ML20004D314
| ML20004D314 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/05/1981 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-0737, RTR-NUREG-578, RTR-NUREG-737, TASK-2.B.2, TASK-TM IEB-79-01B, IEB-79-1B, JPN-81-39, NUDOCS 8106090148 | |
| Download: ML20004D314 (52) | |
Text
POWER AUTHORITY OF THE STATE OF NEW YORK 10 CoLUMaus CincLE NEW YORK. N. Y.10019 e212) 397-6200 GroRott.sERRY ortRavine orricsR TRUSTEES JOHN W. BOSTON JOHN S.DYSCN
" '*""" p EslDEN NOCEDURES
""""'U""""'"
CEORGE L. lNGALLS JOSEPH R. SCHMIEDER wsC E CMalRMAN RES CENT S CM4EF LICHARD M.FLYNN co err i. wiLLoNzi " L7,'o7ei^Z...D..?
June 5, 1981 rREoERiC= R. CLARK JPN-81-39 *',%'.'."'"''
THOMASR FREY Director of Nuclear Reactor Regulation ",'0,"',;"',',',*,'"'
U. S . Nuclear Regulatory Commission Washington, D. C. 20555
" (b Atte:ntion: Mr. Thomas A. Ippolito, Chief [
Operating Reactors Branch No. 2 p J S\\
Division of Licensing
( (
Subject:
James A. FitzPatrick Nuclear Power Plant p pti0 3 igBg F dl Docket No. 50-333 # g 49 w* #
E i Post Accident Shielding Analysis G/
NUREG-0578, Item 2.1.6.b Se NUREG-0737, Item II.B.2 g @
References:
- 1. NRC letter, D.G. Eisenhut to all Licensees of Operating Plants, dated October 31, 1980
Dear Sir:
Enclosed is the subject analysis which you requested in Reference 1. This analysis addresses all of your concerns except equipment qualification. As stated in Reference 2, equipment qualification is being addressed in the Authority's on-going efforts in response to I.E. Bulletin 79-OlB.
The shielding analysis contains recommendations to augment shielding in the reactor building to reduce post-accident radiation fields and enhance accessibility. The Authority has reviewed these recommendations and determined that their implementation is not required to achieve compliance with the NUREG-0737 guidelines.
Should you or your staff have any questions, please contact us.
ry truly yours, cw -
J." . B ne
" Senior Vice President Nuclear Generation 8106090
?
- r. '
2 1' '
POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT NUCLEAR DESIGN AND ANALYSIS-DEPARTMENT f
POST ACCIDENT t
SHIELDING ANALYSIS '
RESPONSE TO NRC-NUREG-0578 r ITEM 2.1.6.b t
i L
i F
I I
l i
By R.E. Deem -
/ f. ,# Date M ,,c] 2 9,, / 9ff ,
Supervising Nuclear Engineer {
-Nuclear _ Design and Analysis l I
James A. FitzPatrick Nuclear Power Plant Post Accident Shielding Review Table of Centents Item Page NRC Position / Clarification 1 Objective 2 Scope of Review 2 Radiation Level Guidance 3 Dominant Source Systems 3 Shielding Source Terms 4 Dilution Volumes 5 Assumptions 6 Shielding Analytical Procedure 7 Analytical Results 8 Discussion of Results by Elevation 10 Recommendations 15 Conclusions- 16 Tables I - IX
. Figures 1 -_12 1 References
~ Appendix A Appendix B
e -
9 DESIGN REVIEW OF PLANT SHIELDING FOR PLANT ACCESS AND SYSTEMS WHICH MAY BE USED IN POST ACCIDENT OPERATIONS NRC Position:-
-With the assumptions of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core halogens, 100% of the core noble gases, and 1% of the core solid fission products are contained in the primary coolant) , each licensee shall perform a radiation and shielding design review of the areas around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrumentation areas, in which personnel occupancy may be unduly limited or safety-related equipment unduly degraded by the radiation fields during the post-accident operations of these systems. Each licensee shall provide for adequate access to vital areas and protection of safety related equipment by design changes, increased permanent shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.
e f
NRC Clarification:
Assure adequate shielding of radiation from systems outside containment which may contain primary coolant or gases to assure access to vital areas or equipment.
Field run piping is also to be evaluated where necessary (NRC letter of October 30, 1979). In the case of depressurized reactor coolant, the noble gas component of the radiation source may be excluded (NUREG-0 73 7) .
SHIELDING REVIEW AND ANALYSIS Objective:
To investigate the adequacy of plant shielding against radiation from sources which are outside the primary containment, with the aim of determining area dose rates and location specific dose rates where access may be deemed necessary during post-accident conditions.
Scope of Review:
With the above assumptions and from consideration of post-accident operations and maintenance requirements, the following areas and equipment locations were considered to be of primary interest in the shielding analysis and review:
- 1. Control Room
- 3. Primary Coolant System Sampling Stations
- 4. Motor Control Centers 1 __
e e.
Scope of Review: (Continued)
- 5. Electrical Control Panels
- 6. Manual isolation and control valve locations for vital equipment
- 7. Chemistry Laboratory and counting room
- 8. Areas where emergency maintenance may be required, such that non-repair of the equipment may re.sult in an additional burden on vital equipment used for mitigation purposes or result ,
in an increased radiological hazard to personnel.
- 9. General access areas which may be required during post-accident conditions.
Radiation Level Guidance:
Areas requiring continuous occupancy: less than 15 mR/hr. ,
(i.e. , Control Room)
Areas requiring possible frequent access: less tnan 100 mR/hr. (i.e.,
Radwaste Control i Panels)
Other areas which may require access for emergency ,
maintenance or operational needs: shielding as required to keep exposures less than 10 CFR Part 20.
Systems considered to be the dominant sources:
- 1. HPCI -
High Pressure Coolant Injection System
- 2. LPCI - Low Pressure Coolant Injection System (one mode of the RHR system)
- 3. RHR -
Residual Heat Removal System
- 4. CSS -
Core Spray System
- 5. CAD -
Containment Atmosphere Dilution System (including Torus /drywell purge & vent system)
- 6. RCIC - Reactor Core Isolation Cooling System
- 7. SGT -
Standby Gas Treatment System 1
Systems considered to be the dominant sources: (Continued)
- 8. Primary coolant sampling lines and other radioactive lines smaller than 2
-inch nominal outside diameter.
Where larger radioactive piping ot equipment is in close proximity, such chat it is the dominant source (i.e., smaller lines are less than 10% of the source) , the sampling lines ~are not considered.
Shielding Source Terms:
Pressurized Liquid Systems Noble Gases: 100% of core-inventory Halogens: 50% of core inventory Remaining Fission Products: 1% of core inventory Depressurized Liquid Systems Halogens: 50% of core inventory Remaining Fission Products: 1% of core inventory Containment Air Noble Gases: 100% of core inventory Halogens: 25% of core inventory g
i A thorough review of all design-basis accident blowdown analyses indicate that the primary-system will be depressurized to below 200 psia in under 25 minutes.
Therefore, the depressurized liquid system source is utilized for all analyses performed during the mitigation phase of the accident where a liquid radiation source is present.
Dilution Volumes:
Liquid source volume dilution:
Primary System 80,000 gallons Torus 750,000 gallons Condensate Storage Tanks 200,000 gallons Total - 1,030,000 gallons Containment air source volume dilution:
Drywell Air Volume - 154,476 cubic feet Torus Air Volume -
113,089 cubic feet Total - 267,565 cubic feet A thorough review of plant technical specifications and system descriptions indicates that at the onset of an accident, the initial supply of water to the emergency core cooling systems (HPCI and RCIC) is provided by the Condensate Storage Tanks (CST). With these systems operable, as they would be for any type of design-basis accident (large pipe break, small pipe break or f ailed-open safety relief valve) , the CST volume would be injected into the primary _ system within 43 minutes.
Consequently, the dilution effects of the CST are not factored into the analysis before 45 minutes into the mitigation phase of the accident.
5-
e 3
Assumptions:
The following assumptions were made for this review, in light of the Lessons Learned from the TMI-2 accident:
- 1) . - No primary containment (drywell and torus) entry will be made during the mitigation phase of the postulated accident.
- 2) . Containment isolation is in effect immediately after the accident commences; main steam lines are isolated and the main turbine has tripped.
- 3) . The Reactor Water Clean-Up System is isolated immediately after the accident commences.
4). No radwaste processing of high-level primary fluids will take place during the mitigation phase.
5). The length of the mitigation phase is taken to be six (6) months. This time will also serve as the limit for integrated airborne dose rates in vital areas.
- 6) . For system activity calculations, an instantaneous release of core radioactivity in the percentages given, as specified previously, is utilized. Concentrations will be assumed uniform at t=0 and no credit taken for decay (prior to t=0) or diffasion.
- 7) . The status (mode) of various plant systems with regard to overall plant conditions is taken into account where access to vital areas or emergency operations or maintenance is required.
(.-
Assumptions: (Continued)
- 8) . Primary coolant system sampling lines are only considered to be a significant source when personnel access is required to be in close proximity to them.
9). The airborne whole body radiation dose co..tribution from postulated primary containment design leakage was not included as a source in the results of the sheilding review. (Table VII and Figures 1 through .' 2)
Shielcing Analytical Procedure:
The activity concentrations for the sources used were taken from Reference 4 and are shown in Tables I, II, and III. These sources were then converted to a multigroup energy source using the ACTGEN computer code (Reference 6, Appendix A) for various times during the mitigation phaseaof the accident. The energy sources are shown in Tables IV, V, and VI. The QAD-QC computer code (Reference 5, Appendix B) wau then t 'ed to determine the dose rates for various :ations and areas throughout the plant. When thet. is a dose point location that involvec: multiple sources, the dose rate contribution from each source is calculated and the contributions summed to determine the location dose rate. No credit for shielding from other equipment or obstructions was taken and all piping was assumed to be bare (no insulation) in the models used in the analysis.
Air dose rates in the Control Room were evaluated using the'0-2 hour X/O and breathing rate throughout the six (6) month period. The Standby Gas Treatment System is assumed operating and charcoal filter efficiencies for halogen removal are taken as 0.9 for the Standby Gas Treatment and Control Room intake systems.
The primary containment design leak rate of 0.5%/ day was also assumed. rur the Technical Support Center, similar assumptions and parameters were used, except that no emergency charcoal filters for halogen removal are assumed to be available at the air intake.
Air d-3e rates in the reactor building were
.alculated assuming the reactor building ventilation system is isolated and operating in the recirculation mode. The Standby Gas Treatment system is also in operation. There are only roughing filters on the reactor building ventilation system and they are assumed to provide no filtration for the reactor building atmosphere.
The DRAGON-3 computer code (Ref. 8) was used to calculate these air dose rates and all integrated dose rates.
Analytical Results:
Utilizing the above assumptions and procedure, dosa rates were calculated at selected locations in the plant. Table VII depicts the dose rates calculated at these locations. These dose point locations are shown on Figures.1 through 12. Also shown on these figures are the primary radiation sources considered in the analysis above the 272 ft. elevation in the Reactor Building.
n o--
Analytical Results: (Continued)
To' determine area dose rates within the Reactor-Building, the two hour radiation source was utilized (Table V and VI). With the conservative assumptit.- of uniform mixing at the onset of the accident aqd other assumptions, it is felt that the two hour source wouldJbe representative of the maximum radiation fields encountered in the Reactor Building. The area dose rates are shown on Figures 1 through 12 according to the following criteria:
Area Designation Dose Rate A..... Less than 100 mR/hr.
B..... Greater than 100 mR/hr.
but less than 5 R/hr.
C..... Greater than 5 R/hr.
but less than 20 R/hr.
D..... Greater than 20 R/br.
Air dose rates and the 6 month integrated doses for the Control Room and Technical Support Center are shown in Tables VIII and IX. Table X depicts the air dose rates in the Reactor Building, which are not included on Figures 1 through 12 or on Table VII.
9 -
l +
Discussion of Results By Elevation Reactor Building,-Elevation 227' (Figure'l)
The Cresent Area is inaccessible for the duration of the mitigation phase (6 months) of the postulated' accident. This is due to the high concentration of safety related pumps and systems piping which contain primary coolant. An operational review of the equipment in this area indicate that access is not vital.
Rea; tor Building, Elevation 272' (Figure 2)
The major source contributors on the 272 ft, elevation (Reactor Building Elevation 272' Figure
- 2) are the containment and core spray piping, the Reactor Building / Torus vacuum breaker, and the torus purge line. All areas directly above the torus are influenced by the large volumetric liquid and gaseous source contained in the torur.
Despite the attenuation afforded by the 2 ft.
Concrete floor, the dose rate due to the torus is approximately 86 R/hr at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The gaseous portion of the torus source is the predominant dose contributor from the torus. This is due to the self-shielding characteristic of the liquid volume source. Consequently, since the high energy gaseous source decays rapidly, the torus dose rate contribution to the 272 ft. elevation is no longer significant after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (i.e., 1.4 R/hr). Consequently, access to certain areas of this elevation will improve substantially after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 10 -
p Reactor Building, Elevation 300' (Figure 3)
Sources which contribute to the dose rates on the 300 ft. elevation (Figure 3) are the core and containment spray lines. The west side of this elevation is relatively inaccessible due to the presence of the exposed core and containment spray lines in this area. The existing sampling station located in the southeast quadrant of this elevation is not readily accessible due to the close proximity of the RHR head spray piping. All contributing sources on this elevation are volumetric liquid sources. Accessibility to this elevation will be enhanced as the mitigation phase progresses, but not as rapidly as would be the case if gaseous sources were involved. The sources, however, do not in any way hamper accessibility requirements to the MG set room. It is envisioned that this area will provide the primary access route from the Administration Building into the Reactor Building, if required, on an emergency basis.
Reactor Building Elevation 326' (Figure 4)
The major sources contributing to the dose rates on the 326 ft. elevation (Figure 4) are the portions of the drywell vent and purge lines up to the outermost containment isolation valve, as well as the RHR head spray piping. The dose contributions due to the core spray and containment spray piping directly below this elevation on the west side of the Reactor Building t
e l
l Reactor Building-Elevation 326' (Figure 4) (Continued) have been-included in the stated dose rates. The results show that limited access to the 600 volt switchgear and MCC's on the west side of this elevation can be achieved on an emergency basis if required.
Reactor Building Elevation-344' (Figure 5)
The dose contributor on the 344 ft, elevation is the normal ventilation filter train of the Reactor Building ventilation system. When high radiation signals from the building's exhaust ducts trigger the intake and exhaust butterfly valves to close, the Reactor Building ventilation system goes into a recirculation mode. During post-accident operation, these filters could become a source of radioactivity due to the buildup of airborne particulate activity on the filters. This activity buildup would result from the conservatively assumed 0.5%/ day primary containment design leakage rate. However, since these filters are intake roughing filters, their ability to entrain particulate activity is quite limited. . Consequently, from a design basis analytical standpoint, the area immediately surround!.ng these filter banks is conservatively assumed to buildup to, but not exceed 20 R/hr.,
after the accident. There are no other significant dose contributors on this elevation of.the Reactor Building.
t
- 12 -
Reactor Building Elevation-369' (Figure 6)
.The 369 ft. elevation contains no major radioactivity sources as a. result of the postuleted accident. Therefore, this elevation is accessible during the mitigation phase. The removable shield above the. reactor cavity on this elevation provides sufficient shielding to preclude the drywell from being a dominant source.
General For all of the subject elevations in the Reactor Building, the drywell was not a contributing source since the thickness of the shield wall surrounding it is more than adequate.
Other radioactive sources and areas investigated included the halogen buildup on filters of the Standby Gas Treatment System, exterior areas of the plant, and the accessibility to the temporary post-accident primary.
system sampling station located in the Radwaste Building.
Variation in dose rates from halogen buildup accumulated on the Standby Gas Treatment System filters is shown on Figure 2 (B to D zone accessibility) . The dose rates in this area peak approximately 2 weeks after the onset of the postulated accident, assuming the system is in continuous operation subsequent to t=0.
4
~ _
Y-
,g.
3,-
. -Exterior. areas of.the. plant which would become ,
radiological hazards are also depicted on' Figure 2.
This;is due' totally to the radiation sources in the
. Crescent Area below. Even with 3, feet 10 inches of concrete shielding, the heavy concentration of radiation ,
sources-in the Cresent Area make these exterior areas an ,
accessibility zone B, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the onset of the accident. >
1 The temporary post-accident Primary. Coolant Sampling ;
~ System, located:in the Radwaste Building was analyzed to L
determine accessibility.- The only sources in this area are the sampling lines themselves. Results indicate '
that the dose rate would be apprximately 11 R/hr six
~
feet from.the sampling station 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the onset of the accident with 0.25 inch of lead on the horizontal :i sections of the sampling lines near the sample sink. At ,
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,.and with the same constraints, the dose rate is :
8 R/hr. Without the shielding, the dose rates would be 19 R/hr and 14 R/hr, respectively.
t The permanent location of the Post-Accident Primary Coolant Sampling System, represented by Detector #10 on Table VII and Figure 3, is adequate. The radiation !
sources in the Reactor Building < provide no major i contributing dose to;this. location to prohibit accessibility.
L I
\
Recommendations:
Results of the shielding analysis and review have
' identified certain items in the plant that present radiological problems from the standpoint of accessibility and possible operational. requirements.
Implementation of the following proposed modifications would enhance accessibility in the plant during the mitigation phase of the accident.
Areas which were identified exterior to the plant to be a radiological hazard during the mitigation phase of the accident should be administrative 1y controlled. An administrative procedure should be developed which addresses this situation if it is not currently addressed in existing administrative procedures.
The isolation valve on the torus purge line, located in the northwest quadrant of the Reactor Building on the 272 ft. elevation, should be relocated to an area below the 272 ft. elevation floor, or shielded up to the existing isolation valve with the equivalent of one foot of concrete. This modification would reduce the relatively high background radiation level in the radiochemistry laboratory (Detector #3 on Table VII and
. Figure 2) , and allow access early in the mitigation phase of the accident to MCC's 162 and 143 located near the air lock on that elevation.
15 -
0.
.~
Recommendations: (Continued)
Another item of concern is the Reactor Building / Torus vacuum breaker located in approximately the same area of the reactor building. This vacuum breaker should also be relocated or shielded with the equivalent of one foot of concrete. Emergency access to the east side of the reactor building on this elevation would be enhanced at later times during the mitigation phase of the accident, if this large volumetric radiation source is significantly reduced. At 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of the accident, the dose rate contribution from this piece of equipment is approximately 35 R/hr. If left as it now exists, it will replace the torus as the dominant source contributor in this area and when added to the other contributing sources, would hinder emergency BCCess.
Conclusions:
The analysis and shielding review performed in this report meets the intent and commitment requirements of NUREG-0 578, NUREG-0737, and other associated NRC guidance concerning post-accident plant shielding and access requirements. Thorough usage of design documentation (i.e., drawings, technical specifications, etc.) has been made to substantiate the shielding calculations. Verification of plant specific data was made by walkthroughs and discussions with plant operating personnel. Postaccident access requirements have been developed in conjunction with requirements specified by plant operating personnel. These post-accident access requirements have been based upon emergency situations which may necessitate entry into the Reactor Building after-a postulated accident.
. . 1 Conclusions (Continued)
With implementation of the two permanent shielding or relocation modifications, access to the 272 ft.
elevation of the reactor building under emergency conditions would be improved. Radioactive sources have also been identified for possible application of temporary shielding to plant personnel if access to these aree.s is deemed necessary. Use of temporary shielding would be dependent however, on duration of access and plant conditions at that time, from both an operational and radiologfcal viewpoint. In all cases, both interior and exterfor to the Reactor Building, access will be administratively controlled. Health physics personnel will monitor and possibly survey all areas requiring access.
The Control Room will have a safe and low radiation level at all times. The Technical Support Center will also meet the 15 mR/hr criterion. Airborne radiation dose rates calculated for the Reactor Building could seriously hinder access. However, these results are based on an assumed conservative 0.5%/ day primary containment leakage rate. It is felt that this is a highly conservative leakage rate and that the Standby Gas Treatment System would greatly reduce the airborne radioactivity concentration in the reactor building if a more realistic leakage rate is considered. Inleakage of air to the Reactor Building of approximately 6,000 CFM
-during operation of the Standby Gas Treatment System would also constitute a substantial additive dilution volume, thereby further reducing airborne radioactivity concentrations.
-r -
v .
Conclusions (Continued)
As a result of this shielding-analysis, in conjunction with'the proposed _ modifications, it is felt that the
. Reactor Building has been made as radiologically accessible as is practical. Additional modifications to further reduce the calculated dose rates would necessitate extensive redesign of the major. plant systems with little radiological benefit.
TABLE I Pressurized Liquid Source
- Isotope Activity uCi/cc Noble Gases Kr-83m 1.87E3 Kr-85m 4.05E3 Kr-85 1.87E2 Kr-87 7.49E3 Kr-88 1.06E4 Kr-89 1.2SE4 Xe-131m' l.12E2 Xe-133m 1.25E2 Xe-133 3.49E4 Xe-135m 1.06E4 Xe-135 6.13E3 Xe-137 2.93E4 Xe-138 2.74E4 Halogens Br 9.36E2 Br-84 1.47E3 Br-85 2.03E3 Br-87 3.44E3 I-131 9.05E3 I-132 1.31E4 I-133 1.50E4 I-134 1.94E4 I-lSo 1.53E4 I-136 6.26E3 Particulates Se-81 9.359E0 Se-83 9.359E0 Se-84 9.359E0 Rb-88 1.059E2 Rb-89 1.373E2 Rb-90 1.684E2 Rb-91 1.745E2 Rb-92 1.684E2 Sr-89 1.372E2 Sr-90 1.559El Sr-91 1.804E2 Sr-92 1.932E2 Sr-93 2.306E2 Sr-94 2.245E2 0
At reactor power level of 2436 MW(t) l
[
4 h
TABLE'I (cont'd)
Isotope Activity uCi/cc Y-90 1.620El i Y-91m 1.121E2 Y-91 1.932E2 i Y-92 2.058E2
.Y-93 2.370E2 Y-94 2.493E2 Y-95 2.816E2 zr-95 2.918E2 i Zr-97 2.867E2 Nb-95m 5.811E0 Nb-95 2.918E2 Nb-97m 2.867E2 Nb-97 3.046E2 ,
Mo-99 3.123E2 l
-Mo-101 2.918E2 Mo-102 2.739E2 Mo-105 1.059E2 Tc-99m 2.739E2 Tc-101 2.918E2 Tc-102 2.867E2 Tc-105 2.058E2 Ru-103 2.557E2 Ru-105 2.183E2 .
Ru-106 1.185E2 Ru-107 1.246E2 Rh-103m 2.493E2 [
Rh-105m 4.608E2 ,
Rh-105 1.871E2 Rh-106 1.246E2 Rh-107 1.310E2 '
Sn-127 1.620E1 Sn-128 2.688El i Sn-130 5.811El .
Sb-127 1.932E2 Sb-128m 2.816El Sb-129m 5.171El i Sb-130 1.185E2 Sb-131 1.559E2 Sb-132 1.684E2 Sb-133 1.185E2 Te-127 1.932El :
Te-129m 8.115E0 Te-129 4.915El Te-131m 3.865El Te-131 1.497E2 Te-132 2.611E2
- Te-133m 1.871E2 L Te-133 1.121E2
, Te-134 2.918E2 ;
1 l
a- ,
'l TABLE I-(cont'd)
Isotope Activity uCi/cc Cs-137 2.245El Cs-138 3.174E2-Cs-139 3.123E2 Cs-140 2.688E2
.Cs-142 9.984El Ba-137m 2.06El Ba-139 3.13E2 Ba-140 2.92E2 Ba-141 3.05E2 Ba-142 2.44E2 La-140 3.12E2 La-141 3.05E2 La-142 2.69E2 La-143 2.56E2 Ce-141 3.05E2 Ce-143 2.62E2 Ce-144 2.1222 Ce-145 1.6922 Ce-146 1.31E2 Pr-143 2.56E2 Pr-144 2.19E2 Pr-145 1.69E2 Pr-146 1.37E2 Nd-147 1.06E2 Nd-149 6.87El Nd-151 3.31El Pm-147 4.18El Pm-149 8.74E0 Pm-151 3.56El Sm-153 5.13El
'l i
I 1
L I
TABLE II De-Pressurized Liquid Source
- Isotope Activity uCi/cc Hnicgens -Br-83 9.36E2 Br-84 1.47E3 Br-85 2.03E3 Br-87 3.44E3 I-131 9.05E3 I-132 1.31E4 I-133 1.50E4 I-134 1.94E4 I-135 1.53E4 I-136 6.26E3 Particulates Se-81 9.359E0 Se-83 9.359E0 Se-84 9.359E0 Rb-88. 1.059E2
-Rb-89 1.372E2 Rb-90 1.684E2 Rb-91 1.745E2 Rb-92 1.684E2 Sr 1.372E2 Sr-90 1.559E2 Sr-91 1.804E2 Sr-92 1.932E2 Sr-93 2.306E2 Sr-94 2.245E2 Y-90 1.620El Y-91m 1.121E2 Y-91 1.932E2 Y-92 2.058E2 Y-93 2.370E2 Y-94 2.493E2 Y-95 2.816E2 Zr-95 2.918E2 Zr-97 2.867E2 Nb-95m 5.188E0 Nb-95 2.918E2 Nb-97m 2.867E2 Nb-97 3.046E2 Mo-99 3.123E2-Mo-101 2.9 1 8E2 Mo-102 1 '39E2 Mo-105 1.0S9E2 586-0067
TABLE ~II (cont'd)
Isotope Activity uCi/cc Tc-99m 2.739E2 Tc-101 2.918E2 Tc-102 2.867E2 Tc-105 2.058E2.
Ru-103 2.557E2 Ru-105 2.188E2 Ru-106 1.185E2 Ru-107 1.246E2 Rh-103m 2.493E2 Rh-] O 5m 4.608E2 Rh-105 1.871E2 Rh-106 1.246E2 Rh-107 1.310E2 Sn-127 1.620El Sn-128 2.688El Sn-130 5.811El Sb-127 1.932E2 Sb-128m 2.816El
-Sb-129m 5.171El Sb-130 1.185E2 Sb-131 1.599E2 Sb-132 1.684E2 Sb-133 1.185E2 Te-127 1.932El Te-129m 8.155E0 Te-129 4.915El Te-131m 3.865El Te-131 1.497E2 Te-132 2.611E2 Te-133m 1.871E2 Te-133 1.121E2 Te-134 2.918E2 Cs-137 2.245El Cs-138 3.174E2 Cs-139 3.123E2 Cs-140 2.688E2 Cs-142 9.984El Ba-137m 2.06El Ba-139 3.13E2 Ba-140 2.92E2 Ba-141 3.05E2 Ba-142 2.44E2 La-140 3.12E2 La-141 3.05E2 La-142 2.69E2 La-143 2.56E2
.Ce-141 3.05E2 Ce-143 2.62E2 Ce-144 2.19E2 Ce-145 1.69E2 Ce-146 1.31E2
'0 At reactor level-of 2436 Mwt.
g ,. . , ,
f TABLE II (cont'd) i Isotope Activity uCi/cc Pr-143 2.56E2 i Pr-144 2.19E2 {
Pr-145 1.69E2 L Pr-146 1.37E2 I Nd-147 1.06E2 1 Nd-149 6.87El i Nd-151 3.31El i Pm-147 4.18El t Pm-149 8.74E0 !
Pm-151 3.56E1 [
Sm-153 5.13El
{
r i:
I t
- i. !
t- [
I
- At reactor power level of 2436 Mwt a
A I
y e 1 i 1
-t i
i s
J w -- e , ..-n + ,-r,--- ~ ~ . ,. ., + o + +- w g,a
. . . ~ _ _ _ _. _, _ _ - . . - _ .
TABLE III Containment Atmosphere Source k
Isotope Activity uCi/cc NSble~ Gases Kr-83m 9.6502 r Kr-85m 2.08E3 '
Kr-85 9.65El ;
Kr-87 3.85E3 -
Kr-88 5.46E3 Kr-89 6.43E3 !
Xe-131m 5.78El Xe-133m 6 43El Xe-133 1.79E4 Xe-135m 5.46E3 Xe-135 3.15E3 Xe-137 1.50E4 Xe-138 1.41E4 i
Halogens Br-83 2.41E2 '
Br-84 3.77E2 -
Br-85 5.23E2 !
4 Br-87 8.84E2 '
I-131 1.33E3
~ I-132 3.38E3 I-133 3.86E3 i I-134 4.98E3 1-135 3.94E3 i I-136 1.61E3 '
i i
i B
5 b
o.
0 TABLE IV :
Pressurized Liquid Source TOTAL SPECIFIC ACTIVITY - (Mev/cc-sec) AT ENERGY (Mev)
TIME .40 .80 3.30 1.70 2.20 2.50 3.50
.5hr .3.423E8 2.266E9 1.065E9 9.622E8 5.821E8 5.936E8 1.140E8 1.0hr 3.350E8 1.792E9 9.042E8 7.942E8 3.501E8 4.829E8 9.502E7 !
2.0hr 3.262E8 1.240E9 6.776E8 5.306E8 2.306E8 3.099E8 5.230E7 t
8.0hr- 2.960E8 4.642E8 2.663E8 1.961E8 6.254E7 4.764E7 7.874E6 j 24.0hr 2.380E8 2.256E8 5.015E7 5.083E7 6.148E8 1.928E6 1.469E5 4 days 1.450E8 6.350E7 2.631E6 1.617E7 4.264E5 9.840E5 0.0 14 days 5.120E7 2.898E7 8.875E5 9.072E6 5.014E4 5.820E5 0.0
'30 days 1.118E7 1.853E7 3.927E5 3.830E6 1.649E3 2.447E5 0.0 90 dayc- 3.119ES 9.826E6 1.257ES 2.118E5 0.0 9.486E3 0.0 180 days 9.465E4 4.850E6 9.430E4 6.332E4 0.0 7.239El 0.0 1
_ _ _ _ _ . - - ~ - - -
I' e
TABLE V Depressurized Liquid Source TOTAL SPECIFIC ACTIVITY (Mev/cc-sec) AT ENERGY (Mev)
TIME _
.40 .80 1.30 1.70 2.20 2.50 3.50
.5hr 1.235E8 1.956E9 1.053E9 8.806E8 1.080E8 1.061E8 1.132E8 1.0hr 1.192E8 1.577E9 8.932E8 7.260E8 9.380E7 7.840E7' 9.502E7 2.Onr 1.161E8 1.093E9 6.690E8 4.788E8 7.570E7 2.670E7 5.230E7 8.0hr- 1.112E8 4.173E8 2.644E8 1.844E8 2.994E7 1.250E6 7.874E6 <
24.0hr 1.037E8 2.164E8 5.011E7 5.061E7 5.542E6 1.119E6 1.469ES 4 days 7.862E7 6.348E7 2.631E6 1.617E7 4.264E5 9,840E5 0.0 14 days 3.431E7 2.898E7 8.875E5 9.072E6 5.014E4 5.820E5 0.0 30 days 8.922E6 1.853E7 3.927E5 3.830E6 1.649E3 2.447E5 0.0 90 days 3.156E5 9.826E6 1.257ES 2.188E5 0.0 9.486E3 0.0 i
~180 days 9.463E4 4.850E4 9.430E4 6.332E4 0.0 7.239El 0.0
?
[
t
[
s
, , , - , --*g
o TABLE VI Containment Atmosphere Source i
TOTAL SPECIFIC ACTIVITY (Mev/cc-sec) AT ENERGY (Mev) r
.4L
. TIME .80 1.30 1.70 2.20 2.50 3.50 I
.Shr 1.382E8 6.408E8 2.945E8 3.227E8 2.712E8 3.005E8 5.218E7 1.0hr 1.355E8 4.897E8 2.540E8 2.618E8 1.557E8 -2.453E8 4.536E7 ,
2.0hr 1.303E8 3.258E8 1.872E8 1.595E8 9.907E7 1.578E8 2.575E7 !
8.0hr 1.103E8 1.068E8 5.941E7 4.826E7 2.425E7 2.390E7 4.029E6 ;
24.0hr 8.509E7 4.570E7 1.163E7 8.301E6 1.532E6 4.165E5 7.564E4 4 days 5.187E7 7.984E6 6.588E3 8.564E4 6.935E2 7.ll4E3 0.0 j 14' days 1.695E7 2.022E6 0.0 2.732E1 0.0 0.0 0.0 ;
'30 days 3.148E6 5.ll3E5 0.0 0.0 0.0 0.0 0.0 ;
90 days .l.318E4 2.957E3 0.0 0.0 0.0 0.0 0.0 180 days 1.057El 1.300E0 0.0 0.0 00 00 00 t
4 I
.m., . - - -
._ , . . _ - . _ . - --y-.
TABLE VII Reactor Buildirq Dose Rates (R/hr.) .
Detector No. ,
_Dcse Pt. Time 1/2 hr lhr 2hr 8hr 24hr 4 days- 14 days 30 days 90 days 180 days l 9.40 6.2 4.0 1.0 .15 0.0 0.0 0.0 0.0 0.0 2 1.06 .64 .40 .083 .007 0.0 0.0 0.0 0.0 0.0 3 .97 .64 .40 .083 .007 0.0 0.0 0.0 0.0 0.0 4 14.3 9.45 5.93 1.30 .18 .04 .02 0.0 0.0 0.0 5- 75.11 49.4 28.3 6.82 6
.87 .16 .07 0.0 0.0 0.0 13 2.6 87.0 55.0 11.8 1.35 .19 .09 0.0 0.0 7 54.8 38.0 25.8 0.0 9.06 3.14 1.02 .47 0.0 0.0 0.0 8 907.8 617.5 417.7 142.5 119.9 37.5 17.1 -- -- --
9 31.6 22.3 15.5 6.0 2.19 .73 .34 .16 .06 .026 10 .15 .103 .071 .027 .010 .003 .002 .001 .0002 .0001 11 21.2 15.0 10.37 4.06 1.55 12
.51 .233 .110 .039 .019 198.0 139.7 97.1 37.9 14.5 4.8 2.31 1.05 13 105.5 .368 .169 74.4 51.7 20.2 7.7 2.2 1.15 .54 .19 14 1.50 1.06 .09
.75 .293 .112 .038 .018 .008 .003 15 1072.5 755.8 525.2 .001 205.1 77.6 26.0 11.7 5.5 2.0 .94 16 5.97 4.0 2.6 .82 .26 .08 .02 .004 0.0 0.0 17 1.95 1.28 .80 .17 .057 0.0 0.0 0.0 0.0 0.0 18 1.44 .962 .637 .2 .06 .02 .0 06 .001 0.0 0.0 19 18.4 12.9 9.0 .35 1.4 .47 .22 .10 .04 .02 20 4.56 3.2 2.3 .87 .35 .12 .06 .04 .01 .004
.. .Q:,
TABLE-VIII ACCIDENT' AIR DOSE RATES IN
.THE CONTROL ROOM Time After Accident-(hr) Dose Rate (mrem /hr)
Gamma Beta 0.5 .001 . 013 1 .003 .037 2 .008 .095 8 .022 .40 24 .021 .60 96 .006 .30 336 .002 .08 720 .0002 .01 6-Months continuous occupancy Integrated Dose (rem)
Tyroid Gamma Beta 2.6 .002 .11
TABLE IX ACCIDENT AIR' DOSE RATES IN TECHNICAL SUPPORT CENTER (TSC) AREA Time ~After Accident-(hr) Dose Rate (mrem /hr)
Gamma Beta ,
0.5 .0035
.07
.1 .0075 .15 1 2
.013 .27
'8 .022 .62 i 24 .017
.69 96 .004 .32 ,
336 .001 .09 6-Months Continuous occupancy
-Integrated Dose (rem)
Tyroid Gamma Beta 28 .002 .12 i ., , y- _ - .- y
TABLE X ACCIDENT AIR DOSE RA7ES IN
- THE REACTOR BUILDING Time After
' Accident (br) Dose Rate (rcm/hr) H Gamma Beta f
0.5 5.30 E+01 1.60 E+02 1 8.20 E+01 2.50 E+02 ;
2 1.10 E+02 3.70 E+02 8 1.30 E+02 6.20 E+02 24 7.90 E+01 5.90 E+02 i 96 2.60 E+01 2.70 E+02 336 8.00 E+00 7.70 E+01 720 1.50 E+00 1.40 E+01 2160 8.70 E-03 1.90 E+00 4320' 2.00 E-03 1.20 E+00 i F
5
l l
nh' 1 ;
l i;i y
-; let -
I
- . Ih.l s :
fc !
.li yf;.
%sy
'. I 50
'agt
. , e' a' . .
lE*
yE l 1/
F a
?.tTM?.h !* ' .,:. r, ..
at i l' l?l- 3 I. f, .$$
~> f)- r f . 's A 555 Y I, 4 '
I
/ _ ./x T l !' k
. - .. _y-. :
l
"!l f. d; .l,i a
g h tune i
i q
isij 1o.
ll $l4ll e .
.' Y~ -.' Ay .~.. ' . $I
' l
,, i .
I-*d lig
.x,.
l *
- l . .
" " ' ' ' I D '(I @~ Yi ,.
.s g N y , ..' . gp
. .j ;)., hkx l f
- . g oc % %
N j ,/ . ' .
\ \. I , s in RQ l
.i 's f
- a ;.193 ' ' 0> h, t N
s
\ \ V V y /g ylOl j gj N fe !,- ! .
. rl Q' s' 6 , ,
s 7
- 2 s l ,.
ll A7 w s x4 p r %4- > ( :gs y .
"! M e m-n ma a, c
pa f 4
- 3.;)l,, p&;3 a . a > 4, a,,x,
,Lj
) '
C C
\ 3; a
. K
- f; r,
C} t s . ..
5 Ill lj) ... '
s .
\
s t-- - ,
. .3 i
... 2.
'" o '
~. J l j!;7
\ i i
\
1
- y. . __
s - ,.
's
- s. '
.J h,
- d ;,)
'ly'
!, i .' /./
\ '
ll;g G l N /
gf ,}
. :1) ..
h20 t!!
m,i;am ;
p hg~i~ t3.. x Ngw, nKSn )1 jj d ~I! ; 1 2.
u j
,s , +w _ a -- u -- aa, - - - - , - - - - , - - -
O o
i s
!n a P00R ORIGINAL. b 9 0 W g*ma "
i=
{ i[
1
=oo ! i!
r
. smuo l i 1:D-t y!:
l
~
sl . .
o ,m
,~.
8 hE-].;. :\-h r
e I j
i g
' N*!
., il V .
"p
= ;
j s i
~ ~
Cm ~
l
~~ " .
-,= u ,
1~!g i
'~'
!l.
.8 j' ; "D%"-- -- I"" - '""
i""w3w ,
,lb 31 3 *I gy
(. *
- j l 1 g.;; -
7
' pgJE.E,E m= r-- '
p8 0l g
,j Q
y
, ,, e g a amm := g! 3 l
_ _i .Q 3
d %@ j gj [.!j;" ,1
,j /N [ Tf.rsg. [','8 .
I J si h[jp14 ,rf r %.:
8 g il,s e,y 7.~
i M _m l
i
- d 5+ ' 4 ,=o 11 l
i i
- I @ ,L'
- ,1 pits ~; !!g d w -
- r= --
-g i , ya j l!.l.,A g :,,. -
re+n.__._ a
- ,[ -y xla2 jh n-3 x_ =g_=- g Zi J,.y 9 4, , -l E ]~ =$y' , . a'
)j ; i %,g fi ..
%. _ ,4. - ,.x a '
[3 (
!! t i, pg . .. ,,-
g'-,.
y O h .t t[
\
- n- o -
f g,
- 4 .pg b y g. ,, gm2, n c ,1g, . e .. y ;
9%, !
v
, v , , .o t r. :
a
"" ~
..if,j
~
yQ * '
O .)l t { - - .
s] -Q: g -
men-- _ =
3 f,sl@:iv.7,a.M- p 1
i.
N . \
f .. id45 %a9 ,e. M_
l
.. w m
-- m. f ~_----
v ,yl @, ,
w !u+.ca<
u L, .. ' I /
l L l y jl .
A@ $. ; ;;$^=M'8_
a.: 7 (' a) :,
1
na l ne
. .' O L
7 7 i SETM e . - -
r
... ,, M MS'* _- _
\, $. ,. ,
f oen==
8
.. .. l j
efN mo .
f'M.tG'M T' s.
" CD-~ 'h @ *gm 3i ;
LQ I " '
n */d CON 7' F## i I
B- j ***--
V e LINE 'A' l
-,p ' '
o
W ' Q . n y:~- J
, I .. e, 1,
_ , " amen. ggg e W 2 RHR(wngy) d.. hi f, . f7
= =8 ==Ql u
jggy
, .m
'I @
le - '
um . - e,4 b@:Eg i .- .D 1 -
1 A < 100 ~ U na 1 6x1 v :.
L.m - 0- r ! w. ouT' B 4 SR/HR 5--?
y??
j 1.'~~' m -- w-,
un
-n -
a c < go p/u
,, w .'Y f A + -
Q,; :- y-l,.h .
k
- x. , ,&
con ua y
~
y y gg gjp,fg 3-~ ,
aans umasun h D
wp thg/
l j ,?
s-( .1 <-
[
w* t.- AM T-L
[- m
. --+
pgy g w \}. < W;
. =*
-[i i t. ; . f, -;;cw. - ~ -
r 4e'
... ..j1 ; $ ; 9 ~
tk '
K_t_.qu d- @ll2
.1
! - -r %
+
s-1 w.m. t.
- d. (w -
.~-
_.1, p
i-a D ""
'C:lll2
.,;;scs
]hs . . . . . 7 . . . . . . J
A R .y.
'Y c ansee
( h"'5m -
PT'l
~ i_ h a,b h ,
RHp cont spaA1 *e* J -ph ;
- y. +.u ---
m- _ . _ . , . _ .
f
__6 [ >
l p -
7.-a q z_ $=_,r-Q p_.;n m . _-j r i i g\__4_,y__
r= u-- , - s g % f-- .
3 p--
..y a p . - . - -F r-l l
~g" xQ-- 5' . h q'i_
I 1
N'.'.".
t m". . A a E'U- 4 =*
. f
- i s_ ....
m 4
n 1"" p. ..r _
g _ . . .. m
) _
~ ' ...
..~.?.0 %*
( MTM ,
REACTOR BUILDING PLAN-U. 300'-#
JAMES A. FIT 2 PATRICK asOCLE AR POWER PLANT J j,
~
,.,, POWER AUTHORITY OF Tb4E STATE OF NEW TORE
~ #
j PLAN EL. 500' O".
t l.
Y. ...
,~..- 6 i '
L
{ ar t'. _
re
& ne _ ans '_ ,,e i ., e l
l --
"L. . j j
~~~~'
"A qB DET 20%
-7,, 5
% < 1 m' W;;,- C -
- i a ,,,. O,
-.- -. . / y i
es --
- -*-' O ~~ - A DETl' m'%[ k
.... u 1 s
.sa- m --f_.) /
L y , , , ,
RHR, HEAD SPRAY.
g := '
LJ I
wesrt
- ~
C -
- l i '
E "8da, mB
- V s P.F... q) i '
i 3, N?"l M'
. 0, e.s e. w
- m .--
i " .~j . e .
%p . . .
- 5. - w =- - . cp-- -
_3 ,.
- e. ...
g
- g. % ., , m_
n- c!- :::: i - - - -
g . . . . .
- g.,g.,v, ,,,..., ,-
3 '. F-f' ~ ' *.5 g DET 18 ggy 22 '
42L g., .,.,.,.,
l ,'-hN / ~ e:..fr, %H
. .. 3 0 -
iJ A < /Co #/N4
'L 5
T,.k: ,_ Mgj
.- m,nC . .,_
g _
3(t ,
= (f. ,'
f's
. F c h /l _
l
't 3 a = ..,.
B < WNR C < 80 A /N4 52?M %-e ..k'M -v ;,,.; m D > atPh'n sOh '-h Os
%.:,,\
- f I
/'y .
Alsa
- O
- .e s
p
, '% .? O .m i . s:: . - u , s .t. .w . (_j I
y .e.e <t
, , h *,,. . , .. , . - - -
. *c. .h M ' 9e .- l., D- -
i j 3 ru, t e <D d'5 r
l,- :D.@
l @
s __ du B =_ .- - .
m i
== s
'. t s
f i
DET O i e '~-
~
9 g
' yr* N '
%. : y
N Nam tocx' p
4;t' ;' ~" }
i CONT. PURGE ' ,f,,,.
f --
N 1
A
(
j - ' ,Pm +
l H.AN EL. 326*9= y '
mEAC70R BUE.Dess PLAtt EL326'-9" g JAase 1 A FITZPATENCE GRACLE Aft POWJR PLafft l g
- Posth AUTHORITY OF TM STATE OFIEW M
- q. 4- . . . 9... q ... O -
_. . u r_ _ _ _ . _ _ _ __ ._ _ .- . _ .r7 r v W T 4 e a L[ ;.___
9 _a_r -?- tLs. __A ** r . L._ _ . e E _
et E. . er .__* ~
j'. l
- r. . _j. '
- w -
-**, j ' 6y " **
'- (
y f g 1. .
9 if~fl3.
+
~
i
'l I g A -
p ,,**s -
g i _. - - -
A~ ~ - i -- - =h - -- " 7- '.V'** ~
' .-:= . ' - - -
! A '~~'~' " . L.:-_ J ' ' ~' !
)! -,J ! G' y- DC._ --
g 7:;!!a.p.
j
! f yi l' '9 s .A !
__m _.
-i'?.... t .,,,..g; -- j' r
.=.-=n t ~ ~ -...-
,h j 08 'T .
4 -=--- - /~ A~ -
~ . - -
n' XEY s _o ..
,' ~
.s
',a._
,e t
IQ:
a...-
-c~-~~-~ _1. . .
~.y
. .~,.%.,
~.
_j .
A < lcc =R/HR a < 5 p/xa
. s. ..
.i ; , *;~= ~.
'- ,o _ .. jyp
. . 4-
. . . 's. ..-
.' .__,..- ,. D > 20 A /N4 T c
N
=M B ) ' , 73~ i K, .. , -
1
/
O c
@ +g % 7 [ = e l...-.[ " ~] %
d,E 'T ._ d .. ~ A 5i 4
..9 8
- i. A .. -." -.2 C3 i
1 ['. ~g I
,c.. t' mammmun
.. t g g PL N EL 344'-6" lt='.
km. .
ne e RE ACTOR BUILDNG PLAN EL.344'-4*
JAMES A F172f% TRICK teUCLEAR POWER PLANT POWER AUTHORITY OF THE STATE OF teEW 10fBE k
agrw M a A n '
w _.-4,.es
=
=
y' nn nnimuUn UI\luilii -
o N w
II ss X ,
ll
=
i i N o( R4 6 ' !!! -
.iill,ji l
.. .. e n t,
- nsu
o s m&& *'
g -
X5 d-l5f' 3.,14 ils ; ; yvv^ w =X E:h.jll 11Ebn! sm o n d 22
!llQ' ;g!]l g '
s mu g g
i !!!!!.gN . r lj--) g .
! g Illl,n,!.g!ii6i u " @ g53 l i .y-s
,i EWII I I II ll = I
! !!! i i I
?If , Iu 1.( 11 \- .! ,
t - - ' - -y
, .r ' -, -r 7 ----.
. p) ; ,
E 1 .$1 \
eh1P.
- N\
! 4 { '
i c E l
_. s 1 Jt , a"(G - .
w q, '
,- : .- l-j
'4 h,Il l -j $,.'Q , E ,
1 90;3-
.i s g
n: 11 i -
y- - -
e,
-T i
-g A(
+
i .
e i i i i i...
_ :;p i. g% .
1 _. l'
=f l li!
k
~
N
._ _i _4y ... .
j 2
= 3,
! M i ,
,p A, ,'1 '
/x +"-1 r 1 i t
a!
b, [4:_l !:
ye
_3 w
-r * -1 3 ,, 1 g .
b hi.
7 4
. o' o
,, / ?hf )4 lNa' f!!. b'!, 54 _ g7 't---
m f { [ gl 13 .
l . k i V l.
3
.i l - -.
. . . .i _
0 ' q(;
I d ' _./ / _4. l i
s m
[,
i t{ l :! >3i.,1!i\i i 1
fg 6
S .l s g !
ll N ',- @. s
.i .
< l
., (
. i I.
'I ; h.[ A ' ".1 i i- ,1]
a
! o s l -,
nt a
4- Q_&,~3L'p. '! f.m w\.,_, .
i J lrilT--_ .I ,
a
>~
- j i!! -!< ij jll i, 1, n ile I! -g /*1 ;
< i 7
._8 01
. ~
t il g.p *n "s.q:j.igU % h
,rqu 2 + ke 1
n ' ,,.)JM
- p#u d Jjf'A..If l'
._ [Ti.k*A! i e 1, . l 31 4h __
Y1_f. L- Ib _ __rh$ _
, m 4_._
-- Q
" i
>=
i,-- .
g -
-=. s .- o vr-----.h. ,. ,, ,. , ,. .
l- ,g
- e e o
- e , a s**W .p sww t
Ji J'{-
, l l
u b
- l. Ng I h, a -
=
8 !
Id > '
l l
HA-l?
l -
s ag l h > < i l
I%'f.. ] l 3 ))Ql i
] -M
db
] .
l l b ., >
y
' ' ~ '
- ,0.-$ ' .~-$,*;. y ~
I l e&*-. CC 5 .E it i p .:
'..w.Jh a . ie.
j P d g l
I '
I l l7+-,c- u --
r- . . - .
n _, ___, R_ . - - _ . ._ ,
,g.gg..--
--- 1%l'
( N p . v c.
- , . c ;, n.
!; h !! fl. _ .. kl !',[ I ,
_N s
, , _ g ',# j i;if" x L
) ,
hl !r ; s L x.
J;Il
- :-w .
, 1 %s_iJl![-
F z,,...
d"'
a b; ,
_\!
I.
,J 4 l --s s.: @-
, . , %,.,g.j
( -'".ff, -
- ~.w'."'." ., .jpI.0;f;3fr ,. }r S I, h.:g,3.
43 l _N_. ,
- u. . . . e i
..I l/ l. m II l! !!.,., !.' ,$j {N, bl I s.~ . l
- 9. 9.:
N . ;.' I
( , *d 'i l l fl _i n.3 .,.
7 i i
i ...
I
,. r _ . . .a
- 1
- l ,l g v
&I$4 ll ; ' 5 ' ' **.
fll! 5;?'_
'$$E 8 ?l% %.l l
^
l t
I /
m n..<!
j I/ / N%m Nl } .N6i-U
+_:. .e. .
l ' ,
j
~f '4Mb \ I
~~
S
] jll' I
ll =
w Ir&-e g L k- , ,[,
f- g 3 g 4 l
1 i I ,
i
- g 5 , ._
', s ~
. E** ' n -- M.
r-=-e h)<.4 S'll f;' E )
II
, I
. r=
. _ _. % y ,"M, : 1 7 .
, i .
_ I #I , k___4F: .
m t__ b,
% O
? ..
n.
M 8
S Sh u Ly
- 9!b E!!
i !>
g" s:s il ey *-s o srd I .i II l l 11
= n !1; s ,. ,
q .q .
.l tg u+
r w 4 I, I -
p n
- s. , 4g ; _g .irm l
.r ...
% .a- ..
tm l %(
i l- [
.a -
! .-le L_ _. O a
, v -
g -r-
[ {~~ l i .. l,5 ib7'l. g \ : / ) -
[
4r. -- li
- j.- 4 ., e p
p-n i
%_.t s, e
- s. 1g.f_.
.a ,
g _g .---; e i 5
a
- 8 8l ji. x .
l( $ .
- I 74 x .,. u -t;- '
)'
t-- t g; .o 7_ihp o .- }!
._ qm_w. -
i, ;:
i j Vdo- m t .
r e$ .,
r..
j
'?
~
i l ( _ s>> lI . .,
li
'.jf.,p; i-E io
.i
- I
,). i 8
[- _
L, /. 2 rx -~, ,
]. .j 1- _< , 1 ..
in a- m m-e q = y , y _ j) j i j'
g j ,s.
.- 4 g,t p e ,
,j *. 3 I, d
lg f: .,,
i I g ~-:
-llhg gm. _ :.."I .. s\ .s a
s 1 -
c ~m ....Lll s ! - (i:
.Al)ri U O~ ] [ 3~ ~
M
\
1 -
t 2-
$1 E? [o, lIl r- T
- - I
' -- ~
i d + l- ~ is i J e-g 1 ln Il.
1 -r --
9 1.i 3
=
,, y 1.. -
_- y .,
j y . I
, , ~., , i il II i r
. s . i ..
-_ I 3' I __b _. i
1 5I
-a u
o
$56g
. l t ii a c
s :
c 4
fD "
I E
- I
+p 5 .
-1 i Ia
. s d Y ,
r~=
lf ,i
~
- p. .r..-
3ra _,
- -=-
r2 ::.r- :
- w -
_l a 9.e 7 3] 2 [L, .ll ;
.-si -- .! .a.\x-..... :.g !
!l1E ,
J: T
-,... '\
~ ~ ~
k }
. _ _ . -mi i
. - + --. '. .. .L ._. >
I
[.7 Il ! I i r,..,_ __ g 2- a $, r.I. .
w w@
a c.--e _ ..
l
- .: - A
~ '
!, '" s i .! ;
r1 mm ;
== E I
j'i l-i
- g. ' yid '- -
g mt-c
" "v< s4ux.-
' l
,jr -
g 3 i .' - .
e E . _. .-%a p
~ ~ ~ ~ ;
~ ^
l .g,-
[__.. j.
4.
- , I l
~
l lg i I x - -% :.
~,
kr-l I o,
= ;-
_.y..- .
N N' i - -ks.-ty n * . . . . ..
J.,
,j,,
, 3--
C I f
, _- ._r- _- . _ .
b t n
- l%]
. - - - , - -l P00R ORIGINAI t
i
,5
{ iggn* p m'a -
' L1 -
2 gl a.
=y h
! ;9J-- p-
,, s, a- =
kg a e p!]; fh
-~
!a I ,
5
, g; i
& / . t gg yJ E a.g *n:-
1 -
~ v i u ~y e .
' ~;a u g'( w!fj f y ' -:
a > s.
a tuf
- ~ . ~. J, <
. a.
a n
~
i o #f5. , .
I c. .I j!' .l.0 N EM $[
I : f I-L %- wg11, j.. ..
h Y ll:
-~
.t Y*j li A g 5
2 "L
\
i t pI --
-4 7-;
iw:.g;- '-
e -.- - ->r_t c :.a -
l 7 '.4.t.c. - ,
' ,- m:
. ~
l,-
m 9g -,
o
';2],m 3 8., L 3
r7---
j !
a a.
A- ,ye - ,
=
g$
N l
~
I I hh n
- --wR ; %P-,'}, -
p ~---q> .. g
- ""::' ~ ".'r -
1
) j ? .I 1
-Y ,I$dy '
l i 'i, ;
j'_
"C3 p
' $a$
e ll.Iil
/ .. T iv
-p l !!rt i ls ,
if p./ n. i g 3
- f-
, k '
, , \ .
.i e .b.: . a. !
h I i.:.; C F J fc - fF-J.~ hg n--
~ -- .m.
.]
' L $ll ! I' p_a- :. ; ;
2E
..l ,
If,n,_J jf -
Ju.rw=--;!
jyJ;In .-. M ME : y
- I , r. - 1
.~
- L li- - :s; E': )
iw t
55 pis !
!!nl1pi 1
~ ]w~ rd ~if ,f.
a f ~,
t
- m. I lh [. -N.
~
- I 3" ,t e ','l 8
,, q' ,.)
w ..
. f
%h
e SO 4
. E s I '
)
d ll -- - "
i_q
]a e n h a
,. g 1
e 2 ci "r'saa vt u
" 1g -
5 .
+
t -
E: =!
2
_a - a w -
8 -
1 4
!!:Arlc-- E
- i
@U 2d : I
~'
i
~
4, .g-y c-- y iI :
+3l.a N0!IOff
.J,'.., Li h
.a pin;;
"' _h~
I
_ w@. !u II c - -'
p' I 'g I ; -
y e.
fB:g3l _. ;
.l
,, a
?o . . g r
i ;~ [3 ! Nj! ,
'. L kJ t
, ; h e i i- 1 ie .
L .
r . S. E n
.r iti,r j<>e,i
_ t.-: .
6 ,ls I i e '
N, l<'
,' j i
of ..
f 1
j s:
f / peSM -
. < ,.',' } l
. .... . - E.
)
t
&P - - - - CL _ D 'cN 'N _ .g _ _ _ l w ~
Eq'>-
v I
[i -
'- : 0 t r 8
TF t lis1 st 1 ai l
- - -rf .
n.. .
s- l *e fi r
r r
8' sg 1 i !), / -
6 I -
1 a
! O R E
Is a ? r 8 k mf
g ll .. ~
\
l~ n is R( ar { k n' i ss
"~
'a' i- ,. @l ; ,
1-llii .
x m 84(q= ,
)*q k. stDkk L VVyA
. 1 we -
9... _ s -
u
. . . s , __ e .. .
y . . . . .n
_ M-= ei ..a. @ ***** .
~ p m=
~ , , , e & yl -
=~ .
^p ~ - . , . . . . , . ..
m ki' D.- --
.=......m L. g>_ h N=FN# h* . -
I_-- p E::=. - --- .=.*[w' -.;.; =-~ :-- l:! ' -l 4 0s @_ s r DeTi':r.
7--- :.. :.= g " '
C .:..
e;. rwc^7O". % 8
..........._.k..
- " i lI --,,,..t..
- ,,,,, .g . 4 r;j,,,,U P ,C. . . .;,.,d. . ',T.s u . u - . 4.. . . 7, . . .a,{
g == ,.
,........T ' . .. .. .
j
, , , , , , ,. ,,.,,413,- af, t
_ ~-'-
g ; o ... .., , .
.=.,..............;.....=..j,
-@ j = / ,
Np -- g, q=
- g .g.;
_ J (j
. Mu..~.w.
~
.a y- .
/ .
y: . ~.Y. .1,m s~. a ;i >,a
,,,, a4.. . L.W -
r *I .= w. .
- . - . - - - 2 i ...E g~. fr. g~a, ; ~~ W.Qu f3 1 3. ,M.-Qu. v.. Lpfp:.: -.---.. .d
' - ~ '
e :: Qt_ . .% . . , -----..-
t .
, . - r -._
. ._ ,. m t_. m 1 ar ,
m 6<:.e ..
.2 ~,v.. i.
+ = ._ c m= =-a
~. 3 4... , ;.,e .: , , y
.u: . g
_.r .. e L_ / = - - ,-
.-. . . . .,. .., e g L.7,.
? i , ,
g == g a Iv. W. . .-.. ..- -'. .m&. ,
~~L ,[.,_ , ."" &~
g K,%.. h " ".. . "
). .
m : w._.(f._ ,= == &? g.
l},,,Q4-_'.s* '
,,,, g
._.,____.__~m..
tfa-*?_--*_J>.'k,.
. .n e . n_ _
Y$=
4"
- as g
- r) .v. _";.. - ,
g",'. q . . . . . . .
k -
i" I .'" f-iAL r, S ,
3 .
v i .g %.~sGiur'.I. . _ .--. '
- ]g .I. ::r.. g-'a g g
r :m n
M;]., A.: . . .
=.-.....,_1
.. /
- 1 O, Q ;
.g%
. . . . . . cJ - =. ...
J
.e. g. y ,.
y..
,. y.f{l,f,3g : E. '- j4'- , w9 :. . w .;]_
u w J -- T'-'b c:- y y
.Q
_. ?%f. - e . ,h ,
. . !"_~'
*~
,9 ,
- ~? ~l Y
" #~
J M'$[" -
c,
- g. .. m s gi .
a , - ..- .m - ., ,
O g
@6 PL AN EL 2F2 O' i
Pt.AN EL 250 6 1
g F_ _
d.
% e.s I=. t e .
C;3 .i '5F.ir--
1 ggy
- t :3 .t . -
_j .
===-= ._ _. >
cr,2 ~
"0 ayu:= A < 100 m.R/nn
~ nv-! B < SR/HR 2: J i b'M- Q)veM g,,
1 e e zoa/a4 p pggg/gg nn gg a
EL2 SOL cnw O
- EL172 ir euu t
$ ELmt0*
..om
.1448E5 A FITZPETeign a.nri rag pgegg ptAfff .
L i. .- k MlWER AUTle0AITY OF1tE STATE er GEW vem
.* 4 s.s e-a.
AnRT M.AN EL 284'-O"
' ~
,es>
REFERENCES
- 1) FSAR, J.A. FitzPatrick Nuclear Power Plant.
- 2) NUREG-0578, USNRC, dated 7/79.
- 3) NUREG-0737, USNRC, dated 11/80.
- 4) G.E. Specification No. 22A2703T, " Radiation Sources for BWR Requirements", Revision 1, dated 3/1/73.
L 5)
QAD-QC, "3-Dimensionel Point Kernal Gamma Shielding Code",
RP-100, by Richard E. Deem, dated 7/77. RSIC code Package ccc-401/QAD-QC.
- 6) ACTGEN, " Radioactivity Source Generation Code", RP-106, by :
Richard E. Deem, dated 8/77.
- 7) G.E. " Post Accident Sample Station Activity Source Terms", by H.R. Helmholz, dated 9/80.
8)- SWEC Radiation Protection Code DRAGON 3 " Dose and ,
Radioactivity from Nuclear Facility Gascous Outflows", NU-115, ;
Version 00, Level 01, dated 6/4/76. !
- 9) " Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be
_Used.in Post-Accident Operations", SWEC, S/80.
i-
4 4
w 4 9 %
9 APPENDIX A
,,s>
POWER AUTHERITY JF THE GTATE C F NEW YSRK PROGRAM
SUMMARY
TITLE AUTHOR (S) / SPONSOR (S) REF. N o. l
/CKCl - R1dioactivitb*
Generation CT.ource e Richard D. twn T-I n 6 i PROGR AM STATUS USER MANUAL STATUS ISSUE DATE R EV. DATE Ccmolete Cor.plete f'-77 l PROGRAM PURPOSE, SCOPE, METHOD, INPUT DATA, OUTPUT RESULT PUr'POS E :
This code nrovides for convertino an activity concentration (i.e. pCi/cc) for radionuclides to a multioroun enerny spectrum (i.e. MeV/sec-cc) . *his eneray socctrum can then be used as a source term for shieldina calculations or other dose rate / heating rate calculations.
SCOPE: This procram is limited only by those constraints described in the users manual. i l
j l
ST.T2!nD :
Given time T initial source activities in a control volume at some i o this progran calculates the residual activities at the en,d of one, or several, decay periods accordinn to the appropriate decay enuation (first, second, and third order isotoDic decay are considered). The ef fect of a continuous recirculation-filter on the system can also be determined. t library of more than 100 isotopes is nrovided k with the ability, when desired, to modify any data within the library or add other isotopes of interest. Also calculated is the nanna ray eneroy released, due to radioactive decay, in seven enerev croups.
IID UT : Decay times initial concentrations of the source isotopes, and any alterations / additions to the existino library in- i formation.
OUTPUT: Library data used, decay time, ouri fication code, initial j inventory , alterations / additions to the library data, ca'"ma enerny released in each of the seven eneroy aroups, and various sums of camma enerny released for each isotone aroun and for all contributors.
I I
i f
i t
i PROGRAM NAME SOURCE ORGAN ORGAN. REF. No.
_@21, CDC600 - [VNTJ h________
1
- 4 og COh APPENDIX B
f~ . .
y- o..s POWER AUTHERITY OF THE CTATE OF NEW YORK PROGRAM
SUMMARY
TITLE QADQC - 3 Dimensional Point Kernal AUTHOR (S) / SPONSOR (S) REF.No.
Camma Shielding code Richard E. Deem Rp - 100 PROGRAM STATU3 USER MANUAL STATUS ISSUE DATE R EV. DATE Complete Complete 7/77 PROGRAM FORPOSE, SCOPE, METHOD, INPUT DATA, OUTPUT RESULT PURPOSE: This program calculates the direct beam samma dose rates at points in 3-dimensional space from point, volumetric, and cosine intensity func-tion sources. The source and dase points can be described in either cartesian, cylindrical, or spherical coordinates, while the geometry description is limited to cartesian only.
- SCOPE: This program is limited only by the contraints specified in the users manual.
PallOD: Numerical integration of point sources in 3-dimensional geometry, Volumetric sources ate divided up into a point source distribution ti.en integrated numerically. The cosine intensity function sources are numerically integrated directly. The line-of-sight distance from each point source to the dose point is then calculated. Based on the distance traveled through each material region specified and the chielding characteristics of each material, the unco 111ded gamma ray flux at the donc point is calculated. A buildup factor is then applied to this dose to take into account the scattering component of the dose rate.
INPUT: Specific sources, geometric model, materials data, buildup factors, and flux to dose conversion factors.
OUTPUT: Dose rates with and without buildup in any units desired by the user.
- QADQC is similar in all respects to the QAD-P5 nuclear code RSIC Document ,
(CCC-048/QAD) except that all neutron moment and heating calculations have '
heen eliminated, the nurber of regions and boundaries has been reduced, and various changes to the output formats have been changed. This code was developed for quick and fairly inexpensive direct beam gamma dose calculations.
PROGRAM NAME SOURCE ORGAN. ORGAN. REF. No.
IBM 370 - QADOC CDC6600 - OADF pAgny-
.nrinrminisucaw nArti enuouren_h amcrince Iu m ant-ar e ruruu-