ML20054M702

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Amend 25 to PSAR
ML20054M702
Person / Time
Site: Skagit
Issue date: 07/02/1982
From:
PUGET SOUND POWER & LIGHT CO.
To:
Shared Package
ML20054M699 List:
References
NUDOCS 8207140248
Download: ML20054M702 (46)
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S/HNP-PSAR 7/2/82 O File this instruction sheet in the front of Volume 1 as a h record of changes.

The following information and check list are furnished as a guide for the insertion of new sheets for Amendment 25 into the Preliminary Safety Analysis Report for the Skagit/

Hanford Nuclear Project. This material is denoted by use of the amendment date in the upper right hand corner of the page.

New sheets should be inserted as listed below:

Discard Old Sheet Insert New Sheet (Front /Back) (Front /Back)

CHAPTER 2 2-1/i1 2-1/11 2.2-11/12 2.2-11/12 2.2-12a/ blank 2.5-21/22 2.5-21/22 2.5-28a/ blank 2.5-28a/ blank 2.5-29/30 2.5-29/30 O 2.5-33/34 2.6-3/4 20-27/28 2.5-33/34 2.6-3/4 20-27/28 CHAPTER 3 3.7-5/6 3.7-5/6 3.7-6a/ blank 3.7-7/8 3.7-7/8 3.8-49/50 3.8-49/50 CHAPTER 6 6.4-1/2 6.4-1/2 8207140248 820702 1 Amendment 25 PDR ADOCK 05000522 D PDR

S/HNP-PSAR 7/2/82 Discard Old Sheet Insert New Sheet (Front /Back) (Front /Back)

CHAPTER 7 Figure 7.3-4 (1 of 11) Figure 7.3-4 (1 of 11)

Figure 7.3-4 (3 of 11) Figure 7.3-4 (3 of 11)

Figure 7.3-4 (9 of 11) Figure 7.3-4 (9 of 11)

CHAPTER 9 9.4-7/8 9.4-7/8 9.4-9/10 9.4-9/10 Figure 9.4-1 Figure 9.4-1 CHAPTER 12 12.1-27/28 12.1-27/28 12.1-28a/28b 12.1-28a/28b 12.1-28c/28d 12.1-28c/28d QUESTIONS AND RESPONSES Question 420.1/ blank Question 420.1/ blank O

l 2 Amendment 25

S/HNP-PSAR 12/21/81 p CHAPTER 2.0 THE SITE AND ENVIRONMENTAL INTERFACES 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 Location 2.1-1 2.1.1.2 Site Area 2.1-1 2.1.1.3 Boundary for Establishing Effluent Release Limits 2.1-3 2.1.2 Exclusion Area Authority and Control 2.1-3 2.1.2.1 Authority 2.1-3 2.1.2.2 Control of Activities Unrelated to Plant Operation 2.1-4 2.1.2.3 Arrangements for Traffic Control 2.1-4 2.1.2.4 Abandonment or Relocations of Roads 2.1-5 2.1.3 Population Distribution 2.1-5 2.1.3.1 Population Within Ten Miles 2.1-6

(' } 2.1.3.2 Population Between 10 and 50 Miles 2.1-6

( 2.1.3.3 2.1.3.4 Transient Population Low Population Zone 2.1-7 2.1-8 2.1.3.5 Population Center 2.1-9 2.1.3.6 Population Density 2.1-9 2.2 Nearby Industrial, Transportation 2.2-1 and Military Facilities 2.2.1 Locations and Routes 2.2-1 2.2.2 Descriptions 2.2-2 2.2.2.1 Description of Facilities 2.2-2 2.2.2.2 Description of Products and 2.2-4 Materials 2.2.2.3 Pipelines 2.2-5 2.2.2.4 Waterways 2.2-5 2.2.2.5 Descriptions of Airports 2.2-5 2.2.2.6 Projection of Industrial Growth 2.2-7 2.2.3 Evaluation of Potential Accidents 2.2-8 2.2.3.1 Determination of Design Basis 2.2-8 Accidents 2.2.3.1.1 Explosions 2.2-8 2.2.3.1.2 Potential Accidents Related to Barge 2.2-8 Traffic 2.2.3.1.3 Water Contamination 2.2-9 O

2-1 Amendment 23

S/HNP-PSAR 7/2/82 SECTION TITLE PAGE 2.2.3.1.4 Toxic Chemicals 2.2-9 2.2.3.1.4.1 Chemicals Stored in Nearby 2.2-9 Facilities 2.2.3.1.4.2 Chemicals Transported Near the S/HNP Site 2.2-10 2.2.3.1.4.3 Chemicals Stored On-Site 2.2-12 2.2.3.1.5 Vapor Clouds 2.2-12 2.2.3.1.6 Ground Fires 2.2-12 2.2.3.2 Effects of Design Basis Events 2.2-12a 2.3 Meteorology 2.3-1 2.3.1 Regional Climatology 2.3-2 2.3.2 Local Meteorology 2.3-3 2.3.2.1 Data Comparisons 2.3-3 2.3.2.2 Potential Influence of S/HNP and Its 2.3-3 Facilities on Local Meteorology 2.3.2.3 Local Meteorological Conditions for 2.3-3 Design and Operating Basis 2.3.3 On-Site Meteorological Measurements 2.3-3 Programs 2.3.4 Short-Term Dispersion Model 2.3-3 2.3.4.1 Dispersion Model 2.3-3 2.3.4.2 Determination of Conservative X/O 2.3-5 Values 2.3.4.3 Input Meteorological Data 2.3-5 2.3.4.4 Short-Term Dispersion Estimates 2.3-6 2.3.5 Long-Term Dispersion Model 2.3-6 2.3.5.1 Long-Term Atmospheric Dispersion 2.3-6 Model 2.3.5.2 Input Meteorological Data 2.3-8 2.3.5.3 Long-Term Dispersion Estimates 2.3-8 2.4 Hydrologic 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 2.4-2 2.4.2 Floods 2.4-5 2.4.2.1 Flood History 2.4-5 2.4.2.2 Flood Design Consideration 2.4-6 2.4.2.3 Effects of Local Intense 2.4-7 Precipitation O

2-11 Amendment 25

S/HNP-PSAR 7/2/82 A preliminary analyais of the chemicals identified above was performed to determine which chemicals could pose a hazard to the control room operators. The concentrations of toxic 23 chemicals in the con:rol room are calculated similarly to the analysis presented in Section 2.2.3.1.4.1. The results, shown on Table 2.2-5, indicate that anhydrous ammonia and truck shipments of chlorine could pose a hazard to the 1 24 control room operators if transported near the S/HNP Site. l 25 A survey perf ormed to determine shipment f requencies of 23 chlorine and anhydrous ammonia indicates that chlorine is not shipped by trucks near the S/HNP Site (Refs 17-20).

25 Anhydrous ammonia is shipped only by truck in quantities of 2000-3000 gallons 8-10 times annually (Ref 17). Therefore, 23 only anhydrous ammonia could pose a hazard to the control room operators. A probabilistic analysis of the risk of potential accidents involving truck shipments of anhydrous ammonia near the S/HNP Site has been performed to ensure the requirements of 10CFR Part 100, Section 100.10, are met.

The analysis indicates clearly that the probability of an accidental spill of anhydrous ammonia which results in the exposure of control room personnel to the toxic level shown in Table 2.2-5 is of such low magnitude that an evacuation of the control room leading to an event resulting in a Part 100 release is extremely unlikely.

The probability of exposure of control room personnel to 25 Os toxic levels of anhydrous ammonia is the product of the number of shipments per year passing the Plant Site; the number of spills per truck mile; the number of miles of road near the Plant on which anhydrous ammonia is transported; and the probability that meteorological conditions are such that an anhydrous ammonia gas plume will reach the plant with sufficient concentration to be toxic.

A conservative estimate of the probability of a truck accident,8 2.7 x 10- accidents per milewhich (Ref results 21)in. a spill Other of anhydrous sources ammonia, is indicate an even lesser probability of an accident (Ref s 22, 23). Anhydrous ammonia shipments pass within 5 miles of S/HNP on a 10-mile segment of the shipment route to the 200 East Area of the Reservation. The probability that meteorological conditions will exist allowing the anhydrous ammonia to reach the control room air intake with sufficient concentration to be toxic from an accident on the 10-mile shipment route segment has been calculated to be .06.

Combining the above factors, the resulting risk to control room operators is approximately 1 x 10-7 per year. This probability is based on conservative assumptions and is less than 1 x 10-6, the guideline set f orth in Standard Review Plan (NUREG-0800) Section 2.2.3 for conservative analyses.

O 2.2-11 Amendment 25

S/ENP-PSAR 7/2/82 Therefore, the potential f or an evacuation of the control room leading to an event resulting in a Part 100 release as a result of a spill of anhydrous ammonia during 25 transportation near the Plant Site is negligible, and a spill of anhydrous ammonia does not present a hazard to the Plant. For details of the complete analysis refer to Ref erence 24.

2.2.3.1.4.3 Chemicals stored on-Site. Chemicals stored on the S/HNP Site that could pose a hazard to the control room operators have been identified and are listed on Table ,

2.2-6. l l

The chemicals listed on Table 2.2-6 were analyzed to i determine the impacts of an accidental spill on the control I room habitability using the methodology described in Section 2.2.3.1.4.1. Results of this analysis as shown on Table 2.2-7 indicate that no chemical stored on the S/HNP Site could pose a hazard to the control room operators.

2.2.3.1.5 Vapor Clouds Flammable vapor clouds at the control room air intake are not a problem at the S/HNP Plant. Liquid petroleum gas (LPG), the probable source f or flammable vapor clouds, has a low usage rate at the Hanf ord Reservation areas north of 23 S/HNP. A small quantity of LPG is stored in the 200 West Area, and it is assumed that a small number of delivery vehicles pass by the Plant. There is no LPG stored within 5 miles of the S/HNP. No other potential sources of flammable vapor clouds have been located.

2.2.3.1.6 Ground Fires l Range fires occur in the desert surrounding the S/HNP Site.

There was an average of 12.2 fires per year during a recent 10 year period, and the median fire covered an area of 6 acres. Every 3.3 years there is a fire greater than 1000 acres (Ref 5). The brush-type fires present no problem to the safety-related concrete structures. Landscaping around the Plant provides a firebreak. A perimeter fence and road serve the same purpose. Smoke induction to the control room is prevented by smoke detectors and automatic isolation of the ventilation system.

O 2.2-12 Amendment 25

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S/HNP-PSAR 7/2/82 l

2.2.3.2 Effects of Design Basis Events Based on the inf ormation given in the preceding sections 'f or 23 the S/HNP and the evaluations described in Section 2.2.3.1,

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1 explosions (Section 2. 2. 3.1.1) , barge traffic accidents (Section 2.2.3.1.2), water contamination (Section 2.2.3.1.3), toxic chemicals (Section 2.2.3.1.'4), flammable l 25 vapor clouds (Section 2. 2. 3.1. 5) and ground fires (Section 2.2.3.1.6) do not constitute hazards to the Plant and are 23 theref ore not considered as design basis events.

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l 2.2-12a Amendment 25

S/HNP-PSAR 12/21/81 2.5.2.7 Operating Basis Earthquake 7~

D A peak acceleration of 0.125g or one-half that of the Site SSE is a= signed for the Site Operating Basis Earthquake (Site OBE). A peak acceleration of 0.175g or one-half that of the Design SSE is assigned for the Design Operating Basis Earthquake ( Design OBE) . Unless otherwise specified, references to the OBE in Chapter 2 of this SAR refer to the Site OBE and references in Chapters 3 through 15 refer to the Design OBE.

2.5.3 SURFACE FAULTING All available geologic and geophysical information was evaluated to determine whether any evidence suggested that surface f aulting might occur within 5 miles of the Site.

Available information was supplemented with detailed, site-specific geologic and geophysical surveys extending beyond 5 miles in some directions and concentrated in a 2 mile radius of the Site. These surveys have included ground gravity and magnetic surveys along closely spaced lines and a seismic refraction survey. The results of the geophysical investigations are described in appendices 2K and 2L.

Geologic investigations undertaken specifically to supple-

) ment available information included photogeology, field s' mapping, rotary and core drilling, and stratigraphic analysis. The results of these investigations are described 23 in Section 2.5.1.2 and Appendix 2R.

Geologic and geophysical studies have shown that the basalt bedrock underlying the Site within a radius of at least 2 miles shows slopes with only gentle relief (generally less than 5 degrees). Sedimentary units within the Miocene-Pliocene Ringold Formation which overlies bedrock are generally horizontal or show some minor warping (slopes less than 5 degrees). Sediments overlying the uppermost Ringold Formation (generally considered to be part of the Hanford Forma tion ) are Pleistocene or older in age and contain a refracting horizon of 8,000 ft/sec velocity which is

[ flat-lying within a radius of 2 miles of the Site. This l velocity horizon has also been found to be flat-lying over an area of approximately 28 square miles within the vicinity of the Site. There are no photolinears within the Site Area which are structurally controlled. On the basis of these data, there is no evidence that suggests potential for surface faulting; therefore, Sections 2.5.3.1 through 2.5.3.8 do not apply.

O 2.5-21 Amendment 23

S/HNP-PSAR 7/2/82 2.5.4 STABILITY OF SUBSURFACE MATERIALS Beneath a surficial layer of loose silty sand, the central Plant f acilities are underlain by geologic strata which will l provide suitable f ounding materials. The loose surficial sands have an r;erage thickness of 6 to 8 feet across the Site (mean approximate surf ace elevation 525 f eet) . These sands are underlain by medium dense to dense sands of late 23 Pleistocene age to approximate elevation 490 feet (MSL),

very dense sands and gravels of late Pliocene (?) to Pleistocene age to approximate elevation 320 feet, and lacustrine and fluvial very dense sands and gravels and hard silts and clayey silts of late Miocene to Pliocene age (Ringold Formation) through to basalt bedrock at approximate elevation -200 feet. The present groundwater table is at elevation 400 feet.

A majority of the central Plant structures, with the excep-tion of the Ultimate Heat Sinks and Radwaste Buildings which will be founded directly on the very dense sands below Eleva- 24 tion 490', will be supported directly on structural backfill (see Figure 2.5-15). The fact that the excavation has reached the very dense (pre-Missoula) sands will be verified by a qualified inspector prior to placement of backfill. 25 The Standby Diesel Generator Fuel Storage tanks will be founded in the Missoula sand.

Because of the use of large mat foundations and the nature of the foundatio- materials, there is no possibility of 23 large scale movements associated with bearing capacity failure. Structure permissible total and dif f erential settlements control the allowable bearing pressures.

Permanent settlements will occur under the static loads applied by the surface structures and their equipment. 24 These settlements are not expected to cause adverse effects on the structures and operating equipment.

During design basis earthquake shaking, the surface l 23 structures will be subjected to dynamic pseudo-elastic 24 (recoverable) movements. The dynamic movements are not expected to have adverse effects on the Plant and equipment during the Design SSE.

Because of the great depth of the water table and the nature of the geologic strata at the Site, there is no potential 23 for liquefaction of the structure foundations. In addition, there are no other foundation conditions (e.g. zones of alteration or weathering, collapse features, poorly consolidated strata or soluble zones) which could impact foundation stability.

O 2.5-22 Amendment 25

S/HNP-PSAR 7/2/82 O conservatively correlated with relative density. The latter test will then be used for quality control in conjunction with appropriate criteria f or moisture control.

l 24 A test fill program will be carried out prior to the start of backfilling in order to develop a suitable construction procedure to meet the backfill compaction criteria. The construction procedure so developed will identify the 25 compaction equipment, loose lift thickness, number of passes of the compaction equipment, and the required speed of travel of the compactor.

The results of the test fill program and the recommended construction procedure f or backfill will be documented in a report which will be provided to the NRC within 30 days of the completion of the test fill program.

2.5.4.5.4 Engineering Properties of Backfill Grain size distributions of the clean black medium sands which will be suitable as Category 1 backfill are given in Figures 20 B-3 and 20 B-9. Static and dynamic material properties were evaluated for samples of the backfill

, material taken f rom the test trenches and prepared to a

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relative density of 75 percent (structural backfill will be placed to a minimum relative density of 75 percent) . The maximum and minimum densities presented in Table 2Q B-6 were 23 used to determine the required sample density corresponding to 75 percent relative density.

The static shear strength of 75 percent relative density backfill was evaluated by a program of triaxial compression testing. These data are given in Figure 20 B-39 and summarized in Figure 20-6.

Strain-dependent dynamic elastic moduli data for 75 percent

relative density backfill were generated by cyclic triaxial testing. The dynamic stress-strain data are given in 4

Figures 20 B-35 and 20 B-36, and summarized in Figure 20-7.

Prior to backfill construction, and af ter specific borrow l areas f or structural backf ill have been identified, a 25 ,

testing program will be implemented to verify that the

, engineering properties of backfill discussed above are

appropriate.

2.5.4.5.5 Quality Control Program l 23 Field quality control tests on compacted backfill and the 25 frequency of testing will conform to ANSI /ASME N45.2.5-1978.

2.5-28a Amendment 25

S/HNP-PSAR 7/2/82 k/)

I All field densities will be compared with the maximum l23 s density determined by ASTM D 1557 or ASTM D 2049, whichever 124 results in the higher density, and from that and the comparison between relative density and relative compaction, 23 verification that the relative density of the fill complies with the criteria in Section 2.5.4.5.3 will be obtained.

2.5.4.5.6 Control of Groundwater Because of the great depth to the water table in relation to the depths of required excavation, dry conditions will exist in all excavations.

2.5.4.5.7 Category I Piping and Electrical Duct Banks Category I piping and electrical duct banks will be supported by backfill f or structures, fill f or rough grade, or native material to an elevation of one foot below the bottom of the pipe or duct bank. The material compaction requirements will be in accordance with Sections 2.5.4.5.2 24 and 2.5.4.5.3 respectively. Category I piping and electrical duct banks will not be founded in near-surface

,O loose silty aeolian sands. The quality control program will conform to Section 2.5.4.5.5.

Bedding material will be placed a minimum distance of one f oot below the bottom, one f oot above the top, and five f eet on each side of the pipe or duct bank. The bedding material will be in accordance with Section 2.5.4.5.2, except that 100 percent of the material will pass a one-half inch screen with no more than five percent passing the No. 200 size screen. Compaction of the bedding material will meet the requirements of Section 2.5.4.5.3.

l l Piping will be designed to accommodate the anticipated dif f erential settlements resulting f rom dif f erent supporting 25 l soils (Missoula sand, structural backfill and pre-Missoula sand).

2.5.4.6 Groundwater Conditions i

Groundwater conditions at the Site are described in Sections 2.4.13 and 2.5.4.5.6 and in Appendix 2P.

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l 2.5-29 Amendment 25

S/HNP-PSAR 12/21/81 2.5.4.6.1 Stability The depth of the water table ensures that groundwater will not impact the stability of the safety related facilities (Sections 2.5.4, 2.5.4.1.5, 2.5.4.7, 2.5.4.8 and 2.5.4.10).

2.5.4.6.2 Control of Water Levels and Seepage Because of the great depth of the water table, all structures will be located above the water table and there will be no special requirement for the collection and control of seepage or for the control of groundwater levels.

2.5.4.6.3 Construction Dewatering There will be no requirement to dewater construction excavations (Section 2.5.4.5.6).

2.5.4.6.4 Pe rmeability Permeability values are discussed in Section 2.4.13.1 and in Appendix 2P.

23 2.5.4.6.5 Groundwater Fluctuations Past and projected fluctuations in groundwater conditions are discussed in Sections 2.4.13.1 and 2.4.13.2.

2.5.4.6.6 Monitoring of Wells and Piezometers Monitoring of local wells and piezometers is considered in Section 2.4.13.1 and 2.4.13.2 and in Appendix 2P.

2.5.4.6.7 Direction of Groundwater Flow Direction of groundwater flow is discussed in Sections 2.4.13.1 and 2.4.13.2 and in Appendix 2P.

O 2.5-30 Amendment 23

S/HNP-PSAR 12/21/81 2.5.4.9 Earthquake Design Basis O(,s The maximum ground acceleration corresponding to the Safe Shutdown Earthquake (SSE) and the Operating Basis Earthquake (OBE) for the S/HNP Site have been derived as 0.25 g and 0.125 g respectively, as described in Section 2.5.2.6 and 2.5.2.7. The Design SSE for the S/HNP Plant has been taken as 0.35 g in accordance with Sections 2.5.6.7 and 3.7. The stability of subsurface foundation materials at the Site was therefore analyzed (Sections 2.5.4.7 and 2.5.4.8) using the specified Design SSE maximum ground acceleration of 0.35 g.

2.5.4.10 Static Stability Bearing capacities, settlements and lateral earth pressures for the S/HNP structures, together with the corresponding design methods employed, are presented in Appendix 2Q (Sections 8.2.1, 8.2.2 and 8.3).

Because of the very dense nature of the Site soils and the proposed use of large mat foundations for all the ma]or s tru c ture s , allowable bearing pressure based on failure considerations is not the controlling design factor. Allow-able total and dif ferential structure settlements will O

y,j determine permissible bearing pressures. Foundation settle-ments, evaluated on the basis of the elastic profile shown on Figure 2.5-18, are given on Figure 2.5-19. Foundation 23 loads and elevations for the major structures are also shown on Figure 2.5-19. The estimated settlements are expected to be within the design limits for the Plant structures and operating equipment. The elastic modulus profile (Figure 2.5-18) used to estimate foundation settlements is also summarized in Table 2.5-6. Field and laboratory data used to determine the soil modulus profile are presented in Appendix 20 (Sections 4.0 and 5.0).

Design lateral earth pressures on subsurface structural walls are shown on Figure 2.5-20. The basis for these recommendations is given in Appendix 2Q (Section 8.3). Due to the depth of the water table in the Site vicinity, sub-surface walls will not be sub]ected to hydrostatic loading.

2.5.4.11 Design Criteria Design criteria, methods, references, factors of safety, test data and computer models used in stability studies of all safety-related facilities are given in Appendix 20 and l O 2.5-33 Amendment 23 l

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S/HNP-PSAR 7/2/82 in Sections 2.5.4.1.5, 2.5.4.2, 2.5.4.6, 2.5.4.7, 2.5.4.8 and 2.5.4.10.

2.5.4.12 Techniques to Improve Subsurface Conditions There will be no special techniques employed to improve subsurface conditions at the S/HNP Site.

23 2.5.4.13 Subsurface Instrumentation Subsurface instrumentation for the surveillance of settle-ment of safety-related structures at the Plant will consist of the installation of settlement monitoring points on the f ollowing f acilities: Auxiliary / Fuel / Control / Reactor Building, Turbine Building, Diesel Generator Building, Radwaste Building and Ultimate Heat Sink.

The settlement monitoring points will be installed at the time the f oundation mat of each of these structures is constructed.

Settlement observations will be made and recorded monthly at each of the buildings identified above until at least 90 per-cent of their structural dead load is in place. Thereaf ter 25 and until the end of construction, the frequency of observa-tions may be reduced, but in no case will it be less fre-quent than once every three months. Prior to start of opera-tion, a program of settlement readings and their frequency during operation will be established.

2.5.5 STABILITY OF SLOPES There are no natural or man-made slopes at the Site, the failure of which could adversely impact the S/HNP Plant.

Therefore, Se,tions 2.5.5.1 through 2.5.5.4 do not apply.

23 2.5.6 EMBANKMENTS AND DAMS There are no embankments or dams at the S/HNP Site for flood protection or for impounding cooling water required for the operation of the nuclear power plant. Therefore, Sections 2.5.6.1 through 2.5.6.10 do not apply.

O 2.5-34 Amendment 25

S/HNP-PSAR 7/2/82

. 14. NUREG-0570, " Toxic Vapor Concentrations in the Control Room following a Postulated Accidental Release,"

O- James Wing (June 1979) .

15. Threshold Limit values for Chemical Substances and Physical Agents in the Workroom with Intended Changes for 1981," ACGIB (1981).
16. Fast Flux Test Facility (FFTF) FSAR, HEDL-TI-75001, Amendment 629 (September 1, 1978). 23
17. Personal Communication to T. Grebel, NESCO, from K. Bracken, DOE, " Hazardous Chemicals on the Hanford Reservation" (November 12, 1981).
18. Personal Communication to T. Grebel, NESCO, from J. P. Chasse, WPPSS, (December 10, 1981).
19. Memorandum f rom J. P. Chasse, WPPSS, to File dated September 9, 1981.
20. Memorandum from J. P. Chasse, WPPSS, to File dated November 25, 1981.
21. Arthur D. Little, Inc., A Model Economic and Safety Analysis of the Transportation of Hazardous Substances in Bulk, report prepared for the U.S. Department of fs ,e m

Commerce, Maritime Administration, Of fice of Domestic Shipping, Washington, DC. Report No. COM-74-11271 25 (1974).

22. Personal Communication, including data, to Duane Mathiowetz, Bechtel Power Corporation, from David Joss, Wilson-Hill Associates, DOT (May 3, 1982).
23. W. F. Hartman, C. A. Davidson and J. T. Foley, Statistical Description of Heavy Truck Accidents on Representative Segments of Interstate Highway, SAND-0409 (January 1977) .
24. Letter from Robert V. Myers, Puget Power, to Harold R. Denton dated July 2, 1982.

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2.6-3 Amendment 25

S/HNP-PSAR 12/21/81 References for Section 2.3

1. W. A. Stone et al., Climatography of the Hanford Area, BNWL-1605, Pacific Northwest Laboratory (June 197 2) .

(In some cases data have been updated through 1975 or 1980).

2. NRC, Onsite Meteorological Programs, Regulatory Guide 1.23, U.S. Nuclear Regulatory Commission (February 1972).
3. NRC, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Regulatory Guide 1.145, U.S. Nuclear Regulatory Commission (August 1979) .
4. NRC, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled-Reactors, Regulatory Guide 23 1.111, Rev. 1, U.S. Nuclear Regulatory Commission (July 1977) .
5. NUS, Joint Frequency Distributions of Atmospheric Stability and Wind for the Skagit/Hanford Nuclear Proiect, NUS-3855, NUS Corporation, Rockville, Maryland 20850 (August 1981).
6. J. F. Sagendorf and J. T. Goll, XOOD00 Procram for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, NUREG-0324 (Draft), U.S.

Nuclear Regulatory Commission (September 1977) .

7. NRC, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Regulatory Guide 1.70, Rev. 3, U.S. Nuclear Regulatory Commission l

(November 1978) .

l 2.6-4 Amendment 23

S/HNP-PSAR 7/2/82 1970. In accordance with recommendations given in Seed and O Whitman, 1970, f or backfill with an angle of f riction equal to 35 degrees, the dynamic f orce is approximately equal to the inertia force on a soil wedge which extends behind the wall a distance equal to three-quarters of the height of the wall. The dyntmic force resultant is generally considered to act somewhat above the mid-height of the wall, but a uniform pressure distribution (i.e., resultant force at mid-height of wall) may be employed to represent the dynamic pressure increment. The dynamic pressure distribution for the design basis earthquake is summarized on Figure 20-12.

For large massive structures (such as the auxiliary / fuel /

control / reactor building) which may move out of phase with the ground motions, dynamic pressures on subsurface walls should be established from a soil / structure interaction analysis. There are, however, no subsurface structural walls associated with the auxiliary / fuel / control / reactor building. For structures with significant embedment, such as the Ultimate Heat Sink, dynamic pressures on subsurface walls will be established by soil / structure interaction 25 analysis. The effect of horizontally propagating surface waves (such as Rayleigh waves) will be considered if significant in establishing these pressures.

8.4 EARTHWORK q 8.4.1 EXCAVATIONS With the exception of the upper 6 to 8 feet of loose silty fine sand at the Site, the in-place soils will provide suit-able foundation materials for all the major structures.

Theref ore, there is no requirement to overexcavate beyond planned foundation depths and replace with structural backfill, except where the proposed founding elevations lie above the competent soils. In-place materials which will provide suitable f ounding materials are readily identified as clean black medium sands or, at specific locations, interbedded clean black medium sands and gray silty fine sands. The overlying dark yellowish brown to gray silty fine sands must be removed beneath all structural founda-tions.

Although exact foundation elevations are not currently available, the approximate foundation grades in relation to the in-place soils for all the major structures are shown on Figure 20-5. The deepest f oundations are those of the ultimate heat sinks and radwaste buildings, and these structures will be founded on the very dense sands at or below elevation 490 feet. All other structures will be founded within the overlying medium dense to dense sands, or on structural fill or backfill. The diesel generator buildings, being located adjacent to the more deeply founded auxiliary / fuel / control / reactor buildings, will be founded 20-27 Amendment 25

S/HNP-PSAR 12/21/81 largely on structural backfill. In addition, there will be a requirement at the Unit 2 site, in particular, to remove the surficial material and replace with structural backfill to the diesel generator building foundation grade. It will probably also be necessary to overexcavate and backfill beneath the eastern sections of the turbine buildings (particularly at Unit 2 location), depending upon exact [

founding elevations.

All temporary excavation slopes should be cut no steeper than 1.0 vertical on 1.5 horizontal. Although the light ,

cementation within the sands permits the excavation of steeper slopes, such slopes will slough within a short period of time. The toes of all slopes should be set back from the structures sufficiently to provide adequate working  !

space for backfilling and compacting with heavy equipment.

Chemical or other types of slope protection materials should be applied as necessary to prevent wind erosion, and prepared foundation soils should be protected with mud mats.

Because of the very deep water table, dry conditions will exist in all excavations.

8.4.2 FILL AND BACKFILL All Category I structures, with the exception of the diesel generator buildings, will be founded on the existing in-place soils. Of the central plant structures, only the diesel generator buildings (Category I structures) and portions of the turbine buildings (non-Category I struc-tures) will be founded on structural backfill. Backfill will also be placed around all of the structures to Site grade (tentatively established as elevation 526 feet) .

During excavation, material suitable for structural backfill should be stockpiled and all other material separately stored for use as non-structural area fill. The surficial zone of loose silty fine sand which occurs across the Site will not be suitable as structural fill or backfill. In addition, the upper zone of the underlying material is very silty in parts (interbedded gray silty fine sand and black clean medium sand) and this material should not be used as structural backfill. The black c1(an medium sands (Missoula sands), which occur beneath the upper silty materials and above the very dense Pre-Missoula sediments, will provide suitable structural backfill material and should be appropriately stockpiled during excavation of the deeper i

foundations. The maximum particle size for fill and backfill materials should not be larger than 3 inches. The presence of larger particles may interfere with obtaining 9

. 20-28 Amendment 23

S/HNP-PSAR 7/2/82

("N Subsystems are assumed rigid and their masses lumped into

( the supporting structural system whenever significant coupling between the primary (supporting) system and the secondary (supported) system does not occur. The decoupled subsystems are later analyzed using the response spectra generated at the supporting levels.

3.7.2.4 Soil-Structure Interaction The input motion, as given in Section 3.7.1, is defined at the surface level in the free field. Because the presence of the Plant structures modifies this motion, a soil-structure interaction analysis will be performed.

The major Category I structures (the Auxiliary, Fuel, Control, and Reactor Buildings) will be constructed on a common basemat, approximately 20 ft thick. The soil- 23 structure interaction analysis for these structures will be done with a combined model, using the " lumped parameter" approach. (This approach is also known as the "sub-structure," the " foundation impedance," and the "multistep" approach.) The decision to use this approach is based on the shallow embedment, relative to horizontal size, of the common basemat.

The analysis will consist of the following steps:

1) A free-field soil column analysis is performed, using the input acceleration time history defined in Section 3.7.1 as the surface control motion.

Strain dependency of stiffness and damping will be considered, using an iterative equivalent linear method. The shear strain - shear stiffness rela- 25 tionships will be those shown in Figure 2.5-16.

2) The soil impedance functions are calculated, using the soil stiffness and damping derived from the free-field analysis. Soil layering will be explicitly considered. 23
3) The base and structural responses are calculated using substructuring techniques.

Soil parameter variations will be accounted for by multiple analyses. The soil stiffnesses used in step 2 will be multi-plied by 1.33 to provide upper bound impedance functions, and divided by 1.5 to provide lower bound impedance func-tions. These stiffness variation factors will also be used 25 f or soil-structure interaction analyses using the finite element method. Significant differences from these ranges

h of parameters will be justified and submitted to the NRC f or concurrence.

Q 3.7-5 Amendment 25

S/HNP-PSAR 7/2/82 A finite element analysis will be performed as a confir-mation of the lumped-parameter approach. The analysis will 23 utilize the FLUSH program or other program approved by the H220.6

, NRC. The size of the soil elements and locations of the i transmitting lateral boundary and the rigid base will be chosen in accordance with standard accepted procedures for 24 that program. The structure will be modelled by beam and plane strain elements, with sufficient detail to simulate the dynamic propecties of the system.

The Ultimate Heat Sink, which has significant embedment, will be analyzed using a finite element model. In addition to calculating structural responses, this model will be used 25 to establish the dynamic soil pressures on the walls.

Horizontally propagating surface waves, such as Rayleigh waves, if significant, will be considered in establishing these pressures.

3.7.2.5 Development of Floor Response Spectra Floor response spectra will be developed using time histories of significant support points within the structures. The effects of the three components of ground motion will be combined as recommended in Regulatory Guide 1.122, Section C, Paragraph 2 or 3. Widening of spectral curves is described in Section 3.7.2.9 of this PSAR.

3.7.2.6 Three Components of Earthquake Motion When the maximum response to the three components of earthquake motion have been calculated separately, the maximum response to the total motion will be taken as the 23 square root of the sum of the squares of the component responses.

When three time histories are applied to a model simul-taneously, the maximum responses will be taken as the maximum of the algebraic combinations of the responses to the three components.

3.7.2.7 Combination of Modal Responses Where the response spectrum method of analysis is used, the modal responses will be combined by the " grouping method" described in Section C, Paragraph 1.2.1 of Regulatory Guide 1.92.

3.7-6 Amendment 25

S/HNP-PSAR 7/2/82 3.7.2.8 Interaction of Non-Category I Structures with Category I Structures Non-Category I structures whose collapse could result in the 23 loss of required f unction of Category I structures, equip-ment or systems required f or saf e shutdown af ter an earthquake will be analytically checked to determine that they will not collapse when subjected to a Saf e Shutdown Earthquake.

3.7-6a Amendment 25

S/HNP-PSAR 7/2/82 3.7.2.9 Effects of Parameter Variations on Floor Response C* Spectra The effects of parameter variations on floor response spectra shall be considered by widening the spectra, using the following procedure: 23 Let fj be the structural frequency, which is determined by using the most probable material and section properties in formulating the structural model. The variation in the structural frequency is determined by evaluating the individual frequency due to the most probable variation in each parameter that is of significant effect, such as soil modulus, material density, material stiffness, etc. The total frequency variation, 16fj, is then determined by taking the square root of the sum of squares of a minimum variation of 0.05fj and the individual frequency variation (afj)n, that is:

Afj = ((0.05fj)2 + {(afj)n 2 )l/2 (5-1) n A value of 0.lfj is used if the actually computed value of ofj is less than 0.10fj.

23 3.7.2.10 Use of Constant Vertical Static Factors Constant vertical static factors will not be used for Category I structures.

3.7.2.11 Methods Used to Account for Torsional Effects Generally Category I structures with low eccentricity, such as containment, will be analyzed using 2-dimensional stick models. A static factor will be used to account for this eccentricity in design. H220.7 Structures with significant 3-dimensional properties will be i modelled using finite element models which will account explicitly for torsion.

l To account for accidental torsion in both of the above cases l an additional eccentricity, applied statically, of 15% of I 25 l the maximum building dimension at the level under

consideration shall be assumed for structural design. H220.7 3.7-7 Amendment 25

S/HMP-PSAR 12/21/81 3.7.2.12 Comparison of Responses When different analysis methods are used, a comparison of the responses will be provided for the operating license review.

3.7.2.13 Methods for Seismic Analysis of Category I Dams This project has no Category I dams.

3.7.2.14 Determination of Category I Structure Overturning Moments Overturning moments will be determined using the results of the dynamic analyses. Three components of input motion will be included, as well as a conservative evaluation of vertical and lateral seismic forces.

3.7.2.15 Analysis Procedure for Damping 23 For cases where modal analysis is used, one of three techniques may be used to account for damping in different elements of the models: mass weighting, stiffness we ig h ting , or dissipating energy technique. These techniques will produce composite modal damping values.

Where modal analysis is used for a soil-structure system, Tsai's method (Ref 5) may be esed.

For cases where complex response (frequency domain) analysis is used, damping is considered by forming a complex-valued stiffness matrix.

3.7.2.16 Seismic Analysis of Radwaste Building A modified seismic analysis will be used for the foundation and walls of the Radwaste Building, at least up to a height sufficient to contain the liquid inventory in the building.

This modified analysis will comply with Regulatory Guide 1.143, Revision 1, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants."

O 3.7-8 Amendment 23

S/HNP-PSAR 7/2/82 If thermal stresses due to To and R o are present, O the following combinations are also used:

S = D + L + Ro + To (3.8-13)

H220.19 S=D+Lo + Eo + Ro + To (3.8-14)

S = D + L + W + Ro + To (3.8-15)

No increase in allowable stress is permitted for load combinations ( .8-13), (3.8-14) and (3.8-15),

except as indicated}below.

If the thermal stresses due to To and R o are secondary and self relieving, the value of S may be increased by 50 percent.

The cases of L having its full value or being completely absent are both checked,

b. If plastic design me'thods are used, the following load combinations are considered:

Y = 1.7D + 1.7L (3.8-16)

Y = 1.7D + 1.7Lo + 1.7Eo (3.8-17) H220.19 Y = 1.7D + 1.7L + 1.7W (3.8-18)

The cases of L having its full value or being completely absent are both checked.

If thermal stresses due to To and Ro are present, the following combinations are also to be satisfied:

Y = 1.3(D + L + To+R) o (3.8-19)

H220.19 Y = 1.3(D + Lo+Eo + To + Ro) (3.8-20) 25 Y = 1.3(D + L + W + To + Ro) (3.8-21) 3.8.6.3.2 Load Combinations for Factored Load Conditions l

The following load combinations are considered:

a. If working stress design methods are used, the 23 applicable load combinations are:

(

3.8-49 Amendment 25

S/HNP-PSRR 4/2/82 1.6S = D + Lo + To + Ro +

(Ess or Wt or V) ( 3. 8-2 2) 1.6S = D + L + Ta + Ra + Pa (3.8-23) 1.6S = D + Lo + Ta + Ra + Pa + H220.19 (Yr + Yj + Ym) + Eo (3.8-24) 1.7S = D + Lo + Ta + Ra + Pa+

(Yr + Yj + Ym) + (Ess Of Wt Of Y)

( 3. 8-2 5)

b. If plastic design methods are used, the applicable load combinations are:

Y = D + Lo + To + Ro+ (E ss Of Wt or V) (3. 8-2 6)

Y = D + L + Ta + Ra + 1.5 Pa (3.8-27)

H220.19 Y=D+Lo + Ta + Ra + 1.25 Pa

+ (Yr + Yj + Ym) + 1.25 Eo (3.8-28) f Y=D+Lo + Ta + Ra+Pa+ (Yr +

Yj + Ym) + (Es s or Wt or V) (3.8-29)

In combinations ( 3. 8-2 2) to (3. 8-29) , thermal loads can be neglected when it can be shown that they are secondary and self-limiting in nature and where the material being designed for is ductile.

In combinations (3.8-27), to account for the effect of SRV H220.15 loads on containment internals, the load fctor for L shall be increased to 1.25.

In combinations (3.8-23) through ( 3. 8-2 5) and (3.8-27) H220.14 through ( 3. 8-29) , the maximum effects of Pa, Ta, Rae Yja Y and Y m are used unless a time-history analysis is performeba to justify otherwise.

For combinations (3.8-22) through ( 3. 8- 2 9) strains due to T a H220.19 and the dynamic effects of Wt (tornado missile impact) , Par Y r, Yj, and Ym may exceed the allowables provided there will be no loss of function of any safety-related system.

Whenever strains are permitted to exceed yield due to a certain type of load, the structure is checked to satisfy that its ability to carry other loads is not jeopardized.

3.8-50 Amendment 24

S/HNP-PSAR 7/2/82 6.4 HABITABILITY SYSTEMS b[N The habitability systems which provide for occupation of the control room during the design basis accident include:

a. Control room shielding
b. Provision for food and water storage
c. Kitchen and sanitary facilities
d. Control Room Heating, Ventilating, and Air-Conditioning (HVAC) System
e. Lighting System
f. Fire Protection System
g. Area Radiation Monitoring 6.4.1 HABITABILITY SYSTEMS FUNCTIONAL DESIGN The design bases, system design, design evaluation, tests, and inspections and instrumentation requirements for the systems which comprise the habitability systems are provided in other chapters as follows:
a. Control Room Shielding - Chapter 12, Section 12.1.2

/T b. Area Radiation Monitoring - Chapter 12,Section V c.

12.1.4 ,

Control Room HVAC System - Chapter 9, Section 9.4

d. Lighting System - Chapter 9, Section 9.5.3
e. Fire Protection System - Chapter 9, Section 9.5.1.

This chapter also covers the requirement of emergency safety breathing apparatus,

f. Radiological Analysis - Chapter 15 25 Medical and first aid facilities will be provided for immediate emergency use in accordance with the requirements of OSHA 1910.151 " Medical Services and First Aid".

Inf ormation on f ood and water storage, kitchen facilities, and sanitary facilities will not be available until the time of construction.

Information on sleeping accommodations will be provided in the FSAR.

Periodic pressurization testing will be provided for the control room if such tests are required during operation.

Protective clothing and eye protection will be available t 20 control room operators in the event final dose calculations at the OL review stage indicate the necessity.

V 6.4-1 Amendment 25

S/HNP-PSAR 7/2/82 As indicated in Sections 15.4.9, 15.6.4, 15.6.5, 15.7.4 and Appendix 15A, the radiological effects on the control room personnel that could exist as a consequence of the 23 postulated design basis accidents for the control room design described in Section 9.4 do not exceed the guidelines set by 10 CFR 50, Appendix A, General Design Criterion 19.

As described in Section 2.2.3, anhydrous ammonia and truck shipments of chlorine could pose a hazard to control room operators if transported near the S/HNP Site. As the 25 analysis presented in Section 2.2.3 indicates, there are no truck shipments of chlorine or rail shipments of anhydrous ammonia and the probability of unacceptable consequences resulting from a truck accident involving anhydrous ammonia is acceptably low.

O O

6.4-2 Amendment 25

i EXHRUST ISOLRTION VALVE v008A (TYPICAL FOR V008E, V007F, V007B)

RETURN-EXH. FRN IVK-701R RUNNING _

O _

V RETURN-EXH. FRN 1VK-7019 RUNNING JiS 143A " STANDBY" ENERGIZE MANUAL iSOLAT10N SWITCH SOLEN 010 -

O & SV-V0080

'-- RADIATION 150. SIGNRL FIRE 150. SIGNRL MOTOR OVERLORO LOSS OF POWER _

S/IINP-PSAR ,J/2482 OPEN f s

  • VG08A Z /

@,\

PUGET SOUND POWER & LIGHT COMPANY SKAGlii HANFORD NUCLEAR PROJECT PRELIMINARY SAFETY ANALYSIS REPORT CONTROL ROOM HVAC LOGIC DIAGRAM FIGURE 7.34 (1 OF 11)

AMENDENr 25 v%

w

(

IVS 701 A RECIRC. EXH. AND MAKE UP RIR DAMPERS (TYPICAL FOR B) 1VS 701A RUNNING ENE FIRE ISOLATION SIGNRL = SOL RRO ISOLRTION SIGNAL MANUAL ISOLATION SWITCH pp HS 143R " STANDBY" _

&w S/IINP-PSAR 7/2/82 NORM STAN08Y (MAINT) (MAINT) l CONTROL ROOM HS 143 A LGIZE MODULATES IN010 ---* TD 143A2 B43Al TD 143R3 043R2 ISOLRTION VRLVES, FAN IVK 702 A PUGET SOUND POWER & LIGHT COMPANY l

SKAGIT i HANFORD NUCLEAR PROJECT PRELIMINARY SAFETY ANALYSIS REPORT CONTROL ROOM HVAC LOGIC DIAGR AM FIGURE 7.34 (3 OF 11)

NENEME2FP 75 .

i STRNDBY FILTRATION UNIT FAN IVK- 746A (TYPICAL FOR 8)

_HS-K746A ' START' o _

RAD 150 SIGNAL d FIRE 150 SIGNAL _

HS-K746A 'HUTO' ~

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FSL-126B FLOW ' LOW' RUNNING HS 143A " STANDBY" (SH 3 0F 11) _

_HS-K746 A 'STOP' _

O .

NOTOR OVERLORD d _

LOSS OF POWER _

h

,HANUAL ISOLATION SWlTCH

a S/HN P-PSAR 772/82 STOP RUTO START CLOSE AUTO (MAINT) (SRTA) (MAINT) h CONTROL ROOM CONTROL ROOM HS K746A HS v009A TYPICAL OF HS V010A (15 V 0t1 A

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l PUGET SOUND POWER & UGHT COMPANY SKAGIT I HANFORD NUCLEAR PROJECT ]

PRELIMINARY SAFETY ANALYSIS REPORT CONTROL ROOM HVAC LOGIC DIAGRAM FIGURE 7.34 (9 OF 11)

AMENDMENT 25 6-

, S/HNP-PSAR 7/2/82 permit recirculation of the control room by an A/C

, unit and a return / exhaust f an and filtration of a portion of the air through the standby filtration unit (s). After the' fire has been extinguished, the Control Room HVAC System can be manually changed to the purge mode.

The control room can also be completely isolated 25 by manual operator action.

9.4.1.1.3 Design Evaluation The concentration of radioactivity, which will be assumed to surround the control room after the postulated accident, will be evaluated as a function of the fission product decay constants, containment leak rate, and the meteorology for each period of interest. The assessment of the amount of radioactivity within the control room takes into consid-eration the flow rate through the control room outside air intake duct, and the effectiveness of the standby filtra-tion unit.

Control room shielding design, discussed in Chapter 12, is based on the fission product release to the Containment

/T caused by the design basis LOCA as evaluated in accordance

( ,) with Regulatory Guide 1.3 in Chapter 15. Shielding is provided to ensure that radiation exposures of the control room personnel for the duration of the accident are within the limits specified by 10 CFR 50, Appendix A, Criterion 19.

Redundant radiation monitors will be provided in the outside air intake duct of the control room central A/C l 23 units. Upon detection of a high radiation signal by the monitors, an alarm will be annunciated in the control room, and the control room central A/C unit (s) will be isolated from its source of outside air supply, and the Control Room HVAC System will be automatically transferred to the standby mode of operation. Transfer of the system to the standby mode also may be initiated manually from the control room upon detection of high radiation by an area radiation monitor located within the control room.

The control room standby filtration unit will draw the incoming air through the high efficiency filters, upstream l 23 HEPA filters, carbon adsorbers, and downstream HEPA filters to minimize the exposure of control room personnel to

(~)

s_.

l 9.4-7 Amendment 25

? <

1 1

S/HNP-PSAR 12/21/81 airborne radioactivity in accordance with 10 CFR 20 require-ments. A portion of the control room air can be recircu-9l lated continuously through the filter train for further removal of airborne radioactive particulates from the control room atmosphere. Operation of the standby filtra-tion unit reduces the likelihood that outside air will enter the control room via paths other than through the standby filtration train. The resulting calculated doses for control room ingress, egress, and occupancy will not exceed 5 rem to the whole body or its equivalent to any part of the body as specified in the NRC General Design Criterion 19. A detailed discussion of the dose levels in the control room under standby operation is presented in Chapter 15.

Procedures will be provided for proper use of immediately-available breathing apparatus by the emergency crew. A minimum six-hour supply of bottled air for the emergency crew will be readily available on-Site to allow sufficient time for off-Site delivery of-bottled air for several hundred hours of consumption.

Noncombustible construction and heat and flame-resistant materials will be used throughout the Plant to minimize the likelihood of fire and consequential fouling of the control room atmosphere with smoke or norious vapors. Smoke detectors will be provided in each outside air inlet duct and areas of the control room to detect smoke or noxious vapors in the control room. In the event that detectable smoke or noxious vapors er.ist in the outside air inlet duct, an alarm will be annunciated in the contrel room and 310.22 the HVAC System will be automatically transferred to the l standby mode of operation. If detectable smoke or noxious j vapors exist in the control coom and clearing of the l control room atmosphere should be required, che Control l Room HVAC System, operated in the purge mode, will remove smoke or noxious vapor from the control room at the rate of approximately 15 air changes per hour.

l The Control Room HVAC equipment, ductvork (except the utility exhaust fans and their associated ductwork), and surrounding structures will be of Seismic Category I design. All components of the system will be operable during a loss of normal power, by connection to the Engi-neered Safety Features huses. Redundant components are provided wherever necessary, to ensure that any single f ailure will not preclude adequate control, room ventila-tion, air cleanup, and pressurization. The redundant unit will be automatically started on failure of the operating 23 unit. The Control Room HVAC System failure analysis is presented in Table 9.4-2. -

f l

9.4-8 Amendment 23

I S/HNP-PSAR 12/21/81 The Control Room KVAC System will provide the control room

[}

\ ,/

with an environment of controlled temperature and humidity as shown in Table 9.4-1 for the safety of the personnel and operability of the components.

9.4.1.1.4 Tests and Inspections The HVAC System will be periodically inspected to assure that all normally-operating equipment is functioning properly. Redundant components will be periodically tested to ensure the system availability.

HEPA filters will be manufactured and tested prior to installation in accordance with MIL-F-51068-1970, as modified by AEC Health and Safety Information Issue 306-1971. After installation, HEPA filters will be tested in accordance with ANSI N101.1-1972, Efficiency Testing of Air-Cleaning Systems Containing Devices for Removal of Parti-cles.

Impregnated, activated carbon will be batch-tested by the manufacturer prior to loading into the adsorber bed.

Acceptance criteria will be similar to those described in Sample Specification A-2 of ORNL-NSIC-65-1970. The adsorber will be tested prior to operation and periodically thereafter to verify less than 0.05-percent bypass.

Low efficiency and high efficiency filters will be tested l23 in accordance with the ASHRAE Standards 52-68, Method of Testing Air Cleaning Devices Used in General Ventilation for Removing Particulate Matter.

Fans will be tested in accordance with standards of the Air Moving and Conditioning Association, Inc. (AMCA) 210-1967, Tests Methods for Air Moving Devices. Ductwork will be tested in accordance with the low velocity Duct Construc-tion Standards of the Sheet Metal and Air-Conditioning Contractor's National Association (SMACNA) for leakage during installation and during operation. All systems will be tested and balanced to provide required design air and

. water quantities within a tolerance of plus or minus ten percent.

The amount of outside air required for control room pressur-ization during the standby mode of operation will be determined and verified at the Plant Site. Outside air 310.21 supply flow rate will be adjusted while room pressure with respect to outside is measured. The outside pressure sensing point will be located on the top of the Control Building and the sensing tip will be so constructed as to minimize wind effects.

(

9.4-9 Amendment 23

S/HNP-PSAR 7/2/82 Cooling coils will be tested and rated in accordance with the Air-Conditioning & Refrigeration Institute (ARI)

Standard 410-1972 for Forced-Circulation Air-Croling and Air-Heating Coils.

9.4.1.1.5 Instrumentation Application Automatic temperature, pressure, and humidity controls will be provided for the HVAC system to maintain the space design temperature, pressure, and relative humidity.

Indications will be provided in the control room for the operating status of all fans. Should an operating fan fail, the resultant loss of air flow will actuate an alarm, l 23 automatically start the standby f an, and reposition the dampers to direct the air through the standby fan and associated components.

Instrumentation will be provided for each prefilter, HEPA filter, and carbon adsorber to monitor the corresponding pressure drop locally. Excessive pressure drop will initiate an alarm in the control room.

Redundant radiation monitors will be provided in the outside air intake to monitor radiation levels and to initiate an alarm locally and in the control room and automatically transfer the HVAC System to the standby mode 23 of operation.

Redundant smoke detectors will be provided in each outside l 25 air inlet duct. Actuation of any of these detectors will 23 initiate an alarm in the control room and automatically transfer the HVAC System to the standby mode of operation. l 25 Temperature sensors will be provided in the standby filtra-tion units to indicate carbon temperature of the unit.

Excessive heat will initiate an alarm in the control room 23 and Technical Support Center.

9.4.1.2 Control Building The Control Building Heating, Ventilating, and Air Condi-tioning (HVAC) System will provide a controlled environment for the safety and comfort of the personnel and the oper-ability of the Control Building components. This section will cover the Control Building exclusive of the control 23 room and the Technical Support Center.

0 9.4-10 Amendment 25

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ANALYSIS REPORT

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  • j ' ' T. PalD FIGURE 9.41 AMEN MENT 25

S/HNP-PSAR 12/21/81

(' for the shielding calculations for this system. The shielding will be based on the reactor steam N-16 activities 331.5 in Table 11 1 4 (251 NSSS GESSAR).

12.1.3.8 Fuel Building 12.1.3 8.1 Spent Fuel Transfer and Storage The primary sources in the Spent Fuel Transfer and Storage areas are the spent fuel elements. The spent fuel element sources are discussed in 251 NSSS GESSAR Section 6 12.1 3 2.4.

The isotopic composition of spent fuel in Ci/ watt is given by Table 12.1-20 for 0 decay time. Fuel is transferred af ter 2 days' decay. The average power per assembly is 4.52 MUT . Two assemblies may be present in the transfer tube 331.17 simultaneously. Normally, one-third of the total core of 848 assemblies will be replaced during a refueling oper-ation. The volume of an assembly is 6 8126 x 10 4cc.

("'S 12.1 3.8.2 Fuel Pool Cooling and Cleanup (FPCC) System V:

6 The following equipment will be potential radiation sources due to radioisotopes which leak from the spent fuel and radioisotopes which dif f use from the reactor vessel into the spent f uel pool and are subsequently pumped through the FPCC System:

a FPCC heat exchangers

b. FPCC pumps
c. Associated valves and piping.

The FPCC filter-demineralizers will be located in the Radwaste Building.

The specific activity of the fuel pool water is assumed to be that of seven day old reactor water diluted to a total isotopic concentration of 1.25 x 10-3 Ci/cc. The basis for this assumption is discussed in Section 12.1.2.4.4. The specific emission spectrum for this source is given in Table 331.17 12.1-21. The emission spectrum was obtained based on data presented in Ref 2. The volume of water in the fuel pool is 1 estimated at 75,000 ft3 The isotopic inventory of the

'% f uel pool filter is give n in Table 12.1-22.

, O I

12.1-27 Amendment 23

S/HNP-PSAR 7/2/82 12.1.3.9 Turbine Shine Dose The N-16 present in the reactor steam in the primary steam lines, turbines, and moisture separators can contribute to the Exclusion Area Boundary dose as a result of the high 23 energy gammas which it emits as it decays.

Turbine shine doses are calculated using the SKYSHINE computer program described in Table 12.1-3. Point sources 331'3 are used to represent the ecmponents on the turbine deck.

Table 12.1-15 provides the estimated N-16 inventories of equipment in the Turbine Building. The equipment and piping located above the main turbine deck were included in the turbine shine dose calculation. These are:

a. "4
b. The high pressure turbine
c. A portion of the crossunder piping (100 ft)
d. The moisture separator / reheaters
e. The crossover piping
f. The low pressure turbines The estimated inventory of N-16 is 195 C1. After adjusting for self absorption in the components, the equivalent inventory was f ound to be 117 Ci of N-16. The sources are surrounded by 24'-6" high walls on the north, south, and east and a 31'-0" high wall on the west. The center of the 23 mid-LP turbine is 60'-10" from the east wall and 50'-0" from the north wall. The area enclosed by the walls is 100'-0" in the north-south direction and 204' in the east-west direction.

The expected turbine shine dose at the Exclusion Area 23 Boundary (EAB), which is approximately 1.9 miles from the turbine building, is conservatively estimated to be less than 0.5 mrem /yr. This is the most appropriate point to estimate the dose potentially incurred by members of the general public as a result of the operation of S/HNP because it is closer than the Wye Barricade (approximately 2 miles). H471.2 The Wye Barricade is an access control point of the Hanford Reservation and, in conjunction with other Hanford Reserva-tion controls, serves to prohibit residences or long-term transients from the vicinity of the S/HNP. For this reason occupancy by the public of any point closer than 1.9 miles is expected to be negligible. Nevertheless, for calcula-tional purposes a conservatively high occupancy factor of 5%

may be assuined f or points closer than 1.9 miles. Under such 25 circumstances, the highest expected turbine shine dose at the site boundary (restricted area boundary) is conserva-tively estimated to be 2.5 mrem /yr, based on two unit opera-tion and an availability of 80%.

12.1-28 Amendment 25

S/HNP-PSAR 7/2/82

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) The dose rates due to radiation scattered by air and the walls surrounding the components on the turbine deck were evaluated at several locations within the Turbine Building 331.32 but outside the shield walls. The maximum cur.ulative contribution from these sources of exposure was found to be less than 1.0 mrem /hr. The shield walls will be designed to maintain the Zone II maximum of 2.5 mrem /hr.

12.1.3.10 Field Run Pipe Routing The procedures for routing of field run piping are discussed in Section 12.1.2.3.2.

12.1.4 AREA RADIATION MONITORING See Appendix 1A, Section lA.1 and PSAR Table 12.1-4. 16 12.1.5 OPERATING PROCEDURES

(NI t

The health physics program and access control described in Section 12.3 will ensure that Plant personnel exposures are

' kept as low as practicable during Plant operation and main-l l tenance.

Operating experience of other BWR Plants will be continually evaluated to determine radiation levels present. Any high l

levels in areas not previously considered will be noted.

Doses received by Plant personnel will also be noted.

Procedures will be developed to ensure personnel exposure to l radiation is maintained ALAP in accordance with Regulatory 6 Guide 8.8. A description of these procedures will be provided in the FSAR.

l 12.1.6 ESTIMATES OF EXPOSURE 331.6 12.1.6.1 Anticipated Doses The peak external doses within each area in the Plant will be considered as the maximum dose for which the area is zoned (Section 12.1.2.1). These doses are not expected to occur during normal operation because the Plant shielding is based on maximum coolant activities while the average isotopic concentrations will be considerably less than the maximum. The highest dose rates will occur in Zone V areas s such as inside the drywell, in the turbine-condenser area, and in rooms containing equipment and piping handling highly radioactive fluids.

l 12.1-28a Amendment 25 l

1

S/HNP-PSAR 7/2/82 and in rooms containing equipment and piping handling highly radioactive fluids.

The direct radiation doses to the control room will be less than the zoned maximum dose rate of 0.5 mr/hr because of the 2'-0" thick concrete wall surrounding the control room. The H471.1 Control Building roof is also 2'-0" thick concrete. Total shielding thickness above the control room including the ceiling, is 2'-9" of concrete. The areas of other buildings adjacent to the control room are designated as Zone II, so the dose rate in them will not exceed 2.5 mr/hr. 23 Conservatively assuming this dose rate is due solely to N-16 gamma radiation, the 2'-0" thick concrete wall will reduce the radiation level inside the control room to less than 0.05 mr/hr. Hence, the dose will be well within the design basis of 0.5 mr/hr.

The annual dose to the construction workers employed in Unit 2 while Unit 1 is in operation has been estimated for various points in the Unit 2 construction area; the locations of these dose points are shown on Figure 12.1-17.

Radiation dose to construction workers will be due mostly to 23 the N-16 sources present in the operating Unit 1 turbine system. Dose due to airborne ef fluents f rom Unit 1 will be 331.8 small in comparison with the dose from N-16 sources. The results of these estimated annual doses at ground level are listed in Table 12.1-30.

The doses from turbine shine were calculated in the manner described in Section 12.1.3.9 based on 2,000 hr per yr. The 331.34 resultant dose includes the direct, as well as air scattered, contribution. No credit was taken for the shielding which will be afforded by the partially erected Unit 2 structures. The radioactive wastes will be processed l 10 and stored in the Radwaste Building where shielding will be provided to insure that the dose outside the building will be less than 0.5 mrem /br. With an allowance for distance 331.8 between the Radwaste Building and the Unit 2 construction area, the estimated direct shine dose will be less than 0.01 mrem /hr.

The exposure for Unit 2 construction workers has been estimated based on the following assumptions:

a. The current schedule will be met. 331.36
b. Airborne doses are small compared to turbine shine doses.
c. Doses to personnel in the Unit 2 structures are 10 negligible once the exterior walls and slabs have been fully erected.

O 12.1-28b Amendment 25

S/HNP-PSAR 7/2/82 1

d. Manual laborers spend 80% of their time inside the s/ Unit 2 building and 20% of their time in yard j areas. The contractor's nonmanual personnel spend 70% of their time in the field office and the 331.36 remainder in Unit 2 structures. Construction l management personnel spend 80% of their time in the field office and the remainder in Unit 2 structures,
e. The average dose rate in the yard areas is the average of the dose rates at points A, C, D and E of Figure 12.1-17, or .075 mrem /hr. The average 23 dose rate in the field office is .05 mrem /hr.
f. The availability factor for Unit 1 is 80%.

Exposure to personnel in various categories and locations is summarized in Table 12.1-25 which gives the total estimated exposure to Unit 2 23 construction workers as 258 man-rem.

Exposure data for several operating reactors and for several categories of Plant personnel are listed in Table 12.1-2.

Part A of Table 12.1-2 gives total annual man-rem exposures which indicate, upon examination, that most of the total personnel exposure is received during maintenance, gs g

refueling, and inspection activities. It is expected that the personnel exposures in this Plant will be below those found in Table 12.1-2, Part B.

12.1.6.2 Estimate of Exposure f or Plant Personnel 12.1.6.2.1 General An estimate of the annual exposure that could be received by 331.6 Plant personnel during routine operations and expected main-tenance is made to verify that, under even the worst expected conditions, the exposure will not exceed 1.25 rem / calendar quarter (5 rem / year). Emergency operations and maintenance are excluded fro,a the estimate because the 1.25 rem / calendar quarter can be exceeded under such exceptional conditions.

For purposes of the estimate, routine operations and expected maintenance are defined as:

a. Routine patrol
b. Periodic tests, operations, and jobs (including planned repairing taking place more than once a O year) 12.1-28c Amendment 25

S/HNP-PSAR 7/2/82

c. Control room operations
d. Refueling.

The assumptions used in the estimate are as follows:

a. Units 1 and 2 are operated by three shif ts and f our crews consisting of the following personnel: two Shif t Supervisors, two assistant Shif t Supervisors, two Reactor Operators, five Plant Operators, and one Radiation Protection Technician. The specific function of these persons is discussed in detail in Section 13.1.2.2. The five Plant operators are assumed to be divided as follows: two operators in each unit who patrol Turbine Building, general yard areas and Auxiliary / Containment / Fuel Buildings, and one operator who patrols the Radwaste Building.

Each Plant Operator patrols his assigned area.

General yard areas are divided between the two units because of the distance involved in patrolling the total area.

b. Reactor Operators are not involved in patrols.
c. All mechanical maintenance is performed by the mechan- 331.6 ical maintenance crew. A maintenance crew of two is assumed for all expected maintenance work, so that l exposures for all expected maintenance are shared among the f our two-man crews available f rom the eight personnel in the mechanical maintenance group.
d. All electrical maintenance is performed by the electrical maintenance crew. A maintenance crew i

of two is assumed for all expected electrical l maintenance, so that the yearly electrical main-tenance exposure will be shared among the f our two-man crews available.

e. Radiation Protection Technicians conduct periodic radiation surveys and accompany Reactor Operators and maintenance crews in entries into radiation areas.
f. Maximum dose rates in the various Plant areas are j shown in Figures 12.1-2 through 12.1-15.

The values given in the results are estimated; actual exposures will be different. However, with the exception of work done in Zone V areas with unshielded conditions, reasonable expected maximum dose rates have been assumed for all operations, and the actual exposures are expected to be less than those estimated below.

O 12.1-28d Amendment 25

S/HNP-PSAR 7/2/82 QUESTION 420.1 Summarize changes to the instrumentation and control systems which will be made as a result of the site relocation and confirm that these changes do not involve changes to the safety-related design bases or criteria from that previously submitted by the applicant in the Skagit/Hanford PSAR through Amendment 22. Also confirm that the instrumentation and control system changes resulting from the site relocation do not depend upon advancements in technology beyond the state of the art used for the instrumentation and control systems previous-ly submitted in the PSAR through Amendment 22.

If there are changes to the safety-related design bases or criteria or advancements in technology beyond the state of the art used for the instrumentation and control systems discussed up through Amendment 22 of the PSAR, the changes should be itemized and a discussion provided for each to justify that, with these changes, the General Design Criteria contained in Table 7-1 of the Standard Review Plan (NUREG-0800 ) and IEEE-279 can be met.

RESPONSE

PSAR Amendment 23 proposed anhydrous ammonia detection in

,} the control room air intake, and complete and automatic

,,/ isolation of the control room upon detection of anhydrous ammonia.

25 PSAR Amendment 25 submitted an anhydrous ammonia hazard analysis which indicated that anhydrous ammonia detection is not required and deleted the anhydrous ammonia detection and automatic isolation of the control room.

Other than the anhydrous ammonia detection which was proposed in Amendment 23 and deleted in Amendment 25, there are no changes to the instrument and control systems as a result of the site relocation.

l O

Amendment 25

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