Text

Limited Const Activities Rept
	ML20062H082
Person / Time
Site:	Clinch River
Issue date:	08/31/1982
From:	; ENERGY, DEPT. OF
To:	;
Shared Package
ML20062H081	List: ML20062H077; ML20062H082;
References
	NUDOCS 8208130291
	Download: ML20062H082 (97)
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CLINCH RIVER BREEDER REACTOR PLAN

i LIMITED CONSTRUCTION ACTIVITIES TABLE OF CONTENTS PJ1 Sit

SUMMARY

vil 1.0 I NTR ODUCT I ON....................

1-1 1.1 PURPOSE AND B ACKGROUND............

1-1 g

1.2 SCDPE OF LIMITED CONSTRUCTION ACTIVITIES...

1-2 1.2.1 NUCLEAR ISL AND MAT..........

1-2 1.2.2 REACTOR CONTAINENT BUILDING,.....

CONTAINENT VESSEL SHELL AND CONF I N EENT..............

1-3 1.2.3 STEAM GENERATOR BUILDING.......

1-5 1.2.4 CONTROL AND ELECTRICAL EQUlFENT...

BUILDING...............

1-6 1.2.5 REACTOR SERVICE BUILDING.......

1-7 1.2.6 ADDITIONAL ACTIVITIES.........

1-7 1.3 SCHEDULE FOR LIMITED CONSTRUCTION ACTIVITIES.

1-7 2.0 CRB RP S I TE.....................'

2-1 3.0 DESIGN CRITERI A - STRUCTURES, CDMPONENTS, EQUIPENT AND SYSTEMS....................

3-1 3.1 CONFORMANCE WITH GENERAL DESIGN CRITERI A...

3-1 3.2 C1.ASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTS..................

3-3 3.3 WIND AND TORNADO LOADINGS...........

3-5 l

TMLE OF CONTENTS (Continued)

C Eaga 3.4 WATER LEVEL (FLOOD) DESIGN.

3-7 3.5 M I S S I LE PROTECT I ON..............

3-9 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCI ATED WITH THE POSTULATED RUPTURE OF PIPING.....

3-12 3.7 SEISMIC DESIGN...............

3-13 3.8 DES IGN OF CATEGORY 1 STRUCTURES.......

3-16 3.9 E CHANICAL SYSTEMS AND COMPONENTS......

3-20 3.10 SEISMIC DESIGN OF AND CATEGORY 1 INSTRUENTATION AND ELECTRICAL EQUIPMENT...

3-20 3.11 ENVIRONENTAL DESIGN OF MECHMICAL AND EL ECTR ICAL EQU IPMENT.............

3-21 4.0 REACTOR......................

4-1 5.0 H5AT TRMSPORT MD CONNECTED SYSTEMS........

5-1 6.0 ENG INEERED SAFETY FEATURES.............

6-1 6.1 GENERAL (NOT APPLICABLE) 6-1 6.2 CONTA I N EN T SYSTEMS..............

6-1 6.3 CONTROL ROOM HABITABILITY SYSTEMS 6-3 6.4 CELL L INER SYSTEM...............

6-3 6.5 CATCH PANS..................

6-4 7.0 INSTRUENTATION AND CONTROLS............

7-1 8.0 EL ECTR I CAL POW ER..................

8-1 9.0 AUXILIARY SYSTEMS.................

9-1 C

11

TABLE OF CONTENTS (Continuec)

(

PJER 9.1 FUEL STORAGE AND H ANDL ING...........

9-1 g

9.2 MAINTENANCE..................

9-1 9.3 AUXIL l ARY L IQUID METAL SYSTEM.........

9-2 9.4 PIPING AND EQUIPMENT ELECTRICAL HEATING....

9-2 9.5 INERT GAS RECEIVING AND PROCESSING SYSTEM...

9-3 9.6 HEATING, VENTILATING MD AIR (DNDITIONING (HVAC) SYSTEM,................

9-4 9.7 CH ILLED W ATER SYSTEMS.............

9-4 9.8 IMPURITY MONITORING AND ANALYSIS SYSTEM....

9-5 9.9 SERVICE WATER SYSTEMS.............

9-5 9.10 COMPRESSED GAS SYSTEM.

9-6 9.11 COMUNICATIONS SYSTEM.

9-6 9.12 LIGHTING SYSTEM.

9-6 9.13 PL ANT FIRE PROTECTION SYSTEM.....

9-7

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9.13.1 NON-SODIUM FIRE PROTECTION SYSTEM.

9-7 9.13.2 SODIUM FIRE PROTECTION SYSTEM.....

9-8 1

9.14 DIESEL GENERATOR MJXIL I ARY SYSTEM.......

9-8 9.15 EQUIPMENT AND FLOOR DRAINAGE SYSTEM.

9-8 9.16 RECIRQJLATING GAS COOLING SYSTEM.

9-9

~ i 10.0 STEAM AND POWER CONVERS 10N SYSTEM.........

10-1 11.0 RADIOACTIVE WASTE SYSTEM.........

11-1

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111

TABLE OF CONTENTS (Continued)

G Eaga 12.0 RADIATION PROTECTION................

12-1 12,1 SHIELDING...................

12-1 12.2 V EN T I L AT I ON..................

12-2 12.3 HEALTH PHYSICS 12-3 13.0 CON DU CT OF OPERAT I ON S...............

13-1 14.0 IN ITI AL TESTS MD OPERATION............

14-1 15.0 ACCI DENT AN ALYS I S.................

15-1 15.1 I N TRODU CT ION............

15-1 15.2 REACTIVITY INSERTION EVENTS.....

15-1 15.3 UNDERCOOL ING EVENTS..............

15-1 15.4 LOCAL FA I LURE EV EN TS.............

15-2 15.5 FUEL HANDL ING MD STORAGE EVENTS.......

15-2 15.6 SODIUM SPILLS.................

15-2 15.7 OTH ER EV ENTS.................

15-4 15.A SITE SU ITAB IL ITY SOURCE TERM.........

15-5 16.0 TECHN I CAL SPEC I F I CAT I ON S..............

16-1 17.0 QU AL I TY AS SU RAN E.................

17-1 APPENDIX A - MJREG-0718, "L ICENSING REQUIREMENTS FOR PENDING APPLICATIONS FOR CDNSTRUCTION PERMITS AND MANUFACTURING L ICENSE"..........

A-1 4

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TABLE OF CONTENTS (Continued)

(

l Enge APPENDIX B - HYPOTHETICAL CORE DISRWTIVE ACCIDENT CX)N S I DERAT ION S................

B-1

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INTRODUCTION B-1 B.1 STRUCTURAL MARGIN BEYOND THE DESIGN BASE.

B-1 B.2 THERMAL MARGIN BEYOND THE DESIGN BASE..

B-2 REFEREN S i

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Floure No.

Title Page 1.2-1 Key Plan.......................

1-9

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1.2-2 Nuclear Island Mat - Bottom Lif t (715 feet-724 feet).

1-10 1.2-3 Nuclear Island Mat - Top Lif t (724 feet-733 feet) 1-11 1,2-4 Status of Construction Activities..........

1-12 1.2-5 Containment Vessel Skirt Support Cetall.......

1-13 1.2-6 RCB Cross S ect i on..................

1-14 1.2-7 SGB Cross Section..................

1-15 1.2-8 CB Cros s S ect i on...................

1-16 1.2-9 EEB Cross Sect i on..................

1-17 1.2-10 RSB Cro s s S ect i on.......

1-18 1.3-1 Schedule of Limited Construction Activities.....

1-19 1

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SUWARY

'l The scope of the limited construction activities proposed f or CRBRP is the construction of building f oundations and walls to grade elavation.

This includes the Nuclear Island basemat and the external and internal walls including anbedments and Integral slabs. Construction of portions of the g

containment vessel, cell liners, and embedded condult and piping would be included to the extent necessary to support a logical construction sequence.

Site suitability was established by the NRC 1?sf f in its Site Sultability Renort (SSR) Issued in 1977 and revised in June 1982. The current PSAR shows that site and plant characteristics continue to support the site suitability conclusion.

The general design criteria and regulatory guidelines relevant to the limited construction activities described herein, have been addressed in the CRBRP design.

The review areas (i.e., subjects of regulatory criteria and guidelines) of primary significance f or the plant features to be constructed t

include the following:

o Site Sultability Source Term Mitigation 1

o Reactor Containment Building Design Basis Accident Definition and Mitigation o

Design of Plant Structures to Accommodate Seismic and Design Basis Accident Conditions o

Plant Margins to Accommodate Events Beyond the Design Basis The plant f eatures to be buil t during the limited construction activities (e.g., the Nuclear Island basemat) have been evaluated relative to the applicable criteria in these review areas.

The features are designed and will l

be constructed in a manner that satisfies the criteria.

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Certain plant features will be ef f ected by the construction activities.

Pending resolution of applicable issues in the on going safety review, g

constraints will be put on construction activities to assure that options are not foreclosed f or those systems, structures, and components not covered by this request.

In these areas, the adequacy of the remaining design and operational flexibility has been established to assure that, af ter completion of the limited construction activities, all remaining saf ety review areas can be resolved.

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f LIMITED CONSTRUCTION ACTIVITIES q

1.0 INTRODUCTION

i 1.1 PURPOSE AND B ACKGROUND The Clinch River Breeder Reactor Plant (CRBRP) site has been well charac-terized as a result of extensive physical and environmental studies conducted at the site and its environs since the initial planning phase of the Project in 1972. A comprehensive assessment of the Impacts of the proposed activities on the CRBRP site Indicates that all environmental values will be adequately protected.

I The CRBRP Project submitted an Environmental Report in October,1974, and a Preliminary Saf ety Analysis Report in April,1975. These were filed under NRC Docket No. 50-537. These initial submittals were supplemented with anendments to both documents. Following NRC staf f review, a Final Environmental Statement (FES) was issued by the NRC in February, 1977, (1-1) In which it was concluded, "In the event the applicant is permitted to proceed with site preparation under a Limited Work Authorization, it is the staf f's opinion that the environmental Impacts of such work would not be significant." (1-2) A draf t supplement to the FES was issued in July 1982, (1-5) in which the f ore-l going conclusion remained unchanged.

l In March,1977, the NRC issued a Site Sultability Report (1-3) which contained the following conclusion, "In detennining the acceptability of the CRBRP site, the staf f has considered the following f actors: population density and use l

characteristics of the site environs, including the exclusion area; low population zone, and population center distance; and, physical characteristics of the site, including seismology, meteorology, geology, and hydrology. The staf f concludes that the above characteristics of the Clinch River site are acceptable." (1-4)

In addition, the NRC's June 1982 update of the SSR also 1-1

confirms the March 1977 SSR conclusions. (1-3) in July 1982, the ACRS en-dorsed the NRC staf f's findings on site suitability. (1-6)

{

This document supports the request for approval of limited plant construction activities extending for a period of time following the completion of site preparation activities, it contains the scope, schedule, and sequence of the limited construction activities, it is arranged to be read in conjunction with the CRBRP PSAR and, in Chapters 2 through 17, it is organized to the same outline, chapter for chapter. References to the applicable PSAR sections are contained throughout.

1.2 SCOPE OF LIMITED CONSTRUCTION ACTIVITIES The Nuclear Island Mat (NIM) forms the foundation for the Reactor Service Buil ding (RSB), the Reactor Containment Building (RCB), the Steam Generator Buil ding (SGB), the Control Building (CB) and the Electrical Equipment Buil ding (EEB). Section 1.2 of the PSAR details the plant layout and relationships of the various buildings on site.

(See Figure 1.2-1)

The following paragraphs, Table 1.2-1 and Figure 1.3-1, outline the scope and sequence of the limited construction activities.

1.2.1 NUCLEAR ISL AND MAT (NIM)

Rebar placement will start from the west side of the CB and EEB and move east toward the RCB.

Concurrent with placing rebar in the lower portion of the mat, water stops wil l be Installed.

The concrete work will start on the west side of the CB and EEB.

When the first block in the area under the RCB becomes available work will start in this area. Form work wil l use expanded metal with structural shapes f or E

stab i l i ty.

The structural shapes will be " stripped" af ter the concrete hardens and the expanded metal lef t in place.

Concrete placements will be 4

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1-2

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" checker-boarded" and the area under containment will be given highest 1

priority (Figure 1.2-2).

Concrete placement sequence f or the f irst Ilf t (bottom Ilf t) wil l be as f ol low s:

e 1.

Reactor Containment Building 2.

Outside blocks on the west side of the NIM 3.

Outside blocks on the east side of the NIM 4.

Remaining blocks The f irst top l i f t w il l be f or the outsi de wal l areas. The remaining blocks, except those that abut the contai nment shel l, wil l fol low. The blocks that immediately surround the shell must await l iner erection (Figure 1.2-3).

The top of the NIM is the floor of the building as described in Paragraph 1.2.

With the exception of the RCB, the floor of the buildings will be the top surf ace of the nominal second lif ts.

The concrete will contain embedded floor drains, pipe, conduit, grounding, equipment foundations, miscellaneous embedments and reinf orcing steel dowels f or concrete walls (Figures 1.2-5 through 1.2-10).

t Operations above the level of the second lift on the NIM will follow building by building, generally traveling south to ncrth (Figure 1.2-4).

1.2.2 REACTOR CONTAINENT BUILDING, CONTAINENT VESSEL SHELL AND CONFINEENT When the first lif t of concrete is placed in the blocks under and immediately around the contai nment shel l (hereaf ter called the shel l), support beams wil l be placed on the shell circumference. The first shell rings will be erected using these beams as a base. Bef ore proceeding beyond the second ring (each ring is about 11 1/2 feet high) concrete will be placed in a lift to embed the i

1-3 i

shell skirt and embedded bottom liner supports at elevation 730 feet. (See Figure 1.2-5)

I Af ter the installation of reinforcing steel and embedded bottom liner support, concrete will be placed to elevation 730 feet.

When concrete is placed to 730 feet, the containment bottom liner and leak channels will be Installed.

Construction of the three feet of concrete between the containment liner and the floor of the RCB will start on the south side, proceeding toward the north. Work on the south side includes several lined cells.

When this floor slab is placed it will incl ude: embedding the lower frame of these lined cells, embedded floor drains and sumps, conduit, pipe, miscellaneous embedments, and equipment foundations.

Cell liner segments will be pref abricated into the largest practical sizes in an onsite shop. They will be set in place by one of the large lif t cranes.

Af ter placing concrete to the floor elevation of the RCB, the containment placements will proceed in the following sequence:

1.

Reactor vessel cavity walls.

2.

Walls just inside the containment shell.

3.

Buckling concrete. (outside the shell) 4 Three primary loop cel ls.

5.

South side where various lined cells are located.

6.

Remaining walls, slabs, etc.

Walls and floor slabs will include embedded pipe, pipe and duct sleeves, grounding, equipment onbedments, miscellaneous onbedments, floor drains, and condult.

8 During this period, concrete placed inside containment will reach approxi-mately elevation 780 feet. (See Figure 1.2-6).

G 1-4

4 Shell erection wil l proceed as shown in Table 1.2-1 and Figure 1.3-1.

During this period, the shell will be erected to approximate elevation 807 feet (see Figure 1.2-6).

Confinement walls are integral walls of the adjoinir.g buildings and are treated as such in this document.

1.2.3 STEAM GENERATOR BUILDING Concrete at elevation 733 feet will include equipment foundations f or the Ex-Containment Sodium Storage Tanks, Sodium Dump Tanks and Reaction Products Tanks.

I Work on the SGB will concentrate on the intermediate and auxillary bays.

Walls and slabs will contain embedded pipe, pipe and duct sleeves, grounding, equipment embedments, miscellaneous embedments, floor drains and condult.

Embedments for the catch pans will be Installed in the Loop Cell areas while concrete is being placed between elevations 733 feet and 764 feet.

Catch pans at el ev ati on 733 f eet w i l l be stored i n pl ace f or l ater i nsta l l ati on.

During this period no concrete or structural steel will be erected higher than 806 feet in the Steam Generator Building. (See Figure 1.2-7).

1.2.4 CONTROL BUILDING AND ELECTRICAL EQUIPENT BUILDING Both building walls will be built from the west side to grade elevation with exterior walls constructed first.

The interior walls are concrete. Col umns are concrete or structural steel.

l Floor slabs are concrete placed on steel decking, supported by the columns and 1-5

walls. Concrete wil l include embedded pipe, pipe and duct penetration sleeves, equipment embedments, miscellaneous anbedments, grounding, floor I

drains and condult. (See Figures 1.2-8 and 1.2-9).

1.2.5 REACTOR SERVICE BUILDING Outside wal ls wil l be constructed f irst, followed by the first wall parallel to the outside wall, and intersecting walls and floor slabs between these parallel walls.

This building contains several lined cells, and cells containing sodium or NaK piping, which, if not lined, will have catch pans.

Outside walls and a portion of the slab at elevation 779 feet will be completed during this time period.

Walls and slabs will contain embedded pipe, pipe and duct penetration sleeves, equipment embedments, miscellaneous anbedments, grounding, floor drains and condult. (See Figure 1.2-10) 1.2.6 ADDITIONAL ACTIVITIES Additional work scheduled for this period includes installation of miscellaneous underground piping; grounding and cathodic protection; and placement of structural backfill and lean concrete backfill af ter floors have been constructed.

1.3 SCHEDULE FOR LIMITED CONSTRUCTION ACTIVITIES Figure 1.3-1 shows the schedule for the activities in each of the major structures to be accomplished as part of the limited construction activities, t

4 1-6

I TM3LE 1.2-1

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LIMITED CONSTRUCTION ACTIVITIES A.

NUCLEAR ISLAND f'AT (RG, RSB, G, EEB, SGB) 1.

Placement of rebar 2.

Placement of forms

_g 3.

Placoment of embedded steeI (Equip. foundations, etc.)

4 Placement of concrete 5.

Installation of embedded floor drains 6.

Installation of embedded conduit & grounding B.

CONTAINfENT VESSEL SHELL 1.

Shell rings below grade 2.

Installation of Bottan Liner Plate with Leak Chases 3.

Start installation of Penetrations, Airlock, Equipment Airlock, Ref ueling Hatch C.

REACTOR CONTAINENT BUILDING & CONFINEMENT s

1.

Interior walls and slabs (concrete), including rebar and embedments (sleeves, equipment supports, conduit, pipe, cell liners, etc.)

2.

Buckling concrete including, rebar and penetrations 3.

Confinement concrete, including rebar and penetrations t

D.

STEAM GENERATOR BUILDING 1.

Concrete, including rebar and embedments a.

Rebar b.

Formwork c.

Embedments (sleeves, catch pan, equipment supports, conduit, pipe, etc.)

d.

Concrete 2.

Installation of Structural Steel and Miscellaneous Ironwork 1-7

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E.

CDNTROL BUILDINGS 1.

Concrete, including robar and embedmonts 4

2.

Installation of structural steel and miscellaneous Ironwork 3.

Equipment foundations F.

ELECTRICAL EQUIPENT BUILDING 1.

Concrete, including rebar and embedments 2.

Installation of structural steel and miscellaneous ironwork 3.

Equipment foundations G.

REACTOR SERVICE BUILDING 1.

Concrete, including rebar and embedments 2.

Installation of structural steel and miscelIaneous f ronwork H.

GROUNDING AND CATH 001C PROTECTION 1.

STRUCTURAL BACKFILL, INCLUDING LEAN CONCRETE J.

MISCELLANEOUS UNDERGROUND PlPING O

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BUILDING CONTROL u

BUILDING Q PROCESS PIPING RADWASTE BUILDING y:

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AY GENERATOR REACTOR BUILDING CONTAINMENT

' STEAM REACTOR BUILDING GENERATOR SERVICE BUILDING BUILDING INTERMEDIATE BAY Figure 1.2-1 Key Plan 7 82 25091 af

BOTTOM LIFT - 715'-724' (SHADED AREA INDICATES BLOCKS THAT MUST BE IN PLACE TO COMPLETE THE CONTAINMENT SKIRT)

PLANT NORTH REACTOR CONTAINMENT

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E Figure 1.2-2. Nuclear Island Mat - Bottom Lif t (715*-724*)

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2.0 CRBRP S ITE -( All limited site construction activities will be performed in accordance with the requirenents set forth in Chapter 2 of the CRBRP PSAR. The site characteristics described in Chapter 2 of the PSAR meet the relevant requirements of 10CFR Parts 50 and 100, the CRBRP Design Criteria and regulatcry guidance. The NRC staf f previously concluded and recently reaf firmed that the CRBRP site was acceptable and issued a f avorable Site Sultabil ITy Report (SSR) In support of a Limited Work Authorization for the

Project, in determining the suitability of the CPERP site, the staf f considered the folicwing f actors: populat!on density and use characteristics of site environs, including the exclusion area, low population zone, and population center distance; and, physical characteristics of the site, incl uding sei smology, meteorology, geology, and hydrology.

'( o '( 2-1 s

3.0 DESIGN CRITERI A - STRUCTURES, COMFONENTS, EQUIPENT AND SYSTEMS ( 3.1 CONFORMANCE Wim GENERAL DESIGN CRITERI A General Design Criteria (GDC) for CRBRP were specified by the NRC in a letter f rom R. P. Denise to L. W. Caf fey, dated January,1976. These GDC were also published by the NRC in the CRBRP Site Sultability Report. The conf ormance of 013RP design to the GDC is discussed in Section 3.1 of the PSAR. LIMITED CONSTRUCTION ACTIVITIES Most of the activities listed in Section 1.2 of this report are relevant to this section: the setting of Nuclear Island Mat Structures as well as the provision of interior concrete f or the Reactor Containment Buil ding, the Steam Generator Building, the Control Building, the Electrical Equipment Building, and the Reactor Service Building; and the installation of cell liners and embedded piping. APPLICABLE CRITERI A The activities identified above must conform to the established CRBRP General Design Criteria. EVALUATION The principal design criteria which establish the requirements for water-cooled nuclear power plants are identified in Appendix A to 10 CFR Part 50. Although specifically oriented toward water-cooled reactors, Appendix A of 10 CFR Part 50 is " Intended to provide guidance in establishing the principal design criteria for such other (types of) units." The CRBRP General Design Criteria were developed by the NRC staf f and issued to the CRBRP Project in January,1976. These General Design Criteria, which i 3-1 s

are the principal design criteria for CRBRP, are comparable to those in Appendix A of 10 CFR Part 50, and will assure a level of saf ety comparable to E current generation light water reactor plants. Chapter 3.1 of the PSAR discusses the principal design features that satisfy the CRBRP GDC and in particular, the criteria which impose constraints on the design of plant strutures to be constructed. The critdria that directly affeet plant structures are: Criterion 1 Qual Ity Standards and Records Criterion 2 Design Bases for Protection Against Natural Phenw.ena Criterion 14 Containment Design Criterion 41 Containment Design Basis Criterion 42 Fracture Prevention of Reactor Containment Boundary Criterion 43 Capability for Containment Leakage Rate Testing Criterion 44 Provisions for Containment Testing and inspection PSAR Section 3.1 discusses conformance of the design of the below grade plant structures to these criteria. The Criteria that Indirectly influence the design of the plant structures to be constructed are: Criterion 3 Fire Protection Criterion 4 Protection Against Sodium Reactions Criterion 5 Environmental and Missile Design Bases Criterion 15 EloctrIc Power Systems Criterion 20 Protection System independence Criterion 35 Reactor Residual Heat Extraction Systems Criterion 38 Additional Cooling Systems Criterion 53 Fuel Storage and Handling and Radioactivity Control l l 3-2

i PSAR Section 3.1 discusses tne CRBRP approaches to conformance with each of ( these criteria. The plant structures below grade provide appropriate potential for fire protection by isolating fire zones. The structures of liquid metal containing cells include provisions f or protection against sodium reacti ons. Separation of equipment by the plant structures provide protection 4 against environmental and missile hazards in the plant. The internal structures provide space and separation for redundant trains of both electric power systems and protection systems. The Internal structures provide space and separation f or redundant trains of reactor residual heat removal systems and additional cooling systems. The below grade structures provide space and support f or f uel storage and handl ing equipment. CONCLUS ION t The General Design Criteria used are those developed by the NRC and published by the NRC in the CRBRP Site Sultability Report. These GDC wil l assure, when satisfied, a level of safety comparable to current generation Light Water Reactor Plants. The structures to be constructed conform to the applicable CRBRP General Design Criteria. The structures provide appropriate space, support and separation f or plant f eatures necessary to assure conf ormance with other CRBRP GDC. i 3.2 CLASSIFICATIONS OF STRUCTURES, SYSTEMS, AND COMPONENTS Section 3.2 of the PSAR discusses selsnic classification and quality group (saf ety) classif ications of CRBRP structures, systems and components. l LIMITED CONSTRUCTION ACTIVITIES l Most of the activ!tles listed in Section 1.2 of this report are relevant to this section: the setting of Nuclear Island Mat Structures as well as the provision of interior concrete for the Reactor Containment Building, the Steam Generator Building, the Control Building, the Reactor Service Building and the ( 3-3

Electrical Equipment Building; and the installation of cell liners and embedded piping. 4 APPLlCABLE CRITERI A Plant structures to be constructed must conform to: (A) The CRBRP General Design Criteria 1 and 2; and (B) 10 CFR 100 Appendix A as it relates to structural and component design to withstand the safe snutdown earthquake and remain f unctional. EVALUATION Those OERP plant structu. es which are included in the limited construction activities have been classified and designed in accordance with the requirenents of Regulatcry Guide 1.29, " Seismic Design Cl assi f ication". The Reactor Containment Buil ding, Steam Generator Building, Control Buil ding, Reactor Service Building, and Electrical Equipment Building are classified as Seismic Category i Structures. The cell liners are designated as Seismic Category 1 Engineered Saf ety Features. The embedded piping which is important to safety has been classified Saf ety Class 1, 2 or 3 in accordance with Regulatory Guide 1.26. Section 3.2 of the PSAR discusses the detailed application of Regulatory l Guides 1.26 and 1.29. i l CONCLUSION The Iimited construction activities satisfy the requirements that structures and components important to saf ety be designed to withstand the ef fects of ( 3-4 l

I earthquakes without loss of capabil Ity to perf orm their saf e1y functions, and ( should be designed, f abricated, erected, and tested to quality standards cunmensurate with the importance of the saf ety f unction to be performed. There are no systems or components clar.sified as 1E that are included in the a limited construction activities. 3.3 WIND AND TORNADO LOADINGS t Section 3.3 of the PSAR, " Wind and Tornado Loadings", detail s the design bases for the plant structures with respect to potential wind and tornado loadings. LIMITED CONSTRUCTION ACTIVITIES The activities in Section 1.2 related to this section are: o Nuclear Island Mat o Containment Vessel ShelI o Interior and outer bull ding wal Is APPLICABLE CRITERI A The Iimited construction activities must comply with the relevant requirenents of GBRP Design Criterion 2 concerning natural phenomena in conformance with Appendix A to 10CFR50. EVALUATION Seismic Category I structures must withstand the ef fects of the design wind load, design tornado wind load and associated missiles and must not be rendered incapable of performing their necessary safety functions by the f ailure of any structure or component not designed for tornado wind loads. l ( 3-5 l l

Wind design criteria in CRBRP meet the requirements delineated in ANSI A58.1 and ASCE Paper No. 3269. C CRBRP structures have been designed with suf ficient margin to prevent structural damage during the most severe wind or tornado event. The Spectrum A of tornado-generated missiles established in the SRP Section 3.5.1.4 (November 24, 1975) has been used in the CR3RP design. The procedure of transforming the impacting missile load into an equivalent static load was f ormulated f rom the paper by Wil liamson and Alvy, " Impact Ef f ect of Fragments Striking Structural Elements", November 1973. PSAR Section 3.3 demonstrates compliance with the relevant requirenents of CRBRP Design Criterion 2 by identifying: o The bases f or sel ecting the wind, tornado wind, and associated missiles generated by the tornadic winds, used in the design; o The bases f or determining site-related wind and tornado parameters; o The procedures (in accordance with ANSI A58.1) used to transf orm the wind velocity and tornado parameters into ef fective loadings on structures; o How f ailure of structures, not designed f or tornado loads, will not af fect the capability of tornado-protected Seismic Category 1 structures to perf orm necessary saf ety functions; o How CRBRP Design Basis Tornado characteristics and structures protected against tornados are consistent with Regulatory Guides 1.76 and 1.117. Since the structures to be constructed are below grade, modification of above grade portions of these structures would not be foreclosed in the event that changes in wind or tornado design bases occurred. 3-6

CONCLUSION t The structures to be constructed meet the requirements of CRBRP Design Criterion 2 with respect to the capability of the structures to withstand design wind loading, design tornado wind loading and tornado missiles. i 3.4 WATER LEVEL (FLOOD) DESIGN Section 3.4 of the PSAR, " Water Level (Flood) Design", detail s the design bases for the plant structures with respect to flood protection and analysis procedures f or flood and hydrostatic loads identif ied in PSAR Section 3.8, " Design of Category 1 Structures." LIMITED CONSTRUCTION ACTIVITIES The activities lIsted in Section 1.2 related to this section are: o Nuclear Island Mat o Containment Vessel Interior and outer building walls o APPLICABLE CR1TERIA The Iimited construction activities must be determined to comply with the relevant requirements of CRBRP Design Criterion 2 concerning natural phenomena and 10CFR Part 100, Appendix A, Section IV.C relating to protecting structures, systems and components important to safety from the ef fects of floods, tsunamis and seiches. EVALUATION As described in PSAR Section 3.4, the maximum design flood level for CRBRP is at least six feet below plant grade, thus precluding the need to consider wave 3-7

action dynamic l oads. Hydrostatic head is considered as a structural load because it will not be relieved by utilizing a drainage and pumping system E around the foundations of structures. (CRBRP is not subject to tsunamis and seiches because of its inland location). PSAR Section 3.4 demonstrates compliance of the CRBRP design with the relevant '? criteria by identify Ing: o Seismic Category 1 systems and equipment requiring flood protection; o The methods utilized f or establishing the flood elevation design bases in accordance with Regulatory Guide 1.59, " Design Basis Floods f or Nuclear Power Plants;" o The means by which flooding of saf ety-related systems or components is prevented; o The analytical procedures utilized to develop flood and groundwater loadings as inputs to the design of selsnic Category 1 structures; and o That the CRBRP structures have been designed to provide protection from flooding in accordance with Regulatory Guide 1.102. " Flood Protection f or Nuclear Power Plants." CONCLUSIONS The 1Imited construction activities cornply with the relevant requirements of CR3RP Design Criterion 2 and 10CFR Part 100, Appendix A, with respect to protection of structures, systems and components important to safety from the ef f ects of floods, tsunamis, and seiches, and with respect to the capabil ity of structures to withstand the ef fects of the highest groundwater on flood l evel. i 4 l 3-8

3.5 MISSILE PROTECTION i Section 3.5 of the PSAR, " Missile Protection", provides the design bases f or structures designed for missile protection. LIMITED CONSTRUCTION ACTIVITIES The structures to be constructed which are relevant to this section are: Steam Generator Building walls and slabs o o Control Building walls and slabs o Reactor Service Building walls and slabs o Reactor Containment Building interior walls and slabs o Conf inement concrete APPL ICABLE CRITERI A The portions of the plant structures to be constructed mue+ comply with CRBRP General Design Criterion 2, " Design Basis f or Protection Against Natural Phenomena," and Criterion 5, " Environmental and Missile Design Basis." These plant structures must withstand the ef fects of Internally generated missiles, without loss of capability to perform their safety function. EVALUATION 1. Internally Generated Missiles (outside containment) The plant structures to be constructed and which are important to safety have been designed to withstand the ef fect of internally generated missiles that may result from equipment f ailure in accordance with Regulatory Guide 1.115. Internally generated missiles have been identifled and structures and components important to saf ety wil l be protected by missile-proof structures and barriers. (See PSAR Sections 3.5.2.1, 3.5.2.2, and 3.5.4). j 3-9 l

2. Internally Generated Missiles (Inside c,ontainment)

\\

The plant structures to be constructed and which are important to safety have been designed to withstand the ef fects of Internally generated missiles in accordance with Regulatory Guide 1.115. Sources of missiles inside containment have been identified and safety related structures and components will be protected by providing missile-proof structure, physical separation of redundant systems, and missile barriers. (See PSAR Section 3.5.2.1.2, and 3.5.4). 3. Turbine Missiles There are no plant structures or components to be constructed in this phase which might potentially be exposed to turbine generated missiles. 4 Missiles Generated by Natural Phenomena There are no plant structures or components to be constructed in this phase which might potentially be exposed to missiles generated by natural phenomena, since the Iimited construction activities are confined to structures below grade and full protection is afforded by structures above grade. 5. Site Proximity Missiles (including Aircraf t Hazards) There are no plant structures on components included in this phase which might potentially be exposed to site proximity missiles (including aircraf t hazards), since the limited construction activities are confined to structures below grade and f ull protection is af forded by structures above grade. ( 3-10

6. Structures, Systems, and Components to be Protected f rom External ly Generated Missiles There are no plant structures or components to be constructed in this phase which might potentially be exposed to externally generated missiles, since the limited construction activities are confined to structures below grade and full protection is afforded the structures above grade. 7. Barrier Design Procedures The structures, barriers, and shields that must withstand the ef fects of environmental and natural phenomena have been designed in accoraance with 093RP General Design Criteria 2 and 5. Suf ficient thickness of concrete or soll is provided to prevent perf oration, spalling, or scabbing of the barriers in the event of missile impact. (PSAR Sections 3.5.4.4.1 and 3.5.4.4.2). Acceptable empirical equations are used to estimate the missile penetration into concrete (PSAR Section 3.5.4.1). However when concrete is not used, a minimum of 8 feet thick soll barrier is employed. I The Stanford Equation and the Ballistic Research Laboratory Formula are used in analyzing the adequacy of the steel containment vessel (PSAR Section 3.5.4.3). The structural response to the missile impact loading is evaluated based on equivalent static loading of the impact. (PSAR Section 3.5.4.5). CONCLUSION The design of the portions of the structures to be constructed complies with the relevant requirements of CRBRP Design Criteria 2 and 5 with respect to misslie protection. Margin exists in the design and added protection can be provided if minor variations in the missile design bases occur. Options te provide protection against missiles above grade elevations are not f oreclosed. 3-11

3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCI ATED WITH THE POSTULATED g RUPTURE OF PlPING SecTion 3.6 of the PSAR discusses CRBRP design f or protection against dynamic ef fects associated with tr.e portulated rupture of piping and the design bases. = LIMITED CONSTRUCTION ACTIVITIES Most of the activities listed in Section 1.2 of this report are relevant to this section: the setting of Nuclear Island Mat Structures as well as the provision of intericr concrete for the Reactor Containment Building, the Steam Generator Building, the Reactor Service Building, and the installation of cell liners. APPLICABLE CRITERI A The structures to be completed in this phase must conform to CR3RP General Design' Criterion 5 " Environmental and Missile Design Bases" as it applies to protection against dynamic ef fects associated with postulated pipe ruptures. Structures included in the limited construction activities must be designed to provide protection against postulated pipe ruptures, including appropriate separation. EVALUATION The plant structures to be constructed have been designed in conformance to Branch Technical Position ASB 3-1 (formerly APCSB 3-1) and the guidance contained in J. F. O' Leary letter 7-12-73, appropriately applied to CRBRP. The principle approach used is to provide separation between moderate and high energ/ systems and essential systems. The structures included in the limited construction activities are arranged to provide for such separations, with structural barriers between each of the three heat transport loops, and between moderate and high energy systems and essential systems. Margin exists 4 3-12

f in the design and added protection can be provided (such as additional fixed restraints, shields, etc. ) If variations in the pipe rupture design bases ( occurs. CDNCLUSION The design of the plant structures to be constructed conforms to the relevant requirements of CRBRP General Design Criterion 5 with respect to providing protection against the dynamic ef fects associated with postulated pipe ruptures. Options to modify systems designs to reduce hazards from piping leaks are not foreclosed by the Iimited construction activities. 3.7 SEISMIC DESIGN Section 3.7 of the PSAR, "Selsmic Design", describes the bases and require-ments for the design of components, systems and structures designated as Category 1 as well as the methods of analysis to assure that the applicable requirements are satisf ied. LIMITED CONSTRUCTION ACTIVITIES Most of the activities listed in Section 1.2 of this report are relevant to i this section: the setting of Nuclear Island Mat Structures as well as the provision of Interior concrete for the Reactor Containment Building, the Steam Generator Bullding, the Control Bullding, the Electrical Equipment Bullding, and the Reactor Service Building. APPLICABLE CRITERI A Compilance with the provisions of CRBRP Design Criterion 2 and Appendix A to 10CFR Part 100 relating to earthquakes is required for Category 1 structures. 3-13

EVALUATION S 1. Seismic Design Parancters The plant structures and components important to safety have been designed to withstand the ef fects of earthquakes in accordance with CR3RP General Design Criterion 2 and 10CFR100, Appendix A. The SSE and OBE have been selected based on ilstorical data for the site and surrounding area. (See PSAR Sections 2.5, " Geology and Seismology" and 3.7-A paragraphs 3 and 4, " Site Description," Operating Basis f or Saf e Shutdown Earthquakes"). The CBE anc SSE design response spectra for the site are linearly scaled f rom the spectra in Regul atory Guide 1.60 as shown in PSAR Section 2.5. Based on the design response spectra, three statistically independent acceleration time histories were developed for the analysis of the plant which represent the three mutually orthogonal independent earthquake motions. The response spectra of the artificial time histories f ully envelope the design response speMra. The design response spectra (or input motions) are applied at the foundation level. (Sections 3.7.1.1 and 3.7.1.2). Critical damping values are specified consistent with Regulatory Guide 1.61 (Section 3.7.1.3). A description of the foundation rock and fin i and calculation of foundation springs are provided in PSAR Section 3.7.1.6. 2. Seismic System Analysis Analysis of Category I structures, systems and components will use either the time history method or the response spectrum method. Soll (rock) structure Interaction will be represented by equivalent springs and masses in l umped mass f ormulations. Three degrees of f reedom are assumed f or horizontal motion: translational, rotational, and torsional. One degree of dynamic vertical freedom is assumed. The number of masses considered I in each model will be such that additional masses or degrees of freedom 3-14

i will not result in more than a 105 increase in responses. Al ternatively, the number of masses or degrees of freedom will be taken equal to twice the number of modes with f requencies less than 33Hz. The number of dynamic response modes considered will be those such that inclusion of additional modes will not result in more than a 10% increase in responses. Significant non-linearities and hydrodynamic ef fects will be considered in the model s. The relative displacements of interconnected components are imposed in the most unfavorable manner to satisfy imput motions out of phase f rom each other. (See PSAR Sections 3.7.2.1.1, 3.7.2.1.2 and 3.7.2.3). Simplified analysis may be used for structural systems (1) for which support motions are represented by response spectra; and (2) which are of a regular nature. The equivalent static force is taken as 1.5 times the peak acceleration of the supports times the weight of the system, incl uding contained l iquids. (See PSAR Section 3.7.2.1.2 and Section 3.7A, paragraph A.2). Uncoupling of systems and subsystems is permitted when justified (PSAR Section 3.7A, paragraph 6.5). Soll (rock) structure Interaction models represent the structure, and f supporting materials down to the foundation of the NIM. The analysis methods and assumptions are discussed and justified in PSAR Section 3.7.1.6. Floor response spectra are based on the direct Integration of the coupled equations of motion. Three directional components of seismic motion are considered with the ef fects combined by the rule of square root of the sum of the squares. (PSAR Section 3.7.2.1.1). Modal responses will be combined using the square root of the sum of the squares method when response spectrum analysis is used. For closely spaced modes, an absolute sum is used or the methods delineated in Regulatory Guide 1.92 may be used (PSAR Section 3.7A, paragraph A.1.3). 3-15 1 l

Seismic Category I structures, systems and components wil I be protected f rom damage by f ailure of Sel smic Category ll and ill structures, systems g and components under SSE loads (PSAR Section 3.7A, paragraph 6). Variations in properties of foundation materials are consicered in generating design response spectra, in addition, the peaks of the floor an response spectra are widened in a i 10% band (PSAR. Section 3.7.2.1.1). i Composite damping of structures with dif ferent elements will be calculated based on the modes of vibration f or a fixed base condition with modal damping ratios evaluated by a weighted average considering the relative strai n energy (PS AR Section 3.7.2.1.1). The most severe combination of loads (earthquake moment, shear and vertical) are considered in determining overturning moments for Category I structures (PSAR Section 3.7.2.13). The occurrence of 5 OBEs and 1 SSE is assumed. Fif ty earthquake maximum peak cycles are assumed for the OBEs and 10 maximum peak SSE cycles are assumed. Seismic Instrumentation will be provided for the plant, but it will not be Installed or otherwise constrained by the limited construction activities. CONCLUS ION The plant seismic design and analysis parameters meet the requirements of CRBRP Design Criterion 2 and Appendix A to 10CFR Part 100 with respect to the capability of the structures to withstand the ef fects of earthquakes. 3.8 DESIGN OF CATEGORY I STRUCTURES Section 3.8 of the PSAR, " Design of Category l Structures", details the load combinations and methods of analysis used f or the design of these structures I in compliance with applicable regulatory guidance and Industry codes. 3-16

f LIMITED CONSTRUCTION ACTIVITIES I The structures to be constructed include the following Category I structures: o Nuclear Island Mat t o Containment Vessel Shell External and Internal Building Walls o APPLICABLE CRITERIA The limited construction activities must comply with the requirements of 10CFR50 and the CR3RP design criteria. The applicable criteria are: 1. 10 CFR 50, 50.55a and CRBRP Design Criterion 1 as they relate to the contai nment, its internal structures, other saf ety-related structures and seismic Category I f oundations being designed, f abricated, erected, and tested to quality standards commensurate with the importance of the safety function to be perf ormed. 2. CRBRP Design Criterion 2 as it related to the design of the containment, its Internal structures, other saf ety-related structures and seismic Category I foundations being capable of withstanding the most severe natural phenomena such as winds, tornadoes, floods, and earthquakes and the appropriate combination of all loads. 3. CR3RP Design Criterion 5 as it relates to the containment, its internal structures, other saf ety-related structures, and seismic Category I foundations being capable of withstanding the dynamic ef fects of equipment f ailures including missiles, pipe whipping and discharging fluids. 4 CRBRP Design Criterion 14 as it relates to the capability of the containment to act as a isaktight membrane to prevent the uncontrolled release of radioactive ef fluents to the environment. 3-17

5. 013RP Design Criterion 41 as it relates to the containment and its internal structures, other saf ety-related structures and seismic g Category I foundations being designed with suf ficient margin of safety to accommodate appropriate design loads. 6. Appendix B to 10CFR Part 50 as it relates to the quality assurance a criteria for nuclear power plants. EVALUATION Section 3.8 of the PSAR describes CRBRP compliance with the above requirements in the design and construction of the Seismic Category I structures. Section 3.8.1 of the PSAR Identifies that the requirements f or concrete containments do not apply to the CRBRP Containment. Section 3.8.2 describes the design requirements of the steel containment. Section 3.8.3 describes the design requirements of the concrete and structural steel Internal structures of steel contai nment. Section 3.8.4 describes the design requirements of other seismic Category i structures. Section 3.8.5 describes the design requirements of foundation and concrete sup ports. The design of the portions of the seismic Category I structures to be constructed meet the relevant requirements by conforming to the following items: o Design and analysis of structures to be constructed comply with the requirements of ACI 349 (as endorsed in Regulatory Guide 1.142), AISC and ASME Section lil, Divisions 1 and 2. C 3-18

loads and load combinations and structural acceptance criteria are o described in PSAR Sections 3.8.3.3, 3.8.4.3 and 3.8.5.3. These loads and load combinations are consistent with the Standard Review Plan Section 3.8.3, e, b and c, and Section 3.8.4 Subsection 11, paragraphs a, b and c. Mechanical splices and reinf orcing coinply with Regulatory Guide 1.10, o 1.15, and 1.142, (PSAR Sections 3.8.3.2.3 and 3.8.3.6) except that the CRBRP splice performance test program allows production and sister splices to be tested in accordance with ASPE-fll Division 2,1975. Concrete placement complies with Regulatory Guides 1.55, 1.136, and 1.142 and is described in Sections 3.8.3.6 and 3.8.4.6 of the PSAR. Design of concrete structures compiles with Regulatory Guide 1.142 and o is described in Section 3.8.4 of the PSAR. Material used in the above structures are described in Sections o 3.8.3.6 and 3.8.4.6 of the PSAR. inspection and testing of the steel containment vessel welds below o elevation 816 feet is described in PSAR Sections 3.8.2.5.2 and 3.8.2.5.5, and compiles with the requirements of Regulatory Guide 1.19. Masonry walls (SRP 3.8.4, Appendix A) are not included in the limited o construction activities. t The design of the cell liners is in accordance with the requirements o and criteria identified in Section 3.8-B and are designed and 1 l constructed as Engineered Safety Features as described in PSAR Section 6.4. The cell liners are used as f ormwork during the construction of the RCB and RSB building structures. l l 3-19 l l {

CONCLUSIONS S The design of Category I structures has been performed according to established criteria derived from industry codes and NRC regulatory guices in of f act at the time of initial design (at the time of contract). The designs meet the relevant requirements of Section 50.55a and Appendix B of 10CFR Part 50, and CRBRP Design Criteria 1, 2, 5,14 and 41. 3.9 EOiANICAL SYSTEMS AND COMPONENTS Section 3.9 of the PSAR provides a description of the dynamic system analysis and testing associated with mechanical systems and components, in addition, a discussion is presented of the design criteria for ASE Code Classes 2 and 3 components and for components not covered by the ASE Code. During the Iimited construction activities, only minor components (e.g., sections of embedded piping) of auxiliary systems will be permanently Installed. The remaining components in the systems are suf ficiently flexible to assure that, given the embedded pipe sections, they can be adapted to meet system final requirements. Therefore, no Section 3.9 review areas are cause fcr withholding authorization of the limited construction activities. 3.10 SEISMIC DESIGN OF CATEGORY 1 INSTRUENTATION AND ELECTRICAL EQUIPENT Discussions of the seismic design, analysis and testing procedures and the restraint messures f or Category 1 Instrtsnentation and electrical equipment are provided in Section 3.10 of the PSAR. During the limited construction activities, no instrumentation or electrical equipment will be installed. Therefore, no Section 3.10 review areas are cause for withholding authorization of the 1imited construction activities. 0 ( 3-20

3.11 ENVIRONENTAL DESIGN OF MEOiANICAL AND ELECTRICAL EQUIPENT t Section 3.11 of the PSAR discusses the environmental qualification basis and progran for saf ety-related electrical equipment. During limited construction activities, no permanent saf ety-related electrical equipment sublect to environmental qual if Ication w!! I be installed. There are no Sect! on 3.11 review areas which are cause for withholding authorization of the limited construction activities. I 3-21 9-_, ~ -,,,--------v- -m__, -..--m

f. 4.0 REACTOR 4 ( Chapter 4.0 of the PSAR describes the reactor and discusses its mechanical, nuclear, and thermal and hydraulic design. During the limited construction activities, no permanent part of the reactor will be installed. There are no Chapter 4.0 review areas which are cause for withholding authorization of the limited construction activities. ( i a 4-1 i n y-,-- --r ~ n,---.---,.-. c..

5.0 HEAT TRANSPORT AND CONNECTED SYSTEMS ( Chapter 5.0 of the PSAR describes and discusses the design of the CRBRP heat transport systems. It also describes and discusses connected auxillery systems necessary for the operation or protection of the heat transport sy stems. During the limited construction activities, only piping penetrations for the auxillary systems will be permanently Installed. The remaining components of such systems are suf ficiently flexible to assure that they can be adapted to meet f Inal requirenents. Therefore, no Chapter 5.0 review areas are cause for withholding authorization of the limited constructed activities. d a 5-1

t 6.0 ENGINEERED SAFETY FEATURES 6.1 GENERAL (NOT APPLICABLE) 6.2 CENTAINPENT SYSTEMS Section 6.2 of the PSAR discusses those CRBRP riant' features that are required to mitigate the ef fects of postulated design basis accidents and which are accordingly designed as Engineered Safety Features (ESFs). For each system or component identified as ESF, the design basis requirements, design description, and design evaluation are provided. i LINITED CONSTRUCTION ACTIVITIES The limited construction activities listed in Section 1.2 of this report which are relevant to this section are the containment vessel and the confinement building. Activities include: Setting Nuclear Island Mat o Setting Containment Vessel Shell o Setting Reactor Containment Building concrete and confinement o l concrete. APPLICABLE CRITERI A The structures to be constructed must conform to CRBRP General Design Criteria 14 " Containment Design" and 41 " Containment Design Basis", and to 10 CFR 100. l 6-1

EVALUATION O o RCB Design Basis Accident A conservative design basis accident has been selected to challenge containment vessel integrity. The event selected is a postulated ag rupture of the largest storage vessel in-containment, with the storage vessel containing its maximum volume of sodium. The resulting sodium f ire consumes al l oxygen in-contai nment, thereby generating maximum pressures and tanperatures f or containment vessel assessment. The containment vessel is designed to indefinitely maintain f unctional integrity following this worst case sodium sp!II/ fire design basis event. Detailed discussion of the design is provided in PSAR Section 3.8.2. o Site Sultability Source Term The containment vessel and containment / confinement annulus f iltration system have been designed to comply with 10CFR Part 100 guidelines for radiological doses. Discussion is provided in Section 15.A of the PSAR. The design basis events have been adequately selected to satisfy the General Design Criteria, and the Engineered Safety Feature Containment has been designed to accommodate the ef fects of the design basis events with margin. CONCLUSION The Containment has been designed as an Engineered Safety Feature. An appropriate Design Basis Event has been established and the Containment has been designed to accommodate it. The design of the containment structures included in the limited construction activities conforms to the relevant requirements for Engineered Safety Features and CRBRP General Design Criteria 14 and 41. ( 6-2

I 6.3 CONTROL ROOM HMITMILITY SYSTEMS Section 6.3 of the PSAR, " Control Room Habitabil ity Systems", provides a description of design bases, evaluation and instrumentation requirements of the control room habitabil Ity systems. LIMITED CONSTRUCTION ACTIVITIES There are no activities under the scope of work identified in Section 1.2 which form part of control room habitability system. No control room habitability system components will be installed into the plant during the projected phase of construction; therefore there are no Section 6.3 review areas which constitute cause for withholding authorization of the limited construction activities. 6.4 CELL LINER SYSTEM Section 6.4 of the PSAR, " Cell Liner System", provides the design bases and the design evaluation of this system. LIMITED CONSTRUCTION ACTIVITIES Some of the limited construction activities listed in Section 1.2 of this report are relevant to this section. The placement of cell liners will occur in conjunction with the provision of Interior concrete for the Reactor Containment Building and the Reactor Service Building. EV ALUATION The capability of this system to meet the applicable criteria for safety-related equipment is discussed in Sections 3.1, 3.2, 3.7 and 3.8. Cell liners will establish the plant response to sodium spills by not allowing NaK or sodium / concrete Interactions in inerted cells. The design size and atmosphere defines the parameters for the analysis of the consequences of leakage or rupture of Na or NaK containing components. 6-3

The cells to be constructed aro large enough that they will not significantly di constrain the design of components to be placed in them. Installation of cell liners still allows flexibility in the arrangement of the equipment placed in them and does not foreclose the possibility of system design modifications. ^? CONCLUS ION The cell liner design is adequate to prevent NaK or sodium / concrete interactions and support the design of other saf ety-related plant f eatures. 6.5 CATCH PAN Section 6.5 of the PSAR, " Catch Pan", provides the design bases and the design eval uation of these components. During the limited construction activities, no catch pans will be installed, only set in place to ease construction. There are no Section 6.5 review areas which are cause for withholding authorization of the limited construction activities. 4 6-4

t 7.0 INSTRulENTATION AND CONROLS ( Chapter 7 of the PSAR, " Instrumentation and Controls", provides the description of the saf ety and non-saf ety related instrumentation and controle. During the limited construction activities, no Instrumentation and control equipment will be Installed. Therefore, no Chapter 7.0 review areas are cause for withholding authorization of the limited constru'ction activities. I t t a i T i 7-1 t t l

8.0 ELECTRIC POWER ( Chapter 8 of the PSAR, " Electric Power", provides the design bases and a description of the Plant Electric Pcwer System (Off site and Onsite) which consists of power supplies and power distribution systems. LIMITED CONSTRUCTION ACTIVITIES During the limited construction activities the following installationd influence the Electric Power System: Reacter Containment and Reactor Service Bul! dings o Embedded Cenduit and Grounding ~~= 14-1~~

~~t 15.0 ACCIDENT ANALYSIS t~~

~~15.1 INTRODUCTION~~
PSAR Section 1.1 provides the criteria for assessing f uel performance. Protection System functions for each Chapter 15 accident are also discussed. In addition, the effect of recent design change on accident evaluations in Chapter 15 are identifled. None of the Iimited construction activities involve the f uel, protection systems or the design changes discussed in Section 15.1. Therefore, there are no Section 15.1 review areas which would constitute cause for withholding authorization of the limited construction activities. 15.2 REACTIVITY, INSERTION EVENTS PSAR Section 15.2 discusses a spectrum of reactivity insertion events and the adequacy of the reactor shutdown system. None of the components considered in these analyses are to be Installed or erected as part of the lImited construction activities. No alternatives to mitigate the postulated accidents will be foreclosed by these activities. Therefore, there are no Section 15.2 review areas which constitute cause for withholding authorization of the i Imited construction activities. 15.3 UNDERCOOLING DESIGN EVENTS PSAR Section 15.3 discussas the spectrum of undercooling events and demonstrates the adequacy of the reactor shutdown systems. None of the components considered in these analyses are to be installed or erected as part of the Iimited construction activities. No alternatives to mitigate the postulated accidents will be foreclosed by these activities. Therefore, there are no Section 15.3 review areas which constitute cause for withholding authorization of the lImited constructon activities. i 15-1
15.4 LOCAL FAILURE EVENTS S PSAR Sectior.15.4 discusses a spectrum of local f ail ure events and demonstrates that, even if such f ail ures were to occur, they could lead only to minor disturbances confined within the assembly in which the f ailure occurred. 89 None of the canponents considered in these analyses are to be instai!ed or erected as part of the limited construction activities. No alternatives f or limiting or mitigating local f ail ure events wil l be f oreclosed by these activities. Theref ore', there are r.o Section 15.4 review areas which, if unresolved, would constitute cause for withholding authorization of the limited construction activities. 15.5 FUEL HANDLING AND STORAGE EVENTS The f uel handling and storage events described in the PSAR, Section 15.5, involve f uel assembly handling and accidents with the reactor vessel and ex-vessel transf er machine. None of these canponents is involved in the limited construction activities. No alternatives for preventing or mitigating ref ueling accidents will be precluded by the limited construction activities. Theref ore, there are no Section 15.5 review areas which, if unresolved, would constitute cause for withholding authorization of the limited construction activities. 15.6 SODIUM SPILLS Sodium spill accident analyses are discussed in Section 15.6 of the PSAR. LIMITED CONSTRUCTION ACTIVITIES The limited construction activities which af fect postulated design basis sodium spill events are installation of the containment floor liner and shell segments and installation of cell liners. 4 15-2
( ACCEPTANCE CRITERI A 4 The ef fects of liquid metal spills and fires must be considered in the design of these structures in accordance with CRBRP GDC 4 and 41. EV ALUATION Postulated design basis sodium or NaK spill events from all of the CRBRP sodium and NaK systems have been analyzed to detennine resulting temperature, pressure, and aerosol conditions f or integration into the nuclear Island buildings designs. The results of these analyses clearly show that the temperatures and pressures calculated for those cells to be constructed in the RCB, RSB and SGB are well below design limits. In addition, calculated radiological consequences f rom the postulated sodium accidents are significantly below the guicelines of 10CFR100. The analyses indicate the design embodies considerable margin to accommodate modest changes in assumed accident conditions that might be necessary for resolution of Construction Permit review areas. CONCLUSION The structures to be Installed during the limited construction activities have been designed with appropriate consideration of the ef fects of liquid metal spills and fires. The designs of the structures are in compliance with CRBRP GDC 4 and 41. Design margins and options provide additional confidence that there are no Section 15.6 review areas which, if unresol ved, woul d constitute l cause for withholding authorization of the limited construction activities. 15-3 11
15.7 OTHER EVENTS S Section 15.7, "Other Events", of the PSAR provides evaluations of the events which do not f all under categories of accidents evaluated in other sections of Chapter 15, " Accident Analysis", of the PSAR. LIMITED CONSTRUCTION ACTIVITIES The plant features to be Installed during the limited construction activities which are important to the accident evaluations of Section 15.7 are the containment floor liner and shell segmen'ts, the Nuclear Island Mat and external walls and the cell liners in cells containing NaK piping. ACGPTANCE CRITERI A The CRBRP design, including the plant f eatures to be constructed, must limit the calculated of fsite dose consequences f rom the postulated accidents to a level below the guidel ine val ues of 10CFR100. This is to be achieved in accordance wIth CRBRP GDC 2 " Design Bases for Protection Against Natural Phenomena" and 53 " Fuel Storage and Handling and Radioactivity Control." EV ALUATION The only Section 15.7 accidents that involve the plant features to be constructed are (1) the events imposing loads on the foundation structures (fuel shipping cask drop and conventional rires, flood and storms), (2) the events involving leakage of NaK, and (3) those events releasing radioactivity in the RCB. The evaluation of the plant structures with respect to natural phenomena such as floods is provided in Section 3.4 The ability of the plant structures to limit the consequences of conventional fires is discussed in Section 9.13.1. The adequacy of the cell liners is discussed in Section 3.8. E 15-4
t The analysos presented in PSAR Section 15.7 demonstrate that the containment
(
boundary as designed assures that guideline limits of 10CFR100 are not exceeded for any of the postulated events. CONCLUSION t The plant structures which are part of the limited construction activities are properly designed to mitigate the postulated accidents in accordance with 10CFR100 and CRBRP GDC 2 and 53. 15.A SITE SUITMILITY SOURCE TERM PSAR Section 15.A provides an analysis of the Site SultabilIty Source Term (SSST). LIMITED CONSTRUCTION ACTIVITIES Limited construction activities which could be involved in mitigating the SSST are the placement of the containment floor liner, shell segments, and placement of confinement concrete. AC PTANCE CRITERIA The structures to be constructed must support overall plant and site compl Iance wIth the gut del ines of 10CFR100. Spect f Ical ly, the postuiated site suitability source term (SSST) must be mitigated so that the calculated of f-site dose consequences do not exceed 10CFR100 guidelines. The details of the source term and appropriate dose limits were specified in NRC's letter to the Project dated May 5,1976, and in the CRBRP Site SultabilIty Report. li 15-5
i l i EVALUAT10N 9 PSAR Section 15.A summarizes the detailed analysis of the consequences of the SSST and shows that the CRBRP design, including the structures mentioned above, reduces the of f-site doses to a level below the limits identified by NRC. o i 4 CONCLUSION The structures to be placed in the limited construction activities include the capability to mitigate the SSST _In accord with the guidelines of 10CFR100. 1 i i I d i 15-6 1
t 16.0 TECNNICAL SPECIFICATIONS -( PSAR Chapter 16 contains preliminary indications of appropriate CRBRP Technical Specifications, LIMITED CONSTRUCTION ACTIVITIES a The limited construction activities which could inpact the plant Technical Spect fIcations are construction of the containment fIoor Iiner and shel I and constructlon of celi IIners. AC PTANCE mlTERI A The portions of the containment and cell liners to be constructed must provide support for the appropriate surveillance activities for these safety-related features. EV ALUATION i The lower containment shell and floor liner will be installed with leak test channels for surveillance of the containment. No special hardware is provided ( for surveillance of the cell liners, since surveillance will be perf ormed from within the cells which allows access to all areas of the liners. CONCLUSION The lImited construction activities wilI include appropriate features to support perf ormance of survell lance activities for saf e1y-related systems. i 16-1 1 --n
T 17.0 QUALITY ASSURANCE .( Chapter 17 of the PSAR, " Quality Assurance", describes the program of plans and actions to assure the quality of parts of the Clinch River Breeder Reactor Pl ant. These parts are those structures, systems, and components whose t satisf actory perf ormance is required to prevent accidents that cause undue risk to the health and safety of the public or to mitigate the consequences of such accidents if they were to occur. LIMITED CONSTRUCTION ACTIVITIES Activities important to safety, Identifled In Section 1.2 will be carried out and documented under the f ull Quality Assurance Program described in Chapter 17 of the PSAR. EVALUATION The CRBRP Quality Assurance Program complies with the Quality Assurance requirenents of Appendix B to 10CFR Part 50, Paragraph 50.55a (except as noted in PSAR Section 3.2.2 regarding Saf ety Classifications) and Paragraph 50.55(e). The Project's method of complying with 19CFR 50 Appendix B is through the implementation of its overall Quality Assurance Program developed in accordance with RDT F2-2 " Quality Assurance Program Requirements." The Progran also complies with supporting standards RDT F1-2 " Preparation of System Design Description," RDT F1-3 " Preparation of Unusual Occurrence Reports," and RDT F3-2, " Cal ibration Program Requirements". l ( i I 17-1 4
The various NRC Regulatory Guides, although not established as requirements for the CERP Quality Assurance Progre, are applicable in principle and 3 Intent. The progra accepts practices which comply with these guides and f ul fill the l ike requirements of RDT F2-2. Maximum recognition is made of quality assurance practices described in the Regulatory Guides and other nationally recognized codes and standards as part of the implementation of the e CFBRP Qual Ity Assurance Progre. CONCLUSION The CRBRP Quality Assurance Progra meets the requirements of Appendix B to ICCFR 50. 4 1 l 4 17-2 i I l "e" - - - -, _,_
t APPENDIX A -( NUREG-0718 - LICENSING REQUIREENTS FOR PENDING APPLICATIONS FOR CONSTRUCTION PERMITS AND MANUFACTURING LICENSE t The evaluation and application of the requirements delineated in NUREG-0718 to CBBRP is provided in Appendix H of the PSAR. APPLICABILlTY TO LIMlTED CONSTRUCTION ACTIVITIES The NUREG-0718 items applicable to the limited construction activities under the scope of work shown in Section 1.2 are as f ol Icws: o item I.C.5 Procedures f or Feedback of Operating, Design and Construction Experience o item i1.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Saf ety Equipment f or Post-Accident Operation o item 11.B.8 Rulemaking Proceeding on Degraded Core Accidents o item li.J.3.1 Organization and Staf fing to Oversee Design and Construction 4 A-1
EVALUAT10N O GBRP project provisions to satisfy items I.C.5, ll.B.2 and ll.J.3.1 of NUREG-0718 are discussed in PSAR Appendix H. The procedures f or incorporating experience galned In operating, design and construction of other nuclear f acilities have been appropriately defined. Plant shielding was assessed in e design reviews. The organization and staf fing to oversee design and construction has been thoroughly considered by Project Management and has been found to be adequate. The relationship of item II.B.8 to the Iimited construction activities is discussed in Appendix B of this report. CONCLUSION lt is concluded that the requirements of NUREG-0718 relating to the Ilmited construction activities have been appropriately considered. Furthermore, with respect to requirements for specific hardware provisions, the features to be constructed can be augmented as required in the future. s k. \\ \\ I l I 4 A-2 4
t ( APPENDIX B HYPOTHETICAL CORE DISRUPTIVE ACCIDENT CONSIDERATIONS i INTRODUCTION The limited construction activities will involve the lower part of the Reactor Continment Buil ding. Certain aspects of this area contribute to the overall plant margins to accommodate Hypothetical Core Disruptive Accicents (HCDAs). However, many plant features not impacted by the limited construction activities provide margins to accommodate HCDAs. This Appendix discusses the ways in which the features included in the limited construction activities impact plant margins. The bases are provided for proceeding with the limited construction activities prior to completion of the NRC review of overall CRBRP plant margins. This Appendix addresses the content of NUREG 0718, ll.B.8 to the extent that the plant capabilities to accommodate a degraded core are impacted by the limited construction activities. Section B.1 addresses the structural margin to accommodate HCDA energetics. Section B.2 addresses the thermal margin to accommodate the long term ef f ects of HCDAs that result in melt through of the reactor vessel. B.1 STRUCTURAL MARGIN BEYOND THE DESIGN BASE (SM3DB) Postulated accident sequences in f ast reactors involving 1) material motions in the core to the extent that there is potential for reactivity to approach or reach prompt criticality, and/or 2) whole-core melting and relocation, are termed Hypothetical Core Disruptive Accidents (HCDAs), it is theoretically possible that this material motion could have the net ef fect of causing ( Increases in reactivity and power level suf ficient enough to result in substantial energy releases. l B-1
The CRBRP design includes suf ficient diversity, redundancy and reliability of g key saf ety features to assure that such accident sequences are extremely improbable and can thereby be excluded from the design base accident spectrum. To assure an acceptably low occurrence rate, potential HCDA initiatcrs have been identified and the f eatures necessary to prevent HCDA initiation have q been incorporated into the CRBRP design. Included in the design features are diverse and redundant reactor shutdown systems and shutdown heat removal systems which lower the probability of an HCDA initiating event sequence to such an extent that It is not reasonabIe to consider an HCDA as belng wIthin the design base. Although HCDAs are not part of the design base for CRBRP, assessments of HCDA consequences have been perf ormea to assure f urther that the risk to the public heal th and saf ety is acceptably low. healistic assessments of HCDA progression paths, including best-estimate analysis and consideration of uncertainties, indicate that the core responds in a non-energetic manner. (See CRBRP-GEFR-00523, "An Assessment of HCDA Energetics in the CRBRP Heterogeneous Reactor Core.") Because the hypothesized initiators involve f ailure of both the reactor shut-down systems, perrranent shutdown of the reactor requires f uel removal f rom the For each accident progression, shutdown is found to result from ejec-core. tion of fuel from the fuel rods and dispersal from the core region. Application of fundamental principles which are applicable following loss of f uel assembly geometry indicate there is little or no potential for energetic recriticalities. Development testing by both NRC and the Project is in place to provide additional confirmation of the fuel disperssi mechanisms leading to permanent subcrIticalIty. Assessments using conservative assumptions regarding design parameters and HCDA physics, which include pessimistic Doppler and materials worths as well as variations in the degree of fuel removal and blockage, also indicate termination in a non-energetic manner (i.e., permanent shutdown or a stabi-I lized, near nominal power level). Only when pessimistic assumptions, well beyond those appropriate for a realistic assessment, are invoked does the B-2
. t' assessment predict energetics. An evaluation of the HCDA energetics potential ( In the CRBRP heterogeneous reactor core produced only one case out of 23 wnere energetics was predicted. This case assumed arbitrary midplane failure with a reactivity insertion ramp rate of 104/sec and resulted in a value of 33 MJ of work-energy when the f uel vapor was Isentropically expanded to the f ree volume I within the reactor vessel. In order to f urther reduce the risk to the public from HCDAs, prudent Structural Margins Beyond the Design Base (SMBDBs) have been incorporated into i the design to mitigate core energetics. The energetics used as a basis to establish the ShBDB requirements can be characterized as the werk potential (approximately 100 MJ) upon fuel vapor expansion induced sodium slug impact with the reactor vessel head. (An i alternative characterizing value for the Isentropic f uel vapor expansion to atmospheric pressure is 661 MJ for the specified SMBDB pressure-volume relationshio.) The results of the assessments done for the earlier homogenous core (CRBRP-GEFR-00103, "An Analysis of Hypothetical Core Disruptive Events in the Cl inch River Reactor Plant") have been compared to those for the heterogeneous core (CRBRP-GEFR-00523). An examination of similar cases in the studies shows that the heterogeneous core des!gn is less sensitive to a range of phenomenological ossumptions than the previous homogeneous core design. Specifically, in the assessment of the previous homogeneous CRBRP core (CRBRP-GEFR-00103) It was concluded that the SMBDB wculd provide a substantial saf ety margin. By using the same energetics basis for the heterogeneous core a f urther level of conservati sm is provided. Analysis has been perf ormed to demonstrate that the structural integrity of the primary boundary is maintained under SMBDB loads. Hydrodynamic / structural computer codes model the dynamic response of the reactor system to the assumed Isentropic expansion of fuel vapor. Current mathematical models were anployed to calculate the pressure pulse transmission in the primary heat transport system, the cover gas system and the overflow and makeup system. Specific SMBDB requirements have been identified and are discussed in Sec'lan 5 of B-3
CRBRP-3, Volume 1, " Hypothetical Core Disruptive Accicent Consicerations in CRBRP, Energetics and Structural Margin Beyond the Design Base." g Component accommodation of the SEDB requirements is assured by analysis, and tests when appropriate, to demonstrate that the component has met the SECB acceptance criteria provided in Section 5.3 of CRBRP-3, Volume 1. SECB e acceptance criteria consist of stress, strain, and leakage limits for tho reactcr coolant boundary. Results of the analyses provided for the reactor coolant boundary components in Section 5.4, CRBRP-3, Volume 1, Indicate that the SEDB loads can be ' accommodated. Because of the cceplex nature of the reactor vessel structural response to HCDA loads, particularly in the reactor vessel head region, several scale-model experiments have been perf ormed. The scale model tests of the reactor vessel and head have been used to obtain better understanding of the dynamic response and to verify adequate margin against structural f ailure. These tests have consistently confirmed that the reactor vessel and head have significant margins to ensure that integrity of the reactor coolant boundary components wiII be retained under postuiated Ioadings. Further, the tests confirmed the adequacy of scaling laws and the adequacy of computer codes to predict the response of components to S208 loads. Limited consiceration has been given to the essessment of energetics in excess of 100 MJ. Scoping analyses of energetics considerably beyond the defined structural margin of 100 MJ at sodium slug impact indicate that the major head components (e.g., rotating plugs) would remain Integral (i.e., not become missiles). Consideration of non-Isentropic phenomena that would occur during vapor expansion would significantly reduce the predicted heaa loads, ef fec-tively resulting in additional margin capability. In addition, the radiolog-Ical consequences have been shown to be Insensitive to a wice range of assumed material releases through the reactor vessel head. LIMITED CONSTRUCTION ACTIVITIES I The limited construction activities do not include the installation of any reactor or core assembly components. in addition, the reactor vessel itself B-4
X and the vessel's steel support ledge will not be installed during the limited a construction activities. However, the activities do include two items which could impact the mitigation of higher structural loads. They are: o The containment vessel shell o Concrete walls of the reactor cavity below ' essel ledge mounts v The containment provides pressure capability to accommodate greater leakage than the required limit for SMBDB. Short term containment integrity would not be challenged by releases of sodium and gases that significantly exceed the current SMBDB acceptance criteria established by the Project. The reactor vessel ledge supports downward and upward loads for the SMBDB. The design presently meets the SMBDB requirements as shcwn through conserva-tive analysis. The construction of the reactor cavity concrete walls below the vessel support ledge mounts does not foreclose design options for further SMBDB mitigation since SMBDB margins for the reactor cavity concrete walls (275 margin from ASME Code Section ill f aulted criteria using elastic analysis) could accommodate variations in SMBDB loads. CONCLUSION HCDAs are beyond the design base, and realistic assessments indicate HCDA sequences result in a non-energetic termination. The reactor coolant boundary is designed with adequate structural margins to accmmodate energetic HCDA dynamic loads without loss of structural Integr i ty. Thus, the CRBRP w i l l be constructed so that the risk to the public f rom en HCDA is acceptably low. Since the existing design features provide suf ficient margin to accommodate somewhat higher SMBDB loads, it is acceptable to proceed with construction. B.2 THERMAL MARGIN BEYOND THE DESIGN BASE (TM3DB) Although some HCDA sequences leave the reactor core in a damaged but coolable condition there is a possibility of core melting and thermal penetration of B-5
the reactor vessel ano guard vessel. To assure the residual risk from such events is low, features have been aaded to the design of CRBRP to provide g thermal margin to mitigate the consequences. This margin to accommodate a core melt accident has been termed the Thermal Vargin Beyond the Design Base (TMBDB) and is assessed in CRBRP-3, Volume 2, " Hypothetical Core Disruptive Accident Considerations in CRBRP, Assessment of Thermal Margin Beyond the gp Design Base." To evaluate the adequacy of the plant thermal margin, the release of the entire core, blankets and primary sodium into the reactor cavity was considered. Requirements have been placed on plant components and structures to assure that containment integrity would be maintair.ed without venting until evacuation procedures coulc be implementea. Features are al so incl uded to provide long term mitigation of HCDA consequences. The principal features that provide the thermal margin beyond the design base are: Flow passages at the lower end of the guard vessel support skirt to o provide unimpeded flow of sodium and to pennit dispersal of the core debris across the entire reactor cavity floor. o Reactor cavity-to-containment vent system to relieve the reactor cavity atmosphere pressere as it is heated by the spilled sodium. o Reactor contai nment building (RCB) vent and purge systems to limit long term pressure buildup in the RCB and to allow the RCB to be purged by outside air to control hydrogen concentration in the RCB atmosphere. I An annius cooling system to cool the RCB steel shell and the concrete o confinement building. o A containment cleanup system to reduce the radiological consequences of vented material s. E l Instrumentation to enable the operator to follow the course of the o B-6 I t
( accident and to make decisions on the operation of the TMBDB t features. Based on the evaluations cons!dering these features, it is concluded that the radiological consequences of a hypothetical core disruptive accicent would be acceptable, considering the extremely low probability of such a condition. The existing design provides protection for a range of sodium concrete reaction rates and total penetrations typical of those observed in tests to date. Specifically, for total penetration of up to about two inches into the reactor cavity floor, contalment Integrity with no venting or operater action is provided f or a period of some 36-hours. Thereaf ter, for this same case, structural Integrity of the containment above the nuclear island mat is provided for at least 8000 hours (extent of analysis) with controlled venting and purging. Sensitivity studies have also demonstrated that even for 12-inches of penetration there would be no requirement to vent the containment for at least 24-hours. 4 Further parametric analyses have been conducted to examine the consequences of postulated sodium concrete reaction sequences f ar in excess of any ever observea in tests to date. The extreme upper limit of the range studied includes a penetration rate of 7-inches per hour for 3-hours followed by 1-inch per hour for as long as the sodlum renains in the reactor cavity. These studies have demonstrated that even under these postulated conditions, the structural capability of the containment and its ability to f unction could be ensured by design modifications (with venting at 10 hours). Namely, there would be a need to strengthen the design of the pipeway cell floor and at the Interf ace between the floor and the wall in the reactor cavity. The radiological consequences of the above parametric sequence with containment venting at 10 hours have been analyzed and are acceptable. Developmental work is continuing in order to more completely understand the sodium-concrete reaction phenomena. The results f rom the development work can be incorporated into the THBDB analytical model in the f uture if there is substantial deviation from the range of conditions already considered. If the data should Indicate that sodium-concrete reactions are more severe than those B-7
1sicered in recent tests and the TMBDB consequences with the present design are found unacceptable, f allback options to accommodate or prevent the gp reactions can be implemented on a time scale compatible with the reacter cavity construction schedules. Experimental data f or sodium-concrete reactions obtained to date have been an produced using calcitic limestone aggregate for the concrete. Aaditional testing is underway to determine the applicability of these data to concrete using dolanitic limestone aggregate. The latter material is readily available at the CRBRP site. If this material is found to be different enough to com-promise the present conclusion of the TNBDB evaluations, calcitic limestone aggregate can be used f or those structures potentially exposed to sodium during TNBDB. Preliminary testing will be completed by September 1,1982 to confirm the acceptability of the dolomitic aggregate. LIMITED CONSTRUCTION ACTIV ITIES During the limited construction activities, the following construction relevant to TMBDB will be started, o Placement of the concrete for the Nuclear Island Mat o installation of the containment floor liner plate o Installation of the containment fill slab o installation of the RCB Interior walls (not including pipeway cell floors) o Installation of the containment vessel vertical wall rings o Installation of confinement concrete l 4 B-8
(
( lt is not envisaged that any of these areas would be impacted by new information relating to sodium-concrete reactions, except the floor / wall interf ace of the reactor cav!ty. CONCLUS ION Although the likelihood of a core meltdown is so low as to not be considered within the design base, features have been included into the design that provide long term mitigation of HCDA consequences. Containment Integrity with no venting or operator action is provided for a period in excess of 24 hours. With controlled venting, structural Integrity above the basemat is proviced f or at least 8000 hours. Sensitivity studies have been conducted over a range of postulated sodium-concrete reaction rates and have demonstrated that a ,g substantial margin of capability exists beyond the base case. Indeed the present margin bounds all renaining uncertainties with the exception of the reactor cavity floor / wall Interf ace. if determined appropriate, fallback designs can be implemented in the areas of exception prior to construction. On the basis of the foregoing, it is acceptable to proceed with limited construction activities. l ^f 'l ( l B-9
( REFERENCES ( 1-1 Final Environmental Statement related to construction and operation of Clinch River Breeder Reactor Plant, U.S. Nuclear Regulatory Ceumission, Of f ice of Nuclear Reactor Regulation, Docket No. 50-537, NURh5-0139, February 1977. ( 1-2 lbid, Chapter 9, Alternatives, pg. 9-23. 1-3 Site Sultabil ity Report by the Of f ice of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission in the Matter of the Clinch River Breeder Reactor Plant, Docket No. 50-537, March 4,1977. Revision, June 11, 1982. 1-4 lbid, Section 1.B, Summary Concl usions, pg.1-7. 1-5 Draf t Supplement to Final Environmental Statement related to construc-tion and operation of Clinch River Breeder Reactor Plant, Docket No. 50-537, NUREG-0139, Suppl ement No.1, Draf t Report, July 1982. 1-6 Letter f rom P. Shewmon, ACRS, to Dr. N. J. Pal ladino, NRC, dated July 13,1982. 'l l f i t I ( ,}}

ML20062H082

Contents

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SUMMARY

1.0 INTRODUCTION

15.1 INTRODUCTION

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ML20062H082

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SUMMARY

1.0 INTRODUCTION

15.1 INTRODUCTION

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