and Iodines) 10 Fuel Pool Decontamination Factor 100 Iodine Noble Gases 1

Filter Efficiency for looines, percent 95 TABLE 15.4 (Revised)

RADIOLOGICAL ACCIDENT CONSEQUENCES Exclusion Area 0-2 Hour Dose (Rem)

Accident limiting X/Q VaTue*

Thyroi d, Wnole Body Loss-of-Coolant 1.9 x 10' seconds per cubic meter 150 20

-3 Fuel Handling 1.9 x 10 seconds per cubic meter 27 2

Required by a site in order to meet Regulatory Guide 1.4 guideline doses (150 ren thyroid and 20 rem whole body) for loss-of-coolent accident.

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16.0 MANUFACTURING CONDITIONS _

in the Safety Evaluation Report we stated that certain limitations or conditions are required during the manufacture, outfitting and testing of the flo2. ting nuclear plant at the nanufacturing site to assure integrity and acceptable perforr ance of a

safety-related features subsequent to nuclear operation and for the service life of the plant. To meet these requirenents the applicant has propored in its letter of March 16, 1976, manufacturing conditions related to manufacture, outfitting and testing of the floating nuclear plant. These conditions include those aspects discussed in Section 16.0 of the Safety Evaluation Report. We have reviewed these conditions and limitations and find them to be acceptable and require them to be incorporated in the license to manufacture. We consider this matter resolved.

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18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Comittee on Reactor Safeguards (the Comittee) completed a partial review of the application for a license to manufacture eight standardized floating nuclear plant units at the 188th meeting of December 4-6, 1975. A copy of the Con-mittee's interim report dated December 10, 1975 is attached as Appendix B.

Our response to these coments and recomendations are described in the following paragraphs.

(1) The Comittee recommended that further consideration be given to methods for the assessment of probabilities for given accident events, such as those involving ships. The guidelines used by the staf f in determining whether potential acci-dents in the vicinity of a site are to be considered as design basis events are discussed in Regulatory Guide 1.70.8 " Additional Information-Nearby Industrial, Transprtation and tiilitary Facilities" Design basis events external to a nuclear plant are defincJ as those accidents which have a probability of occurrence on the order of about 10 ' per year or greater and have consequences severe enough to atiec the safety of the plant to the extent that 10 CFR Part 100 exposure guide-lines c u ld be exceeded.

For an appiicant's consideration in determining design basis events, the staf f has identiff?d s?veral accident categories with the categ vies being basst upon the effect that a particular type of accident could have on a plant. The accident cPegories include explosions, flannable vapor clouds, toxic chemicals and fires.

The probability of occurrence of each category from all potential hazard sources (transportation, industrial, military facilities) is considen d in @ternining whether or not a particular category of accident need be cunsidered a design basis event.

Usually in a site review there are several different kinds of hazardous naterial facilities or activities to be considered. The floating nuclear plant at an offshore site is unique in that the majority of the accident category hazards are related to shipping. However, for estuarine or riverine sites, industrial or military activities could well be the principal source of an accident category hazard. Thus the accident categories considered in determining whether or not an event will be consikred a design basis event are the same as those considered for land based plants recognizing that for offshore sites shipping accidents will likely be the largest contributor for each of the accident categories.

The overall objective of the review in this area is to determine which accident effects, if any, should be included in the plant design. Determining that the probability of a type of accident, such as a ship accident, exceeds scne guide-line value does not in itself give the plant designer sufficient information.

The designer also needs tu know the potential ef fects on the plant produced by the accident. Thus, determining and specifying accidents in terms of accident categories which produce particular ef f ects upon the plant has been the general approach followed in the review of the floating nuclear plant and is similar to the review for land based plants.

(2) The Committee indicated that they wished to be kept inforned on the thatters of containnent shell buckling and the design basis tanker explosion. These r atters are discussed in Section 3.6.1 and 2.10.2, respectively, of this Supplement.

(3) The Comittee indicated that they wished to review the design and analysis of the energency core cooling system and the upper head injection system-These systems are being evaluated by the staf f in cooperation with the Comittee on a generic basis with Westinghouse. The results of our evaluation will be implenented in our review of the floating nuclear plant design which incorporates systems of similar desiga. Our evaluation and conclusions will be included in a supplement to the Safety Evaluation Report.

(4) The Comittee noted areas wherein it wishes to be kept informed. These areas included turbine-generator alignment, hull-coupled vibrations and stresses as,a-ciated with platform towing operation. These matters s.re discussed in Sections 10.2 and 3.11 of this Supplerent.

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(S) The Connittee noted that it wished to be kept informed regarding the location and range of instruments for determining the nature and course of any accidents. This matter is discussed in Section 7.5 of this Supplement.

(6) The Committee indicated that it wished to be kept informed on the matter of verification of structural design of the floating nuclear plant. This natter is discussed in Section 3.11 of this Supplement.

(7) The Connittee stated that consideration should be given in design to the possible provisions for redundant nocring systems. This matter is discussed in Section 2.6 of this Supplement.

(8) The Comittee reconrended further review of the design features that are intended to prevent the occurrence of fires and to minimize the consequences to safety-related equipment should a fire occur. This matter is discussed in Sections 8.4 and 9.5 of this Supplement.

(9) The Comittee stated that it reserves judgmcnt on the generic liquid pathway study that is currently being performed by the staff and applicant until it has had an opportunity to review and evaluate the relevant infornation. We will keep the Committee informed on this matter.

(10) The Co rlittee suggested that analyses be made of any possible increases in the protection of public health and safety which may be obtained by an increase in containnent design pressure. The staff is performing a study, as a part of its environmental review of the floating nuckar plant concept, to compare the environ-mental consequences of a large accident at a land-based rcattor and a floating nuclear plant. The results of our study as appropriate, will be considered in our design requirements for the floating nuclear plant. We will keep the Comittee informed on this matter.

(11) The Connittee suggested that additional attention be given to means for protecting the critical wave and splash zone arei of the platform where repair or renewal may not be practical under the anticipated operating ccoditions of the floating nuclear plant. This matter is discussed in Section 3.11.3 of this Supple ent.

(12) The Comnittee stated that it believes that special consideration should be given to confernance with "as low as reasonably achievable" criteria. This matter is discussed in Section 12.0 of the Supplerent.

(13) The Connittee indicated that the review of the floating nuc'aar plant design for features that could reduce the possibility and ccnseque,ces of sabotage should be continued. The staff considers the design conservatisms provided in the floating nuclear plant for protection against design b. sis accidents also reduce the chance that ara act of sabotage could result in jeopardizing the health and safety of the public. However, the staff wiii continue to review the provisions for protection against sabotaae in applications that utilize the floating nuclear plant design.

(14) The Committee recomended that further attention be given to the possibility of extended loss of offsite power due to natural events or other caun and the potential ircact of this possibility on the requirements for emergency AC power.

This matter is discussed in Section 8.2 of this Supplement.

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20.0 FINANCIAL QUALIFICATIONS In the Safety Evaluation Report, we stated that we would report the results of our evaluation in a supplement to the Safety Evaluation Report. Our evaluation is presented below.

20.1 Intrcduction The Nuclear Regulatory Commission's regulations relating to financial data and information required to establish financial qualifications for applicants for manu-facturing licenses appear in Section 50.33(f) of 10 CFR Part 50 and Appendices C and M to 10 CFR Part 50.

The applicant, Oftshore Power Systems, has applied for a license to nanufacture eight floating nuclear power plants. The license is sought for a period of fourteen years beginning no earlier than January, 1977. No other Nuclear Regulatory Commission permits or licenses have been issued to or applied for by the applicant in connection with the manufacture of these plants. The purchasers of the plants are responsible for obtaining the necessary Nuclear Regulatory Corr 11ssion construction permits and operating licenses. Assaming each purchaser obtains the necessary permits and licenses in a timely manner, plant correccial ooeration should follow completion of manufacture by no more than eighteen months.

Offshore Power Systems is an unincorporated joint venture of Westinghouse Electric Corporation and Westinghouse International Power Systems Company, Inc., Westinghouse International Powar Systems Company, Inc., is a wholly-owned subsidiary of Westinghouse Electric Corporation. Westinghouse Electric Corporation owns 99 percent of Offshore Power Systems, and Westinghouse International Power Systems Company, Inc. Owns the renaining 1 percent. An assessment of the financial qualifications of Offshore Power Systems to undertake the proposed manufacturing activity is essentially an assessrent of the financial qualifications of Westinghouse Electric Corporation, since the one percent interest owned by Westinghouse International Power Systems Company, Inc. does not include an obligation to contribute capital to the venture.

Westinghouse Electric Corporation is a large, diversified enterprise and generally regarded as the second largest producer of electrical equipment in the world. Sales in 1974 amounted to $5,798.5 million, 35 percent of which was accounted for by the energy related product lines. Net incone in 1974 was $28.1 million, down sharply from o high of $198.7 million in 1972. This significant reouction in net incone was primarily the result of ron-recurring losses experienced in the sale of its major appliances, mail order and record club businesses during 1974. In 1975. Westinghouse Electric Corporation had sales of $5,862.7 million and net income of $165.2 million, a substantial rebound from the abnormally low 1974 results.

20.2 Pricing Policy and Manufacturing Cost Estimates 20.2.1 Pricing Policy The applicant has submitted a breakdown of the price of a floating nuclear power plant based on the May 1975 proposal to the Federal Energy Administration. The Federal Energy Adninistration proposal represerts the rcst recent pricing policy for such a plant. The unit cost estinate has been itemized as follows:

Unit Cost (dollars in millions)

Structures and improvenents 5 80.7 Reactor plant equipment 133.3 Turbine generator plant 129.2 Accessory electric equipment 39.0 Miscellaneous power plant equipment 15.4 Transmission facilities 10.5 Platform structures and specifically related systems 18.8 Testing (multi-systens) 2.1 Total Cost per Unit 5 435.0 28

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The $435 nillion price per unit is a base price in January 1975 dollars, subject to escalation upward or cownward as manufacturing costs fluctuate. The base price per unit will be escalated as follows:

Base Price Escalation Index Employed Escalation Base 55 percent labor Average Hourly Earnings in January, 1975 Shipbuilding & Repairing Industry - Bureau of Labor S ta ti s tics 35 percent material Steel Mill Product Index -

January, 1975 Bureau of Labor Statistics 10 percent profit & over-Gross National Product Implicit First quarter, 1975 head allowance Price Deflatcr - U.S. Department of Comerce The escalation provision will enable Westinghouse Electric Corporation to raintain the financial intcgrity of the pricing policy in this venture. This is most irportant when one considers the potential impact future inflation could have on the manufacturing costs during this lengthy future period.

The manufacture of an individual floating nuclear power plant will not conrence until an order has been e.ed for the plant. At present, the applicant has an order from the New Jersey Publ

'rvice Electric and Gas Company f'r four plants.

20.2.2 Manufacturinn Cost Estimates d Sources of Funds The estimated manufacturing costs (including the nanuf acturing facility) for the eight floating nuclear power nlants is $3,287.5 million. The applicant submitted an itemization of the estimated nanufacturing costs, including a detailed breakdown of the cost estimate of the manufacturing facility. This financial information was submitted with a request that it be accorded proprietary treatment. The staff reviewed the applicant's request pursuant to the provisions of 10 CFR 2.790.

Based on this review, the staff concluded that the applicant's justification confcrmed to the criteria for proprietary treatment and, consequently, granted the request.

Through 1978, case requirements of $531.7 million will be provided frcn a con-tinuation of internally generated funds ($343.8 million) and from funds provided by Westinghouse Electric Corporation ($87.9 million). After 1978, the cash requirements of $2,755.8 nillion will be provided by internally generated funds. The $87.9 million represents the maxinun investment to be provided by Westinghouse Electric Corporation.

The funds required to attain the maximum investment will be provided fron the following sources:

Amount in Source of Funds millions of dollars Ratio Sale of Interest Bearing Long-term Debt 5 21.0 23.9 percent Minority Intest 2.2 2.5 Preferred Stock 1.0 1.2 Comnon Stock 25.3 28.7 Internally Generated Funds 38.4 43.7 3 87.9 100.0 percent The cash requirements generated by internally generated funds represent progress pay-ents to be made by the purchasers. These progress pay ents will be in accordance with a payment schedule that is negotiated at the time of purchase.

Revenues from units sold are expected to cover the cost of manufacturing the units, amortization of the nanufacturing facility, interest on money borrowed, and any other cost applicable to the project. While total cash requirements can be projected, any meaningful breakdown of the annual increments of such cash requirements must await firm information on the sale and delivery of the fcur floating nuclear power plants currently being marketed by the applicant.

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20.3 Conclusions Based on the preceding analysis, which included the proprietary data, we have concluded that the applicant is financially qualified to manufacture the proposed eight floating nuclear power plants. Our conclusion is based on a determination that the applicant has reasonable assurance of obtaining the funds necessary to carry out the manufacturing activity for which the license is sought.

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21.0 CONCLUSION

S In Section 21.0 of the Safety Evaluation Report, we stated that we would be able to make certain conclusions upon favorable resolution of the outstanding matters set forth in Section 1.10 of the Safety Evaluation Report. We have discussed these matters in this supplement and indicated a favorable resolution for each matter except for a single issue discussed on Section 1.10 of this supplement.

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APPFNDIX A r0NTINUATION OF THE CHRONOLOGY OF REGULATORY RADIOLOGICAL REVIEW OF FLOATING NUCLEAR PLTNTS l_-8 September 18, 1975 Letter to Offshore Power Systems requesting additional infomation.

Septen.ber 18, 1975 Letter from United States Coast Guard advising that Captain C. E. Mathieu has been transferred and that Comander John Deck III has taken his place.

September 25, 1975 Meeting with Offshore Power Systems to discuss generic liquid pathway study.

September 30, 1975 Letter to Offshore Power Systems sumarizing results of July 15, 1975 meeting regarding environmeatal impact of postulated accidents associated with the NEPA.

October 3,1975 Letter from Offshore Power Systems transmitting Final Report entitled " Evaluation of Hazards to a FNP from a Coastal ranker Accident Near the Plant," dateo September 20, 1975 with attached graph.

October 9, 1975 Meeting with Offshore Power Systems to discuss technical issues rela tive to the liquid pathway generic study.

October 9 and 10, 1975 Meeting with Offshore Power Systems to discuss implemen-tation of generic liquid pathway study.

October 17, 1975 Letter from Offshore Power Systems transmitting Report No. SA-1000-14A96, "Results of Design Overspeed Turbine Missile Strike Probability Calculations on Vital Areas of the FNP Usino the MIDAS Code."

Octnber 20, 1975 Letter to Offscore Power System > requesting additional financial information.

October 20, 1975 Letter from United States Coast Guard providing connents on United States Coast Guard Plan Lists.

Oc tobe r 20, 1975 Letter from United States Coast Guard providing comrents on the Platform Hull Drydocking Equivalency document.

October 20, 1975 Amendment No. 19 provides additional information con-cerning safety related cooling water temp.

October 24, 1975 Meeting to discuss Offshore Power Systems proposal to increase site design envelope maximum basin water ten-perature fro 85 to 95 degrees Fahrenheit (PDR Section 2.7.3).

October 29 and 30, 1975 ACRS Subcomittee Meeting.

November 4, 1975 Amendment No. 20 provides additional concerning shield building annulus.

November 7, 1975 ACRS full comittee meeting.

November 10, 1975 Letter to Offshore Power Systems granting withholding of Control Rod Drive Mechanism analysis.

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November 10, 1975 Letter to Offshore Power Systems granting withholding of financial information.

November 13, 1975 Meeting with Offshore Power Systens to discuss generic liquid pathway sttdy - discussior, of liquid transport models.

December 1, 1975 Meeting with Offshore Power Systems to discuss design basis tanker explosion.

December 2, 1975 Meeting with Offshore Power Systems to discuss generic liquid pathway study.

December 3, 1975 Letter from Offshore Power Systems furnishing informa-tion concerning asymetric loadings on reactor pressure vessel support.

December 9, 1975 Meeting with applicant to discuss generic liquid pathway study.

December 10, 1975 ACRS letter.

December 12, 1975 Letter to Offshore Power Systems transmitting ACRS letter.

December 12, 1975 Meeting with Offshore Power Systems to discuss basis tanker explosion.

December 17, 1975 Letter from Of fshore Power Systems transmitting Of fshore Power Systems Report No. RP-9991-16A50, " Operating Basis Wind for U. S. Atlantic and Gulf Coastal Locations."

December 18, 1975 Letter from Offshore Power Systems transmitting Offshore Power Systems report regarding hazards to a floating nuclear pcwer plant fron a coastal tanker accident near the plant.

December 18, 1975 Letter from Offshore Power Systems transmitting Offshore Power Systems report regarding containment shell buckling criteria.

December 18, 1975 Amendment No. 21 provides additional information con-cerning plant design report.

January 13, 1976 Letter fron United Stat 2s Coast Guard regarding fire tests of weirs for external fire protec. tion.

January 16, 1976 Letter from Offshore Power 3ystens transmitting report regarding wind tunnel study of wind forces.

January 23, 1976 Letter from Offshore Power Systems transmitting report on design for air blast loading.

January 30, 1976 Letter from Offshore Power Systems transmitting report regarding hazards to a floating nuclear power plant frc.n a coastal tanker accident near the plant.

February 5, 1976 Meeting with Offshore Power Systems ta discuss technical issues relative to liquid transport modeling and scheduling problen.

February 17, 1976 Letter from United States Coast Guard regarding report on plant design.

February 23, 1976 Meeting with Offshore Power Systems to discuss contain-ment shell buckling criteria and air blast loads resulting from the design basis tanker explosion.

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APPENDIX 11 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION W ASHINGToN, D. C. 20555 December 10, 1975 Honorable William A. Anders Chairman U. S. Nuclear Regulatory Comission Washington, DC 20555

SUBJECT:

IN1ERIM REPORT ON FLOATING NIELEAR PIRIP

Dear Mr. Anders:

During its 188th Meeting, December 4-6, 1975, the Advisory Comittee on Beactor Safeguards completed a partial review of the applicatica of Off-shore Power Systems for a license to manufacture eight standardized Float ~

ing Nuclear Plant units in a shipyard-like facility located on Blount Island in Jacksonville, Florida. We Comittee had previously reported to the Comission on its review of the concept of a Platform Mounted Nuclear Ibwer Plant in its report of November 15, 1972.

In addition, the Comittee has had discussions of the Floating Nuclear Plant (FNP) concept in connection with the Atlantic Gencrating Station site review on which the Comittee reported on October 18, 1973. % e manufacturing facility site was visited on October 29, 1975 and the project was considered at a Subcomittee Meetir.g on October 29 and 30, 1975, in Jacksonville, Florida. We project was also considered during the 187th Meetirq of the Comittee in Washington, D.

C., November 6-8, 1975. During its review, the Comittee had the benefit of discussions with the Nuclear Regulatory Commission (NRC) Staff, the U. S. Coast Guard, and rspresentatives and consultants of Offshore Ibwer Systems. We Comitee also had the benefit of the documents listed.

h e FNP will make use of the Westinghouse RESAR-3 Consolidated version four-loop pressurized water nuclear reactor having a core power output of 3411 MW(t). %is reactor design is similar to that utilized at the Catawba Nuclear Station thits 1 and 2, reported on by the Comittee in its report of November 13, 1973. he scope of the FNP design includes the Nuclear Steam Supply Systm (NSSS) and the Balance of Plant (BOP).

% e complete system, which is to be mounted on a large floating platform, represents a standard unit which is beirg designed for use at sites which fall within an envelope of parameters or specifications. We plant design includes specific requirements for major components, piping systems, and other information recessary to ensure that both the NSSS and BOP are designed to protect the system from site--related hazards.

Application of the FNP concept will require an evaluation of each site to confirm its acceptability within the given envelo w.

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Honorable William A. Anders December 10, 1975 With respect to the site envelope, the Comittee recomends that further consideration be given to methods for the assessment of probabilities for given accident events, such as those involving chips.

Rather than treat each potential accident situation as a separate class of event, it may be more appropriate in some cases to evaluate the significance of a given class of event on the basis of the total probability of all events within that class.

We NRC Staff has identified several issues which remain to be resolved.

One pertains to the acceptability of criteria for containment shell buckling, including the behavior of the shell during construction. To be included in the assessment of this issue are the effects of deforma-tion of the containment foundation. Another issue concerns the effects and consequences on the FNP of the explosion ncarby of a petro'eum tanker.

Rese matters should be resolved in a manner satisfactory to thc NRC Staff.

We tomittee wishes to be kept informed.

Evaluation of the Emergency Core Cooling Systen (ECCS) design in accor-dance with Appendix K of 10 CFR Part 50 is also an outstanding issue which has been identified by the NRC Staff.

In this regard, the Comittee has special interests relating to detailed assessments of the upper head injection system, the r2 solution of potential problems with the ice condenser pressure suppression system, c.ad the margins available in the ECCS. We Committee wishes to review the design and analysis of both of these systeis prior to the NRC issuance of a license to manufacture the FNP units.

In the course of its review, the Comittee noted other areas wherein it wishes to be kept informed. % ese include any problems associated with turbine-generator alignment; hull-coupled vibrations (particularly as these relate to the potential of turbine failure and the generation of missiles); analysis of stresses on Aey components associated with platform towing operations; and the location and range of instruments for determining the nature and course of any accidents.

Since the FNP is a novel design requiring unusual structural reliability there is a need to develop plans for verification of structural design and to define the requirements for strain and deforiration measurements, visual inspection during operational testing, and nondestructive inspec-tion of critica'. FNP structures subsequent to operational loadirg. %e Comittee wishes to be kept informed.

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Honorable William A. Anders December 10, 1975 Consideration should be given in design to the possible previsions for redundant mooring systms.

We Comittee recommends that the NIC Staff and the Applicant review further the design features that are intended to prevent the occurrence of fires and to minimize the consequeras to safety-related equipnent should a fire occur. Wis evaluation should include a review of systers for detecting and protecting against fires, both within and outside the plant. Wis matter should be resolved to the satisfaction of the NIC Staff. %e Comittw wishes to be kept informed.

Also to be evaluated are the consequences of, and any safeguards nc, essary to cope with, a major accident which could lead to the dispersal of a significant quantity of radioactive materials into the water surrounding the FNP. We Comittee understands that this iten is being evaluated by the NRC Staff and the Applicant. We Comittee will reserve judgmnt on this item, which is both site and plant related, until it has had an opportunity to review and evaluate the relevant information.

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%e Cor,aittee suggests that analyses be made of any possible increases in N protection of public health and safety which may be obtained by en increase in containment design pressure.

%e Applicant has suggested the use of a coating and a cathodic systen to protect the platfonn against corrosion. W e proposed cathodic systen appears to be suitable for the underwater portion of the platform; however, additional attention should be given to means for protecting the critical wave and splash zone areas where repair or renewal my not be practical under the anticipated operating conditions of the FNP.

Because operating and mintenance personnel my te on board the floating platform for extended periods of time, and because shielding m" be limited due to weight restrictions and limitations on available space, ic is possi-ble that doses and dose rates to personnel on the FNP may te greater than for land-based units. As a result, the Committee believes that special consideration should be given to conformance with the "as low as reasonably achievable" criterion.

ne Comittee believes that the Applicant and the NIC Staff should continue to review the FNP design for features that could reduce the possibility and consequences of sabotage.

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Honorable William A. AnderS December 10, 1975 Se Comittee recomenc'3 that further attention be given to the possi-bility of extended loss of off-site power due to natural events or other causes, and the potential inpact of this possibility on the requirements for emergency AC power.

Generic problems relating to large water reactors are discussed in the Comittee's report dated March 12, 1975. 2e Comittee believes that procedures should be developed to incorporate approved resolution of these items into the FNP.

We Advisory Comittee on Reactor Safeguards M1ieves, that subject to the foregoing and to other applicable matters discussed in its reports of November 15, 1972 and October 18, 1973, the Floating Nuclear Plant units can be constructed with reasonable assurance that they can be operated without undue riak to the health and safety of the public.

me Comittee will caplete its review of tNs application when the necessary additional information has been developed.

Sincerely yours, W. Kerr Gairman B-4

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Honorable William A. Anders December 10, 1975 Paferences 1.

Floating Nuclear Plant (FNP) Plant Design Peport (PDR) Volumes 1-8 2.

Amend:mnts 1 through 17 to the PDR 3.

Safety Evaluation Peport by the Division of Reactor Licensing (DRL),

dated September 30, 1975 4.

&morandum of thderstanding Between the U. S. Coast Guard (USCG) and the U. S. Atmic Energy Comission for Regulation of Floating Nuclear Power Plants, dated January 4,1974 5.

Intter, dated September 5,1975, Offshore Power Systems (OPS) to DRL, transraittirg Westinghouse Report entitled " Valve Reliability, 'Ibtbine Inlet Valves," dated August 1975 6.

Intter, dated September 2,1975, USCG to DRL, providing comrrents on Offshore Power Systems Response to Staft Nsition Concerning Garpy V-Notch Testing of Weldments 7.

Letter, dated August 25, 1975, OPS to DRL, providing additional turbine missile information 8.

Intter, dated August 11, 1975, providing information on Emergency Core Cooling System Mrformnce 9.

Intter, dated August 8,1975, OPS to DRL, transmitting Revision 1 to the Platform Hull Drydocking Dquivalency document 10.

Ietter, dated June 3,1975, OPS to DRL, providing informtion on turbine missile penetration of steel barriers 11.

Iatter, dated May 21, 1975, OPS to DRL, regarding external fire protection system 12.

Intter, dated April 3,1975, USCG to DRL, providing plan regarding detai2F of exterior fire protection 13.

Intter, dated March 10, 1975, USCG to DRL, providing coments on Platform Hull Corrosion Control Plan 14.

Ietter, dated January 30, 1975, OPS to DRL, regarding Ibmaged Platform Stability B-5

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Ibnorable William A. Anders December 10, 1975 References - Continued 15.

Intter, dated January 23, 1975, OPS to DRL, tranmitting report on Control Ibd Drive Mechanism Analysis perfcrmed by Westir., house

16. Irtter, dated January 15, 1975, OPS to DRL, tranmitting reports requested by USOG

17. Intter, dated January 3,1975, USOG to DRL, providing ccxments on fracture toughness testing of hull steel for floating nuclear plants 18.

Letter, dated Decerrber 17, 1974, OPS to DRL, tranmitting errata sheet for report on external fire protection

19. Intter, dated Deceber 5,1974, OPS to DRL, regarding procedures for structural plans 20.

Ictter, dated tbverrber 22, 1974, OPS to DRL, transitting revision to Equivalency Derconstration docL. wnt 21.

Intter, dated tbymber 17, 1974, OPS to DRL, tranmitting four reports referenced in the PDR 22.

Ictter, dated Ibvmber 13, 1974, OPS to DRL, tranmitting nonproprietary report on Emergency Trip Svstms and Ultrasonic Inspection 23.

Intter, dated tbvmber 12, 1974, OPS to DRL, transmitting docurent entitled " Floating Nuclear Plant Platform !!ull Corrosion Control Plan 24.

Ictter, dated !bvaber 12, 1974, OPS to DRL, transmitting Westinghouse Electric Corporation reports 25.

Intter, dated October 16, 1974, USCG to DRL, enclosing August 29, 1974 letter fr m OPS and USCG's October 11, 1974 letter to OPS 26.

Ictter, dated October 8,1974, OPS to DRL, transmitting report entitled

" Wind and Wave Persistence and Ebrecast Irad Times for Fbur Offshore Ircations" R-6 t s DR O

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Intter, dated September 26, 1974, OPS to DRL, tran mitting position on Anticipated Transients Without Scram 28.

Ictter, dated September 6,1974, OPS to DRL, transmitting Westinghouse report entitled "A Comparison of Westinghouse Overspeed Protection to the Requirements of IEEE 279" 29, Intcer, dated August 29, 1974, OPS to DE, tranmitting three Westing-house reports 30.

Iatter, dated August 12, 1974, OPS to DRL, providing clarifying information regarding the calculation of tranmission line reliability 31.

Intter, dated August 8,1974, OPS to DRL, transmitting MIDMi code report 32.

Letter, dated August 7,1974, OPS to DRL, transmitting revision to Emergency Power Equivalency document 33.

Intter, dated August 2,1974, OPS to DRL, transmitting Westinghouse Electric Corporation report on Analysis of the Probability of the Generation and Strike of Missiles Prm A Nuclear 'Ibrbine 31.

Ictter, dated July 2,1974, OPS to DRL, tranmitting document entitled " Platform Inclinations Due to Damage to Any One Side" 35.

Intter, dated July 2 1974, OPS to DRL, transitting revision to Dnergency Power Equivalency document 36.

Intters, dated March 29 and February 25, 1974, USCG to DRL, regarding selection of hull material for Floating Nuclear Plants 37.

Ictter, dated January 7, 1974, OPS to DRL, transmitting information on Anticipated Transients Without Scram 38.

Ictter, dated October 26, 1973, OPS to DRL, providing interim informa-tion regarding Nuclear Plant Arrangement and Ice Condenser Design B-7

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APPEhTIX C U.S. DEPARTMENT OF COMMERCE Natlanal Oceanic and Atmospheric Administration MV!ADNML8tTAL DATA SERVICE l National Climctic Center

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Asheville, N. C. 28801 cate November 5, 1975 Rep!y to Attn of:

D5xl To Bob Kornasiewicz Earl Markee From Harold L. Crutcher Scientific Advisor sub ect:

Fastest Mile 100-Year Return Estimate i

Reference is made to our letter of tiarch 10, 1975 signed by Mr. Bill Brower and to your visit here on November 4,1975.

As indicated in our discussion, we see no need to revise our estimates of 160 and 360 mph for the extreme wind expected value and upper 0.975 probability confidence limit, respectively.

These are for 100-year return values for anywhere along the coast from Corpus Christi, TX to Nantucket, A1A and in the nearby oceanic areas any time.

It might be useful to stress the preliminary tables which we provided to you, which show that a 100-year return value has approximately a 1 in 3 chance of occurring in any 40-year period.

If in the course of future work, which hopefully would include more data, it becomes necessary to adjust the above estimates, you will be notified.

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