ML20003A704
| ML20003A704 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/31/1981 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20003A703 | List: |
| References | |
| RTR-NUREG-0660, RTR-NUREG-660, TASK-3.D.3.4, TASK-3.D.3.4., TASK-TM NUDOCS 8102050449 | |
| Download: ML20003A704 (182) | |
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REVIEW OF CONFORMANCE TO THE TMI TASK ACTION PLAN FOR CONTROL-ROOM HABITABILITY (NUREG 0660, SECTION III.D.3.4)
REVISION A FOR THE FORT CALHOUN STATION OMAHA PUBLIC POWER DISTRICT 4
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Prepared by:
Comoustion ~ Engineering, Inc.
509 9 0 'A0
Abstract This document provides a design review of the Fort Calhoun nuclear plant Control Room with the specific objective of evaluating its conformance to the acceptance criteria of the NRC TMI Task Action Plan position on habitability. The data input and design basis assumptions employed in the calculation of the consequences of postulated accidents were based on the procedures and guidelines required by the Task Plan Position.
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Page 2 ofl1817
Review of Conformance to the TMI Task Action Plan for Control Room Habitability, (NUREG'0660,Section III.D.3.4)
TABLE OF CONTENTS Section Title Pace Number 5
1.0 INTRODUCTION
8 1.1 Statement of NUREG 0660 Position 2.0
SUMMARY
OF RESULTS 11 15
3.0 REFERENCES
4.0 DATA REQUIRED FOR CONTROL ROOM HABITABILITY 17 E'iALUATION 4.1 Site Building Layout and Meteorology 17 4.2 Control Room Characteristics-38 4.3 Control Room Modes of Operation 41 4.4 On-site Storage of Hazardous Chemicals 49 4.5 Off-site Manufacturing, Storage and 54 Transportation Facilities of Hazardous Chemicals 4.6 Technical Specification 77 5.0 ASSUMPTIONS AND INITIAL CONDITIONS 81 6.0 CONTROL ROOM HABITABILITY DESIGN EVALUATION 83 AND ANALYSIS 6.1 Assessment of Hazardous Chemical Materials 84 6.1.1 Worst Case Chemical Accident Release 86 6.1.2 Chemical Transport Accident Probability 88 6.1.3 Control Room Infiltration and Exfiltration 96 Rates 6.1.4 Design Basis Chemical Accident Consequences
-99 6.2 Assessment of _ Radiological Events 125 6.2.1 Airborn Radiological Consequences 126 6.2.2 Control Room Shielding Design Review 170 Page 3 of 181
Section Title-Pace f; umber
- 5. 3_
Emergency Instrumentation and Proced.ure-172 Review 6.4 Control Room Sustained Occupancy Review 174 7.0 PROPOSED CORRECTIVE ACTIONS 178 7.1 Instrumentation and Equipment Recommer:dations -
178 7.2 Procedural and Technical Specification 179 Recomendations J
J Page:4.of 181'
1.0 INTRODUCTION
This report provides a design evaluation and analysis of the Fort Calhoun Control Room envelope for detemination of ccmpliance with the NRC TMI Task Action plan NUREG 0660 position on Control Room Habitability.
The objective of this design evaluation is to detemine the adequacy with which Control Room operators can be protected against the effects of an accidental release of either toxic or radioactive gas thereby.
allowing subsequent appropriate actions as required by General Design Criterion 19 of 10 CFR Part 50.
An analysis of potentially hazardous chemical storage deposits both on and off the plant site and shipment past the site was performed to detemine which hazardous chemicals should be considered as sources of credible design basis accidents. The generic frequency and size guidelines within Regulatory Guide 1.78 were utilized to exclude those chemicals of insufficient quantity to warrant further examination. The first phase of the analysis consists of applying a set of conservative accident assumptions for detemination.of the. Control Room air intake concentration for specific chemicals.
Those chemicals shown to be within acceptable limits under the most conservative conditions were eliminated from further examination.
The remaining chemicals were modeled stochastically for :the sequence of events resulting in an accidental atmospheric release to Page.5 of 181
determine the probability of exceeding toxicity limits at the Control Room air intake.
The design basis chemical accidents were established through comparison of the stochastic analyses with the conservatively low probability of 10 events per year.
Postulated chemical accidents identified as design basis events were then evaluated in detail to determine the resulting consequences on Control Room habitability.
Information. generated as a result of the postulated chemical accident analyses is provided within Section 2.0.
The physical descriptions, means of transportation and description of facilities for onsite, offsite, and transported hazardous chemicals was recorded.
A tabular summary of results for the effects of postulated toxic gas releases is also provided.
Several of the items-listed within the table include the specific chemical examined, the type of accident postulated, the chemical concentration calculated at the Control Room air intake and the type of release mechanism employed.
Radiological analyses addressed the effects of Control Room operator exposures to radiation resulting from radiological design basis accidents.
Exposure from airborne radioactive materials as well as shine dose was examined. An evaluation of the design of Control Room Shielding was performed in order to insure that the Control Room operator shine dose.
during a design basis accident is minimized.
' Control Room operator exposure received from the airborne and ' shine doses.are provided in conjunction with the principal design basis accidents responsible for'the resulting doses.
Page5 o'f 181
The chemical and radiological accident scenarios utilized engineering judgement and specific plant data for parameters dealing with meteorological conditions, release rates of the various chemical plumes and means of transport for both chemical and radiological airborne analyses. A listing of the assumptions employed is orovided in Section 5.0.
Corrective action alternatives are provided to insure that the dose rates and occupational exposures to Control Room Operators are witnin the acceptable guidelines and to mitigate the consequences of the chemical scenarios investi-gated.
Each alternative described herein has been developed with the objective of minimizing personnel exposure to radiation and hazardous chemical concentrations.
Page 7 of 181
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1.1 STATEMENT OF NUREG -0660 POSITION The following is a summation of the NRR's proposed program for implementation of operating plant requirements for Control Room Habitability:
"In accordance with the NUREG 0660 position on Control Room Habitability, licensees shall assure that Control Room operators will be adequately protected against the effects of accidental release of toxic and radio-active gases and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (Criterion 19. " Control Room," of Appendix A.
" General Design Criteria for Nuclear Power Plants," to 10CFRPart50)."
The NRC clarification letter of May 7,1980, to all Operating Reactor Licensees defined the following evaluation criteria for Control Room Habitability of all PWR's :
All Facilities that have not been reviewed for conformance with the following sections of the Standard Review Plan:
2.2.1 - 2.2.2 Identification of Potential Hazards in
' Site Vicinity; 2.2.3 Evaluation of Potential Accidents; 6.4
. Habitability Systems; Page.8 of-181 i
shall perform the necessary evaluations and recomend appropriate modifications to neet Control Room Habit-ability requirements.
The following documents may be used for guidance in performing the required evaluations:
1.
Regulatory Guide 1.78, " Assumptions,for Evaluating the Habitability of a Nuclear Power Plant Control Room Ouring a Postulated Hazardous Chemical Release."
2.
Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release."
3.
K. G. Murphy and K. M. Campe, " Nuclear Power Plant Control Room Ventilation System Design for Meeting 4
General Design Criterion 19, 13th AEC Air Cleaning Conference, August,1974.
The licensees submittal shall include the results of the analyses of Control Room concentrations from the postu-lated accidental release of toxic gases and Control Room operator radiation exposures from airborne radioactive material and direct radiation resulting from design basis accidents.
The toxic gas accident analysis should be performed for all potential hazardous chemical releases-occurring either on the site or within five miles of the plant site-boundary.
Regulatory Guide 1.78 lists the' chemicals most ccmonly encountered in -the evaluation of Control Room Habitability but is not all-inclusive.
Page 9 of 181
The CSA radiation source term should be for the LOCA con-tainment leakage and ESF leakage contribution outside containment as described in Appendix A and B of Standard Review Plan Chapter 15.6.5.
Other CBA's should be re-viewed to determine whether they might constitute a more severe Control Room hazard than the LOCA.
In addition to the accident analysis results.which should either identify the possible need for Control Room modifi-cations or provide assurance that the habitability systems will operate under all postulated conditions to permit the Control Room operators to remain in the Control Room to take appraoriate actions as required by General Design Criteria 19, the licensee should submit sufficient infor-mation needed for an independent evaluation of the adequacy of the habitability systems.
Page 10 of 181
2.0
SUMMARY
OF RESULTS The existing Fort Calhoun station Control Room envelope and habitability systems have been evaluated and analyzed to determine the adequacy with which Control Room operators i
are protected against the effects of an accidental release of either toxic or radioactive gas thereby allowing the nuclear power plant to be safely operated or shut down under design basis accident conditions. The objectives for assuring sustained Control Room occupancy as stated in the NRC TMI Task Action Plan Position, Reference 3.1, and the NRC clarifications on this position, References 3.2, 3.3 and 3.24, have been employed as the acceptance criteria. As a result of this design evaluation the follow-ing conclusions have been reached:
(1) The Control Room ventilation system layout, functional design and leak tightness provides appropriate pro-tection under all of the design basis accidents involving airborne toxic chemicals. The calculated effects of postulated toxic gas releases on the Control Room habitability are provided in Table 2.0.
Several specific recomendations for the installation of con-tinuous chemical monitoring equipment in the ambient intake air ductwork have been made to assure that the Control Room operators have sufficient time to implement protective measures within the guidelines provided in Reference 3.15.
Page 11 of 181
(2) Existing Centrol Room radiation shielding provides appropriate protection under all of the design basis accidents with tne exception of a postulated post-accident release of radioactivity equivalent to that described in Reference 3.18.
Specific radiation sources have been identified for additional structural concrete shielding protection to increase the duration of occupancy to within established limits.
Recom-mendations for shielding design modifications are included.
(3) The Engineered Safety Feature systems, the Control Room ventilation system, and emergency procedures provide appropriate protection against airborne sources of radioactivity under all of the design basis accidents.
Specific procedural modifications have been recom-mended to assure implementation of the required protective measures.
(4) The Control Room envelope has been shown to provide sus-tained occupancy under postulated design basis acaident conditions for the required number.of. operators. Ap-propriate emergency supplies, procedures, instrumentation, and other supporting equipment are provided in sufficient quantity to maintain five plant operators for a period of five days.
Page 12 of.181
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(Sec) sec P -instantaneous Release Ca sitne Comt.unt ton of 75 ton 18 (NO )
Yes (2) 2 rattroa<t tank car casaline Rupt ure of 12 pipeline 3600 (rentane)
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Page 14 of 181
3.0 REFERENCES
3.1 NRC Action Plan Developed as a Result of the TMI-2 Accident, NUREG-0660, dated May, 1970.
3.2 NRC Correspondence "Five Additional TMI-2 Related Requirements to Operating Rer*, ors," D. G. Eisenhut To All Operating Reactor Licensees data May 7, 1980.
3.3 NRC Correspondence " Preliminary Clarification of TMI Action Plan Requirements," D. G. Eisenhut To All Licensees of Operating Plants and Applicants fo-Operating Licensees and Holders of Construction Permits dated September 5, 1980.
3.4 Final Safety Analysis Report for the Fort Calhoun Station.
3.5 ASHRAE Handbook and Product Directory,1977 Fundamentals, Second Printing 1978.
3.6 Regulatory Guide 1.78, Assumptions for Evluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release.
Issued June, 1974 3.7 Regulatory Guide 1.95, Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release.
Issued January, 1977.
3.8 Regulatory Guide 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Wa+er Reacters.
Issued June, 1974.
3.9 Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Con-tainment Following a Loss-of-Coolant Accident.
Issued November, 1978.
3.10 Regulatory Guide 1.24, Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure.
Issued March, 1972.
3.11 Regulatory Guide 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of_ a Fuel Handling Accident in the Fuel-Handling and Storage Facility for Boiling and Pressurized Water Reactors.
Issued flarch, 1972.
3.12 Regulatory Guide 1.77, Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors.
Issued May, 1974.
3.13 Standard Review Plan, Sections 2.2.1-2.2.2, Identification of Potential Hazards in Site Vicinity, Revision 01.
'Page 15 of 181
i 3.14 Standard Review Plan, Section 2.2.3, Evaluation of Potential Accidents, Revision 01, 3.15 Standard Review Plan, Section 6.4, Habitability Systems, Revision 01.
3.16 Standard Review Plan, Section 15.6.5, Loss-of-Coolant Accident Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, Revision 01.
3.17 Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19, K. G. Murphy and Dr. K. M. Campe, '.~3th AEC Air Cleaning Conference, August, 1974.
3.18 Design Review of Plant Shielding and Environmental Qualif.ication for Spaces and Systems which may be Used In Post-Accident Operations OPPD Response to HURE3-0578, Section 2.1.6.b.
3.19 WASH 1238 Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants, December 1972.
3.20 tieteorology and Atomic Energy,1968, USAEC, (TID - 24190).
3.21 Building Effects on Effluent Dispersion from Roof Vents at Nuclear Power Plants, April 1980, EPRI NP-1380 Project 1073-1.
3.22 Dangerous Properties of Industrial Materials Fourth Edition Edited by N. Irving Sax dated 1975.
3.23 Risk Assessment of Large Spills of Toxic Materials' Proceedings of the 1974 Naticnal Conference On Control of Hazardous Material Spills.
Simmons, Erdmann, Naft 3.24 Clarification of the TMI Action Plan Requirenents NUREG 0737,0ctober 1980.
3.25 Behavior of Ammonia In the Event of a Spillage AICHE Manual on Aarnonia Plant Safety Volume 22, 1980.
3.26 Accidental Episode Manual EPA Office of Air Programs Publication APTD 1114 dated January 1972.
3.27 Loss Prevention, Volume 13, AICHE, Using Aqueous Foams to Lessen Vaporation From Hazardous Chemical Spills, E. C.
lormal and H. A.- Dowell, dated 1980.
Pa'ge 16 of 181
~~~~:
4.0 DATA REQUIRED FOR CONTROL ROOM HABITABILITY EVALUATION The Control Rcom habitability systems addre.;. sed in this evaluation include radiation shielding, radiation monitoring, air conditioning, ventilation systems, emergency equipment, and procedures. Data employed in the evaluation of Control Room habitability, including design basis paraneters fo'r radiological or toxic material release within the vicinity of the plant site,are provided.
Control Room characteristics, habitability systems, and corresponding modes of ventilation system operation are described in detail. Storage locations of potentially hazardous chemicals both on-site and within a five mile radius of the Control Room ventilation intake are identified. Potentially hazardous materials transported within the same radius are identified by frequency and mode of conveyance. Analytical development of the consequences of postulated credible design basis accidents is based upon the plant specific data provided in this section. This information was compiled through comprehensive on-site inspections, communication with-appropriate Federal, State and County Governmental Agencies, systems and component design descriptions, and plant-structural and piping layout drawings.
4.1 Site Building Layout and Meteorology Fort Calhoun Unit 1 site layout drawings and site-specific meteorn-logical data are provided within this section. The drawings provide a conceptual view of the transport pathway taken by either airborne Page 17 of 181
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hazardous chemicals or airborne radiological contamination from the point of release to +.he Control Room air intake structure. An aerial photognph of the plant site showing physical terrain and man-made structures within the ir:rrediate site vicinity is also provided.
The relative location of the hazardous material storage sites identified during onsite inspections are contained within Figures 4.1-1 through 4.1 5.
Figure 4.1-1 presents an overhead view locating the onsite hazardous material storage areas, nearest interstate highway, nearest railroad right-of-way and barge channel with respect to the plant structures.
Figure 4.1-2 presents an overhead viaw of the plant structures indicating the closest distance of approach of on-site hazardous i
materials to the Control Pcom air intake.
Figures 4.1-3,.and 4.14 present building arrange =ent drawings indicating the hazardous material storage areas within the Turbine Building. Figure 4.1-5 presents an arrangement drawing of the Auxiliary Building ground. level with the hazardous material storage location indicated. Specific quantities and chemical states of the identified materials are recorded within Section 4.4.
An overhead view of the location of the Turbine Building roof fans, Auxiliary Building ventilation duct, Office Build-ing ventilation and safety relief valves with respect to the Control Room air intake. structure is provided.in Figure' 4.1-6 Page 18 of-181
Site specific meteorological parameters utilized during the design evaluation are presented in Tables 4.1.1 through 4.1.11.
Tables 4.'.1 through 4.1.8 list the observed wind speed and direction frequency distribution expressed in percentage for each of the seven Pasquill Stioility Categories. Table 4.1.9 provides a breakdown of each Pasquili category with respect to conditions of stability and temperature change with height. Table 4.1.10 provides a list of the number of observations for each Pasquill Str.bility Category.
Table 4.1.11 lists the calculated atmospheric dispersion factors (X/Q @ 5%) as a function of distance.
Page 19 of 181-
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0.0 0.02 0.14 0.20 0.16 0.13 0.08 0.07 0.05 0.04 0.07 0.01 0.0 0.0 0.0 0.99 ESE1 0.01 0.05 0.25 0.30 0.29 0.31 0.24 0.17 0.11 0.10 0.12 0.02 0.02 0.04 0.0 2.03 SE 0.02 0.05 0.22 0.38 0.73 0.57 0.58 0.41 0.35 0.24 0.28 0.08 0.02 0.02 0.0 3.96 SSE 0.01 0.08 0.16 0.19 0.29 0.49 0.53 0.64 0.54 0.49 0.59 0.49 0.04 0.01 0.01 4.57 S
0.01 0.05 0.13 0.08 0.12 0.25 0.35 0.44 0.55 0.51 1.11 0.57 0.23 0.13 0.04 4.57 SSW 0.01 0.12 0.10 0.11 0.12 0.10 0.14 0.08 0.07 0.17 0.76 0.47 0.28 0.20 0.24 2.97 SW 0.01 0.07 0.19 0.11 0.11 0.07 0.07 0.07 0.04 0.10 0.13 0.24 0.17 0.04 0.05 1.47 WSW 0.01 0.16 0.12 0.08 0.17 0.07 0.08. o.05 0.11 0.06 0.12 0.08 0.04 0.01 0.0 1.17 W
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Table 4.1.8 JOINT FREQUENCY DISTRIBUTION WIND DIRECTION vs. WIND SPEED IN METERS /SEC FOR OT100 = -INF TO 41NF IN PERCENT SECTOR IS WIND DIRECTION NOT AFFECTED DIRECTION 0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 6.0 7.0 8.0 9.0 SECTOR TO TO TO TO TO TO TO TO TO 10 TO 10 TO TO TO
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Stability Pasquill Temperature Change Classification Catecories with Heicnt (*C/100 m)
Extremely unstable A
aT/az < -1.9 Moderately unstable B
-1.9 < AT/az 1 -1.7
)
Slightly unstable C
-1.7 < AT/az < -1.5 Neutral D
-1.5 < aT/az < -0.5 Slightly stable E
-0.5 < aT/az < l.5 Moderately stable F
1.5 < AT/az < 4.0 Extremely stable.
G 4.0 < aT/az i
i T
i i
2 4
i Page'28 of 181
l Table 4.1.10 TOTAL OBSERVATIONS RECORDED PER j
PA50 VILL STASILITY CATEGORY
~
Pasauill Stability Number of Percent of Valid Category Observations Observations A
13 0.2 B
20 0.2 C
108 1.3 0
3,409 41.0 E
3,410 41.0 F
1,001 12.1 G
346 4.2 TOTAL 8,309 94.9 i
1 4
4 4
Page 29.of 181 m.a_* - - -. _,_ _ _ _ -
A..m..-
L.
i.
Table 4.1.11 ATMOSPHERIC DISPERSION FACTORS (X/0) AS A FUNCTION OF DISTANCE i
Time Period and Probability Distance (m) 0-2 hours 2-8 hours 5%
5%
2.54E-03*
100 200 3.99E-03 7.08E-04 500 9.98E-04 2.21E-04 700 7.30E-04 1.53E-04 800 6.58E-04 1.38E-04 900 5.85E-04 1.22E-04 1000 5.14E-04 1.02E-04 1200 4.19E-04 7.98E-05 1400 3.46E-04 6.48E-05 1500 3.21E-04 5.94E-05 1800 2.68E-04 4.69E-05 2000 2.21E-04 4.07E-05 2200 1.98E-04 3.61E-05 2500 1.73E-04 3.06E-05 3000 1.38E-04 2.42E-05 3500 1.12E-04 1.98E-05 4000 8.74E-05 1.69E-05 4500 7.49E-05 1.47E-05 4800 6.99E-05 1.37E-05 5000 6.49E-05 1.29E-05 6000 5.24E-05 9.38E-06 000
- 4. 30E-05 7.68E-06 8000 3.74E-05 6.45E-06 9000 3.24E-05 5.55E-06 10000 2.80E-05 4.82E-06 20000 5.56E-06 1.24E-06 50000 1.98E-06 4.89E-07
- * *
- GREATER THAN 9.50 x 10-3
- AFFECTED DIRECTION IS ESE FOR ALL DISTANCES LISTED Page 30 of 181
n----._.--
i ADDITIONAL EXCLUSIOM AREA MISSOURI RIVER Y
c- -
C
- 1. REACTOR CONTAINMENT f.
~
,,{
- 2. AUXILIARY 3UILDING
@o @.D
- 3. TURBINE ROO.*.13 SERVICE FPARK 7 ggg,ING
- 4. imAxE SmuCmRE FENCE
~
A. CHLORINE CYLINDER STORAGE AREA g
iI B. SULFURIC ACID STORAGE
\\t TANK O
^
I BOUND RY g
HYDROGEN CYLINDERS NITROGEN CYLINCERS
,\\
ARGON CYLINDERS CARSON DIOXIDE SVilTCHYARD CYLINDERS AREA L.P GAS CYLINDERS c
.,.rACCESS ACETYLENE CYLINDERS FENCE ROAD OXYGEN CYLINDERS D. UNDERGROUND AUXILIARY 3 OILER OIL STORAGE E. WASTE OIL 55 GALLON R AI L%-
DRUMS S UR 3 [
F. DEISEL FUEL STCRAGE J
L TANKS G. GASOLINE STORAGE TANK H. CANOPY 3007.1:
OXYGEfJ CYLINDERS NITROGEN CYLINDERS METHANE CYLINDERS CHICAGO & NORTHWESTERN HYDROGEN CYLINDERS RAILWAY TRACMS L.P GAS CYLINDERS 1.
NITROGEN CYLINDERS U.S. HIGHWAY No. 73 FEET 0
250 500 1000 SCALE Figure 4.11 OMAHA PU3LIC POWER DISTRICT FT. CALHOUN STATION UNIT No.1 SITE PLAN Pace 31 of 131
MISSOURI RIVER
~
_s 1.
REACTOR CONTAINMENT
- 2. AUXILI ARY CUILDING
- 3. TURBINE ROO.'.l a SERVICS DUILDING b
bb 4.
INTAKE STRUCTURE
+
- 5. WAREHOUSE STRUCTURES I LOCATION OF CONTROL AREA
-r ROOM AIR INTAKE 3
A CLOSEST DISTANCE OF FENCE \\
UZ 2
n APPROACH TO CONTMOL
'g ROOM AIR INTAKE FROM*
A. CHLORINE CYLit 'OER STORAGE AREA:
}
~ 330 FEET y
[
B. SULFURIC ACID STORAGE TANK:
=
=
~ 330 FEET j:
C. CANOPY ROOM:
(
7--
~ 330 FEET g
D. UNDERGROUND r
AUXILIARY BOILER OIL STOR AG E:
~ 360 FEET I
I E. WASTE Oll:
s ACCESS
~ 500 FEET ROAD F. DIESEL FUEL STORAGE t
TANKS:
g
[
~ 620 FEET i
G. GASOLINE STORAGE L
TANK:
0
~ 575 FEET H. CANOPY ROOM:
~ 180 FEET l.
NITROGEN CYLINDERS:
~ 200 FEET Figure 4.12 OMAHA PU3LIC POWER DISTRICT FT. CALHOUN STATION UNIT No.1 SITE PLAN Pace 32 of 181
i 7
r CONDENSATE SERVICE WATER TANKS CONTROL AND PIPING AREA ROOM N
@ HALON 1301 DISCHARGE t
I I
@ CARBON DIOXIDE (6 CYLINDERS O ELEVATION 1011')
l LEGEl0 MECHANICAL EQUIPMENT t'
0 3 I
aocu COMPUTE R.
ROOM l 1
2 it I
UPPER PART OF CONTAINMENT i
VENTILATION EQUIPt.1ENT BUILDING AREA Y L
.4 0
0d L
u u
t
~
NEW FUEL SPENT FUEL W#8 STORAGE AREA
[
HOLDUP i
TANKS
~
UPP R PART:
CASK DECON.
AREA l
n n
n r1 n
n r
Figure 4.13 PRIMARY PLANT OPERATING FLOOR PLAN Elevation 1025' Page 33 of 181
'5'~
N l
a L
1 m
m u
TRUCK DOCK
\\.
LOB 3Y F
@ HYDR AZINE, POTASH, Lif,1ESTONE,
[
ALUMINUM SULFATE STORAGE 4
J ROTOR C
@ CdLORINE STORAGE C
h HANDLING
@ SULFURIC ACID STOR AGE AREA TOOL
@ SDDIUM HYDROXIDE STORAGE
@ HALON 1301 (7 CYLINDERS)
ROOM
@ HYDRAZlTIE STORAGE I
@ AUXILIARY BOILER OIL C
(UNDERGROUND TANK)
STAIRS
~
CONDENSER l
C
(
CE i
J @
C Of f
CONDENSER 1_
(
li
(
7 h
GAS STORAGE CAfJOPY V/ATER TREATMErlT PUMPS l
AREA J
E
,-s s
h F E E C'.~t/'.T E R l
T ER
\\
TAfJK HEATERS j
OIL
,,/
RESERVOIR N
_s
-s hl I
/
\\
l PRESEDIMEN.y Gj i TATICfJ j
"! O j TANK
/
g g g
y.;.*f
~/
m L
t.
Figure 4.1-4 TURBINE & OFFICE BUILDING PLAN AT ELEVATICNS 1000' 0",1004' G", L 1007' G" Paqe 34 of 181 L.
> c2' I
c--
SWITCHGEAR AREA I
I J
ELECTRICAL PENETRATION AREA DIESEL i
l GENERATOR I
AREA l CHEMISTRY PIPE
)
AND PENETRATIO HEALTH PHYSICS
'LABORAORYl I
CONTAINMENT BUILDING k
SAMPLING AREA l
ION EXCHArJGE AREA V/ASTE HOLDUP TANK AREA u-l l VIASTE DISPOSAL U
AREA 1
h STORAGE LOCATION OF ARGON AND 90% ARGON /10% METHANE CYLIN Figure 4.15 AUXILIARY DUILDING LAYOUT GROUND LEVEL (1007' ELEV)
Page 35 of 101
,,y r--v--
-r
---trw v
--ew*v-*--w-w ir v-w-7
--*y
>-*r*
s-
--*'-w v
s~Fm 1
~ ' -
- Figure 4.16 f
PLANT VENTILATION WITH RESPECT TO CONTROL ROOM AIR INTAKE I
CLOSEST DISTANCE OF OFFICE BUILDING APPROACH TO CONTROL O
l ROOM AIR INTAKE FROM:
O OFFICE BUILDING EXHAUST VENTS j
1.
OFFICE BUILDING EXHAUST VENTS:
l
~ 240 FEET TURBINE BUILDING
- 2. TURBINE BUILDING VENTILATION:
l f
- 100 FEET l
TURBINE BUILDING VENTILATION (ROOF FANS) i
~20 FEET 4- #'M':^"Y ""
~ 50 FEET I
SAFETY VALVES I
I l
CONTROL ROOM EXHAUST l
AIR INTAKE j
l f
l l
A'JXILIARY L
t'llLDING CONTAINMENT i
STACK i
f l
l
)
AUXtLIARY SUILDING Page 36 of 181
3 n
t n
I i
I m
l ee M.i i
N&
m
- uJ CO C/) C O %
CC o
<:C -
I LaJt.L. N m
LLJ cn c
O Q-I i
I l3 l
i O
a' I
' Ih O
g
,... y_
e 12 t
t ',
\\\\
ir l15 'p 4 ;
1
\\. ^
e
-1
.} : ',
i t
i e
- ?
g l
t,
- )5 5
a
.e '8 i
}
- 3. e..
I s
I 9
t i
5 1
.6 i
1
) #
=
a
(
5 7%
ee'44 i#.4e es.
- eGb4fD 6h*bedD h# r 4m W M M SM Ag3 -. e es meenO 44Ek+-
"W
4.2 Control Room Characteristics The Control Room envelope, consisting of tho 0 : paces requiring con-tinuous or frequent operator occupancy, is shown in Figure 4.2-1.
The Control Room envelope includes the shift supervisors office, computer room, control area, storage areas, toilet facilities and kitchen facilities.
Specific Control Room characteristic parameters are provided in Table 4.2.
The room is designed to maintain a suitable environment for sustained occupancy of at least 5 persons.
Personnel protection equipment is provided in-an emergencv cabinet.
Sufficient bottled air is stored in the Control Room envelope for 30 minutes with an additional 30 cylinders stored within quick access. A detailed description of the Control Room ventilation system is provided in Section 4.3.
Radiation protection is provided by an area monitor. The detector is located where potential hazard exists and where routine access is required.
The detectors are coaxial air-filled ionization chambers of wide dynamic range. Two independently adjustable setpoints are provided for each monitor. - The lower setpoint alam warns that the dose rate has reached an abnomal but safe value.
The upper setpoint warns that the dose has reached or passed the pemissible limit for continued occupancy. The monitor alams and indicates both locally and on _the radiation monitor panel AI-33. The monitor records on the radiation monitor panel..
Page 38 of 181'
Table 4.2 Control Room Characteri., tic Parameters Envelope Dimensions:
Length 78.75 ft Width 66.75 ft Height 19.0 ft 3
Air Volume 99875.0 ft Floor Surface 5256.5 ft2 Closest Distance to Ventilation Intake from:
Containment Building 12 ft Containment Building Purge Exhaust 48 ft Auxiliary Building Exhaust 48 ft Office Building Exhaust 240 ft Turbine Building Exhaust 100 ft Steam Safety Valve Exhaust 19 ft Personnel Capacity:
5 Kitchen Facilities:
Sink, stove, microwave oven, cabinets, utensils Emergency Equipment:
5 personnel dosimeters anti-contaminant clothing 5 air particulate masks 2 radioactive liquid spill kits 2 Eberline E120 radmonitors 2 First Aide Kits 1 Air particulate sampler potassium iodide tablets emergency food Area Radiation Monitor:
Detector Type Victoreen Model ~847-1 Readout / power supply Victoreen Model 845 Range 0.1 to 107 mrem /hr
~
Sensi tivi ty.
80 kev to 3 MeV 10", (as low as 40 Kev)
Response Time 3 to 15 seconds Page 39 of 181
Figure 4.2-1 CONTROL ROOM ENVELOPE AND VENTILATION LAYOUT r
i
_J l
I 5
x t.-
/i H
CONTROL
}
'NJ ROOM s
/
~
~
D
[
l T-u i cE l
ELEVATOR KITCHEN T
Y ROOM-gglg7 COMPUTER SUPERVIS3R TOILET ROOM
" Hlyght OFFICE g
L/
w STAIR HALL STORAGE fj "1i N
/
lf a
NOTE:
STAIRWELL AND ELEVATOR SHAFT NOT INCLUDED WITHIN CONTROL ROOM ENVELOPE l
Page 40 of 101
4.3 Control Room Ventilation and Mode of Opert tion The Control Room is ventilated by an independent system which intakes ambient air through a storm proof roof mounted aluminum penthouse and discharges to the atmosphere through a centrifugal type exhaust fan. A drawing of the relative layout of the control room air intake with respect to the auxiliary, turbine, and office building exhausts is provided in Figure 4.1-6.
The ventilation system maintains the control room at a pressure slightly greater than that of surrounding areas to preclude infiltration. Section 6.1.3 provides a detailed description of the infiltration and exfiltration rate analysis. Table 4.3 provides a description of the ventilation system component design parameters.
The system conditions three individually controlled zones: the main Control Room area, the computer room and the supervisor's office.
In addition to the conventional space conditioning in the-control room area, a part of the air supply is ducted through the control panels and instrumentation cabinets to provide direct cooling of the enclosed equipment. The system is designed to maintain a space temperature of 78 F at 50 percent maximum relative humidity. A humidifier is installed in the computer room supply duct-to maintain this area at a constant relative humidity of 50 percent.
The air conditioning eouipment consists of two multi-zone, water-cooled, package air' conditioning units located in the auxiliary building outside of the control room. Each unit is rated at one hundred percent of the system design capacity. ' The equipment is Page-41 of 181
designed for normal operation at 13,970 cfm total air recirculation with 820 cfm of outside air.1akeup. Tie air conditioning unit con-densers operate with component cooling water. A HEPA and charcoal filter assembly, with a separate booster fan rated at 820 cfm, is installed on the outside air makeup intake to the system. Under normal operating conditions this equipment is bypassed.
The ventilation system operates continuously with one air condition-ing unit operating and the other on standby.
There are four in-dependent modes of system operation which may be selected manually.
The system automatically realigns following a ventilation isolation actuation signal.
The normal made of system operation is shown in Figure 4.3-1.
Under the normal mode of operation the air conditioning ~ unit recirculates room air with ambient air being inducted as makeup for exfiltration and exhaust through fan VA-49. The system _ is under automatic control from the space thermostats and humistats with the heating and cooling coils operable. The air intake HEPA and charcoal filter unit is by-
~
passed.
The Once-Through mode of system operation is shown in Figure 4.3-2.
Under the once-through mode of operation the Control Room is ventilated with ambient air exhausted through fan VA-28.
Nc heating or cooling coils are operating.
Page 42 ' of 181
The Filtered Air Makeup mode of operation is shown in Figure 4.3-3.
Under the filtered air makeup mode of operation the air conditioning unit recirculates room air. Makeup is inducted through the HEPA and charcoal filter unit VA-64 by the Control Room emergency air snoply for VA-63. The normal air intake and exhaust dampers are closed and the exhaust fans secured. The system is under automatic control from the space thermostats and humistats with the heating and cooling coils operable. This mode of operation is used only in the event of high airborne radiological or toxic materials.
In the event of a containment ventilation isolation actuation signal, VIAS, the control room ventilation system is automatically realigned to the Filtered Air Makeup made of operation. A VIAS is initiated by a Safety Injection Actuation Signal, Containment Spray Actuation Signal or a Containment Atmosphere High Radiation Signal, CHRS. The VIAS will thus result indirectly from a Pressurizer Pressure Low Signal or a Containment Pressure High Signal.
The Internal Recirculation mode of operation is shown in Figure 4.3-4.
Under the internal recirculation mode of operation the air conditioning unit recirculates room air. All air intake and exhaust dangers are closed and the exhaust fans secured. The system is under automatic control from the space tnermostats and humistats.
4 Page 43 of,181
[-
c__-_____.
Table 4.3 Ventilation System Comoonent Design Parameters Air Conditioni,g Units VA-46 A/B f4anufacturer Trame f4anufacturing Company Type f4ultizone, Size 30 Unit capacity, each 13,970 cfm Fan motor power 15 PP Cooling capacity 314,000 Btu /hr Heating capacity 160,000 Stu/hr Steam pressure supply 10 psig Refrigerant 55 lbs of R-22 Control Room Exhaust Fan VA-28 Type Low Contour Dyafan - centrifugal
. Capacity 13,150 cfm Speed 350 RPM Fan motor power 2.HP Toilet Room Exhaust Fan VA-49 Same as VA-28 Capacity 200 cfm Control Ronm Emergency Air Suoply Fan VA-63
~
f4anufacturer ILG Ventilating Co.
Type Centrifugal size 20HP Static Pressure Rise 1.2 inches H,0 Capacity 820 cfm HEPA and Charcoal Filter Unit VA-64 HEPA removal. efficiency
. 99.975 for pa: ticles
>0.3 micron Charcoal absorption efficiency 90% for iodine Charcoal. filter cell size 2 inch Page 44 of 181
Figure 4.31 CONTROL ROOM VENTILATION SYSTEM MODES OF OPERATION (A)
NORMAL OPERATION l
VA 48 VA-65 7' \\
VA-28 ~
VA-49 w
x n
n h
NOT OPERATING OPERATING PCV PCV PCV 860C i I
8GOA L -.
I 846 I-
_1 OPEN CLOSED CLOSED VA 64 CONTROL
!b ROOM b!
PCV 8608
~
NOT OPERATING VAG3 PCV OPEN CLOSED BOOSTER FAN 340B RUNNIV 5
8==s PART COILS OPEN ENERGlZED i
Page 45 of 181
Figura 4.3-2 CONTROL ROOM VENTILATION SYSTEM MODES OF OPERATION 1
(B) ONCE-THROUGH VENTILATION VA 48 VA-65 VA-28 VA 49 w
/
s A
n i
OPERATING OPERATING PCV PCV PCV 8 GOC l 1
BGOA I.
.I 846 I I
CPEN CLOSED OPEN VA G4 CONTROL
}<
ROOM M i
PCV i
160B ll
~
NOT l
CLOSED BOOSTER FAf4 3403 CLOSED l
OPEN ColLS DE ENERGlZED Page 46 of 181
--..... \\
Figure 4.3-3 CONTROL ROOM VENTILATION SYSTEM MODES OF OPERATION (C) FILTCAED AIR MAKEUP MODE THIS IS THE DESIGN MODE OF OPERATION FOLLOWING A DBA VA-48 VA 65 VA 28 VA-49 T
s n
h NOT NOT OPERATING OPERATING PCV PCV PCV BG0C i 1
8GOA i I
846 1-I CLOSED OPEN CLOSED VA 64 CONTROL
!b ROOM b!
PCV B603
~
OPERATING v
OPEN BOOSTER FAN 3403 OPEN i
~
~
840A VA-4GA m
!hi L3l PART COILS OPEN ENERGlZED Page 47 of 181
Figure 4.34 CONTROL ROOM VENTILATION SYSTEM MODES OF OPERATION (D)
INTERNAL RECIRCULATION MODE VA-48 VA 65 VA-28 VA 49 j
w
/
x n
n NOT c
NOT Q
OPERATING O
OPERATING PCV PCV PCV 860C i i
8GOA I:
I 846 t i
CLOSED CLOSED CLOSED VA G4 CONTROL
!b ROOM b!
PCV 8608
.[ h NOT j OPERATING VA-63 PCV A
CLOSEDBOOSTER FAN 8408 OPEN l'l PCV 840A VA 4GA g!!f CLOSED COlLS ENERGlZED h
e Page 48 of 181
4.4 On-Site Storage of Hazardous Chemicals The storage locations of potentially hazardous chemicals were identified through a comprehensive program of on-site inspection and interviews with staff personnel from groups responsible for chemical use and storage. These groups included Control Room Operations, Maintenance, Chemistry, Auxiliary Operations and Warehouse.
Bulk storage locations are indicated on the plant arrangement drawings provided in Section 4.1.
Bulk storagr. of those hazardous chemicals nomally used in small quantities is in the warehouse located outside of the security fence. The warehouse is located approximately 200 yards from the Control Room air intake. A list of the chemicals stored within is proviced in Table 4.4.1.
Procedures have been established to limit the individual quantity of these chemicals transported within the plant to below that which is considered hazardous.
Storage is at room temperature in individual packaging. At present, plans are being established to construct a protected storage facility independent of other warehouse operations. A detailed list of the chemicals stored in bulk within the plant exclusion area is provided in Table 4.4.2.
This table, in addition, provides a description of the means of storage, use of the material, and closest distance of approach to the Control Room air _ intake.
Page 49~of 181
1 i
i It must be emphasized that the amounts and type of chemicals i
listed in Tables 4.4.1 and 4.4.2 are only those amcunts -
presently on site. The chemicals listed do not represent limits or constraints on types or amounts within content of the facility license and are subject tu change with needs of the plant.
i i
1 9
i
.Page-;50 of.181'-
Table 4.4.1 Warehouse Storage of Hazardous Chemicals Qty Description Min / flax Unit Paint, Enamel, Federal, Yellow 2/6 qt Paint, Latex Fire Retardent, Plat. Gray 3/12 gal Paint, Latex. White 1/8 gal Paint, Latex, Blue, IP7 Flat 2/12 gal Paint, Pastel, Santern 2/8 gal Paint, Semi-Gloss Wall & Trim Enamel Light PPG 6-650 3/8 gal Paint, 307 Med. BS LT Blue 3/8 gal Paint, DP Blue, 776 Dry Swift Enamel 3/8 gal Paint, Semi-Glass Wall & Trim Enamel Tan 3/8 gal Paint, Med Gray ELL Vitraguard 5/15 gal Paint, Light Base 808 1/5 gal Paint Remover, Dupont, "Expidite" 3/10 qt Paint Remover & Varnish, Stripper 2/4 gal Paint, Spray, Temp 524, Flat Black 6/24 gal Enamel Black Gloss Spray Interior / Exterior Fast Drying 3/12 gal Paint, Spray, Purple Passion or Equiv.
6/36 13 oz Paint, Spray White, for Stenciling of Equip. & Material 6/36 13 oz Primer, Rust-lio, Morris #85 Red 1/3 gal Varnish, Poly-Urethane Clear Interior / Exterior 1/4 gal Reducer, Morris Climatic 1/6 gal Cleaner, Barsol 140 1/4 55 gal Cleaner, Electrical, Lix No.1 20/55 gal Compound-Lapping Vehicle Oil Base 10/30 gal Compound, Para-Bond #1001 Gray 2/6 gal Detector, Leak Snoop Liquid., for Nuclear I 18/72 8 oz Sealer-Clear Crete-D, Indu'.,: rial 5/20 gal Thinner, Epoxy 1/4 gal Preservative, Pentachlordphenol 20/50 gal Fluid, Cutting, "Coolsiaol" 1/6 gal Fluid, EHC, G.E. Ma'.erial 4/6 55 gal-Oil, Cutting Thre..d-Ezy 2/10 gal Grease, Moly 41 5/10 lb Grease, Lubricating Texaco, Marfac Multi Purpose 2 '
12/36 14 oz Grease, High Temo Chevron SRl-NLG1 Grade 2 3/12 14 oz Lubricant Almaguard #3752 1/3 120 lb Lubricant Almapex #1275 10/50 14 oz Acetone, Rea. Mall. 2443 4/10 gal Acetone, Tech, Grade, Mid. Sci. #All5 10/15 gal Acetylene, Code 4301,-Size 5 3/3 133 scf Acetylene, (small) Prod. Code 4603 2/3 160 scf Acic, Acetic Glacial, Rea. Mall. 2504 2/4 gal Acid, Nitric, Rea. Mall.1409 6/12 gal Acid, Hydrochloric 3/6
_6 lb Alchohol-Absolute, Rea. Mall. Ploy Jug. 7019 1/8.
gal Alchohol, Iso-Propyl (Tech).
12/32 gal Argen, High Purity, 99.999"...
2/6
.330 scf Argon, Aero Gas, 99.998", Purity.
3/6 330 scf Argon, Prod. Code 5401
.2/5 330 scf Page.51-of:181.
1
1-t Table 4.4.1 (cont'd)
Qty Descriotion Min / Max Unjijt Argon Code 5402 1/3 330 scf Carbon Tetrachloride AR (Low Sulfur) liall Code 2/5 8 pt Chlorine Gas 1/3 150 lb Glycerol Rea. J. T. Baker 2136 Mall. 509 -
2/4 gal Helium Gas, Size A, 99.995% Purity 1/1 286 scf.
Insecticide Certi-Mist 9240 5/35 gal Methane,10%, Argon 90%, (P-10) 4/8 300 scf Methane Gas, 99% Pure 2/4 300 scf Methanol, Absolute Rea. Mall 3024' 2/8 gal Oxygen (Large) Prod. Code 1101 4/10 150 scf Oxygen (Small) Prod. Code 1102 2/4 80 scf' Petroleum Ether AR 30-60 1/5 gal 2-Propanol, Rea. Baker 9084 3/10 gal
't 8
'l 1
Page 52Loff181
Table 4.4.2 On-site llazardous Materials Storage Distance From Number Release Point to R1terial Chemical State Heans of Storage Usage and quantity Toxicity Limit Cont. Rm Air Intake (mg/m3 @ STP)
(Distance in feet)
' Liquefied Gas
' Cylinders Feedwater 5-150 lb 45 3,30 treatment cylinders Sulfuric Acid Liquid @ STP Above-ground tank Ion-exchange 225,000 lbs 2
330 (sp.gr.l.83:
regeneration 98%' conc.)
liydrogen Pressurized gas Cylinders Generator 54-302 scf.
Asphyxiant 330 operations cy1inders Nitrogen Pressurized gas Cylinders Cover Gas13-300 scf Asphyxiant 330 cylinders i
N;6rogen Pressurized gas Large Cylinders Cover Cas 6-cy11nders Asphyxiant 180 l
(18 ft.long; (pressurized 2 ft. diamater) to 1000 psi)
Argon' Pressurized gas Cylinders Instrument 6-330 scf Asphyxiant 330 Calibration cylinders Carbon Dioxide Pressurized gas Cylinders Fire 6-213 scf 1,840 Discharges within Protection cylinders Control Hoom Methane.
Pressurized gas
. Cylinders Maintenance 10-300 scf Asphyxiant 330 Operations cylinders Liquefied Pressurized Cylinders Chemistry 8-300 scf 3,600 48 y
l'etroleum Cas Liquid 11boratory cylinders i - (n Acetylene Pressurized gas Cylinders
.Haintenance 4-133 scf Asphyxiant 330 Operations cylinders 63 Oxygen.
Pressurized gas. -Cylinders Maintenance 10-150 scf 330 cn Operations cylinders lielium Pressurized gas Cylinders Chemistry
'4-286 scf Asphyxiant 48 Laboratory eylinders
4 Table 4.4.2 (cont'd)
On-site llazardous liaterials Storage Distance From Number Helease Point to tiat erial Chemical State Means of Storage Usage and Quantity Toxicit M imit Co n t-. Rm Air Intake (mg/m3 G STP)
(Distance in feet)
Boiler 011 Liquid @ STP Underground tank Auxiliary 18,000 gal.
Asphyxiant 360 Boi1er Fuel
'Wute 011 Liquid @ STP 55 gal. drums N/A 60-55 gal, drums Asphyxiant 500 Diesel Oil Liquid @ STP Above-ground tanks Maintenance 900 gal.
Asphyxiant 620 Operations
'Caroline Liquid @ STP Above-ground tank Fuel for on-300 gal.
3,600 575 site vehicles 90% Argon /
Pressurized gas Cylinders Instrument 3-330 scf Asphyxiant 48 10% tiethane Calibration cylinders llalon 1301 Liquefied gas Cylinders Fire 1-131# cylinder 4.3x10 Discharges with (with Nitrogen Frotection 6-271# cylinder Control Room
@ 360 psi) 7-280# cylinder Ilydrazine Liquid @ STP 55 gal. drums Primary 2-55 gal. drums 2.6 240 (35% concentra-conlant tion) additive Sodium Liquid @ STP Tank in pit lon exchange 190,000 lbs.
4 240 Ilyd roxide (up.gr.1.52:
regeneration 50% conc.)
Aluminum Dry powder 50 lb. bags Feedwater 10-50 lb. bags 240
- .o Sulfate Treatment a
. Potash Dry powder 50 lb. bags Feedwater 5-50 lb. bags 240 Treatment o
. Limestone Dry powder.
50 lb. bags Waste-binding 5-50 lb. baga 20(as airborne 240 operations particulate)
Propane Pressurized gas-Cylinders Maintenance 4-300 scf Asphyxiant 330 Operations cylinders
4.5 Off-Site t!anufacturing, Storage and Transportation Facilities of Hazarcous Chemicals Identification of potentially hazardous materials stored or transported within a five-mile radius of the plant was performed tnrough a compre-hensive survey of the area. Applicable Federal, State, County and Local Government agencies were surveyed along with the industrial sector.
Interviews were conducted with responsible personnel at each identified facility. Descriptions of the nature and extent of the activities conducted at the-various facilities, the distances from the point of location to the Centrol Rocm air intake and the amount and chemical form of the hazardous materials employed have been obtained. The transportation mechanisms which have been surveyed include barge, pipeline, raiiroad, airline, and vehicular traffic.
Characterization of the shipping medium includes the cicsest distance of approach to t'le Control Room air intake, the frequency of transport, the quantity, and the chemical form of' the material.
4.5.1 Offsite Manufacturing and Storage Facilities A total of nine major permanent offsite storage facilities were identified within the five-mile radius. A description of each is provided in this section..
Page 55 of 181
h Liquid anhydrous ammonia and liquid nitrogen fertilizer are stored and transported from the Agrico Chemical Company. The storage capacity for the liquid anhydrous ammonia is two tanks that hold 25,000 tons each and one tank that holds 35,000 tons. These three tanks are refrigerated, have double walls, 2 feet of insulation, concrete floors supported by pilings 80-90 feet deep and are located above ground.
The ammonia is held at -32 F and a pressure 0
of less than l ~ psi.
There are also two tanks of 30,000 gallons each which are kept at 250 psi.
P, elative storage amounts of liquid anhydrous ammonia are seasonally dependent.
The company receives the ammonia via two pipelines and distributes the product via truck, rail, pipeline and occasionally barge within an approximate 200-mile radius of the storage location. The two pipelines that feed the storage plant are the Mid-American Line and the Gulf Central Line. Barge, truck, and rail are sometimes utilized to transport Lae annonia to the storage location. Agrico Chemical servic' Jpproximately 180 truc!s per day with an average load capacity. of approximately 25 tons per. truck.
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Page 56 of 181
= -
The Agrico Chemical Company is located 1.8 miles northwest of the U
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Fort Calhoun site, 320, clockwise from north.
There are two propane storage facilities in the area, Fairway Propane and Rogert 011 and Propane.
Fairway Propane stores approximately 18,000 gallons of liquid propane in one above ground tank. The tank is equipped with a safety relief valve, set at 200 psi, to guard against overpressurization. Distribution is in the local area and amounts to 10,000 gallons per day in the winter and 10,000 gallons per week in the summer.
Distribution is via truck with a maximum lead capacity of 2,300 gallons. This company receives its supplies via Nebraska Route 133 and U.S.30 in 8,500-9,000 gallon shipments. They also store 30,000 gallons of anhydrous ammonia in two tanks. One tank holds 18,000 gallons and the other tank holds 12,000 gallons.
These tanks will be close to capacity in the spring and of low capacity during the remainder of the year.
These tanks are also protected by relief valves. set at 200 psi.
Fairway Propane is located 2.4 miles from the Fort Calhoun site, 3350, clockwise from north.
Page 57 of 181
Rogert Oil and Propane stores a maximum propane capacity of 18,000 gallons within one tank which is protected by a pressure relief valve set at 300 psi. Normal pressure during the winter is 50 psi and in the summer 200 psi. The company usua'ly operates between half full and full tank capacity.
Distribution is approximately 1,000 gallons per day in the local area, winter loads usually being heavier than sumer loads.
Rogert is supplied with propane via trucks traveling on U. S. 30 in 9,000 gallon shipments.
Rogert also maintains a 20,000 gallon supply of fuel oil. Tanks are diked in case of spills and have shut off valves in case of fi re. Distribution averages 500 gallons per day in the sumer and 2,000 gallons per day in the winter.
Restocking of fuel oil comes via U. S. 73 in 7,200 gallon shipments. The Rogert facilities are located on two separate sites.
Location #1 is 4.3 miles west-northwest of the Fort Calhoun site and has 6 tanks for gasoline and fuel oil. Average year-round capacity is approximately 13,000-15,000 gallons.
Location #2 is 3.6 miles west-northwest of the Fort Calhoun site and contains the 18,000 gallon ti.nk of liquid propane.
The Fort Calhoun Stone Company transports limestone and agricultural lime processed at their site. Approximately 20 tons of explosives-are stored at the east quarry.. There is also a 20 ton storage 0
capacity at the west quarry located 3.1 miles,142 clockwise from the Fort Calhoun site. Minimum quantities are stored there except during the peak quarrying period in the spring.
The explosive used is Iremite with the brand name Irecc. They are supplied by truck Page 58 of 181
via U. S. 73 in 20 ton shipments.
Their proce:";res comply witn Mine Safety & Health Regulations as well as applicable state regulations.
Anhydrous ammonia is stored and distributed by the Washco Feed and Supply Company. fiaximum storage capacity is 20 tons in one above ground tank equipped with a safety relief valve. Storage amounts are seasonally dependent, with more in the spring and less in the fall. Approximately 200-300 tons are used between April and May.
They receive and distribute by truck.
The Washco Feed and Supply Company is located 3.2 miles from the Fort Calhoun site, 306o, clock-wise from north.
Fuel oil and gasoline are stored and distributed by the Taylor Oil Company. They have four above ground tanks, two of which hold 8,000 gallons each, and two which hold 6,000 gallons each. Annual average capacity is approximately 3/4 full. Shipments are in 8,000 gallon quantities and distribution is made by truck.with a maximum capacity of 9,300 gallons. The Taylor Oil Company is located 3.7 miles from the Fort Calhoun site, 2960, clockwise from north.
Fuel oil and gasoline are stored and distributed by Bohs Service (Conoco). Bohs can store a maximum of 35,000 gallons in five above ground tanks, however, they only have 10,000 to 12,000 gallons on site at any one time.
All tanks are equipped with safety. relief valves.
Supplies are received via truck in 8,000 gallon shipments.
Distribution is by tank truck with a maximum capacity of 1,200 gallons, but is usually only half. full. Bohs 0
Service is located 3.5 miles from the Fort Calhoun site, 291, clockwise from north.
page 59 of 181
Diesel fuel and gasoline are stored under ground by Foley 011 (Mobil Service Station). Approximately 100-200 yards from the service station, gasoline, diesel fuel and heating oil are stored. Foley 0
011 is located 3.1 miles from the Fort Calhoun site, 293, clockwise from north.
4.5.2 Pipelines There are six pipelines in the five mile radius area. The Northern Natural Gas Company supplies the city of Blair. This line operates at a maximum pressure of 500 psig. Natural gas is carried and no other gases or fluids are planned.
Installation was completed in 1931, and the 6-inch line was buried at an average depth of 4 feet. There are four valves in the line, the nearest isolation valve is located 11.32 miles west of Blair. The closest point of approach to the Fort 0
Calhoun site is located 4.4 '.siles, 299, clockwise from nortt..
The Mobil Pipeline installed in 1942 traverses the five mile area, but
~
has no local outlet.
The 6-inch pipe is buried to a depth of 30 inches and is rated at.1440 psig maximum test pressure.
This line carries refined petroleum.
It is not anticipated that the line will be used for any other fluid or gas. The line is never operated at higher than the normal pressure of 700 psig. There are three isolation valves in the line, the nearest one is located on highway 73-75'less than 2 miles from the plant site. The closest point of approach to the Fort Calhoun site is located 1.3 miles, 2700,. clock-wise from north.
Page 60 of 181
The Williams Brothers Company maintains one 8-inch and one 12-inch line in the five mile radius area.
These lines carry refined petroleum products and ageous fertilizer solutions, depending on the time of year, fertilizer is in greatest demand during the spring.
The 12-inch line was installed in 1950 and the 8-inch line was in-stalled in 1946.
The 12-inch line operates at a 1300 psig pressure and the 8-inch line at 1150 psig. Neither line operates at a higher than normal pressure for storage nor is it anticipated that any other fluid or gas will be used in the lines..Both lines have gate-type isolation valves. The closest point of approach for either of the Williams Brothers Company pipelines to the Fort Calhoun site is located 0.7 miles, 2620, clockwise from north.
The Mid-American Pipeline was installed in 1971. This is a 4-inch line buried to a depth of 42 inches. The pipeline is rated at 1,320 psi but usually operates at.a pressure of 700 psi or less. This line carries liquid anhydrous amonia to the Agrico Chemical Company.
It is not anticipated that the line will be used for any other fluid or gas. -The closest point of approach to the Fort Calhoun site is located 1.8 miles, 320, clockwise from north.
The 6-inch Gulf Central Pipeline, which carries liquid anhydrous ammonia, was installed in 1969 and buried to a-depth.of 36 inches.
This line operates at a maximum pressure of 1,440 psig (usually lower) and supplies Agrico Chemical Company.
It is r.ot anticipated' that the line will be used for any other fluid or gas. The closest point of approach to the Fort Calhoun site is located
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t 1.8 miles, 320, clockwise-from north.
Page 61 of.181
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4.5.3 _ Barge Traffic The plant site is located adjacent to the Missouri River with the con-denser cooling water intake structure located on the river bank.
There is seasonal barge traffic on the Missouri.
During the 1979 Navigational season, the following material was shipped between Sioux City and Rulo; (1) chemicals and related products (Benzine, paints, fertilizer), 363,752 tons (2) non-metallic minerals (salt and phosphate rock), 92,451 tons (3) there were no petroleum products transported via barge during 1979, but for the 1980 Navigational season, up to May 30, 1,302 tons of petroleum products had been shipped.
The navigable river channel passes in front of the safety-related raw water intake structure with barges passing as close as 75 feet from it.
The channel width averages about 300 feet with a total river width of about 600 feet. The river configuration inherently limits the type of traffic on it. The barges are as large as 50 feet by 295 feet, and there may be as maay as four of these in a tow, and two tows per group, depending on whether they are being moved upstream or downstream, or if they are empty or full. The type of towboat utilized depends on the load, with units employed as large as 3000 to 4000 horsepower. There is no limi towever, and bigger boats may be used.
There are four barge terminals in the five mile radius area. The Fcrt Calhoun Stone Company terminal is used for the loading of rock for river channel stabilization and is located five miles from the plant at 1340, clockwise from north. The Consolidated Blenders teminal is used for the loading of processed alfalfa products and is located 2.2 Page 62 of.181
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miles from the plant at 3320, clockwise from north. Anhydrous ammonia is infrequently shipped and received at the Agrico Chemical Company teminals.
One terminal-is located 1.8 miles 0
from the plant at 320, clockwise from north, and the other is 0
2.5 miles away at 337, clockwise from north.
The last time either Agrico terminal received anhydrous ammonia wa., in 1976. The last time they shipped anhydrous ammonia out by barge was in 1979, and that was only one shipment.
The barges used to ship the anhydrous arrnonia are 50 feet wide and 295 feet long and are 'of steel construction.
Each barge is equipped with two storage cylinders in its cargo wall.
Each cylinder is 18 feet 9 inches in diameter and is capable of holding 497,250 gallons of anhydrous ammonia. The liquid anhydrous ammonia is normally carried at a*.mospheric pressure and -280F. However, the storage cylinders are designed for a pressure of 50 psig, are pressure tested to 75 psig and are equipped with four renndant safety valves set at 10 psig to prevent overpressurization of the tanks.
,Page 63 of 181
i 4.5.4 Aircraft Transoort The nearest airport of significance to the plant is the Eppley 0
Airfield, which is located 17.8 miles from the site at 147 from riorth. This airport has three concrete runways, the primary runway, 14R/32L being the longest. Large commercial and light aircraft use the airport. The Eppley Airfield outer marker route is utilized for approaches only, there are no holding patterns in the 5 mile radius of the plant.
There are three Victor air corridors within the 5-mile radius (159, 159W, and 172) which handle aviation traffic through the area. Each corridor is eight nautical miles wide. These infrequently travelled routes average 20 flights per day for V159, 1 flight per day for V159W (which handles overflow from V159), and 5 to 6 flights per day for V172 (which goes toward Fremont). Two of the corridors, V159W and V172 pass over the 0
immediate plant site area; V159W runs 230 at 0 miles and V172 0
runs 173 at 0 miles. Victor air corridor 159 has its closest approach at 90 and approximately 3.3 miles out. Victor air corridor 159 is a florth-South route between Omaha and Sioux City / Sioux Falls. Radar Approach Control Facility at Offutt Air Force Base controls this traffic as well as the overflights-up to 10,000 feet. Altitudes of operation in'the air corridors will depend on weather conditions and traffic congestion, but generally iall between 3,000 feet and 10,000 feet. A-landino field also exists at Blair for VFR (Visual Flight Rules) flights only. The Blair Intersect _is used by 2 to 3 planes per day.
Page 64 of 181
W Offutt Air Force Base, the nearest military installation,is located 27.8 miles out at 1630 from north. This base is the headquarters station of the Strategic Air Command.
No bomber aircrafts are based at Offut Air Force Base, only tanker, reconnaisance, and general service aircraft.
Federal Regulation limits the identity and quantity of potentially hazardous materials which may be transported by aircraft. Title 49 of the Code Of Federal Regulations provides a detailed list of the maximum quantities of potentially hazardous materials which may be transported by aircraft.
This list, rather than specific flight invoices, was employed as the basis for this review.
Thus a complete and conservative representation of the -
materials transported by aircraft was employed.
(
Page 65 of 181
4.5.5 Railway Transport There is one active railroad route within the 5-mile radius. This route is owned by the Chicago and florthwestern Railway and travels primarily east to west from Norfolk, Nebraska to Missouri Valley, Iowa, through Blair Nebraska. This route handles two local freight trans pwer day in addition to a number of transconinental trains.
This track has it's closest point of approach at approximately 2.2 miles from the plant site.
An existing north / south route traveling from Omaha, Nebraska, through Blair, Nebraska has recently been abandoned and is given no further consideration in this report.
Table 4.5.5.1, Section A, provides a list of hazardous commodities shipped past the site in a 12 month period on the transcontinental trains of the Chicago and Northwestern railway. Many shipments were c.arried in " piggy-back" cars which are semi-trailers loaded onto flat railway cars. Many of these trailers carried a
" Hazardous Commodity" designation, without specifying the type or amount of the commodity. When individual commodities were listed, weights were seldom given. Within Table 4.5.5.1, an entry such as:
Butane Flamable Gas 2TR means that in the course of one year, 2 trailers containing Butane were found, but no weights were given. There may have been only a small amount on each trailer, the rest of the trailer being given to other cocredities. A trailer labeled " Dangerous Commodity - Corrosive", may contain as little as one gallon of
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Page 66 of 181
l corrosive material. Weights given are the net weights of the commodity.
Table 4.5.5.1, Section B, provides a list of the hazardous materials shipped on the Chicago and Northwestern Railroad, on the two daily local trains through Blair, fiebraska. These items were shipped during a 12 month period, on the two daily trains which travel between Missouri Valley, Iowa and florfclk, Nebraska. Many times the items were being shipped in semi-trailers loaded onto flat cars.
In such cases, weights for individual connodities were seldom available. The weights given are the net weights of the connodity.
Page'67'of.181
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Table 4.5.5.1 Annual Hazardous Material Transported by Freicht Train Section A: Chicago - Northwestern Transcontinental Transport Number and Type of Material Description Shioments Quantity Acetone Flammable Liquid 6 TR Flammable Liquid 1 TR 136 pounds Flammable Liquid 1 Tankcar 59 tons Acids (types not Corrosive 15 Tankcars 1,073 tons specified)
Corrosive 9 Boxcars 428 tons Corrosive 1 0F Boxcar 54 tons Corrosive 1 Covered Hopper 118 tons Corrosive 1 TR Corrosive 1 Tankcar 97 tons Acids:
Acetic (glacial)
Corrosive 1 TR 50 pounds Hydrochloric Corrosive 6 TR Hydrochloric Corrosive 1 Container Methyl Butyric Corrosive 1 TR Monochloracetic Corrosive 1 Container Phosphoric Corrosive 3 TR Phosphoric Corrosive 1 TR 114 pounds Phosphoric Corrosive 1 TR 114 pounds Phosphoric Corrosive 1 TR 76 pounds Phosphoric Corrosive 1 Covered Hopper Phosphoric Corrosive 2 Tankcars Not Available Nitric Corrosive 1 TR 58 pounds Nitric-Corrosive 1 TR 62 pounds Nitric Corrosive 1 TR Sulfuric Corrosive 1 TR 100 pounds Sulfuric Corrosive 1 TR Acrylonitrile Flammable Liquid 1 TR Adhesives Dangerous Liquid 1 TR Alkali No Information 1 Tankcar 27 toas Liquid Alkaline Hazardous Commodity 1 TR Amine Alcohol Flammable Liquid 57 Tankcars 4,303 tons Alcohol Flammable Liquid 5 TR Alcohol Flammable Liquid 1 Boxcar Not Available Alcohol Flammable Liquid 1 0F Boxcar 54 tons
- Semi-trailers loaded onto flatcars Page 68 of_181 j
Table 4.5.5.1(cont'd)
Number and Type of Material Description Shioments Ouantity Ammonia Hazardous commodity 2 Tankcars 145 tons (Anhydrous)
Armonia Hazardous commodity 2 Tankcars 196 tons (Anhydrous)'
Ammonia Hazardous commodity 1 Covered hopper 92 tons
@nhydrous)
Ammonia Hazardous commodity 1 Tankcar 92 tons (Anhydrous)
Ammonium Corrosive liquid 1 TR Hydroxide Ammoniun Nitrate Oxidizer 3 Covered hocoer 247 tons Ammonium Nitrate Oxidizer 1 TR Asbestos No Information 1 Covered hopper 49 tons Batteries Corrosive 22 TR Benzene Combustible liquid 2 Tankcars 74 tons Benzoyl Peroxide Hazardous commodity 1 TR 42 tons Benzoyl Peroxide Hazardous commodity 1 TR Bromide Poison 1 TR Bromide Cyanogen Poison 1 TR Methyl Bromide Toxic 1 Tankcar 57 tons Boiler cleaner Combustible liquid 1 TR Butane Flammable gas 2 TR Butylamine Flammable liquid 1 TR Carbon No Information 4 Tankcars 376 tons Carbon Dioxide Non-flammable gas 1 Tankcar 94 tons Carbon Dioxide Non-flammable gas 20 TR Carpet Cement Flammable liquid 2 TR Caustic Soda Hazardous commodity 1 Tankcar 47 tons Caustic Soda Hazardous' commodity 4 Tankcars 124 tons Caustic Soda Hazardous commodity 1 Tankcar 42 tons Cyanides:
Cyanide Poison B 1 TR 70 pounds (plating solution)
Cyanide Dry mixture 1 Container 1100 nounds Cyanide Poison 1 TR Cyanogen Poisonous liquid
-1 TR Copper Cyanide No Information 1 TR 1744 pounds Zinc Cyanide No Information 1 TR 90') pounds Sodium Cyanide Poison S 1 TR Sodium Cyanide Poison B 1 TR 25 pounds
. Sodium Cyanide Poison B 1 TR Chlorides No Information 16 Tankcars 1013 tons Chlorides No Information
_5 Boxcars 222 tons-Chlorides No Information 5 Covered hopper 336 tons Chlorides
-No Information 1 Covered hopper 63 tons Chlorides No Information 1 Hopoer-open 99 tons top Cement Flammable liquid 1 TR Cement F1ammable liquid
-1 TR Chloroform No Information.
~1 Covered hopper 94 tons Dichlorofluoro-Hazardous commodity 1 Tankcar 143 tons-methane
- Semi-trailers loaded onto flatcars Page 69 of 181
Table 4.5.5.1 (cont'd)
Number and Type of Material Descriotion 55fements Guantity Ether Hazardous co=odity 2 TR Ether Hazardous comodity 1 TR S counds Ether base fuel Hazerocus comodity 1 TR Ethanol Hazardous comodity 3 TR Ethylene Glycol Hazardous comodity 1 Tankcar 32 tons Ethylene Flamable gas 3 Tankcars 200 tons Ethylene Flamable gas 2 Tankcars 159 tons Ethylene Flamable gas 7 Tankcars 623 tons Ethylene Flammable liquid 1 TR Explosives:
Amunition Small Arms Powder 1 Container Ammunition Small Arms Powder 1 TR Amunition Hazardous comodity 1 TR Flares Class C Explosive 10 TR Fuses Flar=able solid 1 TR Fertilizer No Information 43 Covered hopper 4493 tons Fertilizer No Information 4 Covered hopper 369 tons
' Tankcars 195 tons Z
Fertilizer No Information Fertilizer
% Information 1 DF Boxcar 53 tons Compressed gas Hazardous comodity S TR Comoressed gas No Information 1 Container Compressed gas No Information 1 TR 325 pounds Compressed gas No Information 1 TR Gas No Information 1 Tankcar 67 tons Glycol Fla mable liquid 1 Tankcar 67 tons Hexane Fla:r:rable liquid 5 Tankcars 90 tons Hexane Flar=able liquid 2 TR Hydrogen Peroxide Oxidizer 7 Tankcars 339 tons Hycrogen Hazardous cor=odity 1 Tankcar 47 tons Hydrogen Hazardous comodity 2 Tankcars 69 tons Insecticide Flamable liquid 1 TR Ink Flamable liquid 6 TR Liquor (distilled) Hazardous comodity 1 Tankcar-87 tons Lacquer Thinner Flammable liquid 1 TR 100 counds Lacquer Corrosive liquid 1 TR Lacquer Flammable liquid 1 TR 450 pounds Lacquer (solvent)
Flammable liquid 1 TR 450 pounds Linseed Oil Combustible liquid 1 Tankcar
- 27. tons Linseed Oil Combustible liquid 1 Tankcar 27 tons Methylisocyanates Fla mable liquid 1 Tankcar 50 tons Methylisocyanates Flammable liquid 3 Tankcars 144 tons Methanol Flammable liquid 2 TR Methanol Flammable liquid 1 TR 72 pounds Pethanol Flamable. liquid 1 TR 93 pounds Methyl Ethyl Flamable liquid 1 Container Ketone Mathyl Ethyl Flammable liquid 1 TR.
15 pounds Ketone Methyl Ketone Flamable liquid 1 TR 360 counds Methyl Ethyl Flamable liquid 1 TR
- Semi-trailers loaded onto flatcars.
Page 70 of 181
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r
Table 4.5.5.1 (cont'd)
Number and Type of Material Description Shioments Ouantity Methyl formamide Hazardous commodity 1 TR 88 counds Methlyacetylene Hazardous commodity 2 TR Propadiene Stabilizer Naptha Flammable 1 TR Naptha Flammable 1 Tankcar 37 tons Nitrates Oxidizer 3 Boxcars 151 tons Nitrogen No Information 3 Tankcars 96 tons Nitrogen Dangerous liquid 3 Boxcars 96 tons Nitrites Hazardous commodity 2 TR Cobalt Nitrate Oxidizer 1 TR 0xygen Non-flammable gas 15 Boxcars 1127 tons Oxygen Non-flammable gas 6 Tankcars 449 tons Oxygen Non-flammable gas 1 TR 0xygen Liquid 2 Boxcars 147 tons Paint Flammable liquid 14 TR Peroxide Hazardous commodity 6 Tankcars 326 tons Potassium Caustic 2 TR Hydroxide Petroleum Products No Information 55 Tankcars 3438 tons Petroleum Products No Information 12 Tankcars 922 tons Petroleum Products No Infcrmation 5 Tankcars 231 tons Propane, L.P. Gas Hazardous commodity 21 TR Plating solution Poison 1 TR 14 pounds Poison Flammable & Poison 1 Tankcar 32 tons Poison Flammable & Poison 1 TR Poison Flammable & Poison 1 Tankcar 32 tons Poison B Flammable & Poison 1 TR Poison B Flammable & Poison 1 Tankcar 50 tons Poison B Flammable & Poison 1 Tankcar 57 tons Resin Flammable Liquid 3 TR Resin Flammable Liquid 1 Container Silver Ni+. rate Oxidizer 1 TR Sodium Hyfroxide Corrosive 5 TR Sodium Hydroxide Corrosive 1 Container Sodium Hydrogen Caustic 6 TR Sul fate Sodium Hydrogen Caustic 1 TR Sodium 0xidizer Hazardous commodity-1 Tankcar 68 tons Sodium Potassium Hazardous commodity 5 TR alloy Sodium Hazardous commodity 4 TR Sodium Salicylate Hazardous commodity 1 TR Sodium Sulphydrate Hazardous commodity 1 Tankcar 42 tons Sodium Monoxide No Information 1 TR Sul fur Hazardous commodity 1 Container Sulfur Hazardous commodity
'l Boxcar.
57 tons Syruo Flammable liquid.
1 TR Titanium Hazardous commodity 45 Tankcars 2838 tons Tetrachloride Titanium (metal)
No Information 1 Soxcar 47 tons
- Semi-trailers loaded onto flatcars Page 71 of 181
Table 4.5.5.1 (cont'd)
Number and Tyoe of Material Descriotion Shinments Ouantity Triazenetrion Oxidizer 1 TR 70 pounds Toluene Flammable liquid 1 Tankcar 22 tons Toluene Flammable liquid 4 TR Vinyl Toluene Hazardous commodity 1 Tankcar 65 tons Xylene Flammable liquid 5 TR Miscellaneous Fire Extinguishing No Information 1 TR Chemicals Calcium (metal)
No Information 1 TR 7 pounds Solvent Flammable liquid 1 TR 32 pounds Ferric Chloride No Information 1 TR Solution Cleaning Cot. pounds Hazardous commodity 17 TR Book Matches Hazardous commodity 1 TR Cigarette Lighters Hazardous commodity 2 TR Starting Fluid Flammable liquid 2 TR Section B: Chicago-Northwestern Local Transport Number and Type of Material Oescriotien Shioments Ouantity Acid Corrosive 9 Tankcars 547 tons (types not specified)
Phoschoric Acid Corrosive 1 Tankcar 97 tons Phosphoric Acid Corrosive 1 Covered hopper 99 tons Alcohol Flammable liquid 27 Tankcars 1724 tons Ammonia (anhydrous)Hazardouscommodity 40 Tankcars 3406 tons Ammonia (anhydrous) Hazardous commodity 11 Covered 1.040 tons hoppers Ammunition Hazardous commodity 1 Covered hooper 100 tons Ammunition Hazardous commodity 1 Boxcar 20 tons Batteries Corrosive
.13 Boxcars 431 tons Calcium Carbide Hazardo9s commodity 6 TR Caustic Soda Corrosive 3 Boxcars 94 tons Caustic Soda Corrosive 2 Boxcars 99 tons Caustic Soda Corrosive' 2 DF Boxcars 98 tons Caustic Soda Corrosive 1 Covered 54 tons-hopper Chlorides No Information 15 Tankcars 1452 tons Chlorides No Information 2 Covered hoppers 150 tons Chlorine Hazardous commodity 8 Boxcars
.Not Available Ethylene Flammable gas 1 Tankcar
. Not Available Fertilizer Hazardous commodity 69 Tankcars 5662 tons-Fertilizer Hazardous commodity 10S Covered 8750 tons-hooner-Fertilizer Hazardous commodity 13 Boxcars
.735 tons Fertilizer Hazardous commodity _
1 DF Boxcar 57 tons Fertilizer Hazardous commodity 5 Tankcars 437 tons.
Fertilizer Hazardous commodity 52 Covered 4299 tons.
hoppers.
- Semi-trailers loaded onto flatcars Page 72 of 181
i Table 4.5.5.1 (cont'd)
Number and 4
Type of Material Description Shipments Ouantity I
Insecticide Flammable liquid 1 DF Boxcar 39 tons Insecticide Flammable liquid 1 Tankcar 93 tons Kerosene Combustible liquid 2 Tankcars 84 tons Methane Hazardous commodity 4 Tankcars Not Available Methanol Flammable liquid 1 Boxcar 47 tons Methanol Flammable liquid 1 Tankcar 57 tons Methyl No Information 3 Tankcars 253 tons Methol Hazardous commodity 1 Tankcar 95 tons Napthalene Flammable liquid 3 Tankcars 165 tons Oxygen Non-flammable gas 17 Tankcars 2137 tons Oxygen Non-flammable gas 3 Boxcars 223 tons Paint Flammable liquid 4 Boxcars 113 tons Paint Flammable liquid 1 Boxcar 52 tons Petroleum Products: No Information 155 Tankcars 11,557 tons (Propane, L.P. Gas, Gasoline, Fuel 011)
Petroleum Products: No Information 96 fankcars 5146 tons Lube Oil No Information 3 DF Boxcars Not Available Lube Oil No Information 4 Boxcars Not Available l
Page 73 of 161
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4.5.6 Vehicular Transport Two US and two state highways carry traffic within the 5 mile radius: US 30, US 73, State 133, and State 91. Estimates of present-day traffic riows are indicated in Table 4.5.6.1.
Truck traffic is approximately 10 percent of the total. Of the highways described above, US 73 is nearest to the site (0.7 miles).
US 73 is infrequently traveled, and it is expected that the nature of the materials transported by truck would be mainly agricultural products or similar to that transported by rails.
No shipments of hazardous materials in large quantities were reported by those chemical manufacturing or distributing companies in the Omaha metropolitan area which were surveyed. Table 4.5.6.2 provides a listing of distributors who utilize the four highways near the plant and the accompanying amounts of cargo and shipment frequencies. Hazardous shipments of similar nature to that transported by railroad may be shipped by trucks in smaller quantities.
i Page.74 of 181
Table 4.5.6.1 Traffic Flow or. Major Hichways Within A 5-Mile Radius of the Fort Calhoun Station Unit 1 Site Route Averace D-lily Total Traffic Averace Truck Traffic US 30 5,330 vehicles / day 850 trucks / day US 73 6,180 vehicles / day 425 trucks / day State 133 3,370 vehicles / day 340 trucks / day State 91 1,455 vehicles / day 140 trucks / day Page 75 of 101
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4
_Ipb}eA,LC 2 Potentially Hazardous.*;oterial Tran; ported by Truck Past the f ort Callioun Station Unit I Site Distributor Material Quantity Shipment Frequency Route Central Trasportation Co.
Anhydrous Annonia 17 tons / load 500 shipments From Blair (Agrico) to Norfolk, Nebraska (pressurized gas)
(April-May) all points in Nebraska George Brothers Inc.
Liquid Fertilizer 17 tons / load 6 shipments From Blair (Agrico) to Sutton, Nebraska per year all points in Nebraska Petroleum Transport Anhydrous Ammonia 17 tons / load 100 shipments From Blair (Agrico) to Service (pressurized gas)
(April-May)
Iowa Council Bluffs, Iowa Wynne Transport Serv.
- 1) Anydrous Ammonia
- 1) 17 tons / load 400 shipments From Blair (Agrico) to Omaha, Nebraska.
- 2) Gasoline
- 2) 7,400 gallons / (total) all points in Nebraska load Ward Transport Inc.
- 1) Jet Fuel
- 1) 7,400 gallons /1) 41 shipments per
- 1) Omaha to Denver i
Pueblo, Colorado load year
- 2) Liquid Fertilizer
- 2) 17 tons / load 2) 15-20 shipments
- 2) Fremont to Blair (Agrico) per year (mainly spring time)
Dilts Trucking Anydrous Ammonia 17 tons / load 220 shipments
- 1) 200 shipments from Blair Crescent, Iowa (pressurized gas)
(April-August) to Iowa
- 2) 20 shipments from Omaha tc Blair
. Petroleum Carriers Co.
Anydrous Ammonia 17 tons / load 30 to 40 shipments From Blair to all points in g
Sioux Falls, S. Dakota (pressurized gas) per year Nebraska
[
Wheeler Transport Service
- 1) Fertilizers:
- 1) 17 tons / load For both materials:
From Omaha to Blair (Agrico)
Omaha, Nebraska (Anydrous Annonia -
1 shipment
- m
. S, pressurized gas)
(Nov.-March)
(Nitrogen solution-1-10 shipments liquid)
(April-0ctober) m
- 2) Gasoline
- 2) 7400 gallons /
load
4.6 Technical Soecifications The applicable plant specific technical specifications have been prepared and implemented to limit environmental conditions in the Control Room under both nomal and post accident operations.
The instrumentation in the reactor protective system and the 0
engineered safeguards system were designed and tested for 120 F.
If the temperature of the Control Room exceeds 120 F the reactor will be shutdown and the conditions corrected to preclude failure of components in an untested envirorcent.
If the Control Room air treatment system is found to be inoperable, there is no immediate threat to the Control Pcom and reactor operation may continue for a limited period of time while repairs are being made.
If the system cannot be repaired within seven days, the reactor is shutdown and brought to a cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The HEPA filters and charcoal adsorbers in the Control Room ventilation system are installed to reduce the potential intake of radioactive iodine to the Control Room. The in-place tests are performed periodically to confim system integrity and perfomance. Laboratory carbon sample testing is performed to demonstrate the met'hyl iodine removal efficiency for expected accident conditions.
Pressure drop testing across the combined HEPA filters and charcoal adsorbers in tne air treatment systems will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.
Operation of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate.
operability and remove excessive moisture build-up on the adsorbers.
If significant painting, fire or chemical release occurs such that the HEPA filters or charcoal adsorbers could become contaminated from-the fumes, chemicals or foreign materials, testing will be performed to confirm system performance.
Page 77 of 181
Demonstration of the automatic and/or manual initiation capability will assure the system's availability. Specifically, the plant technical specifications applicable to Control Room habitability are stated as follows:
(1)
If the c'. trol Room air temperature reaches 120 F, immediate U
action shall be taken to reduce this temperature.
If the temperature cannot be reduced to below 120 F in four hours, the reactor will be placed in a hot shutdown condition.
(2)
A therrNee' er must be in the control room at all times.
t (3) All areas of the plant which have safety related instrumentation will be observed during hot functional testing to determine local temperatures and monitored during operation if normal plant ventilation is not available.
(4) From and after the date that the Control Room air treatment system is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such circuit is sooner made operable.
If these conditions cannot be met, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(5) Equipment and sampling tests shall be conducted as specified in Table 4.6.
The specified intervals may be adjusted to accommodate normal test schedules.except that the interval shall riot exceed 1.25 times the specified interval.
Page 78 of 181-
Table 4.6 Minimum Frecuencies for Lnec<s. Calibrations, And Testing of control Room HaDitability Instrumentation Ana Equipment Surveillance Channel Description Function Frecuency Surveillance Method Control Room Ventilation a.
Test 18 months a.
Check damper operation for DBA mode.
b.
Test 18 months b.
Check control room for positive pressure.
Control Rcom Thermometer a.
Test 18 months a.
Compare reading with calibratedthermogeter.
If not within ! 2 F, replace.
Area Radiation flonitor a.
Check Daily a.
flormal readings observed and internal test signals used to verify instrument operation.
b.
Test Mnnthly b.
Detector exposed to remote operated radiation check source.
c.
Calibrate 18 months c.
Exposure to known radiation source.
Charcoal and HEPA Filters Each refueling shutdown not to In-Place Testing **
exceed 18 months or cfter every Charcoal adsorbers and HEPA 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation or filter banks shall be leak after each complete or partial tested and shall show >99%
replacemr.. of the charcoal Freon (R-11 or R-12) aiid adsorber/HEPA filter banks, or cold 00P particulates re-after any major structural moval, respectively.-
maintenance on the system housing and following significant painting, fire or chemical releases in a ventilation zone communicating with the system.
Frecuency Laboratory Testing **
Charcoal and HEPA Filters Prior to initial loading in filter a.
Initial batch tests of unit.
activated charcoal shall show >90% radioactive methyT iodide removal when tested under conditions of >g % relative humidity, 05
>123,F, 0.05 to 0.15 iEg/m# inlet methyl iodide-concentration and at a face velocity of within 120% of system design.
Page_79 of 181
Table 4.6 (cont'd)
Channel Descriotion Surveillance Method LacoratoryTestinc=*(cont'd)
Charcoal and HEPA filter Each refueling shutdown not to b.
Activateo cnarcoal cells exceed 18 months or after every shall be replaced or 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system nperation tested. The test results and following significant shall show>90% methyl painting, fire, or chemical iodide removal when release in a ventilation zone tested under conditions communicating with the system.
of >oS% relative humidity,
>12dFandwithint20%
of system design face velocity.
Overall System Oceration Ten hours every month.
a.
Eacn circuit snall be operated.
At least once per plant oper-b.
The pressure drop across ating cycle.
the combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 6 inches of water at system design flow rate.
At least once per plant oper-c.
Fan shall be shown to ating cycle.
operate within 1107. design flow.
At least once per plant oper-Automatic and manual initiation ating cycle, of the system shall be demon-strated.
00 Tests shall be performed in accordance with applicable sections of ANSI N510-1975 Page 80 of 181
5.0 ASSUf!PTIONS AND IflITIAL C0tiDITIONS The evaluation of post-accident Control Room habitability re-quires that certain generic assumptions be made with respect to initial plant operating conditions, systems and equipment operational status, and offsite accident response.
This section provides the basic assumptions and initial conditions employed in this evaluation.
(1)
The initial operating condition of the plant is dependent upon the accident under consideration. All systems and equipment identified as vital to mitigation or recovery of the plant from an accident are assumed to be avail-able and in compliance with the plant Technical Speci-fications. The Control Room is assumed to be fully staffed with the ventilation system in the normal mode of operation.
(2)
The initial-indication of the consequences of offsite accidents is received from plant monitoring equipment.
No credit for initial warning from offsite agents is assumed.
(3) tio credit is assumed for mitigation of the consequences of offsite accidents prior to their impact on the plant site.
(4)
The response to accidents relies entirely on the systems -
and equipment available on site. tio credit is assumed for support from outside agents.
Page 81 of--181
i (5)
No credit is assumed for seasonal variations in the transportation frequencies and stored quantities of i
offsite hazardous chemicals.
(6)
Adequate separation is assumed between highly reactive chemicals and catalysts to prevent the production of a species not normally present.
(7)
Maximum anticipated quantities are assumed present in each storage location.
(8)
No credit for radioactive decay is assumed during transport from the release point to the Control Room i
j atmosphere.
i 4
i 1
l 4
.Page 82 of 1811
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6.0 C0!1 TROL ROOM HABITABILITY 'DESIGil EVALUA' TION'AtlD ANALYSIS An evaluation of the Control Room habitability is provided herein with respect to the assumption of a postulated release of airborne hazardous chemicals or a post accident operator exposure to direct radiation and airborne radioactive contam-ination. The evaluation was conducted through the use of (1) on-site inspection, (2) governmental and industrial surveys, (3) the identification of potentially hazardous chemical storage and transport, (4) the identification of potential design basis accidents with radiological consequences, (5) a stochastic and probabilistic analysis of the consequences of postulated ha:ardous chemical accidents, (6) the analysis of post accident control room operator exposure to radiation and toxic chemicals, and (7) an assessment of sustained Control Room habitability systems, equipment, and procedures. This evaluation is the determining factor of the recommendations presented in Section 7.0 on the types of corrective actions needed to facilitate Control Room habitability.
The objective of this design evaluation is to detennir.e the
~
adequacy with which Control Room operators will be appropriately protected against the effects of an accidental release of either.
hazardous chemical or radioactive gas as required by General Design Criteria 19 of 10 CFR Part 50. The acceptance criteria used in the evaluation are stated in Standard Review Plans'(SRP) 2.2.3 and 6.4.
Design basis events are defined in SRP 2.2.3.
as those postulated accidents resulting from the presence of hazardous materials for which the expected rate of occurence Page 83 of 181
_l
^
~
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~ ~ ~ ~ ~ ~ ~ ~ ~
1 for exposures in excess of the established guidelines is
~7 estimated to exceed 10 ever.'.s per year..
SRp 6.4 provides the numcrical guidelines for exposure to toxic or radioactive gases, exposure to direct radiation, and evaluation criteria for systems and equipment.
6.1 Assessment of Hazardous Chemical !!aterials The analysis of post accident toxic chemical exposure to Control Room operators includes an evaluation of each potential release point along the pathway by which airborne material may enter t
the Control Room. The potential release points include storage locations both on and off the plant site, chemical pipelines, railroads, air, barge, and truck traffic.
Due to the number of potential releases identified in Sections 4.4 and 4.5, a three step approach was taken to evaluate the consequences of a chemical accident occuring at each given point. The objective of this approach was to eliminate from further consideration those accidents for which no adverse effects on Control Room operators could be determined.
This approach consisted of worst-case, probabilistic, and detailed design bases accident analyses.
The models used to define toxic chemical releases and the subse-quent meteorological dispersion conform to current design basis accidents models concerning radioactive dispersion and dose calculations.
The worst case calculation was performed emphasizing the guidance provided in Regulatory Guide 1.78, Reference 3.6, and other Page 84 of 181
~.
conservative assumptions to determine the consequences of a postulated accident for each chemical present at a given release point along the transport pathway.
The objective of this calculation was to eliminate those postulated accidents which under the most conservative set of assumptions could be identified as having no adverse effect on Control Room operators.
For transported materials such potential accidents are considered as having a zero probability of exceeding the toxicity ilmits in the Control Room.
The probabilisth calculation was then performed on those postulated transport accidents which were identified as having a non-zero probability of exceeding toxicity limits.
This analysis determined the probability of an accident occuring in a manner which would result in adverse consequences.
The calculation employed site specific meteorology, recorded transportation frequencies and government or industry estimates of the probability of occurrence for a given accident. The objective of this calculation was to eliminate those postulated transport accidents for which the probability of exceeding airborne toxicity limits is less than 10 events per year. Those postulated transport accidents which exceeded this probability criteria were defined as being among the design basis for recommended corrective actions.
A detailed calculation of the consequences of the postulated accidents which have been identified for corrective action was performed using realistic assumptions and site specific meteorology.
The calculation _was performed with the objective of determining Page 85 of 181
- ~ ~
whether the intent of Regulatory Guide 1.78 could be met under present plant design conditions.
This cair.ulation employed thermodynamically detemined chemical release rates, the control room infiltration and exfiltration rates calculated in Section 6.1.3, and detailed atmospheric plume dispersion models witn building wake corrections.
6.1.1 Worst-Case Chemical Accident Release The worst-case calculation for ha:ardous chemical release from stationary storage locations was performed to determine if any combinations of conservative analytical assumptions would result in the concentrations exceeding the established toxicity limits.
The stationary sources considered in the calculation include those identified in Section 4.4 as being present on-site in excess of 100 lb. quantities and those id,entifed in Section 4.5 as being present offsite within a five mile radius of the plant.
Two types of industrial accidents are considered for each source including the maximum concentration resulting from an accidental instantaneous release and.the maximum concentration / duration resulting from a safety relief valve or other realisti: slow leak. The evaporation or other mass transport following the release is based upon the physical properties of the chemical, the storage medium, the geometry of the resulting chemical distribution and appropriate heat transfer parameters.
Releases ~
were. assumed to occur under ambient air temperature and Pasquil,1
~
Stability Catagory F with the wind blowing directly toward the control room at a speed of 1 m/s. The meteorological methods
.used to model chemical' dispersion with building wake correction Page 86 of 181
conform to current design basis accident models concerning radioactive dispersion and dose calculation.
The results of this analysis of stationary sources indicate that all of the chemicals investigated could be eliminated from further consideration with the exceptions of hydrazine, sulfuric acid, and chlorine stored on-site and the anhydrous ammonia stored within the plant vicinity by the Agrico and Washco Chemical Companies.
A survey of transported hazardous chemicals was performed to determine which chemicals were transported with sufficient frequency inside a five mile radius of the plant to be considered frequent. Shipments are defined as being frequent, Reference 3.19, if there are 10 per year for truck traffic, 30 per year for rail traffic, or 50 per year for barge traffic.
The transportation sources considered in the calculation are the pipeline, barge, aircraft, railroad and truck sources identified in Sections 4.5.2 through 4.5.6.
The result of this analysis of transportation sources indicates that all of the chemicals investigated could be eliminated from further consideration with the exceptions of vinyl chloride, titanium tetrachloride, gasbline, alcohol and anhydrou: amonia.
Various chemical species of alcohol have been identified in.
transport. Therefore, the most limiting species, methanol is selected for further consideration. All of these chemicals are observed to be transported by railway. Anhydrous ammonia is observed.to be transported in addition by pipeline and truck. Gasoline is observed to be transported by truck.
Page 87 of 181
6.1.2 Chemical F ansport Accident Probability The chemical transport accident probability analyses were performed on those postulated accidents identified as having a non-:ero probability of exceeding toxicity limits. The chemicals inciuded in these analyses are vinyl chloride, titanium tetrachloride, gasoline, methano.1, and anhydrous a monia. The objective of the analyses was to elininate those postulated transport accidents for which the probability of exceeding airborne toxicity limits is less than 10-7 events per year.
The transportation mechanisms for which probability analyses are required are limited to truck and railway.
Figure 6.1.2.1 provides an overhead view locating the highways and railroad right-of-way existing within the five mile radius of the plant site.
Two types of transportation accidents are considered for each source including the maximum concentration case resulting from accidental instantaneous release and the maximum concentration /
duration case resulting from a slow continuous leak.
The evaporation and mass transport following the release is based upon the physical properties of the chemical, the transport container design, the geometry of the resulting chemical dis-tribution and appropriate thermodynamic parameters.
The meteorological cethods used to model' the chemical Page 88 of 181
dispersion conform to current design basis accideat models for radioactive dispersion and dose calculations. The probability of a given postulated accident resulting in toxic concentrations in excess of the established l'Wa within the Centrol Room is a function of a number of independent stochastic variables. These variables include (1) the frequency with which a given chemical is tran: ported within the site vicinity, (2) the accident probability for a given mode of transport, (3) the probability for a given accident to result in a major chemical release.
(4) the prevailing wind speed, (5) the prevailing wind direction, and (6) the atmospheric stability category.
The frequency with which a given chemical is transported within the site vicinity is provided in Section 4.5.
The accident probability for a given mode of transport is pro-vided in Reference 3.19 as a function of the distance traveled and the severity category. The total probability per mile for all severity categories is 8.1x10-7 for rail and 1.6x10-6 for truck.
Conservative estimates of the percentage of total accidents involving rail tank cars which result in a major chemical release le calculated to be 14 from the data pro-vided in Reference 3.23.
This percentage was assumed to be conservative for accidents-involving truck transport also.
The probability per one mile segrant for accidents in which Page 89 of 181
~
~
l a major fraction of the transported chemical is released 1
was calculated to be lx10-7 for rail and 2x10-7 for truck.
The total probability per one mile segment was found by multiplying these values by the number of vehicles carrying a given chemical along that segment per year.
The probability that meteorological conditions at the time of a postulated accident correspond to that which results in exceeding toxic limits was calculated from the data pro-vided in Section 4.1 for the stochastic variables of wind speed, direction, and atmospheric stability category. This data was computer processed to calculate the probability of exceeding a given dispersion factor (X/ ) under all possible Q
combinations of the three stochastic variables. A probability distribution function was calculated corresponding to each of the ten representative release points located at one mile segments along with the transportation routes identified.
Figure 6.1.2.1 indicates each of the release points employed. To re<%ce the mathematical computations, the probabilities for a small range of dispersion factors are sumed for those comb ~ nations of the stochastic variables whose calculated dispersion factor falls into the range.
Figure 6.1.2.2 provides an example of the probability distribution for exceeding a given dispersion factor at the plant site
- from a release point located on the railroad right-of-way 2.2 miles from the Control Room intake. An independent Page 90 of 181.
calculation is perfonned to determine the limiting dispersion factor for each representative release ' point which corresponds to the toxicity limit being reached in the Control Room atmosphere, This value is located on the probability distribution function yielding the probability of equaling or exceeding the limiting dispersion factor.
The probability of having a toxic limiting accident on a given one mile segment of the transport route is equal to the probability per segment for accidents resulting in a major chemical release multiplied by the probability of equaling or exceeding the limiting-dispersion factor for that segment. The total probability of having a toxic limiting accident along the entire route within the plant vicinity is equal to the summation of the probabilities for each one mile segment. This sequence of calculations was performed for each of the identified chemicals with their corresponding transportation routes. The results of the calculations are provided in Table 6.1.2.1 for truck transport and in Table 6.1.2.2 for railroad transport.
Summari::ing the results shows that transportation accidents involving the chemicals anhydrous ammonia, vinyl chloride, and titanium tetrachloride exceed the frequency probability of 10-7 events per year and must be considered for design basis accident analysis.
Page 91 of_181
Table 6.1.2.1 PROBABILITY OF T0XICITY LIMITIllG TRUCK TRAtiSPORT ACCIDEllT Chemical Ammonia Transport Route US 73 US 133 US 30 F (S)
- 508 227 300 F (A)
- 1.0x10'4 5.5x10-5 6.0x10-5 (X/g){
4.5x10-9 4.5x10-9 4.5x10-9 Segment F (X/g)TL,%
F (X/q)TL,%
F (X/g)TL J l
0.14 0.30 0.30 2
0.14 0.17 0.21 3
0.11 0.17 0.26 4
5.55 0.14 0.88 5
10.73 0.02 2.86 6
2.92 0.02 2.86 7
2.93 2.08 8
1.31 2.08 9-0.30 10 0.70 Total Probability ** 2.6x10-5 4.5x10-7 7.5x10-6
- F (S) = the number of trucks oer year using the transport route F (A) = p'robability per segment for accidents. with major chemi:a1 release (X/g)TL = toxic limiting dispersion factor, s/m F (X/q)TL= probability of equaling or exceeding the limit 1ng dispersion factor
'** Total Probability =
I F (X/q)TL x F (A) events _per year--
Page 92 of.181 H
TABLE 6.1.2.2 PROBABILITY OF T0XICITY LIMITING RAILRAOD TRANSPORT ACCIDENTS Vinyl Chloride Chemical tic 14 Methanol Leakage Burning Gasoline Ar:nonia F (S)
- 45 105 53 53 51 341 F (A)
- 4.5-6 1.05-5 5.3-6 5.3-6 5.1-6 3.41-5 (X/Q) k
- 3.36-5 3,49-4 N/A N/A 1.32'4 N/A (X/Q)k N/A N/A 2.95-8 2.75-10 N/A 9.17-10 Segment F(X/Q)
F(X/Q)T U F(X/Q) g F(X/Q)T u F(X/Q)TL'.' F(X/Q)TLJ 1
.07 0
0 4.94 0
3.25.
2
.08 0
.00 2.56 0
1.58 3
.15
.02
.02 3.63
.02 2.35 4
.43
.02
.03 4.08
.02 2.35 5
.40
.05
.19 10.66
.06 3.58 6
.53
.04
.21 10.05
.05 6.00 7
.75 0
0 9.48
~0 6.63 8
.07 0
0 4.21 0
2.22 9
.08 0
0 2.09 0
.99 10
.43-
.02
.08 1.18
.03 0.50-Total Probability ** 1.34" 1.58-8 2.81-8 4.32-6 8.95-9 1.16-5
- F(S)
= the number of railroad cars per year using the given railway route F(A)
= probability per segment for accidents witn major chemical release (X/Q)C
= toxic limiting dispersion factor for continuous release, s/m -
L (X/Q) g
= toxic limiting dispersion. factor for puff release, s/m
- Total Probability = I.F(X/Q)TLx F (A) events per year Page 93 of 181
Figure G.1.2.1 f.
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F T. C ALHOUN i
FACILITIES A - ROGERT OIL & PROPANE B - FAIRWAY PROPANE C - AGRICO CHEMICAL CO
'D - FORT CALHOUN STONE CO E - WILKINSON MANUFACTURING
- CHICAGO & NORTHWESTERN RAILROAD (florth/Scuth Track Abandoned)
PIPELINES
-WILLIAMS 3ROTHERS (8"& 12")
- - MID AMERICAN P!PELINE CO (MAPCO - 4")
--- GULF CENTRAL (6**)
- NORTHERN NATUR AL GAS CO (G")
i MO3tl OIL CORPORATION (G")
M DENOTES LOCATION OF ISOLATION VALVES X
CORRESPONDS TO RELEASE SEGMENTS Par'e 94 of 181 J
Figure 6.1.2.2 PROB A BILITY DISTRIBUTION FUNCTION 0
l x 10
,1 1
W
-1 1 x 10 5
k 1 x 10-2 u?
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CONTINUOUS RELEASE FROM
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-3 RAILWAY ACCIDENT LOCATED l x 10 i
2.2 MILES FROM CONTROL ROOM INTAKE
,g c<n s
1 x 10-4
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-7
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DISPERSION FACTOR XIQ, SECIM
1 6.1.3 Control Room Infiltration and Exfiltration This section describes the method employed to determine the Control Room air infiltration and exfiltration rates.
Descrip-tions of the Control Room ventilation system modes of opera-tion and the room design characteristics are provided in Section 4.2 and 4.3 respectively. The infiltration rate is determined for use in the analysis of contaminant in-leakage under the internal recirculation mode of ventilation system operation. The exfiltration rate is determined to ensure that the ventilation system is designed with sufficient capacity to maintain the Control Room at a positive pressure for conservative conditions in both the normal and filtered air makeup modes of operation.
The Control Room leakage pathways are the same for both in-filtration and exfiltration with variations only in the mode of ventilation system operation. Pathways included in the determination are envelope penetrations for piping, wiring, and ventilation ducts, leakage through the concrete walls, doors, and ventilation dampers. The rate calculations were performed using the metnods and assumptions given in Reference 3.5, the ASHRAE Handbook of Fundamentals and in Reference 3.6, the NRC Regulatory Guide 1.78.
Walls were.specifically modeled using the data for plaster covered brick and concrete. Doors,
dampers and penetrations were modeled using the data for doors and frames hung in masonry.
Dampers were modeled using data Page 96 of 181
modified from that for poor fitting window frames in masonry.
An additional assumption was made for personnel ingress and egress.
The dependence of air flow rate on differential pressure was detemined from the assumptions in Regulatory Guide 1.78, Reference 3.6.
For the calculation of infiltration to the unpressurized Control Room envelope, a one-eighth inch water, gauge pressure differential was considered for all leak pathways with the exception of several closed dampers under the internal recirculation mode of operation. During this mode of operation, the recirculation fan will create an increased pressure differential across the closed upstream dampers.
It was assumed that these dampers were exposed to a three-inch water gauge pressure differential. For the calculation of exfiltration from the pressurized Control Room envelope, a one-quarter inch water gauge pressure differential was considered for all leak pathways.
The calculated Control Room infiltration and exfiltration rate data is provided in Table 6.1.3.1 for all leakage pathways. The total infiltration rate into an unpressurized Control Room is 775 cfm. The total exfiltration rate out of a pressurized Control Room is 460 cfm.. Equipment in the makeup air supply system is designed to deliver 820:cfm into the Control Room under both the nomal and the filtered
~
air makeup modes of operation. Based on the calculated Pace 97.cf 181
rate of exfiltration, this flowrat5 is sufficient to maintain a positive pressure in the Control Room envelope.
Table 6.1.3.1 Control Room Leakage Rates Infiltration Exfil tration -
Leakace Pathway ft / min ft / min Wall 77
-121 Doors 126 198-Ducts and Other Penetrations 14 22 Dampers 548 109 Ingress and Egress 10 10 Total 775 cfm 460 cfm Page98off181
i 6.1.4 Design Basis Chemical Accident Consequences The analysis of post-accident toxic chemical exposure to Control Room operators includes a detailcd evaluation of the six identified design basis accidents (DBA) which could potentially result in exceeding established limits for air-borne concentrations.
The intake of airborne toxic chemicals into the Control Room ventilation system subjects operators to exposure through skin contamination and inhalation.
Each of the design basis accidents analyzed will result in airborne toxic chemicals being transported to the Control Room ventilation intake from release points identified in Section 4.
The methodology used for performing the Control Room habitability analysis with respect to post-accident chemical exposure is based on that described in References 3.6, 3.7, and 3.15.
Design basis accident parameters are based on actual data obtained at the Fort Calhoun Station and assumptions contained in the referenced Regulatory Guides. The chemical releases resulting from the DBAs and assumptions upon which they were-based are provided for each accident.
~
Specific. site related meteorological data, described in Section 4.1, has been -utilized along 'with the methodology of References 3.17, 3.20, and 3.21 in determining appropriate cloud dispersion
~Page"99 of 181-t___
7.
factors (X/g) for DBA releases. The Control Room ventilation system and its mode of operation are described in Sections 4.2 and 4.3.
The description provided therein, coupled with ventila-tion system filtration data and Control Room infiltration rates calculated in Section 6.1.3, has been used to determine consequent airborne chemical removal rates.
6.1.4.1 Sulfuric Acid Storage Tank Rupture The evaluation of the toxic chemical concentration in the Control Room atmosphere from a rupture of the sulfuric acid storage tank is based upon an assumed instantaneous spill of its entire con-tents. Sulfuric acid is stored on site in large quantities for use in regeneration of ion exchange. resin which is used for feedwater treatment. The acid is stored in an above ground tank, located in Figure 4.1-1, outside of the Office Building at ambient temperature. The possibility of a catastrophic tank rupture is remote due to the protection offered by the building.
The spill resulting from the sulfuric acid is conservatively assumed to cover the entire paved surface between the Office Building and the Cooling Water Intake Structure.
The resulting evaporation of the spilled liquid produces a continuous plume releas'e of the sulfuric acid in the direction of the Control Room ventilation system intake. The evaporation rate is calcu-lated using Patsak's formula for mass transport, vaporization based upon an assumed. solar heating sufficient to maintain -
0 the liquid temperature at 100 F, and a time period sufficient for two-phase equilibrium to be reached.
Page 100 of 181
The potential toxic consequences resulting from the occurrence of a postulated storage tank rupture have been analyzed using the assumptions and conditions listed below and the parameters provided in Table.6.1.4.1. The results of this evaluation indicate that under conservative conditions the sulfuric acid concentration in the Control Room-two minutes after the plume reaches the intake penthouse is 1.1 mg/m3 and in time reaches 3
a maximum value of 67 mg/m. The two minute concentration is well below the limit established in Reference 3.6.
The cor-rective action for this postulated accident is to provide detection of the accident prior to the acid plume reaching the ventilation system intake. This provides a two minute warning for operators to realign the ventilation system to the filtered air makeup mode of operation before toxicity limits are reached. Sulfuric acid removal by the charcoal filters in the system is qualitatively categorized in the-high capacity level by Reference 3.5.
Additional operator protection is provided by the use of respirators. The result is to reduce the inhaled concentrations to below the levels established in Reference 3.6.
These corrective action me:sures will be addressed in Section.7.2.
Assumations and Conditions It is assumed that the meteorological conditions are stable
~
with the wind direction directly from the location of the
. spill to the Control Room intake. Building wake correction Page.101 of-181
for the plume dispersion is applied consistent with the as-sumptions provided in Reference 3.20.
Conservativa models have been employed to calculate the evaporation rate taking into account the effects of equiliurium mass transport and solar heating of the liquid.
Uncertainties and Ccnservatisms The uncertainties and conservatisms in the assumptions used to evaluate the toxic consequences of the sulfuric acid storage tank rupture are as follows:
(1) The liquid spill is assumed to cover a surface area of 2
35,000 ft which results in a conservative estimate of the mass evaporation rate.
(2) No credit is assumed for emergency actions taken to mitigate the consequences of the tank rupture.
(3) The tank is assumed to be full at the time of the rupture.
(4). The initial plume width was conservatively estimated to be 100 ft which is less than the assumed width of the liquid spill resulting in a higher estimated concentration.
(5) No credit it assumed for absorption of the acid into the
' ground.
Page 102-of_181
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-f Table 6.1.4.1
~
PARAMETERSUSEDIttEVALUATINGTHET0XICCHEMICALCONSE0bENCESOF'dSULFURIC ACID STORAGE TANK RUPTURE.
I Parameter Assumotion Reference i Chemical Species H 504 sp.gr.l.83 l
2 985 conc.
Stored Quantity, lbs 225,000 3.4 Toxicity Limit, mg/m3 2
3.6 Distance from Control Room 330' 3.4 Intake, ft 2
Surface Area of Spill, ft 35,000 Evaporation Rate, ib/ min.
75 H2 SO4 aHy kcal/gm 0.987 Vapor Pressure, mmHg 0.01 Width of Initial Plume, ft 100 Cnarcoal Filter Removal High Capacity 3.5 Classification (qualitativel Atmospheric Dispersion Factors:
Wind Speed, m/s 1
3.6 Pasquill Category F
3.6 Page 103 of 181
6.1.4.2 Hydrazine Storage Drum Rupture The evaluation of the toxic chemical concentration in the Control Room atmosphere from the rupture of a hydrazine storage drum is based upon an assumed instantaneous spill of its entire contents. Hydrazine is stored on site in large quantities for use in chemistry control of the secondary coolant. The hydrazine is stored at ambient temperature in 55 gallon drums, located in Figure 4.1-4, within the water treatment area of the Office Building.
The possibility of a drum rupture is reduced through the use of administrative controls which minimize the quantity of hydrazine stored within close proximity to the Control Room intake.
The spill resulting from the hydrazine is conservatively.
2 assumed to cover 85 ft with a depth of one inch. The resulting evaporation of the spilled liquid produces a-continuous release of hydrazine to the Office Building ventilation system. The evaporation rate is calculated i
using Matsak's formula. The exhaust from tne Office Build-ing ventilation system, located in Figure 4.1-6, is assumed to be a point source plume directed toward the Control Room intake.
The potantial _ toxic consequences rerulting from the occurrence of a postulated stovage drum rupture have been analyzed using the assumptions and. conditions ' listed below and the parameters Page 104 of 181
- - - =
provided in Table 6.1.4.2.
The results of this evaluation indicate that under conservative conditions the hydrazine concentration in the Control Room two minutes after the plume reaches the intake penthouse is 0.2 mg/m3 and in time reaches a maximum value of 12 mg/m3 The two' minute concentration is well below the limits established in Reference 3.22.
The corrective' action for this postulated accident is to provide detection equipment at tne ventilation system intake.
This provides a two minute warning for operators to realign the ventilation system to the filtered air makeup mode of operation and'to employ respirators'.and anti-contaminate clothing protection before the. toxicity limits are reached. Hydrazine removal by the charcoal filters in the ventilation system is qualitatively categorized in the satisfactory capacity level by Reference 3.5.
The principal means of operator protection is the respirator.
The use of respirators alone will lower the inhaled concen-trations to below the -levels established in Reference 3.6.
Hydrazine also enters the body through adsorption on skin. -
Therefore anti-contaminate clothing is required. These cor-rective measures will be addressed in Section 7.2.
Assumotions and Conditions It is assumed that the meteorological conditions are stable with the wind direction.directly from the~ Office Building exhaust to the Control Room intake.
Building' wake correction Pace 105 of 181'
for the plume dispersion is applied consistent with the assumptions provided in Reference 3.20.
Conservative models have been employed to calculate the evaporation rate taking into account the. effects of equilibrium mass trans-port.
Uncertainties and Conservatisms The uncertainties and conservatisms in the assumptions used to evaluate the toxic consequences of the hydrazine drum rupture are as follows:
(1) The liquid spill is assumed to cover the largest realistic surface area which results in a conservative estimate of the mass evaporation rate.
(2) No credit is assumed for emergency actions taken to mitigate the consequences of the drum rupture.
(3) The initial plume is conservatively assumed to be a point source which results in a conservative estimate of the~ chemical concentration.
i Page 106 of 181'
Table 6.1.4.2 PARAMETERS USED IN EVALUATING THE T0XIC CHEMICAL C0tiSEQUEI!CES OF A HYDRAZINE STORAGE ORUM RUPTURE Parameter Assumption Reference Chemical Species NH 35". Conc.
24 Stored Quantity, gal.
55 3
Toxicity Limit, mg/m 2.6 3.22 Distance From Control Room 240 3.4 Intake, ft.
Surface Areas of Spill, ft 85 Evaporation Rate, g/s 0.7 Vapor Pressure, mm Hg 30 Radius of Initial Plume Point Source Charcoal Filter RemovaT Satisfactory Capacity -
3.5 Classification (qualitative)
Atmospheric Dispersion Factors Wind Speed, m/s 1
3.6 Pasquill Category F
3.6 4
Page 107 of 181
)
i 6.1.4.3 Anhydrous Ammonia Railroad Transport Accident Postulated post-accident chemical releases involving anhydrous ammonia which have been identified as design basis events include truck transport, railroad transport, and stationary storage offsite at the Agrico and Washco Chemical Companies. The consequences essociated with an accidental release of anhydrous ammonia are in general independent of the initiating event. The design basis event is therefore selected to be that accident which -
results in the highest concentration within the Control Room atmosphere. This has been determined to be the railroad transport accident. Results of the analyses for the remaining events are provided in Section 2.0.
Anhydrous ammonia is transported in a two-phase medium in pressurized railroad tank cars. Two types of accident must be considered. The instantaneous release case result-ing in the maximum initial concentration is a postulated rupture of the tank with a consequent release, per Reference 3.23, of 80 percent of the entire contents in a heavier than air cloud and 20 percent rising in a buoyant cloud. The-continuous release case resulting in the maximum duration is a postulated small puncture or pipe break.
1 i
9
'Page.108 of-181
The instantaneous release from a railroad tank car rupture is conservatively estimated to be a cloud with a 75 foot diameter.
The cloud travels in the direction of the Control Room intake at a speed of 1 m/s. The consequences have been analyzed using the assumptions and conditions listed below and the parameters provided in Table 6.1.4.3.
The results indicate that under conservative conditions the anhydrous ammonia concentration in the Control Room two minutes 3
after the cloud reaches the intake penthouse is.0.17 g/m,
The two minute concentration exceeds the limit established in Reference 3.6.
The continuous release from a railroad tank car puncture 2 inches in diameter is a point source plume with an initial release rate calculated to be 200 kg/s.
The consequences have been analyzed using the assumptions and conditions listed below and the parameters provided in Table 6.1.4.3.
The results indicate that under con-servative conditions the anhydrous ammonia concentration in the Control Room two minutes after the plume reaches the intake penthouse is 0.05 g/m3 and in time reaches a 3
maximum value of 0.41.9/m. The two minute concentration is less than the limit established in Reference 3.6.
?
Pace 109 of 181-
The design objective is to provide Control Room operators with at least two minutes to employ protective measures prior to the toxicity level being reached. The corrective action for the postulated accidents is to provide detection and automatic initiation of such protective measures. The initial action is to automatically realign the ventilation system to the filtered air makeup mode of operation upon detection of low anhydrous ammonia concentrations in the ambient air.
Removal of the ammonia by charcoal filters in the system is qualitatively categorized in the low capacity level by Reference 3.5.
Chemically treated charcoal can be employed to enhance the removal efficiency.
The resulting two minute concentration is below the toxicity level.
Assumptions and Conditions It is assumed that the meteorological conditions are stable with the wind direction directly from the source of the plume to the Control Room intake. Conservative models have :
been employed to calculate the. release ~ rates from the tank taking into account the expansion.and rise of the pressurized gas.
Uncertainties and Conservatives The uncertainties and conservatisms in the assumptions used to evalute the toxic consequences 'of the anhydrous ammonia transport accident are as follows:.
f Page.110.of 181
i I
(1) No credit is assumed for emergency actions.taken to mitigate the consequences of the accident.
(2) The tank car is assumed to be full at the time of the accident.
(3) The initial plume width for the. instantaneous -
release was' conservatively estimated to be 75 ft.
(4) The initial plume for the continuous release was conservatively assumed to be a point source..
(5)
The accident was assumed to occur.at the' closest distance of approach to the Control Room intake.
Page.111 of 131'
Table 6.1.4.3 PARAftETERS USED IN EVALUATING THE T0XIC CHEMICAL CONSEQUENCES OF AN AMM0tilA TRANSPORT ACCIDEflT Parameter i Assumption Reference Chemical Species NH H O 3 2 Transported Quantity, tons 104 3
Toxicity Limit, g/m 0.07 3.6 Distance from Control Room Intake,m 3500 3.4 Instantaneous Release:
Ground Level Release, tons 84 Buoyant Release, tons 20 Diameter of Initial Plume, ft.
75 Continuous Release:
Release Rate, kg/s 200 Puncture Diameter, in 2
Tank Pressure, psig 166 Atmospheric Dispersion Factors:
Wind Speed, u1/s 1
3.6 Pasquill Category F
3.6 Page 112 of 181
6.1.4.4 Chlorine Storage Cylinder Rupture The evaluation of the toxic chemical concentrations in the Control Room atmosphere from a rupture of a chlorine storage cylinder is based upon an assumed instantaneous release of 25 percent of the contents with the balance being vaporized.
and released over an extended period of time. Chlorine is stored on site in large quantities for use in raw water treatment. The chlorine is stored above ground in pressurized ambient temperature cylinders, located in Figure 4.1-1, outside of the Office Building. The possibility of a catastrophic rupture of a cylinder is reduced by the protection offered by the build-ing and restricted ~ personnel access.
The plume resulting from the chlorine is conservatively esti-mated to have an initial width of 4 ft. The plume travels in the direction of the Control Room intake at a speed of 1 m/s.
The consequences have been analyzed using the assumptions and conditions listed in Reference 3.7 and the_ parameters provided in Table 6.1.4.4.
The results indicate that under conservative conditions the chlorine gas concentration in the Control Room two minutes after the plume reaches the intake penthouse is 3
77 mg/m.
The two minute concentration exceeds the limit established in Reference 3.6.
The design objective is to provide Control Room operators with at least two minutes to employ protective measures prior to Page 113 of 181, I
the toxicity level being reached.
The corrective action for the postulateo accident is to provide detection and automatic initiation of such protective neasures. The initial action is to autecatically realign the ventilation system to the filtered air makeup made of operation upon detection of low chlorine concentrations in the ambient air.
3 The resulting two minute concentration is 38 mg/m which is below the toxicity level. The corrective actions will be addressed in Section 7.2.
Assumations and Conditions The calculational methods and assumptions described in Regulatory Guide 1.95, Reference 3.7, were assumed as:
(1) an instantaneous release of 25 percent of the chl&ine' container, (2) an initial cloud dimension assuming expansion of the gas into a spherical ' cloud having a Gaussian concen-tration gradient, and (3) the cloud dispersion was adjusted.
in the vertical direction by assuming uniform mixing between the ground and the height of the Office Building.
It is assumed that the meteorological conditions are stable with the wind direction directly from the source of the cloud to the Control Room intake.
Conservative models have been employed to calculate the release rates from the cylinder taking into account the expansion and rise of the pressurized gas.-
Page 114 of 181
l Uncertainties and Conservatisms The uncertainties and conservatisms in the assumptions used to evaluate the toxic consequences of the chlorine cylinder rupture are as follows:
(1) No credit is assumed for emergency actions taken to mitigate the consequences of the accident.
(2) The cylinder is assumed to be full at the time of the accident.
(3) The initial plume width for the instantaneous release was conservatively estimated to be 4 ft.
]
1, f
Page.115 of.181
I s
Table 6.1.4.4 PARAt:ETERS USED IN EVALUATIttG THE T0XIC CHEMICAL CONSEQUENCES OF A CHLORINE CYLINDER RUPTURE Parameter Assumption Reference Chemical Species C12 Stored Quantity, lb 150 3
Toxicity Limit, mg/m 45 3.7 Released Quantity, %
25 3.7 Distance from Control Room, ft 330 3.4 Diameter of Initial Plume, ft 4
Atmospheric Dispersion Factors:
Wind Speed, m/s 1
3.6 Pasquill Category F
3.6 Page 116'of 181~
6.1.4.5 Chloride Railroad Transport Accident The evaluation of the toxic chemical concentration in :he Control Room atmosphere from a transport accident invo: ving chlorides is based upon an assumed instantaneous releasa and subsequent burning of the entire contents of a railroad car carrying vinyl chloride.
Chlorides are frequently transported on the east / west track of the Chicago Northwestern railroad.
Specific informaticn on the chemical species of transported chloride is unavailable. Vinyl chloride was assumed because it is the most commonly transported species per Reference 3.26, is highly flammable,' and has hydrochloric acid as a product of combustion. Hydrochloric acid has one of the lowest toxicity limits of the chloride compounds.
Carbon and nitrous oxides are also products of combustion.
The potential toxic consequences resulting from the occurrence of a postulated railroad accident involving vinyl chloride has been analyzed using the assumptions and conditions listed be-low and the parameters provided.in Table 6.1.4.5.
The results of this evaluation indicate that under conservative conditions the concentrations in the Control Room two minutes after the plume reaches the intake penthouse is 170 mg/d for carbon dioxide, 69 mg/m3 for hydrochloric acid and 1 mg/m3 for nitrogen dioxide. The two minute concentration for hydrochloric acid exceeds the limit obtained from data provided in Reference 3.22.
Page 117_of-181-
The design objective is to provide Control Room operators with at least two minutes to employ protective measures prior to the toxicity level being reached. The corrective action for the postulated accident is to provide detection and automatic initiation of such protective measures. The initial action is to automatically realign the ventilation systec to the filtered air makeup mode of operation by detection of low hydrochloric acid concentrations in the ambient air. The principal protective measure is the use of respirator protection. The result is to reduce the inhaled concentration to below the levels established in Reference 3.6.
The corrective actions will be addressed in Section 7.2.
Assumotions and Conditions The accidental rupture of the railroad tank car was assume.d to ' result in a fire imediately following the instantaneous release of the entire contents. 'The combustion reaction was assumed to stochiometrically consume all of the vinyl chloride in the presence of 50 percent excess air. The meteorological conditions were assumed stable with the wind direction directly_
from the location of the tank car rupture to the Control Room intake.
Uncertainties and Conservatisms The uncertainties and conservatists in the assumptions used to evaluate the toxic consequences of ~ a chloride transport accident are as follows:
Page 118 of 181
.= -
i (1) No credit is assumed for emergency actions taken to mitigate the consequences of the tank car rupture.
(2) The tank car is assumed to be full at the time of the rupture.
(3) The initial plume width is conservatively estimated to be 135 ft.
i (4) No credit is assumed for absorption of the vinyl chloride into the ground.
I (5) The accident is assumed to occur at the closest distance of approach to the Control Room intake.
h 4
4 4
i
-I Page 119'of 181
.j
TABLE 6.1.4.5 PARAllETERS USED Ill EVALUATING THE T0XIC CHEftICAL CONSEQUEllCES OF A CHLORIDE TRANSPORT ACCIDEllT Parameter Assumption Reference Chemical Species Vinyl Chloride Transported Quantity, Tons 97 3
Toxicity Limit, mg/m Hydrochloric Acid 14 3.22 Carbon Dioxide 1840 3.22 i
flitrogen Dioxide 18 3.22 Distance from Control Room Intakg m 3500 3.4 Diameter of Initial Plume, ft 135 Atmospheric Dispersion Factors:
Wind Speed, m/s 1
3.6 Pasquill Category F
3.6 Page 120 of 181.
.....___.l
6.1.4.6 Titanium Tetrachloride Railroad Transport Accident The evaluation of the toxic chemical concentration in the Control Room atmosphere from a transport accident involving titanium tetrachloride is based upon an assumed instantaneous release of the entire contents of a railroad car. Titanium tetrahloride is frequently transported on the east / west track of the Chicago fiorthwestern railroad.
Exposure of titanium tetrachloride to moisture in the atmosphere results in a chemical reaction with hydrochloric acid and titanium oxide as the end products.
Hydrochloric acid has the limit-ing toxicity level and is therefore employed as the acceptance criteria.
The spill resulting from the titanium tetrachloride is con-2 servatively as.sumed to cover 3500 ft with a depth of four inches. The resulting vaporization of the spilled liquid produces a continuous releare of titanium tetrachloride to the atmosphere. The evaporization rate is calculated using Itatsak's formula for mass transport, vaporization based upon an assumed solar heating sufficient to maintain the liquid temperature at 100 F, and a time period sufficient for two phase equilibrium to be reached. Upon vaporization the chemical reacts with moisture in the air to produce hydrochloric acid and titanium oxide.
The reaction is conservatively assumed to consume 15 percent of the titanium tetrachloride present.
Page 121 of 181
~ ~ * * ~
The potential toxic consequences resulting from the occurrence of a postulated railroad tank car rutpure have been analyzed using the assumptions and conditions listed below and the parameters provided in Table 6.1.4.6.
The results of this evaluation indicate that under conservative conditions the resulting hydrochloric acid concentration in the Control Room two minutes after the plume reaches the intake penthouse 3
is 0.1 mg/m. The two minute concentration is less than the limit determined from data provided in Reference 3.22.
The design objective is to provide Control Room operators with at least two minutes to employ protective measures prior to the toxicity level being reached. The corrective action for the costulated, accident is,to provide detection and automatic initiation of such protective measures..The initial action is to automatically realign the ventilation system to the filter air makeup mode of operation by detection d
of low hydrochloric acid concentrations in the ambient air.
The principal protective measure is the use of respirator protection. The result is to reduce the inhaled. concentration to below the levels established in Reference 3.22.
The cor-rective actions will be addressed in Section 7.2.
Assumotions and Conditions It was assumed that the meteorological conditions are stable with the wind direction directly from the location of the tank car rupture to the Control Room intake.
Conservative models have been employed to calculate the evaporation rate taking 1
Page 122 of.181
into account the effects of equilibrium mass transport and solar heatir.g of the liquid.
Uncertainties and Conservatisms a
The uncertainties and conservatisms in the assumptions used to evaluate the toxic consequences of the titanium tetrachloride transport accident are as follows:
(1) The liquid spill is assumed to cover a surface area of 2
3500 ft which results in a conservative estimate of the mass evaporization rate.
(2) fio credit is assumed for emergency actions taken to mitigate the consequences of the tank rupture.
(3) The tank is assumed to be full at the time of the rupture.
(4) The initial plume width is conservatively estimated to be 30 ft which is less than the a:surmd width of the liquid spill resulting in a higher estimated concentration.
(5) fio credit is assumed for absorption of the chemical into the ground.
(6) The chemical reaction is assumed to consume 15 percent-of the titanium tetrachloride initially present.
'(7) The accident is assumed to occur at the closest distance.
of jproach to the Control Room intake.
Page 123 of 181
~ ' ' ~ '
Table 6.1.4.6 PARAMETERS USED Itt EVALUATIt1G TriE T0XIC CHEMICAL CONSEQUEf!CES OF A CHLORIDE TRANSPORT ACCIENT Parameter Assumption Reference Chemical Species Ti Cl4 Transported Quantity, tons 63 3
Toxicity Limit, mg/m 14 3.2 Hydrochloric Acid Distance from Control Room 3500 3.4 Intake, m 2
Surface Area of Spill, ft 3500 Vaporization Rata, g/s 1490 Ti Cl4 AH, k cal /g 0.04 y
Diameter of Initial Plume, ft 30 Atmospheric Dispersion Factors:
Wind Speed, m/s 1
3.6 Pasquill Category F
3.6 Page 124 of 181'
6.2 Assessment of Radiological Materials One design objective of the Control Room envelope is to ensure that the radiation exposure to operating personnel, through the duration of any one of the postulated design basis accidents, does not exceed the limits set by 10CFR50, Appendix A, General Design Criteria 19. This design objective has been evaluated through conservative analytical modeling of the transport mechanisms of post-accident airborne radioactivity, direct radiation to the Control Room, and the subsequent calculation of Control Room operator exposures.
The radiological consequences of airborne sources have been evaluated using conservative assumptions for postulated accident condi tions. The objective is to define the assumptions such that the consequences obtained from analysis are conservatively en-veloped. The consequences of direct radiation have been evaluated based upon data input from the Auxiliary Building design evaluation of post-accident shielding provided in Reference 3.18.
The criteria employed in this evaluation are the radiation dose equivalent guidelines provided in Reference 3.15 of 5 Rem whole body gama, 30 Rem thyroid and 30 Rem beta skin dose. The whole body gamma dose consists of contributions from airborne radioactivity -
transported into the Control Room, as well as direct shine from fission products inside the Reactor Containment and Auxiliary Buidlings. Conservative analyses indicate that only the postulated !.0CA design basis accident results in a.significant direct shine dose.
Page.125 of '181 l
6.2.1 Airborne Radiological Consequences The analysis of post-accident radiation exoosures to Control Room operators includes an evaluation of eight design basis accidents which result in the release of airborne radioactive material to the environment. The intake of the released radio-activity into the Control Room ventilation system subjects Control Room operators to beta skin and whole body gamma dose in addition to an internal thyroid dose from inhalation of radioactive iodine.
Each of the design basis accidents analyzed will result in airborne radioactive material being transported to the Control Room ventilation intakt from one of three release points:
the containment tuilding leakage, the auxiliary building 1
ventilation duct, or the secondary system safety relief and atmospheric dump valves. The methodology used for perfcrming the Control Room habitability analysis with respect to post-accident radiation exposure is based on that described in references 3.15 and 3.17.
Design basis accident parameters are based on actual data obtained from the Ft. Calhoun Station and assumptions con-tained in the referenced Regulatory Guides. The activity releases resulting from the DBA 's and assumptions upon which they were based are provided for each accident.
Page-126 of 181
Specific site-related meteorological data, described in Section 4.1,has been utilized along with the methodology of Reference 1
3.17 in detemining appropriate cloud dispersion factors (h) for DBA releases. The Contrcl Room ventilation system and its modes of operation have been described in Se'ction 4.2 and 4.3. The description provided therein, coupled with ventilation system filtration data and Control Room exfiltration rates calculated in Section 6.1.3, has been used to determine con-sequent airborne activity removal rates.
6.2.1.1 Fuel Handling Accident in th'e Spent Fuel Pool The evaluation of the radiological consequences to Control Room operators from a fuel handling accident in the spent fuel pool i
is based upon an assumed release of radioactivity should a fuel assembly be dropped or otherwise damaged during handling. The possibility of a fuel handling accident is remote because of the many administrative controls and physical limitations imposed on the fuel handling operations.
For the purpose of analysis, the Auxiliary Building is not assumed to be isolated. This results in a more limiting activity release to the environment. The ventilation system draws 12,800 CFM of-air across the spent fuel pool area; this air is discharged to atmosphere at the plant ventilation discharge duct.
Since all exterior doors are closed during fuel handling operations, this is the only route for the _ release of activity to. the environment.
Page 127 of 181
)
l Should high activity be detected during fuel handling operations, the operator wou'd manually direct the air exhausted from the spent fuel pool area through a charcoal adsorber to remove iodines prior to release at the discharge duct.
The potential radiological consequences resulting from the occurrence of a postulated fuel handling accident in the spent fuel pool have been analyzed using assumptions and parameters that are consistent with Regulatory Guide 1.25 recommendations as listed below. The results of this evaluation indicate that under the most conservative set of assumotions the Control Room operator dose equivalent will not exceed 0.11 Rem gamma whole body,1.7 Rem beta skin and 19.4 Rem thyroid.
These values are well within the limits established within reference 3.15.
Assumotions and Conditions The calculational methods and assumptions described in Regulatory Guide 1.25 were assumed as: (1) the values for the maximum fuel rod pressurization, (2) peak linear power density for the highest power assembly discharged, (3) maximum centerline operating fuel-temperature for the assembly in item (2) above, and (4) average burnup for the peak assembly above.
The fuel handling building post-accident cle
'o unit was modeled '
in the evaluation as a once-through filtrat.w
. stem. The removal efficiency for iodine is consistent with pegulatory Guide 1.25 for a 2-inch charcoal bed. This model is a. conservative one.
Pano 17R.nf.1R1
Uncertainties and Conscrvatisms The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of the fuel handling accident in the spent fuel pool are as follows:
(a) The gas gap activity inventory is conservatively calculated based upon plant operation at full power with 3 cycle burnup.
(b) The number of fuel rods that are assumed to fail in the assembly is a conservative number for the design case.
(c) The Auxiliary Building'sventilation system was modeled as a once-through filtration system, with'a conservative filter efficiency.
4
[Page12.9of181
Table 6.2.1.1 PARAMETERS USED Ifi EVALUATI!!G THE PADI0 LOGICAL C0!iSEQUEtiCES OF A FUEL HA!!DLIllG ACCIDENT Parameter lAsswnptions Reference
. Source Data:
Powe" levei, Mwt 1500 3.4 Radial peaking factor 1.65 3.4 Burnup 3 full 3.4 cycles @ power 80%
plant factor Decay time, hr 72 3.4 fiumber of failed rods 56 Fraction of fission product gases contained 3.11 in the gap region of the fuel rods, %
Kr-85,%
30 Other noble gases, %
10-Iodine,%
10 Activity Release Data:
3.11 Fraction of gap activity released.to pool, 100
~'
Minimum water depth above damaged rods,.ft.
23 Pool decontamination factor for noble gases 1
Release location Auxiliary Building Ventilation Duct w
Page 130 of,181' j
Table 6.2.1.1 (cont'd) i Parameter l Assumptions l
Reference Pool Decontamination Factor for Iodine:
3.11 Inorganic 133 1
Organic 1
Iodine Chemical Form Released to Fuel Bldg:
3.11 Elemental iodine, %
75 Organic iodine, %
25 Filter Efficiency:
3.11 Iodine, elemental %
90 Iodine, organic %
70 Iodine, particulate %
90 Others, %
90 Activity Released to Fuel Pool, Ci:
56 rods Isotope 3
I-131 6.14x10 (a)
I Xe-131m 5.00x10 4
Xe-133 1.24x10 2
Kr-85 2.41x10 Dispersion Data:
Distance to Control Room Intake, ft.
48.0 3.4 (a) Determined from Regulatory Guide 1.25 assumptions and core inventories based en 3. cycle burnup at full power.
-Page.131 of 181
6.2.1.2 Fuel Handling Accident in the Containment Building The evaluation of the radiological consequences to Control Room operators from a fuel handling accident in the containment build-ing is based upon an assumed release of radioactivity should a fuel assembly be dropped or otherwise damaged during handling.
The possibility of a fuel handling accident is remote because of the many administrative controls and physical limitations imposed on the fuel handling operations.
The ventilation exhuast air from the containment building is monitored for radioactivity before release to the atmosphere.
High radioactivity in the ventilation exhaust automatically closes the system isolation dampers. The equipment and per-sonnel hatches are closed by procedure during fuel. handling operations.
The potential radiological consequences resulting from the occurrence of a postulated fuel handling accident in the containment building have been analyzed using the assumptions and parameters listed below. The results of-this evaluation indicate that under the most conservative set of assumptions the 'antrol Room' operator dose equaivalent will not exceed C.001 Rem gama whole body, 0.005 Rem beta' skin and 0.2 Rem thyroid. These values are well within the limits established within reference 3.15.
Paca 132 of 181l
Assumotions and Conditions The calculational methods and assumptions described in Regulatory Guide 1.25 were employed for: (1) the values for maximum fuel rod pressurization, (2) peak linear power density for the highest power assembly discharged, (3) maximum centerline operating fuel temperature for the assembly in item (2), and 4
the i rage burnup for the peak assembly.
Uncertainties and Conservatisms The uncertainties and conservatisms in the assumptions used tc evaluate the radiological consequences of the fuel handling accident in containment are as follows:
(A) fio credit for iodine filtration (except pool decontamination) or mixing within the containment.
(B) All fuel rods in the assembly are assumed to fail (176 fuel rods).
i' The gas gap inventory is conservatively calculated based upon plant operation at full power with 3 cycle burnup.
Page 133 of 181-
i Table 6.2.1.2.
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUEllCES OF A FUEL HANDLING ACCIDErlT IN C0tiTAINiiENT Parameter Assumptions Reference Source Data:
Power Level, Mwt 1500 3.4 Radial Peaking Factory 1.65 3.4 Burnup 3 full power cycles Decay Time, hr 72 3.4 Number of Failed Rods 176 3.4 Activity Release Data:
Fraction of gap activity released to pool,%
100 3.11 Pool Decontamination Factor for Noble Gases 1
3.11 Factor for Iodine Inorganic 133 3.11 Organic 1
3.11 Iodine Chemical Fonn Released to Containment Building Elemental, %
75 3.11 Organic, %
25 3.11 Activity Released to Containment Pool, Ci:
2 Kr-85 7.58x102 Xe-131m 1.57x10
(,)
Xe-133 3.89x104 I-131 1.93x10 Dispersion Data:
Distance to Control Room Intake, ft 48.0 3.4 Wind Speed, m/s 1
3.4 Containment Building Design Parameters:
Same as for LOCA (a) Determined from Reg. Guide 1.25 assumptions and core inventories based on 3-cycle burnup at full power with 1% failed fuel.
Page 134 of-181
6.2.1.3 Gas Decay Tank Rupture The evaluation of the radiological consequences to Control Room operators frem a gas decay tank rupture is based on an unexpected and uncontrolled release of the radioactive contents stored in one gas decay tank. Four tanks are provided for hold-up of the com-pressed gases collected from the reactor coolant system and the liquid waste systec. Each tank is located in a separate, shielded compartment in the Auxiliary Building. Administrative procedures require that the tanks be isolated from each other.
The tanks are designed for normal operation at 100 psig and 140 F.
The tanks contain primarily nitrogen with trace amounts of noble gas.
The potential radiological consequences resulting from the occurrence of a postulated gas decay tank rupture have been analyzed using the assumptions and parameters that are con-sistent with Regulatory Guide 1.24 recommendations as listed below. The results of this evaluation indicate that under the most conservative set of assumptions the Control Room operator.
dose equivalent will not exceed 0.2 Ret gamma whole body,19.0 Rem beta skin and 0.1 Rem thyroid. These values are well with-in the limits established within Reference 3.15.-
i Page 135 of 181
Assumptions and Conditions It is assumed that the plant has been operating at 1500 Mwt with 1% failed fuel for an extended period, sufficient to achieve equilibrium radioactive concentrations in the reactor coolant.
The maximum gas activity would occur after shutdown and coolant degasification.
It is also assumed that the activity released from a gas decay tank rupture is the maximum quantity that can be contained within the tank at its maximum operating pressure of 100 psig. The tank is assumed to rupture and all of the noble gases and iodines are assumed to be released to the atmosphere in a 2-hour period, consistent with Regulatory Guide 1.24.
Table 6.2.1.3 lists the conservative assumptions for waste gas decay tank rupture and waste gas decay tank inventory prior to release.
Uncertainties and Conservatisms The uncertairities and conservatisms in the assumptions used to evaluate the radiological consequences of the Gas Decay Tank Rupture are as follows:
1.
No credit is taken for radioactive decay during transit from RCS to the Gaseous Waste Management System or for holdup in the gas-decay tank.
~
2.
The release rate through the Auxiliary Building ventilation system is assumed to be equal to tank leak rate.
3.
The gas decay tank inventory is conservatively calculated.
based upon plant operation with 1%. failed fuel.-
4 No credit is taken for filtration in the Auxiliary Building.
^
Page 136 of 181
Table 6.2.1.3 ASSUMPTI0tlS FOR GAS DECAY TAf1K RELEASE ACCIDEtlT Parameter Assumption Reference A. Source Data:
- 1. Power Level, Mwt 1500 3.4
- 2. RCS Radioactive Maximum values based on 3.4 Concentrations 1". failed fuel 3.10 3.4
- 3. Decay Time, hrs.
0 3.10 All gases str pped from processing the entire RCS volume are immediately passed to gas decay tank which fails. Accident occurs immediately follow-ing a cold shutdown releas-ing entire tank inventory
- 4. Gas Decay Tank (i) volume, scf 2875 3.4 (ii) pressure,psig 100
- 5. Tank Activity, C1.
Isotope (a) 1-131 2.6x10-23 I-132 4.2x10 2 1-133 2.7x10 3 I-134 1.9x10-9.7x10j I-135 Kr-85m 8.0x10+2 Kr-85 5.2x10
- 4. 3x10++2 Kr-87 3
Kr-88 1.4x10 Xe-131m 5.6x10+2 9.6x10l Xe-133 Xe-135 3.0x10+2 Xe-li%
1.9x10
- 6. Activity Release All gases released frcm 3.10 the Auxiliary Building within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (a) Calculated values based on maximum recctor coolant system specific activity per AriSI l1237 Standard methods.
Page 137 of 181
6.2.1.4 Control Element Assembly Ejection The evaluation of the radiological consequences to Control Room operators from a, ostulated control element assembly (CEA) ejection accident assumes a coincident loss of off-site power at the time of turbine trip. Thus, activity is released via two pathways:
a) From the reactor coolant, through the ruptured CEA drive mechanism pressure housing, to the containment, and to the environment via containment leakage.
b) From the reactor coolant to the steam generator secondary side (primary-to-secondary leakage), and to the environment via mass release from secondary relief and dump valves operated during_ initial cooldown.
The potential radiological consequences resulting from the occurrence of a postulated control element assembly ejection have been analyzed utilizing the assumtpions and conditions listed below. The results of this evaluation indicate that under the most conservative set of assumptions the Control Room operator dose equivalent will not exceed 0.0005 Rem gama whole body, 0.05 Rem beta skin and 8.0 Rem thyroid. These values are well within the limits established within reference 3.15.
Assumotions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Table 6.2.1.4.
-Page 133 of 181
Reactor crolant activity levels are calculated assuming that, prior to the accident, the plant operated with simultaeous 1%
failed fuel and a 1 gal / min primary to secondary leak.
For con-servatism, the secondary system is assumed to contain activity levels of 0.1 uCi/gm I-131 Dose Equivalent Curies.
1.
Activity Release frem Containment The activity available for leakage from containment is based on the equilibrium reactor coolant activity.
The nuclide activity instantaneously available for release from contain-ment is 100% of the noble gases and 25% of the iodines.
The activity available for leakage from containment is assumed to be instantaneously mixed in the containment free volume. Activity is asst.med to leak out at the limit (0.2 vol. %/ day) for the first day and at half this rate for the duration of the accident (1-30 days).
fio credit for iodine removal system operation is taken. The activity in centainment is assumed to decay due to holdup.
After leaking from containment, no radioactive decay or ground deposition is assumed during transit to the dose point.
2.
Activity Release from Secondary System The activity released from the secondary system is the activity released to the atmosphere from the main steam safety (relief) valves and dump valves during the cooldown phase until the shutdown cooling system is placed in operation.
Page 139 of 181
Primary-to-secondary leakage continues 01 spm until shutdown cooling is initiated.
Uncertainties and Conservatisms 1.
Reactor coolant equilibriu= activities prior to the accident are based on 1% failed fuel, which is a factor of two to eight greater than that nomally observed in past p'dR operation.
2.
Steam generator equilibrium activity for both steam generators is assumed to be equal to the technical specification limit. The technical specification limits are conservatively derived based on acceptable offsite doses frem accidents such as the CEA ejection accident.
3.
Loss of offsite pcwer is a conservative assumption.
4.
The containment leakage rate is taken to be the leakage rata at maximum peak pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 50t of this value thereafter.
5.
No credit for the fo8ne removal system was assumed.
The operator could " initiate" containment recirculation at a nominal time of 1800 seconds after the accident to mitigate the consequences of the accident.
6.
The secondary system mass release was maximi:ed by assuming:
- 1) all heat must be removed through the stean generators, and 2) a full. year beginning-of-cycle' case. This is conservative :as (1) mass.(and heat) removal actually-Page 140 of 181 j
-.- w
occurs due to coolant flow through the break to containment and (2) because the full power beginning-of-cycle case results in the worse power transient with regard to secondary system releases.
7.
Reactor coolant activities based on extended operation at 1500 Mat.
r: 3,.. 1A1 n# 101
Table 6.2.1.4 PARAMETERS USED IN EVALUATING RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ACCIDENT Parameter Assumptions l Reference Power Level (Mwt) 1500 3.4 Percent Fuel Experiencing Clad Failure 0
3.4 Primary Coolant Liquid Volume (ft )
6618 3.4 Primary Coolant Liquid Mass (lbm) 412,840 3.4 Steam Generator Liquid 11 ass (1bm) 79,530 Steam Generator Steam Mass (lbm) 4,986 Purification Flowrate 36 gpm 3.4
(=1800lb/hr)
Reactor Coolant Specific Activity, (a) before accident (pCi/gm)
Secondary System' Spec. Activity, 0.1 pCi/gm 4
before accident (pC1/gm)
(a) RCS specific activities estimated per ANSI N237 Standard Methods, with following assumptions employed to yield conservative values:
1.
Adjustment factor of "8" introduced to obtain resalts for operation with 1% failed fuel.
2.
Noble Gases: Reactor coolant removal rate conservatively assumed to be zero.
3.
Iodines: Reactor coolant removal rate conservatively taken_ as-0.039 hr -1 Page 142 of 181-
~
Table 6.2.1.4 (cont'd)
ParameteN Assumptions Reference 3
6 Containment Volume (ft )
1.05x10 3.4 Containment Leak Rate i
(1) 0-24 hr, vol.%/ day 0.2 3.4 (2) 1-30 day, vol.%/ day 0.1 3.8 Percent Coolant Fission Products 100%(noble) 3.12 Assumed to Containment 25%(iodines)
Iodine Removal System Parameters Not Utilized Credit for Radioactive Decay (1) Holdup in containment yes 3.12 (2) in-transit to Dose Point no 3.12
}
Activity Available for Release from (b)
Containment 0 t=0 i
Activity Release from Containment (c)
Primary to Secondary Leak Rate 8640 lb/ day Secondary Mass Release to Atmosphere (rate) 4 (1) Safety Valves, lbm/hr 5.16x10 (d) 5
[
(2) Dump Valves, Ibm /hr 2.27x10 (averaged over t=3.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />)
Steam Generator Decontamination Factor 10 (Water to Steam) i Activity Release from Secondary System (e)
Atmosphere Dispersion Factors Wind Speed, m/s 1
3.6 Pasquill Category F
Notes:
(b) Determined from Reg. Guide 1.4 and 1.7' assumptions.
.(c) Determined from previous item of the table. with credit taken for the removal' mechanisms of decay, iodine removal system and containment leakage. Values represent the activity available for leakage from the containment for 30' days following the accident.
(d) For conservatism mass release rate (average) based on estimated steam releases (secondary).
(c) Based on ANSI N237 Standard.
Page 143 of_181
i 6.2.1.5 Loss of Coolant Accident The evaluation of the radiological consequences to Control Room operators from a postulated Loss of Coolant accident assumes a double-ended rupture of a reactor coolant pipe with subsequent blowdown. The emergency core cooling systems limit the clad temperature to well below the melting point and ensure that the reactor core remains intact and in a coolable geometry, minimizing the release of fission products to the containment. However, to demon-strate that the operation of this nuclear' power plant does not represent any undue radiological hazard to the Control Room operators, a hypothetical accident involving a signi-ficant release of fission products to the containment is evalua ted.
It is. assumed that 100% of the noble gas and 50% of the iodine equilibrium core saturation fission product inventory is immediately released to the containment atmosphere. Of the iodine released to the containment, 50% is assumed to plate cut onto the internal surfaces of the containment or adhere to internal components. The remaining fodine and the noble gas activity is assumed to be immediately available for leakage from the containment.
Page-144 of_181;
.~
1 i
Once the gaseous fission prodact activity is released.a the containment atmosphere, it is subject to various mechanisms of re. oval which operate simultaneously to reduce the amount of activity in the containment. The removal mechanisms in-clude radioactive decay, containment iodine removal system and centainment leakage.
For the noble gas fission products, l
the only removal processes considered in the containment are radioactive decay and containment leakage.
4 The hydrogen purge systen is placed in operation 5 days af ter a LOCA to prevent the hydrogen concentration from exceeding 4.0 vol. 5.
The hydrogen purge system perfoms a 25.0 standard -
3 i
ft / min bleed and feed operation on the containnent atnospisere.
The LOCA source term available fer purging is presented in Table 6.2.1.5.
These source terms are based on the 7egulatory 1_
Guide 1.4 assumptions with credit taken for 5 days of radio-i active decay, containment iodine removal system and leakage.
Hydrogen purge system filter efficiencies for iodines are presented in Table 6.2.1.5.
Both trains of the engineered safety feature systems (ESF) are assumed to ' operate at 100 percent capacity.for the 30 days following the accident. Water in the' containment sumps is recirculated by the high pressure safety injection pumps -and the containment spray pump's. The sump water is as-sumed to contain 50. percent of the core. iodine inventory
~Page 145 of 181-
' " ~ " -
immediately following the accident.
Therefore potential exposures to Control Room operators due to operation of this external recirculation path are evaluated with an assumed leakage pathway to the environment.
The fodine is assumed to be homogeneously mixed with the water in the reactor coolant system and in the containment sump at the beginning of recirculation.
The minimum time interval between the initiation of the injection and re-circulation phases of ESF system operation is 27.4 minutes.
Therefore iodine decay is calculated for this amount of time in order to obtain the correct activity at the time recirculation is initiated.
The dissolved noble gas activity in the recirculation loop is negligible since noble gases are not readily entrained in water and are assumed for accident analysis to be in the contain.nent atmosphere.
An iodine-water partition factor of 0.1%
is used to calculate the amount of fodine which is avail-able for release from the pump rooms.
Activity was assumed to be dispersed instantaneously from the pump rooms to the atmosphere with no further nuclide holdup or decay.
In the event of a loss of coolant accident the Control Room ventilation system automatically realigns to the filtered air, makeup mode of operation.
This mode of operation is assumed to be employed for the duration of the accident.
)
Par a 1dA nr 1R1
The potential radiological consequences resulting from the occurrence of a postulated loss of coolant accident have been analyzed utilizing the assumptions and cenditions listed below. The results of this evaluation have been deternined separately for the accident consequences and for the containment hydrogen purge consequences due to the purge being initiated only after 5 days following the accident.
Under the most conservative set of assumptions the Control Room operator dose equivalent will not exceed 1.8 Rem gamma whole body at the control panels, 62.2 Rem beta skin and 81.0 Rem thyroid due to the accident release from the containment Building, the ESF leakage and the subsequent hydrogen purge.
The results of this evaluation indicate that the-thyroid dose exceeds the limits established in Reference 3.15. ' Corrective actions will be addressed in Section 7.2.
Assumotions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Table 6.2.1.5.
In the evaluation of a LOCA, the fission product release assumptions -
of Regulatory Guide 1.4 were folicwed. The following specific -
assumptions were used in the analysis.
1.
The reactor core equilibrium noble gas and iodine inven-tories are based on long-term operation at 100% of the -
ultimate core power level of 1500 Mwt.
.2.
One hundred percent of the core equilibrium radioactive noble gas inventory is invrediately available for Lleakage from the containment.
3.
Twenty-five percent of -the core equilibrium radioactive iodine inventory :is immediat'ely available for. leakage
-Pace '147 of 181
from the containment.
4.
Of the iodine fission product inventory released to the con-tainment, 91% is in the form of elemental iodine 5% is in r
the fom of particulate iodine, and 4% is in the form of organic iodine.
5.
Radioactive Decay - Credit for radioactive decay for fissicn product concentrations located within the containment is as-sumed throughout the course of the accident. No credit for radioae.tive decay or deposition is taken during transport of airborne activity to the Control Room.
6.
Containment Iodine Removal System - Credit for the re-moval of iodine from the containment building atmosphere is assumed during the course of the accident resulting.
from filtration in the iodine removal system. The system consists of four air handling units; two having filtering capacity and the other two having no filtering capacity.
In the calculation of the radiological consequences an absorption efficiency of 0.9 and operation of only one of the two containment filtering units is assumed.
Credit for organic iodine removal is taken until a reduction of -
50 percent is achieved.
Credit for elemental and parti-culate iodine removal _ is taken for the course of the accident.
7.
The following removal constants for the containment iodine
~
removal systems are assumed in the analysis:
Elemental Iodine-5.14 hr'I 5.14 hr'I OrganicLIodine Particulate Iodine
.5.14 hr'I-8.
The containment is' assumed to leak at 0.2 vol.5/d during the._
first 24 ' hours 'imediately following the accident and 0.1 vol.Vd thereafter.
.Page 148 of.181
Uncertainties and Conservatisms The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of the LOCA are as follows:
1.
The ECCS is designed to prevent fuel cladding damage that would allow the release of the fission products contained in the fuel to the reactor coolant. Operation of the ECCS mitigates the consequences of fission product release as assumed in the analysis.
2.
The release of fission products to the containment is assumed to occur instantaneously.
3.
It is assumed that 50% of the iodines released to the con-tainment atmosphere are absorbed onto internal surfaces of the containment or adhere to inta mal components; however, it is estimated that the removal of airborne iodines by various physical phenomena such as adsorption, adherence, and settling could reduce the resultant doses by a factor of 3 to 10.
4.
The activity released to the containment atmosphere is assumed to leak to the' environment at the containment leakage rate of 0.2 vol.%/d for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.1 vol.%/d thereafter. The initial containment leakage rate is based on the peak calculated internal containment
.Page 149-of 181.
. - _. =.
~
pressure enticipated after a LOCA. The pressure within the coi.tainment actually decreases with time. Taking into account that the containment leak rate is a function of pressure drop, the resultant doses could be reduced i
by a factor of 5 to 10.
5.
Credit for iodine removal by the containment iodine removal system is not taken until 0.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the accident.
Credit for organic iodine removal is not taken after a re-duction of 50% is achieved.
- l Page 150. of 181'-.
~. j
Table 5.2.1.5 PARAi4ETERS USED II: EVALUATI;G THE RADIOLOGICAL C0?tSEQUEf!CES OF A LOSS OF C00LAf;T ACCIDEtli Parameter Assumntions Reference 1
A.
Source Data 1.
Power level, ibt 1500 OPPD FSAR 3.4 2.
Fraction of core activity 3.8 initially airborne in the containment, a) floble Gas 100 b)
Iodine 25 B.
Activity Release Data 1.
Containment Leakage Rate, Vol %/d a) 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.2 3.4 b) 1 to 30 days 0.1 3.8 3.4 2.
Fraction of containment 100 leakage that is unfiltered,".
3.
Credit for containment iodine absorption system a)
Iodine Removal Constants, 3.4 hr -l (1) Elemental 5.14-(2) Organic 5.14 (3) Particulate 5.14 b) Decontamination Factor (1) Elemental 10 (2) Organic 10'
.(3) Particulate 10 Page 151 of 1si-
~
Table 6.2.1.5 (co'nt'd)
I Pa rar.e ters Assumntions Reference c.
Fraction not affected by the filters (1) Elemental 0'
(2) organic 50 (3) Particulate 0
4.
Activity initially. airborne in the containment building,Ci Isotope I-131 1.04 (+7) 1-132 1.53 (+7)
(a)
I-133 2.11 (+7)
I-134 2.28 (+7)
I-135 1.96 (+7)
Xe-131m 2.94 (+5)
Xe-133 8.46 (+7)
Xe-135m 1.71 (+7)
Xe-135 1.51 (+7)
.6.75 (+7)
Kr-85m 1.05(+7)
Kr-85 3.34 (+5)
Kr-87 1.93 (+7)
Kr-88 2.76 (+7) 5.
Initiation of Containment purge for hydrogen reduction, days fol-
.-. 5..
.. _,, _,3. '4 lowing event 5.
Duration of hydrogen purge, days 4d 3.4 7.
Containment airborne activity at time of hydrogen purge initiation, curie Isotope I-1 31 1.47 (+5)
(b) 1-132 0.0 1-133 0.0
'I-134 0.0 1-135 0.0 Xe-131m 2.32 (+5)
Xe-133 4.97 (+5)
Xe-135m 0.0 Xe-135 1.05 (+4)
Xe-138 0.0 Page 152 of 181
- ~
? -
c - - c
3 Table 6.2.1.5 (cont')
Parameters Assumntions l
Reference 7.
(cont'd)
Kr-85m 2.8 Kr-85 3.32(+5)
Kr-87 0.0 Kr-88 0.0 8.
Containment Hydrogen Purge 25 34 Rate, SCFM 9.
llydrogen Purge System Filter 3.4 Iodine Decontamination Factors (1) Elemental 10 (2) Organic 10 (3) Particulate 10 C.
Dispersion Data, Wind Speed 1 m/s 3.6 Pasquill Category F
3.6 2
(a) Determined from Regulatory Guide 1.4 assumptions and core inventories based on 2 cycle burnuo at full power.
(b) Determined from Item (4) of the table with credit taken for the removal mechanisms of decay, iodine removal system and containr.:ent leakage.
Values represent the activity available for leakage from the containment 5 days -
following the accident.
- s. E O n
Page 133.of 131
6.2.1.6 Main Steam Line Break To evaluate the radiological consequences to Control Room operators from a postulated main steam line break (outside containment), it is assumed that there is a complete severance of a main steam line outside the containment with the plant in a hot zero power condition where the transient is initiated
- hortly after full power operation.
It is also assumed-that there is a simultaneous loss of offsite power. The hot zero power condition assures the maximum water inventory in the steam generators and the shutdown from full power (in con-junction with the loss of offsite power) assures the maximum decay heat which must be removed by manual control of the atmospheric dump valve associated with the intact steam generator.
The main steam isolation valves are installed in the main -
steam lines from each generator, downstraam from the safety relief valves outside containment. The.
severance of the main steam line is assumed to be upstream of the main steam isolation valve. A reactor trip is actuated by a low steam generator pressure signal. A main steam isolation signal (MSIS) is actuated to shut-the main steam isolation valves from both steam generatars.
The affected steam generator (steam generator. connected to the severed steam line) blows down completely.. The steam is vented directly, to the atmosphere. 'The atmospheric Page 154 of'181.-
'l i
dump valve of the unaffected steam generator is used to initiate a cooldown of the reactor coolant system 1800 seconds after initiation of the accident. The steam is vented directly to the atmosphere. ibss release from the unaffected steam generator is terminated when the shutdown cooling system is initiated at a reactor coolant system temperature of 300 F.
In this evaluation, a case with an iodine spike caused by the main steam line break accident was evaluated for radio-logical consequences.
Prior to the main steam line break accident the reactor coolant system activity is based on 1% failed fuel. At the initiation of the 14SLB accident, the I-131 equivalent source term (released from' fuel)'is assumed to increase'by a factor of 500 The potential radiological consequences resulting from the occurrence of a postulated main steam line break outside of the containment building have been analyzed using.the-parameters provided in Table 6.2.1.6 The results of -
this evaluation indicate that under the most conservative set of assumptions the Control Room operator dose equivalent will not exceed 0.0001 Ren garna whole bocy, 0.004 Rem beta skin and 26.9 Rem thyroid.
These values are within the-guidelines established in reference 3.15.
Page 155'of=181-
Assumations and Conditions The major assumptions, parameters, and calculational methods used in the design basis analysis are presented in Table 6.2.1.6.
Additional clarificatior is provided as follows:
1.
Reactor Coolant Activity The reactor coolant equilibrium activity is based on long term operation at 100% of the ultimate core power level of 1500 Mwt and 1% failed fuel.
2.
Secondary System Activity The activity.in steam generators is conservatively assumed to be equal to 0.l';Ci/gm dose equivalent Iodine-131 (I-131).
3.
Primary-to-Secondary Leakage The primary to secondary leakage of 1 gal / min.(technical specification limit) was assumed to continue through the affected steam generator at a constant rate until shutdown cooling is initiated.
Page'156 of 181-
Table 6.2 1.6 PARAMETERS USED IN EVALUATIrlG THE RADIOLOGICAL CONSEQUENCES OF A MAI:. STEAM LIriE BREAK ACCIDEllT (IELBA)
Parameter Assumptions Reference Data and Assumptions used:
General:
Pcwer Level, Mwt 1500 3.4 Burnup End of cycle Percent of Fuel Perforated 0
Reactor Ccolant Activity after acci-60 uCi/gm dent iodiae spike caused by accident DEC I-131 Steam Generator Activity before 0.1 uCi/gm accident dose equiv.
I-131 General:
Loss of offsite power yes Credit for radioactive decay in no.
transit to dose point Affected Steam Generator':
Primary-to-secondary leakage rate, 8,640 lb/d (1 gal / min)
Secondary mass realease to atmosphere 233,500 (through severed line), lbm Mass of primary-to-secondary leakage, 490.
lbm Steam generator decentamination factor 1
between steam and water phase Unaffected Steam Generator:
Primary-to-secondary. leakage rate,1b/d 0
Dispersion Data:
Atmospheric dispersion factors 3.6 Wind Speed, m/s 1
Pasquill Category F
L DEC = dose equivalent curies P
57 of 18L m--..
m
i 6.2.1.7 Single Reactor Coolant Pump Shaft Seizure The evaluation of the radiological consequences to Control Room operators from a reacter coolant pump shaft seizure assumes a conservative release of secondary activity as well as reactor coe:2nt leakage. The inventory of iodine and noble gas fission product activity available for release to the environment is a function of the primary-to-secondary coolant leakage rate, the percentage of defective fuel in the core, the percentage of fuel to experience DNS, and the mass of steam leaving the generator.
Conservative assumptions are made for all of these parareters.
Following seizure of a reactor coolant pump shaft, the core flowrate rapidly decreases to the value that would occur with only three reactor coolant pumps operating.
The re-duction in coolant flourate causes an increase in the average coolant' temperature in the core and causes a reactor trip on low flow. The reactor trip produces an automatic turbine trip.
Following the turbine trip, the turbine bypass valves open. The Control Room operator initiates a controlled cooldown using the turbine bypass valves after the reactor trip. Radioactivity enters the secondary sice through an assumed primary to secondary leak. A radiation monitor en the air _ ejector condenser exhaust realigns automatically on hign :ctivity-to
.Page 158 of 181
direct condenser offgas to the filtered auxiliary building ventilation system.
The results of this evaluation indicate that under the most conservative set of assumptions the Control Room operator dose equivalent will not exceed 0.002 Rem gamma whole body, 0.1 Rem beta skin and 5.5 Rem thyroid. These values are well within the guidelines established in reference 3.15.
Assumotions and Conditions The major parameters assumed in the analysis are itemized in Tabic 6.2.1.7.
The following assumptions and parameters are used to calculate the activity releases and offsite doses for a single reactor coolant pump shaft seizure:
(1) The RCS equilibrium activity is given in Table 6.2.1.7.
(2) The steam generator equilibrium activity for both steam generators is assumed to be 0.1 pCi/gm dose equivalent I-131 prior to the accident.
(3) The amount of noble gas activity released is equal to the amount present in the reactor coolant discharged into the secondary side oue to primary-to-secondary leakage. The amount of noble gas activity contained in the secondary
. system is negligible in comparison.
Page 159 of 181 7-
(4) Iodine activity released is based on the equilibrium activity present in the steam generators (0.1 pCi/gm dose equivalent I-131) and the amount of activity present in the reactor coolant due to failed fuel.
(5) 2.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> after the accident, total cooldown' period is reached. No steam and fission product activities are released from the steam generator thereafter.
(6) The total amount of discharge of reactor coolant into the secondary system through leakage.'is assumed to be 1389 pounds.
(7) A post-accident DF of 10 was used in the steam ganerator.
between the water and steam phases.
(8) The primary-to-secondary leakage of 8640 lbm/ day (1.0 gal / min) 1s assumed to be applicable to both steam generators.
.The portion of the noble gas activity from the primary-to-secondary leakage attributed to the steam generators is assumed to be released during the course of the-accident.
(9) The activity released from the condensers is _ imediately-vented to the atmosphere. No credit for radioactive decay for isotopes in transit to dose points.
Uncertainties and ~ Conservatisms The uncertainties and conservatisms in the assumptions used
- to evaluate the radiological consequences of a single reactor coolant pump shaft seizure are.as follows:
Page 160 of 181 -
_j 3..___....
1.
Reactor coolant equilibrium activities are based on 1%
failed fuel, which is greater by a factor of two to eight than that normally observed in past PWR operation.
2.
Steam generator equilibrium activity for coth steam generators is assumed to be equal to the technical specification limit. The technical specification limits are conservatively derived based on accidents such as the SGTR.
3.
Conservative values for both the RCS and gas gap activities were chosen, based upon plant operation at full power with 3-cycle burnup.
4.
2% of the fuel pins in the core were assumed to ex-perience DNB, and all that experienced DNB were assumed to fail.
5.
A conservitive steam generator decontamination factor (DF) of 10 is used.
6.
No credit is taken for radioactive recay of isotopes in transit to dose points.
7.
The meterological conditions assumed at the time of the accident are based on reference 3.6.
The wind speed is 1 m/s in the direction of 'the Control Room ventilation intake with a Pasquill Stability Category. F.
8.
A conservative main condenser mechanical vacuum pump partition factor of unity was assumed for iodines, Page-161 of 181
Table 6.2.1.7 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SIllGLE REACTOR COOLANT PUMP SHAFT SEIZURE Parameter Assumption Reference Source Data:
Power Level, !+.f t 1500 3.4 Fraction Failed Fuel, %
1 3.4 Steam Generator Tube Leakage, lb/d of 8640 equilibrium reactor coolant activity (1 gal / min)
Coincident (existing) Fuel Failure 2
due to DNB, ".
Equilibrium baccadary System Activity 0.1 uCi/gm DEC I-131 Initial RCS Activity Inventory, pCi/gm (a)
Kr-85 8.7_(+1)
Xe-131m 7.1 (-1)
-9.9 (+1)
I-131 2.6 (0)
Gas Gap Activity, C1/ assembly isotope (b)
Kr-85 7.6
(+2)
Xe-131m 5.2
(+3)
Xe-133 5.8
(+4)
I-131 2.5
(+4)
Activity Release Data Steam Discharge, Ib Mass of Steam Dumped 8.0 (+5)
(c)
Discersion Data Atmospheric Dispersion Factors Wind Speed, m/s 1
Pasquill Category F
3.6 Iodine Decontamination Factors for 10 Steam Generators (between water and steamphase)
Notes:
(a) RCS Specific Activities estimated per' ANSI N237 Standard itethods,-with following assumptions employed to yield conservative values:
- 1. Adjustment factor of "8" introduced to obtain results for operation with 1"; failed fuel.
- 2. Noble Gases: Reactor coolant removal rate conservatively assumed to be zero.
- 3. Iodines: Reactor coolant removal rate conservatively taken-as 0.039 hr-I (b) Based on plant operation at full power with 3-cycle burnup;
-(c) Determined from conservative estimate.
Page 162 of 181
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i 6.2.1.8 Steam Generator Tube Rupture The evaluation of the radiological consequences to Control Room operators from a postu'ated steam generator tube rupture assumes a complete severance of a single steam generator tube while the reactor is operating at full rated power and a coincident loss of offsite power at the time of reactor trip. Occurrence of the accident leads to an increase in contamination of the secondary system due to reactor coolant leakage through the tube break. A reactor trip occurs automatically as a result of low pressurizer pressure after the tube rupture occurs. The reactor trip automatically trips the turbine.
The resulting increase in radioactivity in the secondary system is detected by radiation monitors.
The coincident loss of offsite station power causes closure of the turbine bypass valves to protect the condenser. The steam generator pressure will increase rapidly, resulting in steam discharge as well as activity release through the main steam safety valves.
Venting from the affected steam generator, i.e.,
the steam generator which experiences tube ~ rupture, continues until the secondary system pressure is below the main steam safety valve setpoint. At this time, the affected steam generator is effectively isolatea, and thereafter, no steam or activity is assumed to be released from the affected steam generator.
The remaining unaffected steam generator removes core decay heat by venting steam through the main safety valves, Page 163 of 181
atmospheric dump valve, and steam driven auxiliary turbine until cooldown can be accomplished with the shutdown cooling system.
The analysis of the radiological' consequences of a steam generator tube rupture considers the most severe release of secondary activity as well as reactor activity leaked from the tube break. The inventory of iodine and noble gas fission product activity available for release to the environment is is a function of the primary-to-secondary coolant leakage rate,
~
the percentage of defective fuel in the core, and the mass of steam discharged to the environment.
Conservative assumptions are made for all these parameters.
In this evaluation, a case with coincident iodine spike which already exists due to a previous power transient was considered.
The reactor coolant system inventory was assumed to be 60 pCi/gm dose equivalent Iodine I-131. This 60 u Ci/gm is the technical specification limit for full power operation following'an.
iodine spike for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- The potential radiological consequences resulting from the occurrence.of a postulated steam generator tube rupture have been analyzed utilizing the assumptions and conditions-listed below. The results of this evaluation indicate that under the most conservative set of assumptions the Control. Room operator--
dose equivalent will not exceed 0.0008 ' Rem gamma whole body, 0.06 Rem beta skin and 4.3 : Rem thyroid..These values Page 164 of 181
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a
are well within the guidelines established in Reference 3.15.
Assumotions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Table 6.2.1.8.
The following assumptions and parameters are used to calculate the activity releases and offsite doses for a steam generator tube rupture (SGTR):
1.
The reactor coolant system equilibrium activity is 60 pCi/gm DEC I-131.
2.
The steam ' generator equilibrium activity for both steam generators is assumed to be 0.1 uCi/gn. dose equivalent I-131 prior to the accident.
3.
Offsite power is lost; the main condenser is not available for steam relief via the turbine bypass system.
4.
Following the accident, no additional steam and radioactivity are released to-the environment when the shutdown cooling system is placed in operation.
5.
The Control Room ventilation system'is automatically realigned to the filtered air makeup mode of operation.
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5.
There is no main condenser evacuation system release and no steam generator blowdown during the accident.
6.
Only one steam generator is affected.
7.
The amount of noble gas activity releases is equal to the amount present in the reactor coolant discharged into the secondary side folicwing the tube rupture. The amount of noble gas activity contained in the secondary system is' negligible in comparison.
8.
Iodine activity released is based on the equilibrium activity present in the steam generators (0.1 uCi/gm dose equivalent I-131) and the amount of activity present in the reactor coolant discharged into the affected steam generator.
9.
Thirty-minutes after the accident, the affected unit is isolated. No steam and fission product activities are.
released from the affected steam generator thereafter.
10.
The total amount of discharge of reactor-coolant.into the secondary system through the rupture is 42,300 pounds (in 30 minutes).
11.
A post-accident 0F of.10 was used in the steam generator between -the water and steam phases.
12.
The primary-to-secondary leakage of 8640 lbm/d (1.0 gal / min) is assumed to be' applicable to the unaffected steam generator.
The portion of the noble gas activity from the primary-to-secondary leakage attributed to the unaffected steam genera-tor.is assumed 'to be released during the course of the accident.-
o Page 166 of 181 ma
13.
The amount of discharge of steam from the a ffected steam generator is assumed to be 3.5x106 pounds.
14.
The activity released from the affected and unaffected steam generators is immediately vented to the atmoshphere.
No credit for radioactive decay for isotopes in transit to dose points.
Uncertainties and Conservatisms The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of a steam generator tube rupture are as follows:
1.
Reactor. coolant equilibrium activities are based on 1%
failed fuel, which is greater by a facior of two to eight than that nomally observed in past PWR operation.
2.
Steam generator equilibrium activity for both steam generators is assumed to be equal to the technical specification limit. The technical specification limits are conservatively derived-based on a'ccidents such as the SGTR.
3.
- Tube rupture of the steam generator is assumed to be a.
double-ended severance of-a single steam generator tube.
This is a conservative. assumption since the steam generator tubes are constructed of highly ductile materials. The more probable mode of tube failure is one of minor leaks of undetermined origin. Activity in the secondary steam system is subject to continual surveillance, and the Page 167 of 181
accumulation of activity from minor leaks that exceed the limits established in the technical specifications would lead to reactor shutdown.
Therefore, it is unlikely that the total amount of activity considered available for release in this analysis would ever be realized.
4.
The coincident loss of offsite power with the occurrence of the reactor trip following the steam generator tube rupture is a conservative assumption.
In the event of availability of offsite power, the turbine bypass valves will open, relieving steam to the main condenser. This will reduce the amount of steam and entrained activity discharged directly to the environment from the unaffected steam generators.
5.
A conservative steam generator' decontamination factor (DF) of 10 is used in the cooldown phase (release to atmospheric-dump valve).
LPage_168 of-181.
4 Table 6.2.1.8 PAF#tETERS USED Iti EVALUATIriG THE RADIOLOGICAL C0tiSEQUEriCES OF A STEAM GEtiERATOR TUSE RUPTURE Parameter Assumption
{
Reference Source Data:
Power level, Mwt 1500 3.4 Steam generator tube leakage, Ib/d 8640-(1 gal / min)
Equilibrium reactor coolant activity Coincident (existing) iodine spike 60 uc/g dose equivalent I-131 Equilibrium secondary system 0.l'uCi/gm activi ty dose equivalent Activity Release Data:
1-131 Steam discharge, lb Affected steam generator Reactor coolant leakage to 42,300 steam generator (0-30 min) 6 itass of ' steam released 3.5x10 Dispersion Data:
Atmospheric dispersion factors
-3.6 Wind Speed 1
Pasquill Category F
Iodine decontamination factors 10 for steam generators -(between water and steam phase)
F Page'169 ofL 181
)
6.2.2 Control Room Shielding Design Review A detailed design review of existing plant shielding and access control, provided in Reference 3.18, for systems and spaces vital to post-accident operations has been perfomed. This review has identified the Control Room as requi. ing corrective action shielding modifications.
Additional shielding is specifically required for the containment spray recirculation piping to reduce the dose rate field in the direction of the Control Room.
i This piping is located in the eastern corner of the ventilation equipment area on the 1025 ft elevation of the Auxiliary Building.
The referenced shielding design review was perfomed with the assumption of a pust-accident release of radioactivity equivalent to that described in Regulatory Guide 1.4 and 1.7.
An additional evaluation was perfomed with the assumptions corresponding to the postulated design basis accidents addressed in Section 6.2.1.
Conservative estimates indicate that only the postulated Loss Of Coolant accident results in a significant direct radiation exposure to Control Room operators.
Shielding analyses have been perfomed to design _the required thickness of structural concrete. These analyses are based on the clarified post-accident Page 170 of 181-
J radiation source terms provided in Reference 3.24.
The design cbjective is to reduce the whole body gamma radiation exposure of Control Room operators to within 5.0 Rem dose equivalent. The sources of radiatior, considered in the analysis include the containment spray recirculation piping, the containment building, airborne source terms calculated in Section 6.2.1.5 and other sources within the Auxiliary Building as identified in the referenced design review.
The results of these analyses indicate that the whole body gamma raciation exposure-to Control Room operators under post-accident conditions will not exceed 5.0 Rem dose equivalent.
O
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6.3 Emergency Instrumentation and Procedure Review Instrumentation related to Control Room habitability has been reviewed with the objective of ensuring that operators are provided with sufficient information to comply with plant technical specifications and to evaluate the con-sequences of the identified design basis accidents.
Indication of Control Room temperature, relative humidity, area radiation levels and ventilation system status are provided on the instrumentation panels. Local and panel alarms are provided to alert operators to high radiation levels. Alarms are provided on the security
. panels to indicate open access doors _ which could result in the infiltration of hazardous naterials. Existing Control Room instrumentation provides adequate operator protection with the exception of the design basis toxic accidents identified in Section 6.1.4.
Corrective action recommendations for toxic gas detection is provided in Section 7.2.
Emergency procedures elated to Control Room habitability have been reviewed with the ' objective of ensuring that operators are provided with sufficient procedures -to
- be followed in the event of an accidental : release of toxic or radioactive gas. The procedures reviewed include those for accidents involving 'the release of '
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radioactive materialst post-accident operation of the Control Room ventilation system, and forced evacuation of the Control Room. This review has determined that existing procedares provide adequate operating guidelines with the exception of the design basis toxic chemical accidents _ identified in Section 6.1.4.
Corrective action recommendations for procedures to be followed in the event of a toxic gas release are provided in Section 7.2.
t Page 173 of 18i
6.4 Control Room Sustained Occupancy Review The Control Room design has been reviewed for conformance to the T!!I Task Plan position on operator habitability.
The review consisted of a qualitative evaluation of the system design and operation information provided in Section 4 and a quantitative analysis of the postulated design basis accident data provided in Sections 6.1 and 6.2.
Acceptance criteria in each case is provided in the referenced Section Review I'lans.
Specific areas of review and conformance to the acceptance criteria are discussed.
The physical location of the Control Room, described in Section 4.1, is protected with respect to the postulated release points of hazardous materials.
The isolated location of the Control Room within the Auxiliary Building assures that hazardous materials will not enter through corridors or ventilation ducts.
The Control Room envelope, described in Section 4.2, is-serviced by the emergency ventilation system.
The envelope includes all critical areas requiring access in the event of an accident and excludes those areas _not requiring access.
The capacity of the Control Room is sufficient to accorrnodate a minimum of five operators. Sufficient bottled air is stored within the envelope-for 30 minutes with an additional
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30 cylinders stored within quick access. The volume of the envelope is sufficiently large to allow isolation for a period of five days before CO2 levels beccme excessive.
Food, water and medical supplies are sufficient to maintain the operators for a period of five days.
The Control Room envelope is ventilated by an independent system described in Section 4.3.
A single failure of an active component in the ventilation system will not result in the loss of functional performance.
Isolation of the system from ambient air is accomplished by low leakage dampers. Verification of the systems pressurization capability is performed every 18 months.
Technical speci-fications, described in Section 4.6, for operation and testing of the system ar'e designed to limit enviornmental conditions within the envelope and to assure reliable operation of the systems emergency features should they be required.
The physical location of the Control Room ventilation system intake penthouse is provided with sufficient separation from the postulated release points of airborne hazardous' materials.
The analyses performed to evaluate design bases accidents are described in Section 6.1 for radioactive contaminants and in Section 6.2 for toxic chemicals. -The'resulting corrective action recomendations are provided in Section 7 for those postulated Page 175 of 181
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toxic chemical accidents which require early detection. The Control Room envelope is designed, as described in Section 6.1.3, I
to creclude the infiltration of airborne hazardous materials.
The Control Room radiation shielding, described in Section 6.2.2, 4
is provided by structural concrete of sufficient thickness to maintain operator exposure within the acceptance criteria. The Auxiliary Building layout has been designed to eliminate radiation streaming through doors or other apertures. Further radiation protection is provided by monitors appropriately located for early detection of high levels.
Instrumentation related to Control Room habitability, described in Section 6.3, provides the operators with sufficient infor-mation to comply with the plant technical specifications and to evaluate the consequences of design basis accidents on personnel safety and sustained occupancy. Appropriate ?ro-cedures are provided to protect the operators and ensure safe shutdown of the reactor.
The result.of this design review for sustained Control Room occupancy is that, through implementation of the appropriate corrective action alternatives, operators are adequately pro-tected against the effects of.an accidental release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under design basis accident d
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conditions. The Control Room envelope is adequately designed to provide sustained occupancy under postulated design basis accident conditions for the required number of plant operators.
e>e m 997 no 3Ed-
7.0 PROPOSED CORRECTIVE ACTIONS The following corrective actions are provided based upon the results of the design evaluation and analysis of Control Room habitability.
Corrective actions are provided for each case in which the design review indicates that sustained personnel occupancy may be unduly limited or that additional time is required to implement protective measures.
The installation of instrumentation, modifications of plant procedures and equipment, and the addition of radiation shielding proposed in this Section or equivalent alternatives will be made. The proposed corrective actions will insure that the post-accident Control Room operator exposure to either radiation or toxic chemicals will be within established guidelines.
7.1 Instrumentation and Equipment Recommendations ~
Toxic Chemical Monitoring Section 6.1.4 describes several design basis accidents which-require continuous monitoring of the toxic chemical concen-tration in the ventilation system ambient air intake.
Instrumen-tation is required to monitor for sulfuric acid, hydrazine, anhydrous armr.onia, chlorine, and hydrochloric acid.
Redundant toxic chemical detectors will be provided in the intake ductwork.-
The ' detectors will meet the single failure criteria.. Alarms, local readout, and control logic will-be provided to warn the.
operators and automatically realign the ventilation system when the chemicals are' present-in haza~ dous quantities.
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I Ammonia Filtration Section 6.1.4.3 describes the consequences of a design basis anhydrous ammonia railroad transport accident.
Removal of anhydrous ammonia by the charcoal filters presently installed in the ventilation system is qualitatively categorized in the low capacity level. Therefore the two minute concentration exceeds the established toxicity limit. Two alternatives are currently being evaluated to resolve this item.
a)
Provide control logic to automatically realign the ventilation system _to the internal recirculation mode of operation rather then the filtered air makeup mode.
b)
Chemically treat the filter charcoal-to enhance ammonia absorption.
Radiation Shielding Section 6.2.2 identified that additional structural concrete radiation shielding is required for the containment spray recirculation piping to reduce the dose rate field in the direction of the Control Room. This piping is located in the eastern corner of the ventilation equipment area on the 1025 ft elevation of the Auxiliary Building.
7.2 Procedural and Technical Soecification Recommendations Toxic Chemical Protection Procedures will be implemented. for operator actions following indication of airborne chemicals exceeding hazardous quantities by the toxic chemical detectors.
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Airborne Radioactive Contamination Protection (1) The evaluation provided in Section 6.2.1.5 of the radio-logical consequences to Control Room operators from a postulated loss of coolant accident under conservative assumptions could exceed the Standard Review Plan thyroid dose criteria.
Instrumentation is provided to automatically realign the ventilation system to the filtered air makeup mode of operation. Two alternatives are currently being evaluated to resolve this item.
a.
Procedures will be impicmented to monitor the Control Room atmosphere for radioactive iodine. Should iodine be detected in the ventilation system discharge registrar the system will be manually realigned to the-internal recirculation mode of operation. -
h.
Charcoal filtration installed in the recirculation i
ductwork of the ventilation system.
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Technical Soecifications Technical specifications will. be implemented for.f r.im'an frequencies for checks, calibrations and testing of the Control Room toxic chemical conitoring instrumentation.
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