ML20049H265

Chapter 2 to Gessar, Site Characteristics.

ML20049H265
Person / Time
Site:	05000447
Issue date:	02/12/1982
From:	GENERAL ELECTRIC CO.
To:
References
NUDOCS 8202230013
Download: ML20049H265 (66)

Text

. _ . _ . _ _ . - - . . _ . . _ _ _ _ _ _ _ _ . - . . - - -

L

) GESSAR II 22A7007 .l 238 NUCLEAR ISLAND Rev. 0'  !

i  !

1 6 i

!O 4

1 I

i t

]  !

.i -

\

l ,

1 1

• i i CHAPTER-2

' i l SITE CHARACTERISTICS  ;

. (

i .

4 1

• 1 s l I 4

i O  ;

l i

e

> t

! l 4 ,

I i i

I j

i i

F i

O  ;

t 8202230013 820212 PDR ADOCK 05000447 K PDR

. . , . _ _ _ , , _ _ , _ , , , , _ _ _ _ _ __ .__.a.,_. g,_,f

J'  :

GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. O i

i i

I SECTION 2.0 i'

i i CONTENTS  ;

i l f Section Title Page i i

. 2.0 Summary 2.0-1  ;

i i

i i

i TABLES l

Table Title- -Page I

'2.0-1 Summary of GEGSAR II Siting Envelope 2.0-2 r

i t

i ,

d E

I i

i i

,! r t

i r

.! i i i 2.0-1/2.0-il t

i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l

() 2. SITE CHARACTERISTICS 2.0

SUMMARY

A detailed description of all site characteristics is not practical, since this SAR is not based on a specific site location. However, it is possible to define an envelope of selected site-related para-meters which will blanket the majority of potential reactor sites in the conterminous United States. This envelope of site-related parameters establishes the conditions of phenomena which the generic Nuclear Island is designed to accommodate. These characteristics, i

which were picked after a review of values used in recently licensed plants, provide the bases for design of the Nuclear Island. A sum-mary of the GESSAR II sitting anvelope is given in Table 2.0-1.

Variations in chosen site parameters are to be expected. When a specific plant site is selected, a plant-unique set of plant design

() conditions will be established. It is expected that the unique site will have most design conditions lower than (and with a lim-ited number higher than) the GESSAR II siting envelope. Confirming calculations and analyses will be made with the site-unique condi-tions for the Nuclear Island buildings using the loading combina-tions and allowable stresses given in Section 3.8. For the total loading condition'on any structure or system, it is anticipated that the Nuclear Island design will be adequate. If the unique

site conditions indicate that areas of the Nuclear Island design are inadequate, a site-unique design will be provided. The meteor-ological assumptions used in the accident-analyses are defined in Chapters 2 and 15. If the site-unique meteorological conditions vary from the Nuclear Island assumptions, the accident analyses will be repeated to determine acceptability for new offsite doses.

For these cases where the doses exceed allowable limits, design modifications and new analyses will be performed to provide a t

design that meets licensing requirements.

v l

i i 2.0-1

GESSAR II 22A7007 238 WUCLEAR ISLAND Rev. O Table 2.0-1

SUMMARY

OF GESSAR II SITING ENVELOPE Parameter Value Meteorology Extreme Wind 130 mph (30 ft above grade)

Tornad 290 mph maximum rotational 70 mph maximum translational 3 psi /1.5 see pressure change Short-Term Meteorological Pasquill Type F Conditions Temperature Range -40 F to +115 F Hydrology Ground Water Level 2 ft below grade Flood Level 1 ft below grade Maximum Rainfall Rate 4 inches / hour 2

Maximum Snow Load 50 lb/ft Maximum Cooling Water Temperature 100"F Seismology Safe Shutdown Earthquake / 0.3 g horizontal free-field Soil Interaction Characteristics as measured at grade level.

Total of 12 soil-structure interaction cases for all Nuclear Island Buildings (Appendix 3A)

O 2.0-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

() SECTION 2.1 CONTENTS Section Title Page 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1-1 2.1.1 Site Location and Description 2.1-1 2.1.1.1 Specification of Location 2.1-1 2.1.1.2 Site Area Map 2.1-1 2.1.1.3 Boundaries for Establishing Effluent Release Limits 2.1-1 2.1.2 Exclusion Area Authority and Control. 2.1-1 2.1.2.1 Authority 2.1-1 2.1.2.2 Control of Activities Unrelated to Plant Operation 2.1-2 2.1.2.3 Arrangements for Traffic Control 2.1-2 2.1.2.4 Abandonment or Relocation of Roads 2.1-2

'N 2.1.3 Population Distribution 2.1-2

)

2.1.3.1 Population Within 10 Miles 2.1-3 2.1.3.2 Population Between 10 and 50 Miles 2.1-3 2.1.3.3 Transient Population 2.1-3 2.1.3.4 Low Population Zone 2.1-3 2.1.3.5 Population Center 2.1-3 f 2.1-3 2.1.3.6 Population Density l

O 2.1-i/2.1-ii r

I . .. - - . _ _ , . . . - - - - . . . .

GESSAR.II 22A7007 238 NUCLEAR ISLAND- Rev. 0 g

2.1 GEOGRAPl!Y AND DEMOGRAPHY 2.1.1 Site Location and Description ,

-2.1.1.1 Specification of Location i

The standard plant is based on a land site adjacent to or conven-iently close to a body of water, sufficient for either once-through or recirculated cooling for the turbine-condenser and other cooling requirements, or a combination of the two. cooling methods. The specifics of the site location will be provided by the Applicant.

i 2.1.1.2 Site Area Map The map of the site area will be provided by the Applicant.

2.1.1.3 Boundaries For Establishing Effluent Release Limits The distance to site boundaries from plant effluent release points is equal to or gre9'er than those necessary to meet either 10CFR20 or 10CFR100. The minimum boundaries are established by dose rate e calculations in Chapter 15 based on the Design Basis Accidents (DBA). Perimeter fencing or other reasonable means of access con-trol will be provided to exclude the public for radiation protec-tion purposes. The location and means of access control will be described by the Applicant.

I 2.1.2 Exclusion Area Authority and Control 2.1.2.1 Authority 2-

.i The Applicant's legal rights with respect to designated exclusion

, areas-will be c. dressed by the Applicant.

t O

1 2.1-1

- - - - . - . + ..m,.p.... , , ., e,e-,7 ev.p.-- - - . , -, .-o..-,- p-- , -,,, _ _--

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.1.2.2 Control of Activities Unrelated to Plant Operation The only activities within the exclusion area unrelated to plant operation are assumed to be those specifically provided for in the plant operating license. None of these activities will interfere with normal plant operation, nor will they result in doses to the public greater than those allowed by appropriate federal regulations.

2.1.2.3 Arrangements for Traffic Control It is assumed that the plant is served by a local railroad company with access to the exclusion area over a spur owned by the Appli-cant. Only railroad cars consigned to the plant will be brought onto the site via this spur.

An emergency plan and plant security plan is used to prohibit the general public from entering the exclusion area during a plant g emergency. These plans will be submitted by the Applicant.

2.1.2.4 Abandonment or Relocation of Roads Information pertaining to abandonment or relocation of roads will be provided by the Applicant.

2.1.3 Population Distribution

! The safety design basis of the standard plant is independent of population and population distribution beyond the effective site boundaries. Such infcrmation is used for evaluation of the site on considerations which do not relate to plant safety features provided by design.

2.1-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O t' 2.1.3.1 Population Within 10 Miles The population distribution within 10 miles and its basis will be provided by the Applicant.

2.1.3.2 Population Between 10 and 50 Miles The population distribution within 10 and 50 miles and its basis will be provided by the Applicant. .

2.1.3.3 Transient Population Seasonal and daily variations in population and population distri-bution will be described by the Applicant.

2.1.3.4 Low Population Zone The safety design basis of the standard plant is that, as a result of postulated accidents as described in Chapter 15, the exposure of any member of the public is below the guidelines of 10CFR100.

2.1.3.5 Population Center Acceptable population center distances are not related to the safety design basis of the plant; however, the population center distance was assumed to be in acccrdance with the guidelines in 10CFR100.

2.1.3.6 Population Density The cumulative resident population projected for the first and  ;

last year's plant operation will be compared with the acceptable population densities by the Applicant. )

l 1

2.1-3/2.1-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O o

i SECTION 2.2

(

CONTENTS t

Section Title Page 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 2.2-1 2.2.1 Location and Routes 2.2-1 2.2.2 Descriptions 2.2-1 2.2.2.1 Description of Facilities 2.2-1 2.2.2.2 Description of Products and Materials 2.2-1

2.2.2.3 Pipelines 2.2.1 2.2.2.4 Waterways 2.2-1 l 2.2.2.5 Airports ,

2.2-2 1

2.2.2.6 Projections of Industrial Growth 2.2-2 2.2.3 Evaluation of Potential Accidents 2.2-2 2.2.3.1 Determination nf Design Basis Events 2.2-2

()

2.2.3.2 Effects of Design Basis Events 2.2-3 I

O 2.2-1/2.2-11

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

() 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES The safety design bases of the Nuclear Island do not include any special features for the presence of industrial, transportation and military facilities in the environs.

2.2.1 Location and Routes Location and routes of any nearby industrial, transportation, and military facilities will be described by the Applicant.

2.2.2 Descriptions 2.2.2.1 Description of Facilities Descriptions of nearby industrial, transportation, and military facilities will be provided by the Applicant.

O 2.2.2.2 Description of Products and Materials Descriptions of the products and materials regularly manufactured, stored, used, or transported in the vicinity of the plant will be provided by the Applicant.

2.2.2.3 Pipelines Descriptions of nearby pipelines will be provided by the Applicant..

2.2.2.4 Waterways Information on navigable waterways adjacent to the site will be provided by the Applicant.

O 2.2-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.2.2.5 Airports The Nuclear Island is intended for use at sites where the probabil-

-7 per year.

ity of an aircraft impact is 110 It is the responsi-bility of the Applicant to show compliance with this requirement.

If the Applicant's plant is located at a site where this probability

-7 is not 110 per year, the Applicant will provide an evaluation of the consequences of an aircraft crash considering the type and fre-quency of aircraft germane to his site in his SAR. It is believed that the criteria used for aircraft cra:h will include at least 90% of the sites in the United States.

2.2.2.6 Projections of Industrial Growth Projections of the growth of current and new industrial activities in the vicinity of the plant will be provided by the Applicant.

2.2.3 Evaluat-ion _of Potential Accidents 2.2.3.1 Determination of Design Basis Events Determination of design basis events (i.e., probability of occur-rence >10~ times per year and have potential consequences serious enough to affect the safety of the plant to the extent that 10CFR100 guidelines could be exceeded) will be provided by the Applicant for each of the potential accident categories listed below:

(1) explosions; (2) flammable vapor clouds (delayed ignition);

(3) toxic chemicals; (4) fires; 2.2-2

'G"SSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.2.3.1 Determination of Design Basis Events (Continued)

(5) collisions with intake structures; and (6) liquid spills.

2.2.3.2 Effects of Design Basis Events The effects of the potential design basis events of Subsection 2.2.3.1 will be addressed by the Applicant.

5 O

i l

l i

i l

l l

l0 l 2.2-3/2.2-4 I

GESSAR II 22A7007 233 NUCLEAR ISLAND Rev. O Ib SECTION 2.3 (s / .

CONTENTS i

Saction Title Page 2.3 METEOROLOGY 2.3-1 2.3.1 Regional Climatology 2.3-1 2.3.1.1 General Climate 2.3-1 2.3.1.2 Regional Meteorological Conditions for Design and Operat'ing Bases 2.3-1

. 2.3.2 Local Meteorology 2.3-2 2.3.2.1 Normal and Extreme Values of Meteorological Parameters 2.3-3 2.3.2.2 Potential Influence of the Plant and the Facilities on Local Meteorology 2.3-3 2.3.2.3 Local Meteorological Conditions for Design and Operating Bases 2.3-3 2.3.3 Onsite Meteorological Measurements Program 2.3-3

() 2.3.4 2.3.5 Short-Term Atmospheric Diffusion Estimates Long-Term Atmospheric Diffusion Estimates 2.3-3 2.3-4 i

i l

l l

l O

2.3-i/2.3-il

GESSAR II. 22A7007 238 NUCLEAR ISLAND Rev. 0 2.3 METEOROLOGY 2.3.1 Regional Climatology 2.3.1.1 General Climate The general climate for the region will be described by the Applicant.

2.3.1.2 Regional Meteorological Conditions for Design and Operat-ing Bases The safety design basis of the Nuclear Island is that meteorologi-cal extremes, with the exception of a tornado, will have no effect on the continued routine operation of the plant and thus will impose no safety restric'tions. Specific considerations for safety-related structures include:

(1) Any precipitation rate will have no direct effect on the safety of such structures, and the drainage design of the site will be such that no secondary effects on the safety of such structures will occur.

The roof drains for the Nuclear Island buildings will be designed for a precipitation rate of 4 in./hr. Roof parapets on the Nuclear Island buildings will be provided I

with drain openings to the outside of the building to limit the standing' water level on the roof to a maximum of 9.5 inches. This will protect against plugged drains and higher rates o'. precipitation. The Applicant will design the site drainage system for the site-unique pre-cipitacion rate.

(2) The structures are designed for vertical loads of 50 psf of snow or ice, in addition to other design loads, as O described in Subsection 3.8.

2.3-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.3.1.2 Regional Meteorological Conditions for Design and Operat-ing Bases (Continued)

(3) The structures are designed to be unaf fected by lightning strikes.

(4) The structures are designed to withstand wind velocities of 130 mph at 30 ft above plant grade with a velocity distribution and gust factor as described in ASCE 3269 (Wind Forces on Structures).

(5) The safety-related structures and equipment are designed for the Design Basis Tornado described in NRC Regulatory Guide 1.76 for Region I. The characteristics of this tornado are:

Maximum wind speed (mph) 360 Rotational speed (mph) 290 Translational speed:

Maximum (mph) 70 Minimum (mph) 5 Radius of maximum rotational speed (ft) 150 Pressure drop (psi) 3.0 Rate of pressure drop (psi /sec) 2.0 The tornado missiles are discussed in Chapter 3, Section 3.5.

(6) The structures are designed for an environmental temper-ature range from -40 F to +115 F.

2.3.2 Local Meteorology A program of onsite data collection, supplemented by National Weather Service (NOAA) summaries from locations near a specific site, will be conducted. The site assumed in this SAR falls 2.3-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.3.2 Local Meteorology (Continued) within the standard plant design meteorological criteria specified in Subsection 2.3.1.

2.3.2.1 Normal and Extreme Values of. Meteorological Parameters Monthly and annual summaries of meteorological parameters will be provided by the Applicant.

2.3.2.2 Potential Influence of the Plant and the Facilities on Local Meteorology An evaluation of the potential modification of meteorological parameters as a result of the presence and operation of the plant will be provided by the Applicant.

2.3.2.3 Local Meteorological Conditions for Design and Operating O Bases Local meteorological and air quality conditions used for design and operating basis considerations will be provided by the Applicant.

! 2.3.3 Onsite Meteorological Measurements Program i

It is assumed that no result from the onsite measurement program will change the safety design basis of the plant described herein.

l This assumption, for a particular site, will be verified by the i Applicant for all safety-related structures.

2.3.4 Short-Term Atmospheric Diffusion Estimates I

Short-term atmospheric diffusion conditions are assumed to exist f

in conjunction with a hypothetical accident. These conditions are

() represented by low wind speeds and are assumed to persist 2.3-3

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.3.4 Short-Term Atmospheric Diffusion Estimates (Continued) throughout the accident. Meteorological conditions which are assumed to be typical of short-term estimates are presented in Chapter 15. Design basis radiological consequence analyses are based on Pasquill Type F meteorological conditions.

As site data become available, an evaluation of that data will be made and, if deemed appropriate, it will be used in lieu of the conditions defined above. This will include:

(1) objective (estimates of atmospberic diffusion), and (2) calculations (hourly cumulative frequency distributions of relative concentrations).

2.3.5 Long-Term Atmospheric Diffusion Es_timates Loi.g-term atmospheric diffusion conditions will be established at the completion of the collection of onsite meteorological data.

TSese data will be used in evaluating the consequences of normal operational releases from the plant. Equipment design will be compatible with the meteorological data to assure that the average annual exposures are within the regulatory guidelines in existence at the time the plant is licensed. The long-term atmospheric diffusion estimates will include:

(1) objective (estimates of annual average atmospheric transport and diffusion characteristics), and (2) calculations (description of model used to calculate annual average atmospheric dif fusion values) .

O 2.3-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O SECTION 2.4 CONTENTS Section Title Page-2.4 HYDROLOGY ENGINEERING 2.4-1 2.4.1 Hydrologic Description 2.4-1 2.4.1.1 Site and' Facilities 2.4-1 2.4.1.2 Hydrosphere 2.4-1 2.4.2 Floods 2.4-2 2.4.2.1 Flood History 2.4-2 2.4.2.2 Flood Design Considerations 2.4-2 2.4.2.3 Effects of Local Intense Precipitation 2.4-3 2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers 2.4-3 2.4.4 Potential Dam Failures, Seismically Induced 2.4-3 2.4.5 Probable Maximum Surge and Seiche Flooding 2.4-4

() 2.4.6 2.4.7 Probable Maximum Tsunami Flooding Ice Effects 2.4-4 2.4-5 2.4.8 Cooling Water Canals and Reservoirs 2.4-5 2.4.9 Channel Diversions 2.4-5 2.4.10 Flooding Protection Requirements 2.4-5 2.4.11 Low Water Considerations 2.4-6 2.4.12 Dispersion, Dilution and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters 2.4-6 2.4.13 Groundwater 2.4-6 2.4.14 Technical Specification and Emergency Operation Requirements 2.4-7 O

V 2.4-i/2.4-11

GESSAR II 22A7007

-238 NUCLEAR ISLAND Rev. 0 2.4 -IIYDROLOGY ENGINEERING The safety design basis of the Nuclear Island provides that struc-tures of safety significance will be unaffected t, the hydrology defined below.

2.4.1 liydrologic Description 2.4.1.1 Site and Facilities The structures of safety significance will be located on the site such that: (1) the total design is compatible with existing ground water levels up to 2 ft below grade; (2) the flood level associated with the design basis' flood is at or belowlan elevation correspond-ing to approximately 1 ft below plant grade; and (3) the loading on these structures does not include simultaneous flood levels and seismic events.

O The specific description of the site and all safety-related elevations, structures, exterior accesses, equipment and systems from the standpoint of hydrology considerations will be provided by the Applicant.

2.4.1.2 flydrosphere The major hydrologic feature on or near the site is the body of water which provides the ultimate heat sink for the Nuclear Island.

l No upstream or downstream river control structures are present I which will cause either groundwater levels or flood water levels to

! exceed the values given in (1) and (2) above.

I t

l The Applicant will provide a detailed description of all major i hydrologic features on or in the vicinity of the site.

t i

O 2.4-1

GESSAR Il 22A7007 238 NUCLEAR ISLAND Rev. 0 2.4.2 Floods O

The structures of safety significance are designed for a design basis flood, as defined in Regulatory Guide 1.59, up to an eleva-tion 1 ft below plant grade including allowance for the effects of coincident waves and the resultant runup as calculated from site unique parameters.

2.4.2.1 Flood History Date, level, peak discharge and related information for major historical flood events in the site region will be provided by the Applicant.

2.4.2.2 Flood Design Considerations Seismic Category I structures that may be affected by design basis floods are designed to withstand the floods postulated in Section 2.4, using the " hardened" flood protection approach. Through the hardened protection approach, structural provisions are incorpo-rated in the plant's design to protect safety-related structures, systems and components from postulated flooding. Seismic Category I structures required for safe shutdown remain accessible during all flood conditions.

Safety-related systems and components are flood protected either because of their location above the design flood level, or because they are enclosed in reinforced concrete Seismic Category I struc-tures which have the following requirements:

(1) wall thickr sses below flood level of not less than 2 ft; (2) water-stops provided in all construction joints below flood level; O

2.4-2

GESSAR II 22A7007 238 NUCLEAR ISLAND. Rev. O

_2.4.2.2' Flood Design Considerations -(Continued)

(3) watertight doors and equipment hatches installed below design flood level; and (4) waterproof coating.

Additional. flood protection from external sources is discussed in Subsection 3.4.1.

2.4.2.3 Effects of Local Intense Precipitation The effects of local probable maximum precipitation on adjacent drainage areas and site drainage systems, including drainage from the roofs of structures, will be provided by the Applicant.

2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers O Criteria for the PMF conform to NRC Regulatory Guide.l.59. The Applicant will address following topics:

(1) Probable Maximum Precipitation (PMP)

(2) Precipitation Losses (3) Runoff and Stream Course Models (4) Maximum Flood Flow (5) Water Level Determination (6) Coincident Wind Wave Activity 2.4.4 Potential Dam Failures, Seismically Induced The Nuclear Island structures of safety significance are assumed to be located such that the failure of existing impoundments and poten-tial future impoundments will not result in flooding in excess of that described in Subsection 2.4.2. The failure of existing down-stream impoundments will not result in the loss of a dependable 2.4-3

GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 2.4.4 Potential Dam Failures, Seismically Induced (Continued) heat sink or sources of emergency water supplies, such as firewater.

Criteria for seismically induced potential dam failure conform to NRC Regulatory Guide 1.59. The Applicant will address the follow-ing topics:

(1) Dam Failure Permutations (2) Unsteady Flow Analysis of Potential Dam Failures (3) Water Level at Plant Site 2.4.5 Probable Maximum Surge and Seiche Flooding The Applicant will address the following topics:

(1) Probable Maximum Winds and Associated Meteorological Parameters; (2) Surge and Seiche Water Levels; (3) Wave Action; (4) Resonance; and (5) Protective Structures.

2.4.6 Probable Maximum Tsunami Flooding The Applicant will address the following topics:

(1) Probable Maximum Tsunami; (2) IIistorical Tsunami Record; (3) Source Generator Characteristics; 2.4-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

() 2.4.6 Probable kaximum Tsunami Flooding (Continued)

'(4) Tsunami Analysis; (5) Tsunami Water Levels; (6) Hydrography and Harbor or Breakwater Influences on Tsunami; and (7) Effects on Safety-Related Facilities.

2.4.7 Ice Effects Potential icing effects and design criteria for protecting safety-related facilities from ice effects will be provided by the Applicant.

() 2.4.8 Cooling Water Canals and Reservoirs The design bases for the capacity and the operating plan for safety-related cooling water canals and reservoirs will be pro-vided by the Applicant.

2.4.9 Channel Diversions The potential for upstream diversion or rerouting of the source of cooling. water to the site will be addressed by the Applicant.

2.4.10 Flooding Protection-Requirements The static and dynamic consequences of all types of flooding on each pertinent safety-related facility will be described by the Applicant.

O 2.4-5

GESSAR II 22A7007 238 NUCLEAR ISLRWD Rev. 0 2.4.11 Low Water Considerations The following low water topics will be addressed by the Applicant:

(1) Low Flow in Streams; (2) Low Water Resulting from Surges, Seiches, or Tsunami; (3) IIistorical Low Water; (4) Future Controls; (5) Plant Requirements; and (6) IIc a t Sink Dependability Requirements.

2.4.12 Dispersion,._ Dilution _and Travel Times of Accidental Releases of Li_gu_id_ Effluents in Surface Waters The ability of the surface water environment to disperse, dilute, or concentrate liquid releases of radioactive effluents relative to existing or potential future water users will be addressed by the Applicant.

2.4.13 Groundwater The following groundwater information will be provided by the Applicant:

(1) Description and Onsite Use; (2) Sources; (3) Accident Effects; O

2.4-6

-- - ---- - - - - < - - - - - - - ~ ~ .

t

) GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 '

il i 1 l -

2.4.13 Groundwater (Continued) {

(4) Monitoring or Safeguard Requirements; and j (5) Decign Bases for Subsurface Hydrostatic Loading;

2.4.14 Technical Specification and Emergency Operation j Requirements i i i

i Emergency protective measures, if any, designed to minimize the

impact of adverse hydrology-related events on safety-related

facilities will be described by.the Applicant. ,

j -

i

\

l 1

s

)

i I

\

t i

t s

!~

n I

, i 2.4-7/2.4-8 L_ ,_ _ _.- - _ _ . --

l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O I

SECTION 2.5 i

i CONTENTS

{

Section Title Page i 2.5 GEOLOGY, SEISMOLOGY AND GEOTECHNICAL I

ENGINEERING 2.5-1

2.5.1 Basic Geologic and Seismic Information 2.5-1 l 2.5.2 Vibratory Ground Motion 2.5-2

] 2.5.3 Surface Faulting 2.5-3 j 2.5.4 Stability of Subsurface Materials and j

Foundations 2.5-4 l 2.5.5 Stability of Slopes 2.5-5

2.5.6 Embankments and Dams 2.5-5 i

1 l,

l i

i >

i 4

l 2.5-i/2.5-ii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

) 2.5 GEOLOGY, SEISMOLOGY AND GEOTECHNICAL ENGINEERING 2.5.1 Basic Geologic and Seismic Information The detailed geologic and seismic information for the specific site will be gathered, as appropriate, from published reports, maps, knowledgeable sources, surveys, geophysical investigations, borings, trenches and other investigations and documents. The data collected will be presented by the Applicant according to NRC Regulatory Guide 1.70, Rev. 3 regtsirements and will include:

(1) regional geology, and (2) site geology.

The collected information will be used for evaluating the suit-ability of the site for the construction of the Nuclear Island.

() The location of Seismic Category I structures, planning the extent of excavations and backfill, choosing the compaction criteria and other site-related decisions will utilize the collected information.

The design and analysis of the buildings, including their founda-tion mats, and other structural components are independent of the site-unique requirements. The design loads are the envelope of loads associated with a broad spectrum of foundation conditions.

The methods for calculating overturning moments and foundation pressures for the Nuclear Island buildings and their foundation mats are presented in Subsection 3.8.5. For each site selected, analysis will be performed to ensure that the soil-bearing pres-sures are compatible with the site soil conditions.

D O

2.5-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.5.2 Vibratory Ground Motion The information relating to local and regional and site geology as it affects seismological investigations, history of earthquakes, correlation of epicenters with geologic structures, identification of active faults and surface faulting, will be obtained and pre-sented by the Applicant according to the NRC Regulatory Guide 1.70 Rev. 3 requirements. This information includes:

(1) Seismicity; (2) Geologic Structures and Tectonic Activity; (3) Correlation of Earthquake Activity with Geologic Structures or Tectonic Provinces; (4) Maximum Earthquake Potential; (5) Seismic Wave Transmission Characteristics of the Site; (6) Safe Shutdown Earthquake (SSE); and (7) Operating Basis Earthquake (OBE).

The objective of collecting and presenting the above information will be to establish that:

(1) the SSE maximum horizontal free-field ground acceleration at the site as measured at the existing grade level near the Nuclear Island is less than or equal to 0.3 g; (2) the SSE horizontal acceleration ground response spectra are less than or equal to the spectra shown on Fig-ure 3.7-1; and O

2.5-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.5.2 Vibratory Ground Motion (Continued)

(v)

(3) the nuclear power plant does not need to be designed for surface faulting.

The requirements relating to the SSE and OBE in the vertical direction and their relationship to (1) and (2) above are dis-cussed in Section 3.7.1.

The nuclear power plant design discussed in this SAR satisfies the seismic requirements for all sites which meet the conditions described in items (1), (2) and (3) above.

2.5.3 Surfa_cc Faulting Information describing whether or not there exists a potential for faulting at the site will be provided by the Applicant. This will

( ) include:

N./

(1) Geologic Conditions of the Site; (2) Evidence of Fault Offset; (3) Earthquakes Associated with Capable Faults; (4) Investigation of Capable Faults; (5) Correlation of Epicenters with Capable Faults; (6) Description of Capable Faults; (7) Zone, Requiring Detailed Faulting Investigation; and (8) Results ot Faulting Investigation.

A o,

2.5-3

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.5.4 Stability of Subsurface Materials and Foundations The Applicant will define the conditions and engineering properties of both soil and/or rock supporting Seismic Category I building foundations. This will include:

(1) Geologic Features; (2) Properties of Subsurface Materials; (3) Exploration; (4) Geophysical Surveys; (5) Excavations and Backfill; (6) Groundwgter Conditions; (7) Response of Soil and Rock to Dynamic Loading; (8) Liquefaction Potential; (9) Earthquake Design Basis; (10) Static Stability; (ll) Design Criteria; (12) Techniques to Improve Subsurface Conditions; (13) Subsurface Instrumentation; and (14) Construction tiotes.

O 2.5-4

l

, GESSAR II 22A7007

! 238 NUCLEAR ISLAND Rev. 0 1 i t

i 2.5.5 Stability of Slopes The Applicant will provide information concerning the static and 3

dynamic stability of all soil and rock slopes, the failure of which could adversely affect the safety of the plant. This will

! include:

,I ,

(1) Slope Characteristics; i (2) Design Criteria and Analyses; j (3) Logs of Boring; and i

t j (4) Compacted Fill j 2.5.6 Embankments and Dams

)

.i

(~' Information related to the investigation, engineering design, proposed construction and performance of all earth, rock, or rock i fill embankments used for plant flood protection or for impounding, i

cooling water required for the operation of the plant will be_ pro-vided by the Applicant. This will include:

(1) General (Purpose of the Embankment);

2 (2) Exploration;

(3) Foundation and Abutment Treatment; f (4) Embankment; (5) Slope Stability; (6) Seepage Control; (7) Diversion and Closure; j 2.5-5

___ __ . . _ _ _ . _ _ . .D

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 2.5.6 Embankments and Dams (Continued)

(8) Performance Monitoring; (9) Construction Notes; and (10) Operational Noten.

l l

i l

(

O O

2.5-6

k t

~

e G 55# e c f.

\

[

i e

l l s l

f l

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

~

SUMMARY

TABLE OF CONTENTS Chapter /

Section Title Volume 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

1 1.1.1 Type of License Required 1.1.2 Identification of Applicant 1.1.3 Number of Plant Units 1.1.4 Description of Location 1.1.5 Type of Nuclear Steam Supply System 1.1.6 Type of Containment 1.1.7 Core Thermal Power Levels 1.1.8 Scheduled Completion and Operation Dates 1.2 GENERAL PLANT DESCRIPTION 1 1.2.1 Principal Design Criteria 1.2.2 Plant Description O 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility 1

Designs 1.3.2 Comparisons of Final and Preliminary Information 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1 1.4.1 Applicant 1.4.2 Architect Engineer - Nuclear Island Design 1.4.3 Nuclear Steam Supply System Supplier 1.4.4 Turbine-Generator Vendor 1.4.5 Consultants 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1 1.5.1 Current Development Programs 1.5.2 PSAR Commitment Items 1.6 MATERIAL INCORPORATED BY REFERENCE 1 O

iii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7.2 Piping and Instrumentation Diagrams 1.7.3 Abbreviations and Symbols 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1 1.8.1 Compliance Assessment Method 1.9 STANDARD DESIGNS 1 1.9.1 Interfaces 1.9.2 Exceptions O

O iv

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0,

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 2 SITE CHARACTERISTICS 2.0

SUMMARY

l 2.1 GEOGRAPHY AND DEMOGRAPHY l 2.1.1 Site Location and Description 2.1.2 Exclusion Area Authority and Control 2.1.3 Population Distribution 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 1 2.2.1 Location and Routes 2.2.2 Descriptions 2.2.3 Evaluation of Potential Accidents 2.3 METEOROLOGY l 2.3.1 Regional Climatology 2.3.2 Local Meteorology

) 2.3.3 Onsite Meteorological Measurements Program 2.3.4 Short-Term Atmospheric Diffusion Estimates 2.3.5 Long-Term Atmospheric Diffusion Estimates 2.4 HYDROLOGIC ENGINEERING 1 2.4.1 Hydrologic Description 2.4.2 Floods 2.4.3 Probable Maximum Flood (PMF) on S,treams and Rivers 2.4.4 Potential Dam Failures, Seismically Induced 2.4.5 Probable Maximum Surge and Seiche Flooding 2.4.6 Probable Maximum Tsunami Flooding 2.4.7 Ice Effects 2.4.8 Cooling Water Canals and Reservoirs 2.4.9 Channel Diversions

( 2.4.10 Flooding Protection Requirements v

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 2.4.11 Low Water Considerations 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters 2.4.13 Ground Water 2.4.14 Technical Specifications and Emergency Operation Requirements 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING 1 2.5.1 Basic Geologic and Seismic Information 2.5.2 Vibratory Ground Motion 2.5.3 Surface Faulting 2.5.4 Stability of Subsurface Materials and Foundations 2.5.5 Stability of Slopes 2.5.6 Embankments and Dams O

vi

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume i

3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 2 3.1.1 Summary Description  :

3.1.2 Criterion Conformance 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND 2 SYSTEMS . 2 3.2.1 Seismic Classification 3.2.2 System Quality Group Classifications

! 3.2.3 System Safety Classifications 3.2.4 Quality Assurance 3.2.5 Correlation of Safety Classes with

', Industry Codes 3.3 WIND AND TORNADO LOADINGS 2 3.3.1 Wind Loadings 3.3.2 Tornado Loadings 3.3.3 BOP Interface 3.3.4 References 3.4 WATER LEVEL (FLOOD) DESIGN 2 3.4.1 Flood Protection 3.4.2 Analytical and Test Procedures 3.4.3 BOP Interface 3.4.4 References 3.5 MISSILE PROTECTION 2 3.5.1- Missile Selection and Description 3.5.2 Structures, Systems, and Components to be Protected from Externally i

Generated Missiles 3.5.3 Barrier Design Procedures 3.5.4 BOP Interface 3.5.5 References 4

O vii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Jection Title Volume 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITII Tile POSTULATED RUPTURE OF PIPING 2 3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside of Containment 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.3 References 3.7 SEISMIC DESIGN 3 3.7.1 Seismic Input 3.7.2 Seismic System Analysis 3.7.3 Seismic Subsystem Analysis 3.7.4 Seismic Instrumentation 3.7.5 BOP Interface 3.7.6 References 3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES 3 3.8.1 Concrete Containment 3.8.2 Steel Containment 3.8.3 Concrete and Steel Internal Structures of Steel Containment 3.8.4 Other Seismic Category I Structures 3.8.5 Foundations 3.8.6 BOP Interface 3.9 MECIIANICAL SYSTEMS AND COMPONENTS 4 3.9.1 Special Topics for Mechanical Components 3.9.2 Dynamic Testing and Analysis 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures 3.9.4 Control Rod Drive System 3.9.5 Reactor Pressure Vessels Internals 3.9.6 Inservice Testing of Pumps and Valves 3.9.7 BOP Interface 3.9.8 References viii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

SUMMARY

TABLE OF CONTENTS (Continued)

(J)

Chapter /

Section Title Volume 3.10 SEISMIC QUALIFICATIONS OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT (INCLUDING HYDRODYNAMIC EFFORTS) 5 3.10.1 Seismic Qualif# cation Criteria (Including Hydrodynamic Loads) 3.10.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation 3.10.3 Methods and Procedure of Seismic Analysis or Testing of Supports of Electrical Equipment and Instrumentation (Including Hydrodynamic Loads) 3.10.4 Seismic Qualification Tests and Analyses (Including Hydrodynamic Loads) 3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 5

[v) 3.11.1 Equipment Identification and Environmental Conditions 3.11.2 Qualification Tests and Analyses 3.11.3 Qualification Results 3.11.4 Loss of Ventilation 3.11.5 Estimated Chemical and Radiation Environment APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF THE NUCLEAR ISLAND 5 APPENDIX 3B CONTAINMENT LOADS 6,7 APPENDIX 3C COMPUTER PROGRAMS USED IN THE DESIGN OF SEISMIC CATEGORY I STRUCTURES 8 APPENDIX 3D ANALYSIS OF RECIRCULATION MOTOR A"D PUMP UNDER ACCIDENT CONDITIONS 8 APPENDIX 3E DESCRIPTION OF SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 8 APPENDIX 3F DYNAMIC BUCKLING CRITERIA FOR CONTAINMENT VESSEL 8 O

ix

l GESSAR II 22A7007 ,

238 tJUCLEAR ISLAND Rev. O {

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume _

1 APPENDIX 3G PIPE FAILURE ANALYSIS 8 APPENDIX 311 EFFECT OF CONCRETE ANNULUS BELOW ELEVATION (-) 5 FT., 3 IN. ON SEISMIC DESIGN LOADS Af1D BUILDING RESPONSES 8 t

O X

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O r"N SUMMiiRY TABLE OF CONTENTS (Continued)

(v)

Chapter /

Section Title Volume 4 REACTOR 4.1

SUMMARY

DESCRIPTION 9 4.1.1 Reactor Vessel 4.1.2 Reactor Internal Components 4.1.3 Reactivity Control Systems 4.1.4 Analysis Techniques 4.1.5 References s.2 FUEL SYSTEM DESIGN 9 4.2.1 General and Detailed Design Base 4.2.2 General Design Description 4.2.3 Design Evaluations 4.2.4 Testing and Inspection 4.2.5 Operating and Developmental Experience

(,,

7-~) 4.2.6 References 4.3 NUCLEAR DESIGN 9 4.3.1 Design Bases 4.3.2 Description 4.3.3 Analytical Methods 4.3.4 Changes 4.3.5 References 4.4 THERMAL - HYDRAULIC DESIGN 9 4.4.1 Design Basis 4.4.2 Description of Thermal Hydraulic Design of the Reactor Core 4.4.3 DeLcription of the Thermal and Hydraulic Design of the Reactor Coolant System 4.4.4 Evaluation 4.4.5 Testing and Verification 4.4.6 Instrumentation Requirements S 4.4.7 References xi

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 4.5 REACTOR MATERIALS 9 4.5.1 Control Rod System Structural Materials 4.5.2 Reactor Internal Materials 4.5.3 Control Rod Drive Housing Supports 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 9 4.6.1 Information for Control Rod Drive System (CRDs) 4.6.2 Evaluations of the CRDs 4.6.3 Testing and Verification of the CRDs 4.6.4 Information for Combined Performance cf Reactivity Systems 4.6.5 Evaluation of Combined Performance 4.6.6 References APPENDIX 4A CONTROL ROD PATTERNS AND ASSOCIATED POWER DISTRIBUTION FOR TYPICAL BWR 9 4A.1 Introduction 4A.2 Power Distribution Strategy 4A.3 Results of Core Simulation Studies 4A.4 Glossary of Terms 4A.5 References O

xii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

~

/ 'N

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1

SUMMARY

DESCRIPTION 10 5.1.1 Schematic Flow Diagram 5.1.2 Piping and Instrumentation Diagram 5.1.3 Elevation Drawing 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 10 5.2.1 Compliance with Codes and Code cases 5.2.2 Overpressure Protection 5.2.3 Reactor Coolant Pressure Boundary Materials 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.2.5 Reactor Coolant Pressure Boundary and ECCS System Leakage Detection N System 5.2.6 References 5.3 REACTOR VESSEL 10 5.3.1 Reactor Vessel Materials 5.3.2 Pressure / Temperature Limits 5.3.3 Reactor Vessel Integrity

, 5.3.4 References 5.4 COMPONENT AND SUBSYSTEM DESIGN 10 5.4.1 Reactor Recirculation Pumps 5.4.2 Steam Generators (PWR) 5.4.3 Reactor Coolant Piping 5.4.4 Main Steam Line Flow Restrictors 5.4.5 Main Steam Line Isolation System 5.4.6 Reactor Core Isolation Cooling System (RCIC) 5.4.7 Residual Heat Removal System 5.4.8 Reactor Water Cleanup System 5.4.9 Main Steam Lines and Feedwater Piping

, ) 5.4.10 Pressurizer xiii i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 5.4.11 Pressurizer Relief Valve Discharge System 5.4.12 Valves

5. .13 Safety and Relief Valves 5.4.14 Component Supports 5.4.15 References O

O xiv

.GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 6 ENGINEERED SAFETY FEATURES 6.0 GENERAL ll 6.1 ENGINEERED SAFETY FEATURE MATERIALS 11 6.1.1 Metallic Materials 6.1.2 Organic Materials 6.2 CONTAINMENT SYSTEMS 11 6.2.1 Containment Functional Design 6.2.2 Containment Heat Removal System 6.2.3 Secondary Containment Functional Design 6.2.4 Containment Isolation System 6.2.5 Combustible Gas Control in Containment 6.2.6 Containment Leakage Testing

() 6.2.7 6.2.8 Suppression Pool Makeup System References 6.3 EMERGENCY CORE COOLING SYSTEMS 11 6.3.1 Design Bases and Summary Description 6.3.2 System Design 6.3.3 ECCS Performance Evaluation ,

j 6.3.4 Tests and Inspections l 6.3.5 Instrumentation Requirements 6.3.6 References 6.4 HABITABILITY SYSTEMS ll l

{ 6.4.1 Design Basis 6.4.2 System Design l 6.4.3 Systems Operational Procedures 6.4.4 Design Evaluations f

6.4.5 Testing and Inspection f

6.4.6 Instrumentation Requirements 6.4.7 Nuclear Island / BOP Interface t

XV

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENYS (Continued)

O Chapter /

Section Title Volume 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 11 6.5.1 Standby Gas Treatment System (SGTS) 6.5.2 Containment Spray Systems 6.5.3 Fission Product Control Systems 6.5.4 Ice Condensers as a Fission Product Control System 6.5.5 References 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS 11 6.6.1 Components Subject to Examination 6.6.2 Accessibility 6.6.3 Examination Techniques and Procedures 6.6.4 Inspection Intervals 6.6.5 Examination Categories and Requirements 6.6.6 Evaluation of Examination Results 6.6.7 System Pressure Tests 6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures 6.7 MAIN STEAM POSITIVE LEAKAGE CONTROL SYSTEM (MSPLCS) 11 6.7.1 Design Bases 6.7.2 System Description 6.7.3 System Evaluation 6.7.4 Inspection and Testir.g 6.7.5 Instrumentation Requirements 6.8 PNEUMATIC SUPPLY SYSTEM ll 6.8.1 Design Bases 6.8.2 System Description 6.8.3 System Evaluation 6.8.4 Inspection and Testing Requirements 6.8.5 Instrumentation Requirements APPENDIX 6A IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS 11 xvi

22A7007 GESSAR II 238 NUCLEAR ISLAND Rev. O'

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section ' Title Volume 7 INSTRUMENTATION AND-CONTROL SYSTEMS

7.1 INTRODUCTION

(All Systems) 12 7.1.1 Identification of Safety-Related 1 Systems 7.1.2 Identification of Safety and Power  :

+

Generation Criteria 7.2 REACTOR PROTECTION (TRIP) SYSTEM (RPS) 12 7.2.1 Description 7.2.2 Conformance Analysis I

7.3 ENGINEERED SAFETY FEATURES SYSTEM, INSTRUMENTATION AND CONTROL 13 7.3.1 Description

7. 3. 2. Analysis 6

4 l -HPCS -Shield Building Annu us Mixing

-ADS

-LPCS - econdary Cgntain-ment Isolation

- " !b -Primary Containment

-CRVICS Isolation LCS i -MSPLCS -Standby Power

-RHR/ Containment -D-G Support Systems j- Spray -Essential Service

-RHR/ Suppression Pool Water Cooling -ESF Area Cooling

! -Suppression Pool -Pneumatic Supply Makeup

-CB Atmospheric

-Combustible Gas Control Control 1 -CB Chilled Water

-SGTS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 14 7.4.1 Description 7.4.2 Analysis

-RCIC -RHR/ Shutdown Cooling

-SLC -Remote Shutdown j

xvii i

1

., ,.___._m. _ _ , . . . _ , . - = - . . . . , . , _ . , . . , , . . _ . , . . . _ .- . _ , . . . _ _ , - . -

GESSAR II 22A7007 238 KUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

O Chapter / Volume Section Title SAFETY-RELATED DISPLAY INSTRUMENTATION 14 7.5 7.5.1 Description 7.5.2 Analysis

-Nuclenet Control -BOP Benchboard nsol -Supervisory Moni-

-Standby Information toring Console Panel -Display Control

-Rx Core Cooling BB System 7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 14 7.6.1 Description 7.6.2 Analysis 7.6.3 Additional Design Considerations Analyses 7.6.4 References

-Neutron Monitoring -FPFCS h

-Process Radiation -DW/ Containment Monitoring Vacuum Relief

-Refueling Interlocks -Vent & Pressure Control

-Leak Detection

~^

-Rod Pattern Control

-Suppression Pool

-HP/LP System * '# "'

Interlock Mon?itoring

-Recirculation Pump Trip 9

xviii

GESSAR II 22A7007 i 238 NUCLEAR' ISLAND Rev. 0

SUMMARY

TABLE OF CONTENTS (Continued) a

(

Chapter /

Section Title Volume 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 14 4 7.7.1 Description 7.7.2 Analysis 7.7.3 References

! -RPV Instrumentation -Leak Detection t-i -Rod Control & -Rod Block Trip Information -Fire Protection

-Recirculation Flow -Drywell Chiller &

Cooling

-Feedwater Control -Plant Instrument Air

-Performance Moni- -Neutron Monitoring toring System

-Radwaste 7.8 NI/ BOP INTERFACES 14

]

7.8.1 Essential Service Water (Supply)

} System Instrumentation an'd Controls i

7.8.2 Diesel Generator Fuel Oil Transfer

{' System 1

i APPENDIX 7A I&C ELEMENTARY DIAGRAMS :15 1

i l

1 4

l' i

xix 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 8 ELECTRIC POWER

8.1 INTRODUCTION

16 8.1.1 Utility Grid Description 8.1.2 Onsite Electric Power System 8.1.3 Design Bases 8.2 OFFSITE POWER SYSTEM 16 8.2.1 Description 8.2.2 Analysis 8.2.3 Nuclear Island - BOP Interface 8.3 ONSITE POWER SYSTEMS 16 8.3.1 AC Power Systems 8.3.2 DC Power Systems 8.3.3 Fire Protection of Cable Systems O

O XX

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

/~'s

SUMMARY

TABLE OF CONTENTS (Continued) v Chapter /

Section Title Volume 9 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANLLING 17 9.1.1 New Fuel Storage (High Density) 9.1.2 Spent Fuel Storage (High Density) 9.1.3 Fuel Pool Cooling and Cleanup System 9.1.4 Fuel-Handling System 9.2 WATER SYSTEMS 17 9.2.1 Essential Service Water (ESW) System 9.2.2 Closed Cooling Water System 9.2.3 Demineralized Water Makeup System 9.2.4 Potable and Sanitary Water Systems 9.2.5 Ultimate Heat Sink 9.2.6 Condensate Storage Facilities and g Distribution System

(_,/ 9.2.7 Plant Chilled Water Systems 9.2.8 Heated Water Systems 9.2.9 Nuclear Island / BOP Interfaces 9.3 PROCESS AUXILIARIES 17 9.3.1 Compressed Air Systems 9.3.2 Process Sampling System 9.3.3 Floor and Equipment Drainage Systems 9.3.4 Chemical and Volume Control System (PWR) 9.3.5 Standby Liquid Control System 9.4 AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEM 17 9.4.1 Control Room HVAC System 9.4.2 Fuel Building HVAC System 9.4.3 Auxiliary Building HVAC Systems 9.4.4 Turbine Building Area Ventilation System gw 9.4.5 Reactor Building H.AC System L) xxi

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF~ CONTENTS (Continued)

Unapter/

3ection Title Volume 9.4.6 Radwaste Building IIVAC System 9.4.7 Diesel-Generator Buildings llVAC Systems 9.5 OTilER AUXILIARY SYSTEMS 17 9.5.1 Fire Protection System 9.5.2 Communications Systems 9.5.3 Lighting Systems 9.5.4 Diesel-Generator Fuel Oil Storage and Transfer System 9.5.5 Diesel-Generator Cooling Water Syctem 9.5.6 Diesel-Generator Starting Air System 9.5.7 Diesel Engine Lubricatfon System 9.5.8 Diesel Generator Combustion Air ,

Intake and Exhaust System .

9.5.9 Suppression Pool Cleanup System 9.5.10 Nuclear Island - BOP Interface APPENDIX 9A FIRE !!AZARD ANALYSIS 18 s

I i l l' 1

l 6

l l

O Xxii I

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

V Chapter /

Section i Title Volume 10 STEAM AND POWER CONVERSION SYSTEM 10.1

SUMMARY

DESCRIPTION 19 10.2 TURBINE GENERATOR 19 10.2.1 Design Bases Functional Limitations by Design or Operational Charac-teristics of the Reactor Coolant System 10.2.2 System Description 10.2.3 Turbine Disk Integrity 10.2.4 Evaluation 10.3 MAIN STEAM SUPPLY 19 10.4 OTilER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 19 10.4.1 Main Condensers s

s ,

10.4.2 Condenser Air Removal System 10.4.3 Main Condenser Evacuation System 10.4.4 Turbine Bypass System 10.4.5 Circulating Water System 10.4.6 Condensate Cleanup System 10.4.7 Condensate and Feedwater System 10.4.8 Steam Generator Blowdown System (PWR) 10.4.9 Auxiliary Feedwater System (PWR)

L J

xxiii t

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 11 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS 19 11.1.1 Fission Products 11.1.2 Activation Products 11.1.3 Tritium 11.1.4 Fuel Fission Production Inventory and Fuel Experience 11.1.5 Process Leakage Sources 11.1.6 Radwaste System 11.1.7 Radioactive Sources in the Gas Treatment System 11.1.8 Source Terms for Component Failures 11.1.9 Other Releases 11.1.10 References 11.2 LIQUID WASTE MANAGEMENT SYSTEM 19 11.2.1 Design Basis 11.2.2 System Descriptions 11.2.3 Estimated Releases 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 19 11.3.1 Design Bases 11.3.2 Main Condenser Steam Jet Air Ejector Low-Temp RECHAR System Description 11.3.3 RECHAR System Operating Procedure 11.3.4 Radioactive Releases 11.3.5 References 11.4 SOLID RADWASTE SYSTEM 19 11.1.1 Design Bases 11.4.2 System Description 1

0 xxiv l

1 ,

l l GESSAR II 22A7007 1 238 NUCLEAR ISLAND Rev. 0 l

l

SUMMARY

TABLE OF CONTENTS (Continued) 1 l Chapter /

Section Title Volume j l 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING

AND SAMPLING SYSTEMS 19

i 11.5.1 Design Bases i

11.5.2 System Description  ;

11.5.3 Effluent Monitoring and Sampling l I 11.5.4 Process Monitoring and Sampling 11.5.5 Calibration and Maintenance  !

11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 19

{

?

f i

l i

i n

I i

xxv i

?

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OP CONTENTS (Continued)

Chapter /

Section Title Volume 12 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 19 12.1.1 Policy Considerations 12.1.2 Design Considerations 12.1.3 Operational Considerations 12.2 RADIATION SOURCES 19 12.2.1 Contained Sources 12.2.2 Airborne Radioactive Material Sources 12.2.3 References 12.3 RADIATION PROTECTION DESIGN FEATURES 19 12.3.1 Facility Design Features 12.3.2 shielding 12.3.3 Ventilation 12.3.4 Area Radiation and Airborne Radioactivity Monitors 12.3.5 References 12.4 DOSE ASSESSMENT 19 12.5 HEALTH PHYSICS PROGRAM 19 l

l xxvi

1 I

1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i

i SUK%RY TABLE OF CONTENTS (Continued) i '

G Chapter /

i Section Title

! Volume

) 13 CONDUCT OF OPERATIONS 19 i

5 i

i i

i i

}

4 1

v l l

O '

I h

6 l I

i xxvii  !

w ---,y._.,y,-.--w-

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

O Section Title Volume 14 INITIAL TEST PROGRAM 14.1 TEST PROGRAM 20 14.1.1 Administrative Procedures (Testing) 14.1.2 Administrative Procedures (Modifications) 14.1.3 Test Objectives and Procedures 14.1.4 Fuel Loading and Initial Operation 14.1.5 Administrative Procedures (System Operation) 14.2 SPECIFIC INFORMATION TO BE INCLUDED IN SAFETY ANALYSIS REPORTS 20 14.2.1 Summary of Test Program and Objectives 14.2.2 Organization and Staffing 14.2.3 Test Procedures 14.2.4 Conduct of Test Program 14.2.5 Review, Evaluation and Approval of Test Results 14.2.6 Test Records 14.2.7 Conformance of Test Programs with Regulatory Guides 14.2.8 Utilization of Reactor Operating and Testing Experiences in the Develop-ment of Test Program 14.2.9 Trial Use of Plant Operating and Emergency Procedures 14.2.10 Initial Fuel Loading and Initial Criticality 14.2.11 Test Program Schedule 14.2.12 Individual Test Descriptions O

xxviii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

SUMMARY

TABLE OF CONTENTS (Continued) 1 Chapter /

Section Title Volume 15 ACCIDENT ANALYSES 15.0 GENERAL 21 15.0.1 Analytical objective 15.0.2 Analytical Categories 15.0.3 Event Evaluation 15.0.4 Nuclear Safety Operational Analysis (NSOA) 31ationship 15.0.5 Referent as 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 21 15.1.1 Loss of Feedwater Heating 15.1.2 Feedwater Controller Failure -

Maximum Demand 15.1.3 Pressure Regulator Failure - Open 15.1.4 Inadvertent Safety / Relief Valve Opening N s/ 15.1.5 Spectrum of Steam System Piping Failures Inside and Outside of Containment in a PWR 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.1.7 References 15.2 INCREASE IN REACTOR PRESSURE 21 15.2.1 Pressure Regulator Failure - Closed 15.2.2 Generator Load Rejection 15.2.3 Turbine Trip 15.2.4 MSLIV Closures 1

15.2.5 Loss of Condenser Vacuum 15.2.6 Loss of Offsite AC Power I 15.2.7 Loss of Feedwater Flow 15.2.8 Feedwater Line Break 15.2.9 Failure of RHR Shutdown Cooling O

XXiX

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev, O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

O Section Title Volume 15.3 DUCREASE IN REACTOR COOLANT SYSTEM FLOW RATE 21 15.3.1 Recirculation Pump Trip 15.3.2 Recirculation Flow Control Failure -

Decreasing Flow 15.3.3 Recirculation Pump Seizure 15.3.4 Recirculation Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 21 15.4.1 Rod Withdrawal Error - Low Power 15.4.2 Rod Withdrawal Error at Power 15.4.3 Control Rod Maloperation (System Malfunction or Operator Error) 15.4.4 Abnormal Startup of Idle Recirculation Pump 15.4.5 Recirculation Flow Control with Increasing Flow 15.4.6 Chemical and Volume Control System Malfunctions 15.4.7 Misplaced Bundles Accident 15.4.8 Spectrum of Rod Ejection Assemblies 15.4.9 Control Rod Drop Accident (CRDA) 15.5 INCREASE IN REACTOR COOLANT INVENTORY 21 15.5.1 Inadvertent HPCS Startup 15.5.2 Chemical Volume Control System Malfunction (or Operator Error) 15.5.3 BWR Transients Which Increase Reactor Coolant Inventory O

XXX

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

/)

V Chapter /

Section Title Volume 15.6 DECREASE IN REACTOR COOLANT INVENTORY 21 15.6.1 Inadvertent Safety / Relief Valve Opaning 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside containment 15.6.3 Steam Generator Tube Failure 15.6.4 Steam System Piping Break Outside Containment 15.6.5 Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary) - Inside Containment 15.6.6 Feedwater Line Break - Outside Containment 15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND f_s COMPONENTS 21

( ,) 15.7.1 Radioactive Waste System Leak or Failure 15.7.2 Liquid Radioactive System Failure 15.7.3 Postulated Radioactive Releases Due to Liquid Radwaste Tank Failure 15.7.4 Fuel-Ilandling Accident 15.7.5 Spent Fuel Cask Drop Accidents APPENDIX 15A PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS 21 APPENDIX 15B BWR/6 GENERIC ROD WITilDRAWAL ERROR ANALYSIS 21 (3

G XXXi

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3UMMARY TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 16 STANDARD TECHNICAL SPECIFICATIONS FOR GENERAL ELECTRIC BOILING WATER REACTORS 16.1 DEFINITIONS 22 16.1.1 Action 16.1.2 Average Planar Exposure 16.1.3 Average Planar Linear Heat Generation Rate 16.1.4 Channel Calibration 16.1.5 Channel Check 16.1.6 Channel Functional Test 16.1.7 Core Alteration 16.1.8 Critical Power Ratio 16.1.9 Dose Equivalent I-131 16.1.10 E-Average Disintegration Energy 16.1.11 Emergency Core Cooling System (ECCS)

Response Time 16.1.12 Frequency Notation 16.1.13 Identified Leakage 16.1.14 Isolation System Response Time 16.1.15 Limiting Control Rod Pattern 16.1.16 Linear Heat Generation Rate 16.1.17 Logic System Functional Test 16.1.18 Maximum Total Peaking Factor 16.1.19 Minimum Critical Power Ratio 16.1.20 Operable - Operability j 16.1.21 Operational Condition (Condition) 16.1.22 Physics Test 16.1.23 Presutre Boundary Leakage 16.1.24 Primary Containment Integrity 16.1.25 Rated Thermal Power 16.1.26 Reactor Protection System Response Time 16.1.27 Recirculation Pump Trip System Response Time xxxii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 7-~g

SUMMARY

TABLE OF CONTENTS (Continued) b Chapter /

Section Title Volume 16.1.28 Reportable Occurrence 16.1.29 Rod Density 16.1.30 Secondary Containment Integrity 16.1.31 Shutdown Margin 16.1.32 Staggered Test Basis 16.1.33 Thermal Power 16.1.34 Total Peaking Factor 16.1.35 Unidentified Leakage 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 22 16.2.1 Safety Limits 16.2.2 Limiting Safety System Settings 16.B2 SAFETY LIMITS 22 16.B2.1 Bases O 16.B2.2 Limiting Safety System Settings

( ,/

16.3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 22 16.3/4.0 Applicability 16.3/4.1 Surveillance Requirements 16.3/4.2 Power Distribution Limits 16.3/4.3 Instrumentation 16.3/4.4 Reactor Coolant System 16.3/4.5 Emergency Core Cooling Systems 16.3/4.6 Containment Systems 16.3/4.7 Plant Systems 16.3/4.8 Electrical Power Systems 16.3/4.9 Refueling Operations 16.3/4.10 Special Test Exceptions l

N.

[O xxxiii i

GESSAR II 22A7007 238 NUCLEAR ISLAND Revo 0

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

O Section Title Volume 16.B3/4.0 Applicability 16.B3/4.1 Reactivity Control Systems 16.B3/4.2 Power Distribution Limits 16.B3/4.3 Instrumentation 16.B3/4.4 Reactor Coolant System 16.B3/4.5 Emergency Core Cooling System 16.B3/4.6 Containment Systems 16.B3/4.7 Plant Systems 16.B3/4.8 Electrical Power Systems 16.B3/4.9 Refueling Operations 16.B3/4.10 Special Test Exceptions 16.5 DESIGN FEATURES 22 16.5.1 Site 16.5.2 Containment 16.5.3 Reactor Core 16.5.4 Reactor Coolant System 16.5.5 Meteorological Tower Location 16.5.6 Fuel Storage 16.5.7 Camponent Cyclic or Transient Limit O

XXXiv

GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter / 1

' Section Title Volume I

)

17 QUALITY ASSURANCE 4 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 22 17.2 QUALITY ASSURANCE DURING THE OPERATING PHASE 22 l 1

l.

i t

i i

i

!O i

i i

9 xxxv/xxxvi

_ - . - . . _ _ _ - - _ - . -