ML20065M628
| ML20065M628 | |
| Person / Time | |
|---|---|
| Site: | Skagit |
| Issue date: | 10/08/1982 |
| From: | PUGET SOUND POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20065M625 | List: |
| References | |
| NUDOCS 8210210339 | |
| Download: ML20065M628 (101) | |
Text
-_
S/HNP-PSAR 10/8/82 File this instruction sheet in the front of Volume 1 as a record of changes.
The following information and check list are furnished as a guide for the insertion of new sheets for Amendment 27 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 1 Figure 1.2-1 Figure 1.2-1 Table 1.7-1 Sht 7 of 26/
Table 1.7-1 Sht 7 of 26/
Table 1.7-1 Sht 8 of 26 Table 1.7-1 Sht 8 of 26
(
APPENDIX 1A 1A-3/lA-4 lA-3/lA-4 Table 15.1.36-2/
Blank / Table 3.2.1 (cont'd)
Table 3.2.1 (cont'd) 15.1.36-8/15.1.36-9 l
CHAPTER 2 l
l 2.1-1/2.1-2 2.1-1/2.1-2 2.1-3/2.1-4 2.1-3/2.1-4 l
2.1-5/2.1-6 2.1-5/2.1-6 l
Figure 2.1-2 Figure 2.1-2 Figure 2.1-3 (1 of 2)
Figure 2.1-3 (1 of 2)
Figure 2.1-3 (2 of 2)
Figure 2.1-3 (2 of 2) 2.3-5/2.3-6 2.3-5/2.3-6 Table 2.3-2/ Table 2.3-3 Table 2.3-2/ Table 2.3-3 i
l Sht 1 of 2 Sht 1 of 2 Figure 2.4-3 Figure 2.4-3 i
Figure 2.4-17 Figure 2.4-17 Figure 2.4-20 Figure 2.4-20 G
1 Amendment 27 8210210339 821008 PDR ADOCK 05000S22 D
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S/HNP-PSAR 10/8/82
'O Discard Old Sheet Insert New Sheet (Front /Back)
(Front /Back)
Table 15.6-1/ Table 's,6-2 Table 15.6-1/ Table 15.6-2 Table 15.6-3/ Table 15.6-4 Table 15.6-3/ Table 15.6-4 Table 15.6-7/ Table 1s.6-8 Table 15.6-7/ Table 15.6-8 Table 15.6-15/ Table _5.6-16 Table 15.6-15/ Table 15.6-16 through Table 15.6-23/
through Table 15.6-23/
Table 15.6-24 Table 15.6-24 Table 15.6-35/ Table 15.6-36 Table 15.6-35/ Table 15.6-36 Figure 15.6-1 through Figure 15.6-1 through Figure 15.6-4 Figure 15.6-4 Table 15.7-3/ Table 15.7-4 Table 15.7-3/ Table 15.7-4 Table 15.7-5/ Table 15.7-6 Table 15.7-5/ Table 15.7-6 Table 15.7-7/ Table 15.7-8 Table 15.7-7/ Table 15.7-8 Table 15.7-ll/ Table 15.7-12 Table 15.7-ll/ Table 15.7-12 Table 15.7-15/ Table 15.7-16 Table 15.7-15/ Table 15.7-16 Table 15.7-17/ Table 15.7-18 Table 15.7-17/ Table 15.7-18 Table 15.7-19/ Table 15.7-20 Table 15.7-19/ Table 15.7-20 O
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ALTERNATE SOUTH ACCESS ROAD l
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/HNP-PSAR 10/8/82 I'
% r-1 FUEL BUILDING 2 DIESEL BUILDING 3 CONTROL BUILDING 4 SWITCHGEAR BUILDING y
5 TURBINE BUILDING O
6 AUXlLIARY BUILDING P
7 REACTOR BUILDING 8 REFUELING WATER STORAGE TANK 9 CONDENSATE STORAGE TANK s
+O 11 RADWASTE BUILDING 12 SOUTH GUARD STATION (below service bidg) 4Te 1S SERVICE BUILDING 14 SHOP AND WAREHOUSE RAILROAD 40 15 WAREHOUSE YARD
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16 SALLY PORT 17 NORTH GUARD STATION 18 WATER TREATMENT BUILDING
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19 LOW VOLUME WASTE POND 20 TRAINING FACILITY (future) p
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21 CONSTRUCTION OFFICE 22 CONSTRUCTION WAREHOUSE
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23 PARKING
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24 CONTROL HOUSE 14 l UNIT NO.2 25 UNIT NO.1525 KV LINE OPTION A 5
26 UNIT NO.1525 KV LINE OPTION B ll 27 UNIT NO.2 525 KV LINE 28 SUBSTATION l
29 PERCOLATION POND 30 SEWAGE TREATMENT PLANT
~'f 31 COOLING TOWERS
"'l 32 BRIDGE (topowerblock)
_ _ _,I 33 RADWASTE BUILDitjG (future-if required) 34 DIESEL FUEL STORAGE TANKS (underground) o soo 200 sco ooe NORTH N
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PUGET SOUND POWER & LIGHT COMPANY SKAGIT 1 HANFORD NUCLEAR PROJECT
-t-PRELIMINARY SAFETY PREFERREDSOUTH ACCESS ROAD ANALYSIS REPORT PROJECT STRUCTURES AND FACILITIES SITE PLAN - UNITS 1 AND 2 FIGURE 1.21 Amendment 27
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4 TABLE 1.7-1 Sheet 7 of 26 Term Definition Ref erence b.
Engineered saf ety f eature system ( ESFS) consists of those systems, i ncluding essential support systems or components thereof the primary purpose of which during a design basis accident (DBA) wi!! be tos (1) Retain f uel temperatures within design limits by maintaining f uel coolant inventory and temperatures within design limits.
1 (2) Maintain f uel temperatures within design limits by inserting ausiliary negative reactivity.
(3) Prevent the escape of radioactive materials to the environment in excess of 10 CFR 100 Ilmits by isolation of the systems or structures.
(4) Reduce the quantity of radioactivity available f or leakage and its potential f or leakage by purif ication, cle anup, containment heat removal UI and containment pressure reduction.
g (5) Control the concentration of combustible gases in the containment fh systems within established limits, g
M Exclusion Area That area within 1 mile of the !!ne joining the reactor centers as defined by 10 27 to CFR 100.3.
>p Al Fa!!ure The termination of the ability of an item to perf orm its required f unction. Fa!!ures may be unannounced and not detected until the next test (unannounced f ailure), or they may be announced and detected by any number of methods at the instant of occurrence (announced failure).
(ii) 23 Faulted Condition Those combinations of conditions associated with estremely-low-probability, postulated (Limiting Faults) events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the estent that considerations cd public health and safety are involved. Such considerations require compliance with saf ety criteria as may be specitled by jurisdictional authorities.
(gg)
Forced Shutdown A f orced shutdown is def ined as an instance where the Plant is shut down and the g,
reactor cooled to cold shutdown conditions as quickly as possible without violating 2
g Technical Specif ications requirements or damaging any equipment. A f orced shutdown is g,
an unscheduled event.
D Ch Fu9ctional Test The manual operation or thittation of a system, subsystem, or component to verif y that g
it f unctions within design tolerances (eg, the manual start of a core spray pump to Fd y
verify that it runs and that it pumps the required volume of water).
(mm)
C3 3
rt General 9esign Criteria A set of design criteria f or structures, systems, and components important to safety.
00 (GDC) which are given in Appendia A to 10 CFR 50, and provide reasonable assurance that the
'N h
Plant can be operated without undue risk to the health and saf ety of the public.
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TABLE 1.7-1 Shtet 8 of 26 Tara Definition Reference HI tup Heatup begins where achieving criticality ends and includes all actions which are normally accomplished in approaching nuclear system rated temperature and pressure by using nuclear power (reactor critical). Heatup extends through warmup and synchronization of the turbine generator.
(mm)
High Radiation Area Any area, accessible to personnel, in which there exists radiation originating in whole or in part within licensed material at such levels that a major portion of the body could receive in any one hour a dose in excess of 100 mrem.
(m)
Hot Functional Testing This testing is performed prior to loading fuel in reactor. The reactor coolant is raised in temperature to no-load temperature using the heat generated by operation of the recirculation pumps. This condition may be maintained for a considerable period of time (possibly 15 days) while various system controls, instrumentation etc., are checked to ensure their proper operation.
Hot Safe Shutdown When reactor is subcritical by an amount greater than or equal to the margin as Condition specified in Technical Specificatien 16.3.10 and Tavg is 1212*F.
(a)
Hot Standby Condition The plant condition in which the coolant temperature is greater than 212*F, system pressure is less than 600 psig, and the node switch is in startup.
The plant condition in which the reactor is sustained at 50-100 percent of rat"d pressure and a low power level with no electric power being generated. Sufficient control rods are withdrawn to maintain the power level required to hold pressure.
If core decay heat is adequate to hold pressure, the reactor may be held below critical but with suf ficient control rods withdrawn to minimize the time required to return to power operation.
(hh)
Ita2diate Immediate means that the required action will be initiated as soon as practicable cons 1dering the safe operation of the unit and the importance of the required action.
(mm)
In:ctive Components Those components whose operability (eg, valve opening or closing, pump operation or t r i g.) are not relied upon to perform the system f unction during the transients or events considered an the respective operating condition categories.
(b)
Incident Any natural or accidental event of infrequent occurrence and its related consequences which affect the Plant operation and require the use of Engineered Safety Feature systems. Such events, which are analyzed independently and are not assumed to occur simultaneously, include the loss-of-coolant accident, steam line ruptures, steam generator tube ruptures, etc.
A system blackout may be an isolated occurrence or may be concurrent with any event requiring Engineered Safety Feature systems use.
(n)
Incident Detection Includes those trip systems which are used to sense the occurrence of an incident.
(mm)
Circuitry Instrument Calibration An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entare anstrument including actuation, alarm, or trip.
(mm)
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031577 1A.2 Offgas System - Sections Section Page 1.0 Introduction 1.2 General Plant Description 1.2.2 Plant Description 1.2.2.11 Radioactive Waste Systems 1.2.2.11.1 Gaseous Radwaste System 1.2-16 1.10 Nuclear Steam Supply System - Balance of Plant Interfaces 1.10.28 off-Gas System 1.10.2E 1.10.28-6 7.0 Instrumentation and Control 7.1 Introduction 7.1.2 Identification of Safety and Power Generation Criteria 7.1-24 7.7 Control Systems 7.7.1 Description 7.7.1.5 Gaseous Radwaste Control System 7.7 Instrumentation and Control 7.7-36 7.7.2 Analysis 7.7.2.6 Gaseous Radwaste Control Systems Instrumentation and Centrols 7.7-43 11.0 Radioactive Waste Management 11.3 Gaseous Effluent Treatment System 11.3 11.3-11 Appendix 11 A Additional Justificatien for Classification of Effluent Treatment System as Group D llA llA-35 12.0 Radwaste Protection l
12.1 Shielding i
12.1.2 Design Description 12.1.2.1.4.2 Gaseous Waste 12.1-lb 12.1.3 Source Terr.s 12.1.3.5.3 Sources in Gas System 12.1-4 15.0 Accident Analysis 15.1 General 15.J.36 Main Condenser Gas Treatment 15.1.36 System Failure 15.1.36-4 l
O 1A-3
S/HNP-PSAR 10/8/82 lA.2 Offgas System - Tables Table Page 3.2.1 Equipment Classifications (Partial) 3.2-15 and 3.2-21 7.7-48 Gaseous Radwaste Process Instrumentation Alarms 7.7-48 11.3.1 Estimated Air Ejector Offgas Release Rates per Unit 11.3-12 11.3.2 Process Data for the Compact Low Temperature Rechar System 11.3-13 11.3.3 Design Data for GE-Supplied Equipment 11.3 11.3-18 11.3.4 Alarmed Process Parameters 11.3-19 11.3.5 Equip Malfunction Analysis 11.3 11.3-23 11.3.6 Doses at 300 Meters from Failure of Offgas Equipment and Piping 11.3-23a 15.1.36-1 Inventory Activities for Offgas Rechar Equipment 15.1.36 16.1.36-8 15.1-36.2 Deleted lA.2 Offgas System - Figures Figure Page 7.7-15a Offgas System Elementary Erawing 7.7-98 7.7-15b Offgas System Elementary Drawing 7.7-99 7.7-15c Offgas System Elementary Drawing 7.7-100 7.7-15d Offgas System Elementary Drawing 7.7-101 7.7-15e Offgas System Elementary Drawing 7.7-102 7.7-15f Offgas System Elementary Drawing 7.7-103 7.7-15g Offgas System Elementary Drawing 7.7-104 7.7-15h Offgas System Elementary Drawing 7.7-105 7.7-15i Offgas System Elementary Drawing 7.7-106 7.7-15j Offgas System Elementary Drawing 7.7-107 11.3-1 Offgas System (Low Temperature) 11.3-24 Process Diagram 11.3-2 Offgas System (Low Temperature) P&ID 11.3 11.3-28 O
1A-4 Amendment 27
I S/HNP-PSAR 10/8/82
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This page intenticnally 27 left blank l
Amendment 27
161.LE 1.2.1 (Lontinued)
Quality' Quality Group Assurance I
Safety Classi-Require-Setemic d
Friocipal Component cleoe toeatIon fteetton inent cetegory Commmente Y
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Offsee System 1.
Tanke Other T
D N/A N/A (v) 2.
Meet anchangere other T
D N/A N/A (v) 3.
Piping Other T
D N/A N/A (v) (m),(q)
Other T
D N/A N/A (v),(q) 4.
Pumpe 5.
Velves, flow control Other T
D N/A N/A (v),(q) 64 Velves, other Other T
D N/A N/A (v),(m),(q) 7.
Mchnical : nodules, with N/A (v),(m) esfety function Other T.A D
N/A (q) 8.
Fressure vessele Other T.A D
N/A N/A (v) 8 S
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S/HNP-PSAR 12/21/81 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1.1 SITE LOCATION AND DESCRIPTION 2.1.1.1 Location The Skagit/Hanford Nuclear Project (S/HNP) Site is located in the southeast area of the U.S.
Department of Energy's (DOE) Hanford Reservation in Benton County, Washington.
The S/HNP Site is approximately 5 miles west of the Washington Public Power Suppl. System's (Supply System)
Nuclear Project No. 2 (WNP-2)
. tit.
It is approximately 8 miles west of the Columbia Riv..r, 7 miles north of the Yakima River at Horn Rapids Da n, and 12 miles northwest of the City of North Richland.
Pigures 2.1-1 and 2.1-2 show the S/HNP location with respect to roads, highways, rivers, and population centers within the Site Region and Site Area.
The following table lists the approximate geographical coordinates for the reactor containment structure centroids:
23 O
Lambert Coordinates l
Latitude Universal (State of and Transverse Washington)
Unit Longitude Mercator (ft) 1 460 29' 15" N 5150900 m N 422710 1190 26' 4"
E 313200 m E 2268390 2
460 29' 15" N 5150900 m N 422710 1190 25' 51" E 313400 m E 2269290 2.1.1.2 Site Area Figure 2.1-2 shows the S/HNP Site and its topographic features, and the location and orientation of the principal Plant structures.
No public roads or railroads cross the Site.
The S/HNP land requirements consist of the Site and Associated Areas.
The major Project facilities will be located on the Site, and other supporting facilities (e.g.,
a 2.1-1 Amendment 23
S/HNP-PS AR 10/8/82 transmission lines, intake and discharge pipeline, railroad and access roads) will be located on the Associated Areas.
The Site and Associated Areas are depicted in Figure 2.1-3 and described as follows:
The Site will consist of 1200 acres.
Title will be acquired to 640 acres (the owned land) and easements will be obtained for the remaining 560 acres (the easement area).
Owned land will be comprised of Section 33 of Township 12 North, Range 27 East of the Willamette Meridian.
The easement area will be the south half of Section 28, the west quarter of Section 34 and the west half of the southwest quarter of Section 27 of Township 12 North, Range 27 East of the Willamette Meridian.
The Associated Area will be made up of the following easements and totaling approximately 420 acres on land 23 outside of the Site:
Estimated Acres Easement Outside Facility Width Site 1.
Intake and discharge 150 feet 134 pipelines (200 feet at pump-house) 2.
Rail road 100 feet 42 3.
Transmission Lines 600 feet 192 4.
Access Roads a.
North 100 feet 19 b.
South
- 100 feet 17 27
- An alternative access route totaling 33 acres, identified as South Alternative Access Road in Figure 2.1-3, is being considered.
Figure 2.1-3 shows the centerlines for the preliminary corridors (each 1,000 feet wide) in which the final respective easement routes will be selected.
A legal 23 description and final area f or each easement will be provided af ter selection of the final routes.
The raw water pumphouse will be located near the west bank of the Columbia River, approximately 75 feet downstream of River Mile 361.5.
O 2.1-2 Amendment 27
S/HNP-PSAR 10/8/82 Figure 2.1-2 shows the Site Boundary lines and the Plant exclusion area boundary.
The Site Boundary, the Plant 23 property lines, and the restricted area boundary are the same.
The S/HNP exclusion area boundary encloses an area within 1 mile of the line joining the reactor centers.
27 2.1.1.3 Boundary for Establishing Effluent Release Limits The boundary for establishing effluent release limits, in conformance with the restricted area as defined by 10 CFR 20, coincides with the Site Boundary (refer to Figure 2.1-2).
Table 2.1-1 lists the minimum distances to the Site Boundary from the effluent release points (center of each containment).
For purposes of radiation protection and general safety, the area inside the Site Boundary will be under the control of Puget.
The Site Boundary will be fenced.
As described in Section 2.1.3, there are no permanent residences or significant numbers of transients within the exclusion area.
Vehicles will be able to access the restricted area via two roads that pass through normally open gates at the Site Boundary.
If it becomes necessary to prohibit vehicle entry, the gates will be closed and monitored by a guard.
O 23 2.1.2 EXCLUSION AREA AUTHORITY AND CONTROL 2.1.2.1 Authority All of the land within the exclusion area is, at present, owned by the United States of America and managed by the Department of Energy as part of the Hanford Reservation.
Puget is currently negotiating with the Department of Energy to acquire the legal rights necessary to use the Site for the Project and those necessary to determine all activities within the exclusion area, as required by 10 CFR 100.3(a).
l Puget expects to acquire title to 640 acres (the owned land) of the 1200 acre Site and to acquire appropriate easements over the remaining 560 acres (the easement area) of the Site.
The owned land, the land being purchased by l
Puget, is Section 33 of the Township 12 North, Range 27 East of the Willamette Meridian.
The easement area is the remainder of the Site described in Section 2.1.1.2.
l v
l 2.1-3 Amendment 27 l
S/HNP-PSAR 10/8/82 Puget's use of the owned land will be restricted to the construction and operation of nuclear electric generating facilities.
Upon completion of the use of the owned land for these purposes, title to the owned land will revert to the Government.
The Government will retain all mineral rights upon or in the owned land, but will agree not to exercise those rights so long as title to the owned land remains vested in Puget.
Except f or the Substation, all S/HNP structures to be located on the Site will be located on the owned land.
The Substation will be located on the easement area.
The easements to be acquired by Puget over the easement area will include an easement for an access-control peri-meter f ence, thus permitting Puget to f ence the Site boundary and control access to the entire Site, as dis-cussed in Section 2.1.1.3.
In conjunction with purchase of the owned land, Puget 23 expects to acquire f rom the Government the authority to determine all activities within the exclusion area consis-tent with the meaning of 10 CFR 100.3(a), including the authority to remove all personnel and property from the area.
Puget will agree to exercise this authority in a manner so as not to preclude the Government from under-taking any action or activity within the exclusion area that is permissible under the provisions of 10 CFR 100.3(a).
The Government will retain all mineral rights upon or in the exclusion area, but any exercise of these rights will be subject to Puget's above described authority to control all activities within the exclusion area.
There are no easements of record within the exclusion area.
2.1.2.2 Control of Activities Unrelated to Plant Operation There are ne activities unrelated to S/HNP operation within l27 the exclusion area.
2.1.2.3 Arrangements f or Traf f ic Control No public roads, railroads, or waterways traverse the 23 exclusion area.
The S/HNP access roads and railroad (Figure 2.1-2) will be located on easements to be granted to Puget by the Government.
Puget will have the authority to control travel on these facilities within the exclusion O
2.1-4 Anendment 27
S/HNP-PSAR 10/8/82 1
area.
In the event that evacuation or other control of the 27 exclusion area should become necessary, appropriate notice will be given to the DOE-Richland Operations Office for control of non-Puget related activities.
2.1.2.4 Abandonment or Relocations of Roads There are no public roads traversing the S/HNP Site.
2.1.3 POPULATION DISTRIBUTION All population estimates and projections were calculated with the centroid of the S/HNP reactors as the geographic reference.
For the analysis of the population within 10 miles, a house count was conducted in October, 1981.
For the estimate of the population between 10 and 50 miles, data from the 1980 U.S. Census were analyzed for blocks, tracts, and enumeration districts (Ref 1).
Population projections from 1990 through 2030 were based on county forecasts for the states of Washington and Oregon 23 (Refs 2, 3).
For the census years 1990 and 2000, existing i ()N projections were directly employed.
For the years 2010,
' (
2020 and 2030, projections were made following a logic similar to that of the U.S. Census projections through 2030 for the nation as a whole (Ref 4).
It was assumed that i
after the year 2000, stabilization of population growth will gradually occur within a 50-mile radius of the Site.
I For each of the census years from 2010 through 2030, it was estimated that the rate of population increase in each county would decline by one-half of the rate prevalent in the previous decade.
For each county within 50 miles, this procedure results in a stabilized population b the year 2030.
Distribution of population growth was assumed to ne equal throughout each county, with the major exceptions of the Benton and Franklin metropolitan counties whdea are nearest to the Site.
Based on interviews with local plannerE and city officials and review of land use and annexation plans, a number of areas within the Tri-Cities were identified as having high growth potential.
These areas include the Horn Rapids Triangle in Richland, the Horn-Willamette area in West Richland and northwest Pasco near the new I-182 bridge.
Accordingly, for the period 1980-2000 appropriate O
2.1-5 Amendment 27
S/HNP-PSAR 12/21/81 enumeration districts and census blocks were projected to grow at approximately twice the rate of the remainder of the metropolitan area.
Several areas were projected to sustain growth from 2000-2010 but after 2010 all areas were projected to stabilize and generally parallel the overall metropolitan area growth rates.
2.1.3.1 Population Within Ten Miles Figure 2.1-4 shows the estimated 1980 population within a 10-mile radius of the Site for each compass sector at distances of 1, 2,
3, 4,
5 and 10 miles.
As these data indicate, there are no residences within 5 miles.
An October, 1981 house count determined the nearest residence to be approximately 7.5 miles from the Site.
Based on average household data for the area, it is estimated that 357 people reside within 10 miles of the Site - all in southerly to easterly directions.
These 357 residents represent about.13 percent of the approximately 280,000 residents in the 50-mile radius.
Projected population within 10 miles of the Site for the census years 1990-2030 are shown in Figures 2.1-5 through 2.1-9.
As these data indicate, projections are that the 23 population within the 10 mile radius will be 513 in 1990, and 639 in 2000, 683 in 2010, and 691 in 2020.
By 2030, the population within ten miles is estimated at 691, which is a 93.5 percent increase over 1980.
No major land use changes are projected for the Hanford Reservation and population growth is expected to be concentrated in areas which actually had residents in 1980.
The projected age distribution of the population at the midpoint of S/HNP operating life (2010) is presented in Table 2.1-2.
These data are calculated using a cohort survival method which utilized the State of Washington county estimates of age and sex distribution in 2000 as the base.
2.1.3.2 Poculation Between 10 and 50 Miles Figure 2.1-10 shows estimates of the number of persons (N =
278,871) residing within the 10-50 mile radius of the Site in 1980.
As these data indicate, the bulk of the population within 30 miles is concentrated in the Tri-Cities metropolitan areas in the SE and SSE directions from the S/HNP.
2.1-6 Amendment 23
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S/HNP-PSAR 10/8/82
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s For all other meteorological conditions (ie, unstable A, B, or C atmospheric stability and/or 10 meter wind speeds of 6 m/s or greater), plume meander was not considered.
The appropriate x/Q value was chosen as the higher value calculated frcm Equation 2.3-1 or 2.3-2.
~
x 23 2.3.4.2 Determination of tonservative y/O Values Cumulative probability distributions of X/Q values were determined for each of the 16 wind sectors for the Exclusion Arce Boundary (EAB) (1609m) and Low Population l 27 Zone (LPZ) (6437m) distances.
The distributions were structured in terms of probabilities (relative to total hours in all sectors) of given x/O values being exceeded in a given sector. The conservative estimate was determined by selecting the X.'Q values which are exceeded not more than 0.5 percent of the time.
The X/Q values thus determined are applicable for release durations less than or equal to two hours. The annual average value was calculated for ground-level release in accordance with methodology described in Regulatory Guide 1.111, Rev. 1 (Ref 4).
Values for periods of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, 3 days (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />),
and 26 days (624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br />) were obtained by a logarithmic
('S interpolation between the 2-hour value and the annual
( j average in the same sactor. The maximum-value sectorsfor each time period becomes the controllin9 X/Q.value.
However, a direction-independent conservative estimate was used as an additional constraint on the controlling ')p4) 23 value for the conservative accident assessmen't. An'cverall 5th percentile x/O was determined from a direction-independent probability distribution.
This overall 5th percentile value was calculated at the EAB and LPZ i
distances and compared to the direction-dependent l'
conservative estimates.
If the overall 5th percentile l
value (for a given time period) was greater than the maximum direction-dependent value, then the direction-independent value would be used for the sccident assessment.
2.3.4.3 Input Meteorological Data l
[
L Input meteorological data consisted of joint frequency l
distributions (JFDs) of hourly averages of wind speed and 1 l
wind direction by stability class. For computer modeling i
purposes, twelve wind speed groups were used to give good l
resolution at lower wind speeds (Ref 5).
The annual-JFD with the standard 7 wind speed groups is shown in Table
\\<
s 2.3-5 gf Amendment 27
[A
S/HNP-PSAR 10/8/82 2.3-1.
The JFD's were based on two years of data collected nearby at WNP-2.
Occurrences of calms and variable wind directions were distributed by direction and stability class to the lowest wind speed group of the JFD's.
Calms were assigned a speed one-half of the threshold speed of the wind vane.
Winds were based on observations at 33 ft and stability class on observations of delta T (245-33 f t) 23 per Regulatory Guide 1.23 (Ref 2).
2.3.4.4 Short Term Dispersion Estimatas The short-term (X/Q) values are presented by accident period in Table 2.3-2 f or the EAB distance of 1 mile and l27 the LPZ distance of 4 miles during the course of a l23 hypothetical accident.
The 0-2 hour value at the EAB is 1.5 x 10-4 sec/m3; the sector associated with this value is l27 to the SSE of the Plant. The sector of maximum x/Qs stays the same f or the duration of the accident (30 days).
2.3.5 LONG-TERM ATMOSPHERIC DISPERSION MODEL 2.3.5.1 Dispersion Model Dispersion factors (x/Q) were determined using the methodology presented in Regulatory Guide 1.111 (Ref 4) and the NRC computer code XOQDOQ (Ref 6).
The calculations were made f or the Site Boundary and at the 23 standard distances discussed in Regulatory Guide 1.70 (Ref 7).
All releases were assumed to be at ground level.
X /Q values were determined by:
2.032
[
nij (2.3.5-1)
( X/0)D
=
ij NIgjU j x
i where
( X/0)D the average effluent concentration, X,
=
normalized by source strength, Q, at a (sec/mb)f or a given downwind distance, x
direction, D O
2.3-6 Amendment 27
S/HNP-PSAR 10/8/82 TABLE 2.3-2 CONSERVATIVE x/Q VALUES FOR SHORT-TERM (ACCIDENT)
ASSESSMENT AT S/HNP 23 Accident Distance Maximum Sector Period (m) y /Q (sec/m3) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1609 (EAB) 1.5E-4(SSE) 27 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6437 (LPZ) 2.lE-5(SSE) 24 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 6437(LPZ) 1.4E-5(SSE) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 6437 (LPZ) 5.7E-6(SSE)
(3 days) 624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br /> 6437 (LPZ)
(26 days)
Notes:
23 1.
Relative concentrations are for a ground-level release O
to a ground-level receptor including credit for plume meander and building wake effects.
2.
Based on WNP-2 meteorological data for the period April 1, 1974, to March 31, 1976:
33-ft wind and delta T (245-33 ft).
l l
l Amendment 27
. =
S/HNP-PSAR 12/21/81 TABLE 2.3-3 Sheet 1 of 2 ANNUAL AVERAGE ATMOSPHERIC DISPERSION AND DEPOSITION PARAMETERS FOR S/HNP Site Boundary:
Unit 1 Chi /Q Chi /Q
- Decayed, D/
Distance Chi /Q Decayed Depleted (mg)
Dir (mete rs)
(sec/m3)
(sec/m3)
(sec/m3)
N 1150.
1.043E-05 1.040E-05 9.308E-06 4.571E-08 NNE 1175.
8.661E-06 8.632E-06 7.722E-06 4.ll2E-08 NE 1095.
7.276E-06 7.252E-06 6.513E-06 3.177E-08 ENE 930.
9.820E-06 9.780E-06 8.876E-06 3.185E-08 E
910.
8.727E-06 8.699E-06 7.900E-06 3.383E-08 ESE 930.
1.504E-05 1.499E-05 1.360E-05 5.545E-08 SE 1095.
1.400E-05 1.396E-05 1.253E-05 5.311E-08 SSE 1290.
1.012E-05 1.007E-05 8.970E-06 2.780E-08 S
1265.
8.321E-06 8.281E-06 7.385E-06 2.202E-08 SSW 1290.
6.341E-06 6.310E-06 5.621E-06 1.626E-08 23 SW 1325.
4.941E-06 4.918E-06 4.373E-06 1.061E-08 WSW 1125.
5.499E-06 5.474E-06 4.913E-06 1.242E-08 W
1100.
4.439E-06 4.423E-06 3.972E-06 9.598E-09 WNW 1120.
5.175E-06 5.148E-06 4.624E-06 1.106E-08 NW 1325.
4.921E-06 4.896E-06 4.355E-06 1.397E-08 NNW 1175.
9.362E-06 9.334E-06 8.348E-06 3.502E-08 NOTES:
1.
Relative concentrations are for a ground-level release l
to a ground-level receptor, are undepleted and unde-cayed, and incorporate Pasquill-Gif ford dispersion coefficients, building height wake, and open terrain correction factors.
l 2.
Based on WNP-2 meteorological data for tae period l
April 1, 1974 to March 31, 1976:
33-ft wind and l
delta T ( 245-33 f t).
3.
Distances are from the center of each Containment.
1 O
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AREAS INUNDATED BY 1-HOUR THUNDERSTORM PMF
__-'""" L -
a - J" FIGURE 2.4-20 Amendment 27
S/HNP-PSAR 10/8/82 Live loads including floor occupancy loads, L
=
laydown loads due to temporary placement of equipment; nuclear fuel and fuel transfer casks, equipment handling loads, lateral 23 earthfill loads, lateral and vertical sur-charge loads due to transport vehicles; pressure differences due to heating, cool-ing and normal atmospheric changes; roof loads due to snow and impounded rainfall up t 27 6" deep; hydrostatic loads due to compartment flooding.
Loads due to Safety Relief Valve pressures as outlined in Appendix 6C of this PSAR are included.
Operating live loads likely to occur during Lo
=
normal operation.
These are the live loads to be used in seismic analysis and with 23 seismic load combinations.
The operating live load (Lo) is a relatively small fraction of the design live load (L) ; Lo does not include such loads as those due to laydown, maintenance, or temporary cranes or moving equipment.
l27 o
Thermal effects and loads during normal T
=
f()
operating or shutdown conditions, based on
/
the most critical transient or steady state condition.
Pipe reactions during normal operating or Ro
=
shutdown conditions, based on the most critical transient or steady state condi-tion.
3.8.6.1.2 Severe Environmental Loads Severe Environmental. loads are those that could infrequently be encountered during the Plant life.
Included in this cat-egory are:
Loads generated by the Operating Basis Eo
=
Earthquake (OBE).
The earthquake is com-posed of two horizontal and one vertical components and the effects of the three components are combined, based on the square root of the sum of the squares.
Only the dead load (D) and the operating.
23 live load (L ) need be considered in a
evaluating the seismic response forces, bU 3.8-43 Amendment 27
S/HNP-PSAR 10/t'/82 Loads generated by the design wind speci-W
=
fied for the Plant.
3.8.6.1.3 Extreme Environmental Loads Extreme environmental loads are those which are credible but are highly improbable.
They include:
Ess =
Loads generated by the Safe Shutdown l
Earthquake (SSE).
The earthquake is com-posed of two horizontal and one vertical components and the effects of the three components are combined, based on the square root of the sum of the squares.
Only the dead load (D) and the operating 23 live load (L ) need to be considered in a
evaluating the seismic response forces.
Roof load due to volcanic ashfall.
V
=
Effects generated by the design tornado Wt
=
specified for the Plant.
They include loads due to the tornado wind pressure and differential pressures, and also the energy resulting from impact of tornado-generated missiles.
p Design-basis winter precipitation resulting P
=
from a combination of 11.7 in of water frem the 48-hr PMP coincident with 3.8 in. of water 27 equivalent from the 100-year snowoack.
(See Section 2. 4. 2. 3. )
3.8.6.1.4 Abnormal Loads Abnormal loads are those loads generated by a postulated high-energy pipe break accident within a building and/or compartment thereof.
Included in this category are the following:
Design Pressure load within or across a Pa
=
compartment and/or building, generated by the postulated pipe rupture, including the dynamic effects due to the pressure time history and pool-swell phenomena as out-12 lined in Appendix 6C of this PSAR.
Thermal effects due to thermal conditions T
=
a generated by the postulated break and including T,
o 3.8-44 Amendment 27
S/IINP-PSAR 10/8/82 O) 3.8.6.2.2 Load Combinations f or Factored Load Conditionn
(
For thene conditions, which represent extreme environmental, abnormal, abnormal / severe environmental, and abnormal /
extreme environmental conditions, respectively, the strength design method in used and the following load combinations are considered:
D + Lo + To + Ro + (Enn Of Wt or V or P )
27 U
=
p D + L + Ta + Ra + 1.5 Pa (3.8-10) 0
=
!!220.19 D + Lo + Ta+Ra + 1.25 Pa+ (Yr+
U
=
+ 1.25 (Eo or Wt or V or P )
27 Yj + Y )
m p
D + Lo + Ta + Ra + Pa + (Yr + Yj + Y )
U
=
m
+ (Enn Of Wt or V or P )
(3.8-12) 27 p
In combinations (3.8-10), (3.8-11) and (3.8-12), the maximum are considered unions ef f ects of Pa, Ta, R Yir Yr, and Ym at
()
a time-history analysis In perf ormed to justif y otherwise.
For combinationn (3.8-9) to (3.8-12), strains due to Ta and 16 due to the dynamic ef f ects of Wt (tornado minnile impact),
24 Par Y Yi, and Ym may exceed the allowable strains, rs provided there will be no lone of f unction of any saf ety-related nyctem.
In combination (3.8-10), to account f or the ef f ect of SRV
!!220*15 loads on containment internals, the load f actor of L shall be increased to 1.25.
Whenever strains are permitted to exceed yield due to a certain type of load, the structure in checked to natisf y that its ability to carry other loads in not jeopardized.
The cenen of L having its f ull value or being completely absent are both checked.
The effects of tornado-generated dif f erential pressuren and 12 minniles are combined in accordance with BC-TOP-3-A (Ref 1).
3.8.6.2.3 Concrete Temperaturen
(}
The limitations listed below are considered applicable only
(,/
to concrete structural components:
3.8-47 Amendment 27 9
l S/HNP-PSAR 12/21/81 O
a.
The following temperature limitations are for normal operation or any other long-term period.
The temperatures are not allowed to exceed 150'F, except for local areas which may be allowed increased temperatures not exceeding 200'F.
b.
The following temperature limitations are for accident or any other short-term period.
The temperatures are not allowed to exceed 350'F for the interior surface.
However, local areas may be allowed to reach 650*F from steam and/or water jets in the event of a pipe failure.
c.
Higher temperatures than given in items a. and b.
may be allowed in concrete, if test data can be provided to evaluate the reduction in strength.
Such a reduction can be applied to the design allowable values.
Also, evidence will be provided which verifies that the increased temperatures do not cause deterioration of concrete, either with or without load.
3.8.6.3 Load Combinations and Acceptance Criteria for Seismic Category I Steel Structures The following presents a set of load combinations and allowable design limits used for Seismic Category I steel structures.
To assure that the structural integrity will be maintained, limits on the resulting stresses and the required strength capacities are considered for service loads and for factored loads.
3.8.6.3.1 Load Combinations for Service Load Conditions Either the working stress design methods of Part 1 of AISC, l23 or the plastic design methods of Part 2 of AISC will be used.
a.
If the working stress design methods are used, the l23 following load combinations are considered:
S=D+L 12 S=D+Lo+Eo S=D+L+W 9
3.8-48 Amendment 23
S/HNP-PSAR 7/2/82 O
s_/
If thermal stresses due to To and Ro are present, 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 (3.8-13), (3.8-14) and (3.8-15),
except as indicated below.
If the thermal stresses due to To and R are o
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 methods 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 f ull value or being completely absent are both checked.
If thermal stresses due to To and R are present, o
the f ollowing combinations are also to be satisfied:
Y = 1.3(D + L + To + Ro)
(3.8-19)
H220.19 Y= 1.3(D + Lo + Eo+To+R)
(3.8-20) o 25 Y = 1.3(D + L + W + To + Ro)
(3.8-21) 3.8.6.3.2 Load Combinations f or Factored Load Conditions f
The following load combinations are considered:
a.
If working stress design methods are used, the 23 applicable load combinations are:
3.8-49 Amendmerit 25
S/HNP-PSAR 10/8/82 1.6S = D + Lo + To + Ro +
(E or Wt or V or P )
(3.8-22) 27 ss p
1.6S = D + L + Ta + Ra + Pa (3.8-23) 1.6S = D + Lo + Ta + Ra + Pa +
H220.19 (Yr + Yj + Ym) + Eo
( 3. 8-2 4,)
1.7S = D + Lo + Ta + Ra + Pa +
(Yr + Yj
+Y) +
m (Ess or Wt or V or P )
(3.8-25) 27 p
b.
If plastic design methods are used, the applicable load combinations are:
Y=D+Lo+To+Ro + (Ess Of Wt or V or P )
(3.8-26) 27 p
Y=D+L+Ta+Ra + 1.5 P (3.8-27) a Y=D+Lo + Ta + Ra + 1.25 Pa H220.1
+ (Yr + Yj + Y ) + 1.25 E (3.8-28) m o
Y=D+Lo+Ta+Ea+Pa+ (Yr+
Yj + Y )
+ (E or Wt or V or P )
(3.8-29) 27 m
ss p
In combinations (3.8-22) 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 f or is ductile.
In combinations (3.8-27), to account for the effect of SRV loads on containment internals, the load f actor f or L shall H220.15 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 eff ects of Pa, Ta, R Yja Y ae rr and Y are used unless a time-history analysis is performed m
to justif y otherwise.
For combinations (3.8-22) through (3.8-29) strains due to T, H220.19 and the dynamic ef f ects of Wt (tornado missile impact), Par Y
Yj, and Ym may exceed the allowables provided there will rr be no loss of f unction of any saf ety-related system.
Whenever strains are permitted to exceed yield due to a certain type of load, the structure is checked to satisf y that its ability to carry other loads is not jeopardized.
3.8-50 Amendment 27
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S/HNP-PSAR 10/8/82 e.
1 1 FUEL BUILDING 2 DIESEL BUILD;NG 3 CONTROL BUILDING 4 SWITCHGEAR BUILDING 5 TURBINE BUILDING 6 AUXILIARY BUILDING 7 REACTOR BUILDING 8 REFUELING WATER STORAGE TANK 9 CONDENSATE STORAGE TANK 10 ULTIMATE HEAT SINK 11 RADWASTE BUILDING 12 SOUTH GUARD STATION (below service bldg) 13 SERVICE BUILDING 14 SHOP AND WAREHOUSE 4
15 WAREHOUSE YARD Sy 16 SALLY PORT g
4e 17 NORTH GUARD STATION Sg 18 WATER TREATMENT BUILDING N40 19 LOW VOLUME WASTE POND l
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22 CONSTRUCTION WAREHOUSE 23 PARKING 24 CONTROL HOUSE 25 UNIT NO.1525 KV UNE OPTION A 26 UNIT NO.1525 KV LINE OPTION B 27 UNIT NO.2 525 KV LINE 28 SUBSTATION 29 PERCOLATION POND 30 SEWAGE TREATMENT PLANT 31 COOLING TOWERS 32 BRfDGE(topowerblock) 33 RADWASTE BUILDh.G(future-if required) 34 DIESEL FUEL STORAGE TANKS (underground) o ioo no ao ear NORTH Q
n f
SCALE l
SEISMIC SEISMIC TORNADO STRUCTURE CATEGORY CATEGORY RESISTANT I
ll Containment X
X
(
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Diesel Building X
X i
Control Building X
X Radweste Building X
Auxiliary Building X
X Fuel Building X
X p
X SKAG lT / HANFORD NUCLEAR PROJECT Condensate Storage Tank Basin X
PRMIMRY SMm j
Diesel Fuel Oil Storage Tanks X
X ANALYSIS REPORT l
Turbine Building X
SS ROAD Administration Building X
Circulating Water Pump House X
LAYOUT OF l
Mechanical Draft Cooling Towers X
PLANT STRUCTURES Raw Water Pump House X
b
" NOTE: ALL EXCEPT SIDING FIGURE 3A-10 l
Amendment 27
S/HNP-PSAR 10/8/82 1
inlets and exhausts on safety-related buildings will be protected by tornado missile barriers (and if necessary louvers) which will preclude any significant snow, ice or dust from blocking the inlets or exhausts, or any sig-nificant snow, water or dust from entering the air systems.
020.24 The Diesel Generator exhaust will be discharged through 27 exhaust stacks which will be designed to preclude any significant amount of rain, ice, snow or dust from entering 2
or blocking them.
Section 9.2.5.3.7 discusses ice protection for the Ultimate Heat Sink Complex.
3.11-5 Amendment 27
- -... ~. -
4 S/HNP-PSAk 10/8/82
'l permit recirculation of the control room by an A/C unit and a return / exhaust fan and filtration of a
\\s 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.
In the event that the FSAR analysis of the S/HNP of f site hazards identifies the requirement f or an automatic detection and isolation system, this 27 system will be provided in accordance with the criteria of the Standard Review Plan Section 9.4 (NUREG-0800).
I 9.4.1.1.3 Design Evaluation The concentration of radioactivity, which will be assumed to surround the control room after the postulated accident, 1
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 ef f ectiveness of the standby filtra-tion unit.
control room shielding design, discussed in Chapter 12, is J
based on the fission product release to the Containment 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 f or 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 23 outside air intake duct of the control room central A/C 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 rocm standby filtration unit will draw the incoming air through the high ef ficiency filters, upstream l 23 HEPA filters, carbon adsorbers, and downstream HEPA filters to minimize the exposure of control room personnel to 9.4-7 Amendment 27
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-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 tne standby filtration train.
The resulting calculated doses for control room ingress, egress, and occupancy will not exceed 5 rem to the wn le 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 et 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 noxious 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 exist in the outside air inlet 310.22 duct, an alarm will be annunciated in the control room and the HVAC System will be automatically transferred to the standby mode of operation.
If detectable smoke or noxious vapors exist in the control room and clearing of the control room atmosphere should be required, the Control 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, ductwork (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 buses.
Redundant components are provided wherever necessary, to ensure that any single failure 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.
9.4-8 Amendment 23
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 l6 12.1.3.2.4.
The isotopic composition of spent fuel in C1/ 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 MWT.
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 cc.
4
/)
12 1 3 8.2 Fuel Pool Cooling and Cleanup (FPCC System V
The following equipment will be potential radiation sources due to radioisotopes which leak from the spent fuel and radioisotopes which diffuse from the reactor vessel into the spent f uel pool and are subsequently pumped through the FPCC System:
a FPCC heat exchangers l
b.
FPCC pumps c.
Associated valves and piping.
The FPCC filter-demineralizers will be located in the Radwaste Building.
l The specific activity of the fuel pool water is assumed to l
be that of seven day old reactor water diluted to a total l
isotopic concentration of 1.25 x 10-3 C1/cc.
The basis l
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 l
presented in Ref 2.
The volume of water in the fuel pool is estimated at 75,000 ft3 The isotopic inventory of the
()
fuel pool f11ter is given in Table 12 1-22.
N,)
l 12.1-27 Amendment 23 i
S/HNP-PSRR 10/8/82 12.1.3.9 Turbine Shine Dose The N-16 present in the reactor steam in the primary steam llnes, 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 331.3 computer program described in Table 12.1-3.
Point sources are used to represent the components 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.
A portion of the main steam piping (40 ft) b.
The high pressure turbine H471.2 c.
A portion of the crossunder piping (100 f t) d.
The moisture separator / reheaters e.
The crossover piping f.
The low pressure turbines The estimated inventory of N-16 is 195 Ci.
After adjusting for self absorption in the components, the equivalent inventory was found to be 117 Ci of N-16.
The sources are surroundad by 24'-6" high walls on the north, south, and 23 east and a 31'-0" high wall on the west.
The center of the 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 Wye Barricade, which is approximately 2 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 the Wye Barricade is an access control point of the Hanford Reservation and, U471.2 in conjunction with other Hanford Reservation 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 about 2 miles is expected to be negligible.
Nevertheless, for calculational purposes a conservatively high occupancy factor of 5% may be assumed 27 for points closer than 2 miles.
Under such circumstances, the highest expected turbine shine dose at the site boundary (restricted area boundary) is conservatively estimated to be 25 2.5 mrem /yr, based on two unit operation and an availability of 80%.
Ol 12.1-28 Amendment 27
S/HNP-PSAR 12/21/81
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Recovery - A period of time beginning when the Plant reaches a safe shutdown condition, and lasting until the Plant is restored as nearly as possible to its pre-emergency condition.
Site - The area controlled by Puget and within the exclusion area boundary as defined in Title 10, Code of Federal Regulations, Part 100.3(a).
Site Area Emergency - An event at the Plant involving actual or potential major failures of key safety-related equipment which might lead to a potential degraded core situation.
Technical Specifications - The limits, operating conditions, 23 and other requirements imposed by the NRC on S/HNP operation.
Technical Support Center - On-Site facility which provides a location for Puget technical support of the reactor command and control functions of the control room.
TLD - Thermoluminescent dosimeters.
Devices used to measure the level of exposure to radiation.
Unusual Event - An event at the Plant which results in no significant release of radioactive material, but which could s.
lead to a potential degradation in the level of safety of m
the Plant.
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v 13A-5 Amendment 23
l S/HNP-PSAR 10/8/82 l
3.0 SITE DESCRIPTION AND EMERGENCY PLANNING ZONES 3.1 SITE DESCRIPTION The Skagit/Hanford Nuclear Project Site is located in the southeast area of the U.S. Department of Energy's (DOE)
Hanford Reservation in Benton County, Washington.
The Site is approximately 5 miles west of the Washington Public Power Supply System's Nuclear Project No. 2 (WNP-2) unit.
It is approximately 8 miles west of the Columbia River, approx-imately 7.5 miles north of the Yakima River at Horn Rapids Dam, and approximately 12 miles northwest of North Richland.
Figures 1 and 2 locate the Site within the region and 23 identify the general location of the Plant Site with respect to roads, highways, rivers, and population centers within the vicinity.
Figure 2 shows the Plant Site, including topographic features, and the location and orientation of principal Plant structures.
No public roads or railroads cross the Site.
The Site boundary lines are shown in Figure 2.
The Site area boundary, the station property lines, and the restricted area boundary are the same.
The Plant exclusion area boundary is shown on Figure 2.
The exclusion area is that area within 1 mile of the line joining the reactor centers.
27 3.2 EMERGENCY PLANNING ZONES i
The S/HNP Emergency Program provides for emergency planning i
within two Emergency Planning Zones (EPZs).
A plume exposure EPZ, of about a 10-mile radius around the Plant, is I
defined for the purpose of planning for public protective l
actions based upon exposure or inhalation of a passing 23 l
radioactive plume released during an accident.
An ingestion exposure EPZ, of about a 50-mile radius around the Plant, is defined for the purpose of planning for public protective actions based upon ingestion of contaminated water or foods.
l The size of the EPZs have been determined in relation to I
local emergency response needs and capabilities as they are affected by demography, topography, land characteristics, access routes and jurisdictional boundaries, t
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Amendmelit 27
S/HNP-PSAR 10/8/82 TABLE 15.2-1 l
TYPE II TRANSIENT OFF-SITE DOSE 23 l
i Dose Effect (mrem)
Distance (m)
Whole Body Skin Thyroid 1609 (EAB)1 4.90 2.88 2.85 x 10-2 27 6437 (LPZ)2 5.32 x 10-1 3.13 x 10-1 3.11 x 10-3 23 lEAB Exclusionary Area Boundary 2LPZ Low Population Zone i
i i
i e
l Amendment 27
I TABLE 15.2-2 TYPE II TRANSIENT ON-SITE EXPOSURES Organ Evaluated Dose Effect (mrem) 23 Whole Body 39 Skin 545 Thyroid 0.1 Lung 0.8 O
O
S/HNP-PSAR 10/8/82 x
TABLE 15. 2-3 TYPE II S/R VALVE TRANSIENT - PARAMETERS TO BE TABULATED FOR POSTULATED ACCIDENT ANALYSES Realistic Conservative (Conserva-(NRC) tive Engineer.
Assumptions ing) Assumotions i
Data and assumptions used to estimate radioactive source f rom postulated accidents A.
Power level NA 4100 Nwt B.
Burn-up NA NA C.
Fuel damaged NA None D.
Release of activity by nuclide NA Sect 12.2.3 E.
Iodine f ractions MA (1) Orga91c NA 0
(2) Elemental NA 1.0 (3) Particulate NA 0
F.
Reactor coolant activity bef ore NA NA 3
the accident 23 11.
Data and assumptions used to estimate NA r
activity released A.
Containment leak rate NA Infinite (a) 8.
Secondary containment leak rate (t/ day)
NA NA C.
Valve movement times NA 15.2.4.2.2 D.
Adsorption and f iltration ef ficiencies MA (1) Organic todine NA 999 O
(2) Elemental iodine NA 996 (3) Particulate iodine NA 994 (4) Particulate fission products MA 994 E.
Recirculation system parameters NA (1) Flow rate NA (2) Mixing efficiency NA NA (3) Filter efficiency NA NA F.
Containment spray parameters (flow rate, MA NA l
drop site, etc.)
l C.
Containment volumes NA NA B.
All other pertinent data and assumptions MA 15.2.4.2.2
!!!. Dispersion Data NA A.
EAB and LP2 distenets (m)
NA 1609/6437 1
8.
X/Q values in sec/m3 NA 2.8 x 10-5/
27 3.0 x 10-6 IV.
Dose Data A.
Method of dose calculation NA Sect 15.2.4.2.2 B.
Dose conversion assumptions MA Sect 15.2.4.2.2 C.
Activity in containment NA Sect 12.2.3 D.
Doses MA Tables 15.2-2 23 and 15.2-1 I
tal l
Applicable 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after S/R valve transient commences l
I i
\\
Amendment 27
- =. - -
._-,.~ -.
S/HNP-PSAR 10/8/82 TABLE 15.2-4 FEEDWATER LINE BREAK ACCIDENT 23 Distance Thyroid Dose (meters)
(rem)
Conservative Analysis 1609 (EAB) 2.40 x 10-4 27 6437 (LPZ) 3.39 x 10-5 23 Realistic Analysis 1609 (EAB) 1.25 x 10-5 27 6437 (LPZ) 1.38 x 10-6 9
O Amendment 27
S/HNP-PSAR 10/8/82 J
TABLE 15. 2-5 FEEDWATER LINE BREAlt ACCIDENT - PARAMETERS TO BE TABULATED FOR POSTULATED ACCIDENT ANALYSES Realistic Conservative (Conservative (NRC)
Engineering)
Assumptions Assaretions Data and assumptions used to estimate radioactive source from postulated accidents A.
Power level 4100 MWt 4100 Mwt B.
Burn-up NA NA C.
Fuel damaged None None D.
Release of activity by nuclide 15.2.8.2.1.2.2 15.2.8.2.2.2.2 E.
Iodine fractions (1) Organic 0
0 (2) Elemental 1
1 (3) Particulate 0
0 F.
Reactor coolant activity before the accident 15.2.8.2.1.1.2 15.2.8.2.2.1.2 23 II.
Data and assumptions used to estimate activity released A.
Containment leak rate (t/ day)
NA NA B.
Secondary containment leak rate (t/ day)
NA NA i
C.
Isolation valve closure time (sec) 30 30 D.
Adsorption and filtration ef ficiencies (1) Organic iodine NA NA (2) Elemental iodine NA NA (3) Particulate iodine NA NA (4) Particulate fission products NA NA E.
Recirculation systes parameters (1) Flow rate NA NA j
(2) Mining efficiency NA NA (3) Filter efficiency NA NA F.
Containment spray parameters (flow rate, MA NA drop size, etc.)
G.
Containment volumes MA NA H.
All other pertinent data and assumptions None None
.II.
Dispersion Data A.
EA8 and LP3 distances (m) 1609/6437 16C9/6437 8.
X/Q values in sec/m3 1.5 x 10-4/
2.8 m 10-5/
27 2.1 x 10-5 3.0 x 10-6
- V.
Dose Data A.
Method of dose calculation Reference 1 Reference 1 B.
Dose conversion assueptions Reference 1 Reference 1 23 C.
Activity in containment NA NA D.
Off-Site Doses Table 15.2-4 Table 15.2-4 Amendment 27
l TABLE 15.4-7 i
CONTROL ROD DROP ACCIDENT i
OFF-SITE DOSES 23 Dose (rem)
Distance (meters)
Whole-Body Thyroid Conservative Analysis i
1609 (EAB) 2.95 x 10-2 1.29 x 10-2 27 tn I
6437 (LPZ) 1.02 x 10-2 1.99 x 10-2
)
23 Realistic Analysis 4
tn 1609 (EAB) 8.23 x 10-7 6.12 x 10-7 27 l
6437 (LPZ) 2.83 x 10-6 7.22 x 10-6 23 l
J t
}
B te Oss gg O
i 3
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et m
w m
4 N
n i
S/HNP-PSAR 12/21/81 1
TABL9 15.4-8 CONTROL ROD DROP ACCIDENT CRUCIAL VARIABLES Realistic Conservative (Conservative (NRC)
Engineering)
Assumptions Assumptions Power, WWt 4100 4100 Fuel Rods Damaged 770 770 Peaking Factor 1.5 1.0 Released from Each Rod, %
- Halogens 50*
0.32
- Noble Gases 100*
1.8 23 Retained in Reactor Water, %
90 97 Valve Shut Time, sec 5.5 Halogen Carryover Fraction 1.0 0.02 Partition Factor in Condenser 100 100 lu b ne Bu dn L ak day b
b0 Gap activity release, gap activity is 10% of the core activity.
t O
Amendment 23
S/HNP-PSAR 10/8/82 10-'
A I
%y
__m d
10-*
2 2
N
~
b
%g 10-'
l I
I I
I I
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37, 1
EAB 2 3
4 5
6 LPZ 7
DISTANCE FROM PLANT, KM PUGET SOUND POWER & UGHT COMPANY SKAGIT / HANFORD NUCLEAR PROJECT PREUMINARY SAFETY ANALYSIS REPORT CONTROL ROD DROP ACCIDENT (REALISTIC CASE)
WHOLE BODY DOSE, REM FIGURE 15.41 Amendment 27
S/HNP-PSAR 10/8/82 l
10-4
~
ko 10-'
ba 10-'
E
~
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Z 1
I I
I I
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1 EAB 2 3
4 5
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DISTANCE FROM PLANT, KM J
PUGET SOUND POWER & LIGHT COMPANY OKAGIT / HANFORD NUCLEAR PROJECT PREUMINARY SAFETY ANALYSIS REPORT i
O CONTROL ROD DROP ACCIDENT (REALISTIC CASE)
THYROID DOSE, REM FIGURE 15.4 2 Amendment 27
S/HNP-PSAR 10/8/82 10' Z
10-'
=
m 6
- %y 10-8h
=
a S
I I
I I
I I
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EAB 2 3
4 5
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DISTANCE FROM PLANT, KM i
l PUGET SOUND POWER & LIGHT COMPANY SKAGliI HANFORD NUCLEAR PROJECT PREUMINARY SAFETY ANALYSIS REPORT CONTROL ROD DROP ACCIDENI' (CONSERVATIVE CASE)
\\
WHOLE BODY DOSE, REM FIGURE 15.4 3 Amendment 27
S/HNP-PSAR 10/8/82 10-*
Z 10-'
Z
~
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y w
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10-8
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DISTANCE FROM PLANT, KM i
i l
PUGET SOUND POWER & LIGHT COMPANY SKAGIT1 HANFORD NUCLEAR PROJECT PREUMINARY SAFETY ANALYSIS REPORT l
i CONTROL ROD DROP ACCIDENT I
(CONSERVATIVE CASE)
THYROID DOSE, REM FIGURE 15.4 4 Amendment 27 i
S/HNP-PSAR 10/8/82 s.
i s
TABLE 15.6-1 TYPE III AND IV S/R VALVE TRANSIENT PARAMETERS 10 BE TABULATED FOR POSTULATED ACCIDENT ANALYSIS Realistic Conservative (Conservative (NRC)
Engineering)
Assamptions Assaarttens I.
Data and assumptions used to estimate radioactive source from postulated accidents A.
Power level NA 4100 MWt B.
Burn-up NA NA C.
Fuel damaged NA Mone D.
Release of activity by nuclide NA Sect. 12.2.3 E.
Iodine fractions NA (1) Organic NA 0
(2) Elemental NA 1.0 23 (3) Particulate NA 0
F.
Reactor coolant activity before the accident NA NA 11.
Data and assumptions used to estimate activity released A.
Containment leak rate (t/ day)
NA Infinite (8) 8.
Secondary conainment leak rate (t/ day)
NA NA C.
Valve movement times NA NA D.
Adsorption and filtration efficiencies
[\\
(1) Organic todine NA 994
(
(2) Elemental lodine NA 994 i
j
\\
(3) Particulate iodine NA 994 (4) Particulate fission products MA 994 i
E.
Recirculation system parameters i
di Flow rate NA NA (2) Mixing ef ficiency NA (3) Filter efficiency NA NA I
F.
Containment spray parameters (flow rate, drop sise, etc.)
NA NA C.
Containment volume s MA NA I-H.
All other pertinent data and assumptions MA 15.6.1.2.1 24
!!!. Dispersion Data
[
A.
EAB and LP1 distancts (m)
NA 1609'6437 27 B.
X/Q values in sec/mJ NA 2.8 x 10-5/
3.0:10-6 IV.
Dose Data A.
Metnod of dose calculation NA See 15.6.1.2.1 B.
Dose conversion assumptions NA Sect. 15.6.1.2.1 C.
Activity in containment NA Sect 12.2.3 23 D.
Doses NA Tables 15.6-2, 15.6-3, 15.6-4
'88 Applicable 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after S/R valve transient commences 1
i 9
t I
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i Amendment 27 e
i L
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_.,.,-...,.,.,.,....r-_.
,v-v.,-
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S/HNP-PSAR 12/21/g1 TABLE 15.6-2 TYPE III TRANSIENT ON-SITE DOSE Organ Evaluated Dose Effect (nrem) 23 Whole Body 43 Skin yyg O
i O
Amendment 23
i 4
i
]
S/HNP-PSAR 12/21/81 i
TABLE 15.6-3 TYPE IV TRANSIENT ON-SITE DOSE 23 l
Organ Evaluated Dose Effect (mrem) l
)
Whole Body 39 Skin 545 Thyroid 0.1 Lung 0.8 f
i I
l 1
i i
e 4
l a
I i
i i
i I
i t
1 l
1 1
i i
l l
t l
l 4
i Amendment 23 I
(
i i
S/HNP-PSAR 10/8/82 TABLE 15.6-4 TYPE IV TRANSIENT OFF-SITE DOSE 23 Dose Effect (mrem)
Distance (m)
Whole Body Skin Thyroid 1609 (EAB) 4.90 2.88 2.85 x 10-2 27 6437 (LPZ) 5.32 x 10-1 3.13 x 10-1 3.11 x 10-3 23 i
1 O
1 O
Amendment 27
S/HNP-PSAR 10/8/82 l
l TABLE 15.6-7 OFF-SITE EXPOSURE (a)
(CONSERVATIVE BASIS) 23 Dose Effect (mrem)
EAB LPZ Exposure Mode (mrem)
(1609m)
( 6 4 3 7m',
27 Whole Body 9.41 x 101 13.00 Skin 5.52 x 101 7.67 Thyroid 5.47 x 10-1 7.62 x 10-2 1
(a) 350,000 pCi/sec off-gas release rate 0 meter effective release height 23 i
l l
l l
1 l
I l
l l
l i
l l
1.
i l
l Amendment 27 L
--- - --~ -- -
TABLE 15.6-8 INSTRUMENT LINF FAILURE ACTIVITY AIRBORNE IN TIIE CONTAINMENT, CURIES (CONSERVATIVE ANALYSIS)
Isotope i "
1 lir 2 lir 8 Ilr 1 Day 4 Days 30 Days I-131 2.21E-2 1.32 2.63 1.32E+1 1.24E+1 9.53 9.53E-1 23 132 2.79E-1 1.25E+1 1.87E+1 1.36E+1 1.17E-1 5.82E-ll 0
133 1.03E-1 6.0 1.16E+1 4.77E+1 2.80FJ1 2.56 2.54E-9 134 5.44E-1 1.54E+1 1.42E+1 4.27E-1 1.34E-6 0
0 135 1.29E-1 6.97 1.26E+1 3.24E+1 6.22 3.71E-3 0
m Total 1.08 4.23E+1 5.97E+1 1.07E+2 4.68E+1 1.21E+1 9.53E-l Dz 7
8 s
e to E.
3 U
c)
\\
tt N
M Gb H
e O
O
I S/HNP-PSAR 10/8/82 TABLE 15.6-15 INSTRUMENT LINE BREAK OFF-SITE DOSES, REM 23
- Distance, Inhalation Thyroid Dose meters rem Conservative Analysis
'27 1609 (EAB) 6.44 x 10-4 l
6437 (LPZ) 9.46 x 10-5 Realistic Analysis 1609 (EAB) 2.46 x 10-7 27 6437 (LPZ) 6.56 x 10-7 23 i
i 1
l a
i j
l Amendment 27 l
TABLE 15.6-16 INSTRUMENT LINE FAILURE CRUCIAL VARIABLES Conservative (NRC)
Realistic Assumptions Assumptions I
First ten minutes 23 Containment vent rate cfm 6 x 103 6 x 103 Containment air volume ft3 1.76 x 106 1.76 x 106 Vent filter efficiency %
0 0
Iodine plateout factor NA 2
m m
II Subsequent five hours Containment leak rate %/ day
.25
.25 4
Enclosure building leak rate %/ day 100 100 m
Recirculation flow cfm 0
0 Recirculation filter efficiency %
0 0
SGTS filter efficiency %
99 99 III Dispersion Data A.
EAB and LPZ distance (m) lbC9/K437 1609/6437 l27 B.
X/O values in sec/m3 Table 15.6-17 Table 15.6-17 23 c.
G
?,
a a
b 5
i5 O
O O
S/HNP-PSAR 10/8/82 TABLE 15.6-17 ATMOSPHERIC DISPERSION FACTORS 3
(X/Q VALUES IN SEC/M )
23 1
- Distance, Time Period Conservative Realistic meters Hr (5% X/Q)
(50% X/0) 1609 (EAB) 0-2 1.5 x 10-4 2.8 x 10-5 27 6437 (LPZ) 0-8 2.1 x 10-5 3.0 x 10-6 8 - 24 1.4 x 10-5 2.4 x 10-6 23 24 - 96 5.7 x 10-6 1.5 x 10-6 96 - 720 1.6 x 10-6 7.0l) 1(j-7 4 3 i
i t
i i
I Amendment 27 l
l
S/HNP-PSAR 12/21/81 TABLE 15.6-18 STEAM LINE BREAK ACCIDENT (REALISTIC ANALYSIS)
ACTIVITY RELEASED FROM THE BREAK (CURIES)
Isotope Activity I-131 3.lE-l I-132 3.5E+0 I-133 2.2E+0 I-134 7.0E+0 I-135 3.5E+0 23 Kr-83m 2.0E-2 Kr-85m 3.3E-2 Kr-85 1.3E-4 Ke-87 1.0E-1 Kr-88 1.0E-l Kr-89 4.5E-l Xe -131m 9.4E-5 Xe-133m 1.6E-3 Xe-133 4.5E-2 Xe-135m 1.3E-l Xe-135 1.2E-1 Xe-137 5.8E-l Xe-138 4.5E-l O
Amendment 23
3 S/HNP-PSAR 12/21/81 TABLE 15.6-19 ST2AM LINE PREAK ACCIDENT (REALISTIC ANALYSIS)
ACTIVITY RELEASED TO THE ENVIRONMENT (CURIES)
Isotope Activity I-131 1.6E-l I-132 1.7E+0 I-133 1.lE+0 I-134 3.5E+0 I-135 1.7E+0 23 KR-83m 2.0E-2 KR-85m 3.3E-2 KR-85 1.3E-4 i
KR-87 1.0E-1 KR-88 1.0E-l KR-89 4.5E-1 XE-131m 9.4E-5 XE-133m 1.6E-3 Xe-133 4.5E-2 Xe-135m 1.3E-1 Xe-135 1.2E-1 Xe-137 5.8E-1 Xe-138 4.5E-l t
a Amendment 23
.n.-
...,,,.,.,.w.-
S/HNP-PSAR 10/8/82 TABLE 15.6-20 STEAM LINE BREAK OUTSIDE CONTAINMENT OFF-SITE DOSES (REALISTIC ANALYSIS) 23
- Distance, Whole Body Thyroid Dose, meters Dose, rem rem 1609 (EAB) 1.19 x 10-4 1.04 x 10-2 27 6437 (LPZ) 1.30 x 10-5 1.13 x 10-3 O
O Amendment 27
S/HNP-PSAR 12/21/81 TABLE 15.6-21 STEAM LINE BREAK ACCIDENT FISSION PRODUCT RELEASE TO ENVIRONMENT CONSERVATIVE (NRC) ANALYSIS Activity Released Isotope (C1)
I-131 1.5E+0 I-132 1.7E+1 I-133 1.lE+1 I-134 3.3E+1 I-135 1.7E+1 23 Kr-83m 5.7E-2 Kr-85m 1.0E-1 Kr-85 3.9E-4 Kr-87 3.lE-1 i
Kr-88 3.lE-1 Kr-89 1.3E+0 l
Xe-131m 3.lE-4 i
Xe-133m 4.8E-3 O
Xe-133 1.3E-1 Xe-135m 3.9E-1 Xe-135 3.6E-1 Xe-137 1.8E+0 Xe-138 1.3E+0 l
l l
l l
Amendment 23
S/HNP-PSAR 10/8/82 TABLE 15.6-22 STEAM LINE BREAK OUTSIDE CONTAINMENT OFF-SITE DOSES (CONSERVATIVE ANALYSIS) 23
- Distance, Whole Body Thyroid Dose, meters Dose, rem rem 1609 (EAB) 5.59 x 10-3 5.38 x 10-1 27 6437 (LPZ) 7.79 x 10-4 7.50 x 10-2 23 O
O Amendment 27
S/HNP-PSAR 12/21/81 TABLE 15.6-23 STEAM LINE BREAK ACCIDENT (CONSERVATIVE CASE)
CONTROL ROOM PERSONNEL DOSES (1), REM Sources Skin (S)
Whole-Body (7)
Thyroid (2)
Direct Shine Insignificant Immersion 23 Dose 1.9E-5 1.5E-6 1.lE-3 Total Dose 1.9E-5 1.5E-6 1.lE-3 (1) 1000 cfm intake flow, 2000 cfm recirculation flow, filter efficiency of 99% for iodine, 10 cfm unfiltered inleakage for all time periods 3
(2)
Breathing rate of 3.47E-4 m /sec for all time periods I
O Amendment 23
S/HNP-PSAR 10/8/82 O
TABLE 15.6-24 STEAM LINE BREAK ACCIDENT = PARAMETERS TO BE TABULATED FOR POSTULATED ACCIDENT ANALYSES Mealistic Conservative (Conservative)
(NPC)
Engineerlag)
Assumptions Assawrtiens 1.
Data and assumptions used to esimate radioactive source from postulated accidents A
Power level, MWt 4100 4100 B.
Burn-up NA MWt C.
luel damaged None NA D.
Release of activity by nuclide E.
Iodine fractions Table 15.6-21 Table 15.6-19 (1) Organic 0
0 (2) Elemental 1
1 (3) Particulate 0
0 F.
Reactor coolant activity before the accident 15.6.5.5.2.2 15.6.5.5.2.2 Data and assumptions used to estimate activity released 23 II.
Cortainment leak rate (t/ day)
NA NA A.
B.
Sec-idary containment leak rate (t/ day)
RA NA C.
Isolation valve closure time (sec) 5 5
D.
Adsorption and filtration efficiencies (11 Organic iodine MA NA (24 Elemenal iodine NA NA (3) Particulate iodine KA NA (4) Particulate fission products NA NA E.
Pecirculation system parameters RA NA (1) Flow rate NA NA f2) Mixing efficiency NA NA
'3)
Filter efficiency NA NA F.
fontainment spray parameters (flow rate.
i drop size. ect)
NA NA G.
d'ontainnment volumes RA NA H.
All other pertinent data and assumpions None Ncne III. Dispersion Data A.
EAB and LPZ distances X/Q valueh in Sec/m3;(m) 1609/6437 1609/6437 27 B.
EAB 1.5 x 10-4 2.8 x 10-5 LP2 2.1 x 10-5 3,o 10-6
- V.
Dose Date, A.
Mett'ad of dose calculation Regulatory Peference 2 Calde 1.5 B.
Dose conversion assumptions Regulatory Reference 2 23 C.
Activity in containment Guide 1.5 NA NA D.
Off-Sate Doses Table 15.6-22 Table 15.6-20 0
Amendment 27
TABLE 15.6-35 LOSS-OF-COOLANT ACCIDENT OFF-SITE DOSES, REM 23 Dose Model l
Assumptions Whole-Body Thyroid i
2-hr EAB 30-Day LPZ 2-hr EAB 30-Day LPZ I
Conservative Case j
(Reg. Guide 1.3) 2.92 1.16 20.6 12.8 m
i s
I Realistic Case 2.3 x 10-7 1.12 x 10-6 3.31 x 10-6 1.94 x 10-6 27 I
Mechanistic g
Fission Product Distribution 3.02 2.65 25.6 34.2 i
i 1
>a (D
3 Q.
H t
O 3
\\
ft co
\\
a M
a2 M
N 1
.m
S/HNP-PSAR 10/8/82 O
TABLE 15.6-36 LOSS-OF-COOL..JT ACCIDENT PARAMETERS TO BE TABULATED
- FOR POSTULATED ACCIDENT ANALYSES Realistic Conservative (Censervative (NRC)
Engineering)
Assumptions Assa etiens I.
Data and assumptions used to estimate radioactive source from postulated accidents A.
Power level 4100 Mwt 4100 Mwt B.
Burn-up NA NA C.
Fuel damaged 100%
None D.
Activity in containment Table 15.6-25 Table 15.6-27 E.
Iodine fractions (1) Organic 44 It (2) Elemental 914 994 (3) Particulate 54 0
F.
Reactor coolant activity before the accident 15.6.5.5.1.2 15.6.5.5.1.2 II.
Data and assumptions used to estimate activity released 23 A.
Containment leak rate (t/ day) 0.25 0.25 B.
Secondary containment leak rate (t/ day)
NA 100 C.
Valve movement times NA NA D.
Adsorption and filtration efficiencies (1) Organic iodine 994 994 (2) Elemental iodine 994 994 (3) Particulate lodine 996 994 (4) Particulate fassion pr od uct s 994 994 E.
Recirculation system parameters (1) Flow rate NA NA (2) Mixing efficiency RA NA (3) Falter efficiency NA NA F.
Containment spray parameters (flow rate, drop size, etc.)
Section 6.2.3 Section 6.2.3 C.
Containment volumes Table 6.2-1 Table 6.2-1 H.
All other pertinent data and assumptions Table 15.6-32 Table 15.6-32 III. Dispersion Data A.
EAB and LPI distances (m) 1609/6437 1609/6437 l27 B.
X/Q values in Sec/m3 Table 15.6-17 Table 15.6-17 l
IV.
Dose Data A.
Method of dose calculation Regulatory Guide 1.3 Reference 3 B.
Dose conversion assumptions Regulatory Guide 1.3 Reference 3 23 C.
Activity in released to the environs Table 15.6-26 Table 15.6-29 D.
Off-Site Doses Figures 15.6-3 & 15.6-4 Figures 15.6-1 & 15.6-2 Table 15.6-35 15.6-35 1
As applicable to ene event bein, describ d.
1 l
O Amendment 27 l
S/HNP-PSAR 10/8/82 10-'
2
%og m
5 10-' _-=
2 ua m
N S
10-'
O 2
~
44 I
I I
l 1
I I
33, 1
EAB 2 3
4 5
6 LPZ 7
DISTANCE FROM PLANT, KM PUGET SOUND POWER & UGHT COMPANY SKAGIT I HANFORD NUCLEAR PROJECT PREUMINARY SAFETY ANALYSIS REPORT O
LOCA REALISTIC CASE WHOLE BODY DOSE, REM FIGURE 15.61 Amendment 27
S/HNP-PSAR 10/8/82 0
10-*
Z 10-*
~ 5
- %y 1
e 10-'
p g
Z J
1 I
I I
I I
I I
10*
1 EAB 2 3
4 5
6 LPZ 7
DISTANCE FROM PLANT, KM i
PUGET SOUND POWER & LIGHT COMPANY SKAGIT / HANFORD NUCLEAR PROJECT PREUMINARY SAFETY ANALYSIS REPORT LOCA REALISTIC CASE INHALATION THYROID DOSE, REM FIGURE 15.6 2 Amendment 27
S/HNP-PSAR 10/8/82 102 Z
10'
=
m
_ 6 3
- %y 10' NFt W
a O
I I
I I
I I
I 3g, 1
EAB 2 3
4 5
6 LPZ 7
DISTANCE FROM PLANT, KM PUGET SOUND POWER & LIGHT COMPANY SKAGIT / HANFORD NUCLEAR PROJIOT PRELIMINARY SAFETY ANALYSIS REPORT LOCA-CONSERVATIVE CASE WHOLE BODY DOSE, REM l
FIGURE 15.6 3 Amendment 27
{
S/HNP-PSAR 10/8/82 1
I I
i 10' 1
i 1 08
- m
- 5 3
- oe x
10'
^
NR l
1.0 1
EAB 2 3
4 5
6 LPZ 7
DISTANCE FROM PLANT, KM PUGET SOUND POWER & LIGHT COMPANY SKAGIT 1 HANFORD NUCLEAR PROJECT PREUMINARY SAFETY ANALYSIS REPORT O
LOCA CONSERVATIVE CASE INHALATION THYROID DOSE, REM FIGURE 15.64 Amendment 27
i.
S/HNP-PSAR 10/8/82 l
TABLE 15.7-3 OFF-SITE DOSE FROM OFFGAS SYSTEM FAILURE l
(Conservative Analysis) 23
f l
l i
i j
Amendment 27 e--
e--m, ce -
e-.
n-,,-,-,-,
nn--,------r.,rw-,
n,~
...------..w,--,
-w
S/HNP-PSAR 10/8/82 TABLE 15.7-4 OFF-SITE DOSE FROM OFFGAS SYSTEM FAILURE (Realistic Analysis) 23 Dose in rem EAB LPZ (1609 m)
(6437 m)
Whole-Body 6.46E-3 7.04E-4 Thyroid 5.41E-5 5.89E-6 27 Bone 3.24E-4 3.52E-5 Lung 1.30E-3 1.42E-4 G.I.
1.63E-2 1.77E-3 O
O Amendment 27
S/HNP-PSAR 10/8/82 m
Tall.E 15.7-5 CASEOUS RADWASTE (YSTEM FAILi~im PARA."ETERS TO BE TABULATED
- FOR 1'STULATED N ANA!.YSES Realistic Conservative (Conservative (NRC)
Engineering)
Assuwrtions Assawrtions Data and assumptions used to estimate radioactive source from postulated accidents A.
Power Level 4100 MWt 4100 Nwt B.
Burn-up NA NA C.
Fuel damaged None None D.
Release of activity by nuclide Table 15.7-1 251 NSSS GESSAR Table 15.1.36-1 and Table 15.7-2 E.
lodine fractions NA (1) Crganic NA 0
(2) Elemental NA 1
23 (3) Particulate NA 0
F.
Rea: tor coolant activity before 15.6.5.5.2.2 15.6.5.5.2.2 the accident gwg
(
\\
f II.
Data and assumptions used to estimate (j
N activity released A.
Containment leak rate (t/ day)
NA NA B.
Secondary containment leak rate (%/ day)
NA NA C.
Valve movement times NA NA D.
Adsorption and filtration efficiencies NA NA (1) Organic iodine MA NA (2) Elemented iodine NA NA (3) Particulate iodine NA NA (4) Particulate fission products NA NA E.
Recirculation system parameters MA NA (1) Flow Rate NA NA (2) Mixing Efficiency NA NA (3) Filter Efficiency NA NA
- r. Containn nt spray parameters (flow rate, drop size, etc)
NA NA C.
Containment volumes NA NA H.
All other pertinent data and assumptions None None
.11.
Dispersion Data A.
EAB and LPZ distances (m) 1609/6437 1609/6437 27 B.
x/C values in sec/m3 1.5E a 10-4/2.1E-5 2.8 x 10-5/3.0E-6
- V.
Dose Data A.
Method of dose calculation Appendix 15A Appendix 15A B.
Dose conversion assumptions Appendix 15A Appendix 15A C.
Activity in Containment NA NA 23 D.
Doses Table 15.7-3 Table 15.7-4 As applicable to the event being described.
\\%
Amendment 27
S/HNP-PSAR 12/21/81 TABLE 15.7-6 FAILURE OF AIR EJECTOR LINES ACTIVITY RELEASED TO ENVIRONMENT (Realistic Case)
Activity Release Isotope (Ci)
I-131 3.2E-3 I-132 3.2E-2 I-133 2.lE-2 I-134 6.7E-2 I-135 3.2E-2 Kr-83M 3.lE+0 23 Kr-85M 5.5E+0 Kr-85 2.lE-2 Kr-87 1.7E+1 Kr-88 1.8E+1 Kr-89 7.4E+1 Kr-90 1.9E+1 Xe-131M 1.4E-2 Xe-133M 2.5E-1 Xe-133 7.3E+0 Xe-135M 2.2E+1 Xe-137 9.7E+1 Xe-138 7.3E+1 Xe-139 3.2E+1 O
Amendment 23
~
S/HNP-PSAR 10/8/82 TABLE 15.7-7 FAILURE OF AIR EJECTOR LINES OFF-SITE RADIOLOGICAL DOSES (Realistic Case) 23 Distance Whole-Body Thyroid (meters)
(rem)
(rem) 1609 (EAB) 3.15 x 10-3 1.81 x 10-4 l27 6437 (LPZ) 3.43 x 10-4 1.97 x 10-5 23 i
f 5,
r l
l i
I Amendment 27 4
S/HNP-PSAR 10/8/82 O
TABLE 15.7-8 FAILURE OF AIR EJECTOR LINES - PARAMETERS TO BE TABULATED
- TOR POSTULATED ACCIDENT ANALYSES Realistic Conservative (Conservative (NRC)
Engineering)
Assumptions Assumptions I.
Data and as=umptions used to estimate radioactive source from postulated accidents A.
Power level NA 4100 Mwt B.
Bern-up NA NA C.
Fuel damaged NA None D.
Release of activity by nuclide NA Table 15.7-5 E.
Iodine fractions (1) Organic NA 0
(2) Elemental NA 1
(3) Particulate NA 0
F.
Reactor coolant activity before the accident NA 15.6.5.5.2.2 23 II.
Data and assumptions used to estimate activity released A.
Containment leak rate (t/ day)
NA NA B.
Secondary containment leak rate (t/ day)
NA NA C.
Valve moverent times NA NA D.
Adsorption and filtration efficiencies NA NA (1) Organic iodine NA NA (2) Elemental iodine NA NA (3) Particulate iodine NA NA (4) Particulate fission products NA NA E.
Recirculation system parameters (1) Flow rate NA NA (2) Mixing efficiency NA NA (3) Falter efficiency NA NA F.
Containment spray parameters (flow rate, NA NA drcp size, etc)
G.
Containment volumes NA NA H.
All other pertinent data and assumptions NA None 111.
Dispersion Data A.
LAB and LP1 dastances (m)
NA 1609/6437 B.
X/C values in sec/m3 NA 2.8 x 10-5/3.0x10-6 27 IV.
Dose Data A.
Method of dose calculation NA Reference 1 B.
Dose conversion assurptions NA Reference 1 C.
Activity in containment NA NA D.
Off-Site doses 23 NA Table 15.7-7 As applacable to the event being described.
O Antendment 27
S/HNP-PSAR 10/8/82 0
TABLE 15.7-11 LIQUID RADWASTE TANK RUPTURE OFF-SITE DOSES 23 Distance Inhalation Thyroid Dose (meters)
(rem)
Conservative Analysis EAB (1609) 1.71 x 10-3 27 LPZ (6437) 2.38 x 10-4 23 I
Realistic Analysis EAB (1609) 4.21 x 10-5 27 LPZ (6437) 4.54 x 10-6 O
l l
i
' O Amendment 27
S/HNP-PSAR 10/8/82 O
TABLE 15.7-12 LIQUID RADWASTE TANK FAILURE: PA/ MITERS TABULATED TCR POSTULATED ACCIDEhT ANALYSES Realistic Conservative (Conser va t i*ie (NRC)
Engineeringt Assumptions Assur;tioes I.
Data and assumptions used to estimate radioactive source from postulated accidents A.
Power level NA NA B.
Burn-up NA NA C.
Tission products released from fuel NA NA (fuel damaged)
D.
Release of activity by nuclide Table 15.7-10 Table 15.7-10 E.
Jodine fractions (1) Crgante 0.01 0.01 (2) Elemental 0.01 0.01 (3) farticulate 0.01 0.01 23 T.
Reactor coolant activity before the NA NA accident II.
Data and assurptions used to estimate activity released A.
Containment leak rate (t/ day)
NA NA B.
Secondary containment release rate (t/ day)
NA NA C.
Valve movement times NA NA D.
Adsorption and filtration efficiencies NA NA (1) Organic iodine NA MA (2) Elemented iodin, NA NA (3) Particulate iodine NA NA (4) Particulate fission products NA NA 1.
Recirculation system parameters NA NA (1) Flow rate NA NA (2) Mixing efficiency NA NA (3) Filter efficiency NA NA F.
Containment spray parameters (flow rate, NA NA drop size, etc)
G.
Containment volumes KA NA B.
All other pertinent data and assurptions (1) Dilutaon factor afforded by public NA NA waterway (2) Dilution of liquid ingestion NA NA (3) Agaatic life consumed NA NA 111.
Dispersion data A.
EAB and LP: distances (m) 1609/6437 1609/6437 27 B.
X/Q values in sec/m3 1.5 x 10-4/2.1E-5 2.8 x 10-5/3.CE-6 IV.
Dose data A.
Metnod of dose calculation Appendix 15A Appendix 15A B.
Dose conversion assumptions Appendia 15A Arperdir 15A C.
Peak activity concentrations in containment NA NA 23 D.
Doses Table 15.7-11 Tatles 15.7-11 O
Amendment 27
S/HNP-PSAR 10/8/82 TABLE 15.7-15 FUEL HANDLING ACCIDENT OFF-SITE RADIOLOGICAL EXPOSURES (Realistic Analysis) 23 Distance Whole-Body Thyroid (meters)
(rem)
(rem) 1609 (EAB) 5.41 x 10-4 2.97 x 10-4 l27 6437 (LPZ) 3.30 x 10-4 2.05 x 10-4 23 4
l
\\
l l
l Amendment 27
S/HNP-PSAR 12/21/81 O
TABLE 15.7-16 FUEL HANDLING ACCIDENT (Conservative Analysis)
Activity Airborne in Isotope Refueling Building, Ci I-131 2.58E+2 I-132 4.03E-1 I-133 2.92E+2 23 I-134 I-135 4.95E+1 Kr-83M 6.50E-1 Kr-85M 2.82E+2 Kr-85 8.53E+2 Kr-87 4.80E-2 Kr-88 8.33E+1 Xe-131M 2.00E+2 Xe-133M 1.20E+3 Xe-133 5.62E+4 Xe-135 9.93E+3 O
Amendment 23
S/HNP-PSAR 12/21/81 TABLE 15.7-17 FUEL HANDLING ACCIDENT (Conservative Analysis)
Fission Product Released Isotope to Environs, C1, 0-2 hour I-131 2.58E+0 I-132 4.00E-3 I-133 2.92E+0 23 I-134 I-135 5.00E-1 I
Kr-83M 6.50E-1 Kr-85M 2.82E+2 Kr-85 8.53E+2 Kr-87 4.80E-2 i
Kr-88 8.33E+1 Xe-131M 2.00E+2 Xe-133M 1.20E+3 Xe-133 5.62E+4 Xe-135 9.93E+3 l
l l
i l
l i
i Amendment 23
..~. -
.~
S/EMP-PSAR 10/8/82 TABLE 15.7-18 FUEL HANDLING ACCIDENT OFF-SITE RADIOLOGICAL EXPOSURES (Conservative Analysis) 23 Distance Whole-Body Thyroid (meters.1 (rem)
(rem) 1609 (EAB) 8.96 x 10-3 1.34 27 6437 (LPZ) 1.24 x 10-3 1.86 x 10-1 23 O
O Amendment 27
~.
S/HNP-PSAR 12/21/81 O-TABLE 15.7-19 l
FUEL HANDLING ACCIDENT CONTROL ROOM PERSONNEL DOSES (l), REM (Conservative Case)
Skin Whole-Body Sources (B)
(7)
Thyroid (2) 4 Direct Shine 1.4E-2 Immersion Dose 9.0E-2 3.3E-3 8.3E-4 Total Dose 9.0E-2 1.7E-2 8.3E-4 (1) 1000 cfm intake flow, 2000 cfm recirculation flow, filter efficiency of 99% for iodine, 10 cfm unfiltered inleakage for all time periods.
(2)
Breathing rate of 3.47E-4 m3/sec for all time periods.
O lO Amendment 23
S/HNP-PSAR 10/8/82 O
TABLE 15.7-20 FUEL BANDLING ACCIDENT - PARAMETERS TABULATED FOR POSTULATED ACCIDE!rT ANAI.YSES Realistic Conservative (Conservative (NRC)
Engineering)
Assumptions AssJmptions Data and assomptions used to estimate radioactive source from postulated accidents A.
Power level NA NA B.
Barn-up factor 1.5 1.0 C.
Fuel damaged 98 rods 98 rods D.
Release of activity by nuclide lot noble gas, 15.7.4.5.2.2.2 10% iodine, 30% Kr-85 E.
Iodine f ractions (1) Organic 0.254 0
(2) Elemental 99.754 1
23 (3) Particulata 0
0 F.
Reactor coolant activity before the NA NA accident II.
Data and assumptions used to estimate activity released A.
Refaeling building release rate NA loot / day E.
Secondary containment release rate (t/ day)
NA NA C.
Valve movement times NA NA D.
Adsorption and filtraticn efficiencies (1) organic iodine 994 994 (2) Elemental iodine 994 994 E.
Recirculation system parameters (1) Flow rate NA NA (2) Mixing efficiency NA NA (3) Filter efficiency NA NA F.
Containment spray parameters (flow rate, drop size, etc)
G.
Containment volumes NA NA H.
All other pertinent data and assumptions None None i
.11.
Dispersion data e
A.
EAB and LPZ distances (m) 1609/6437 1609/6437 l27 B.
X/O values in sec/m3 Table 15.6-17 Table 15.6-17
'V.
Dose data A.
Method of dose calculaticn Regulatory Guide 1.25 Reference 1 B.
Dose conversion assumptions Regulatory Guide 1.25 Reference 1 23 C.
Activity in Refueling Building Table'15.7-16 Table 15.7-13 D.
off-site doses Table 15.7-18 Tables 15.7-15 O
Amendment 27