ML20236N910
| ML20236N910 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 03/31/1998 |
| From: | External (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20236N893 | List: |
| References | |
| WMG-9803-7025, WMG-9803-7025-R01, WMG-9803-7025-R1, NUDOCS 9807160055 | |
| Download: ML20236N910 (117) | |
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l Attachment 5 i Saxton Nuclear Experimental Corporation Facility Steam Generator & Pressurizer Characterization Report I
l 9807160055 9006241 l PDR ADOCK 05000146 p PDR
. 4 SAXTON STEAM GENERATOR AND PRESSURIZER CHARACTERIZATION l
Report WMG 9803-7025 1
! March 1998 l
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! Prepared for:
l l Raytheon Engineers and Coristructors !
i Prepared by:
WMG, Inc.
16 Bank Street Peekskill, NY 10566 ;
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I l' FOREWORD This report details component characterization work performed by WMG, Inc. to ,
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' support the decommissioning of the Saxton Nuclear Power Plant by Raytheon. This work was performed by WMG, Inc. under Raytheon Subcontract 77581803-49-1.
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I Saxton Steam Generator and WMG Report 9803-7025 l
. Pressurizer Characterization Rev.1 3/98 !
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L TABLE OF CONTENTS
- 1. 0 i N T RO D U C TIO N .. .. .. . .. .. . . .. ... ..... . .... ....... .. . . ..... . .. . . . . . . . . . .. .. . . . . . . . . . . . .
l 2.0 STEAM G ENERATOR CH ARACTERIZATION.............................................. 2 1 l 2.1 Input Parameters.. . .2-1 2.2 Characterization Calculation Assumptions . .2-5 2.3 Characterization Results. . . .
. .2-8 2.3.1 Characterization of Tube Bundle Section. . . . .2-8 2.3.2 Characterization of Channel Head Area.. . . . . . 2-8 '
2.3.3 Dose to Curie Characterization Methodology and Uncertainty . .2-9 2.4 Steam Generator Classification . . .
.2-10 2.4.1 DOT Transportation Ctass.. .. . . 2-11 2.4.2 NRC Waste Form Class. .2-11 3.0 PR ESS U RIZER C HARACTERIZATlON ......................................................... 3-1 l
3.1 Input Parameters.. . .. . . . . .. . . . . .3-1 i
3.2 Characterization Calculation Assumptions . . . . . . .. . . 3-5 3.3 Characterization Results.. . . . .. . . . . . . . . ... . .3-6 1
3.3.1 Characterization of the Pressurizer Shell Section. . . .3-6 '
3.3.2 Characterization of Heater Bundle Area..... . . ... . .3-7 3.4 Pressurizer Classification.. . .. . . . . . . . . . . . . . . .3-7 3.4.1 DOT Transportation Class.. . . . . . . . . . .. . .3-7 3.4.2 NRC Waste Form Class. . . . . . . . .. . .3-7
- 4. 0 R E F E R E N C E S . . .. .. . . . . .. . . . . . . . .. .. . . . . . . ... .. . .. . . .. . . . .. . . .. . . . . . . . . . .. . . .. . . . . . . . . . . . . . . . .. . . . .:
APPENDICES APPENDIX A - GPUN Provided Data t
l l APPENDlX B - MegaShield" / Microshield Cases l
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. I l Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 ii l
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LIST OF TABLES Table Title Page 2-1 Saxton Steam Generator Physical Characteristics.. . .. 2-2 2-2 Saxton Steam Generator Isotopic Distribution.. . . . 2-3 2-3 Saxton Steam Generator Additional Swipe Data in uCi/100cm' .2-4 2-4 Saxton Steam Generator External Survey Data in mR/hr.. .2-6 2-5 Saxton Steam Generator Contamination. .. .. . . .2-9 3-1 Pressurizer Physical Characteristics... . . . . . . . . . 3-2 3-2 Saxton Pressurizer Isotopic Distribution...... .. . . . . 3-3 3-3 Saxton Pressurizer Survey Results in mR/hr.. .. ..... 3-4 LIST OF EXHIBITS Exhibit Title Page l 1 Saxton Steam Generator DOT Classification Summary. .. . 2-12 L 2 Saxton Steam Generator NRC Classification Summary. . .. . .2-13 3 Saxton Pressurizer DOT Classification Summary ... . . . . .... 3-8 l 4 . Saxton Pressurizer NRC Classification Summary. . . 3-9 l-i
. Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 iii -
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1.0 INTRODUCTION
This report presents the work performed by WMG, Inc. to characterize and classify the Saxton Project Westinghouse Vertical Steam Generator and Pressurizer for Raytheon Engineers and Constructors. The steam generator and pressurizer were characterized to determine radionuclides activities and classified according to DOT criteria in 49 CFR 173 for transportation as radioactive material, and according to NRC criteria in 10 CFR 61 for disposal 'as low level radioactive waste. This report presents the input parameters, methodologies and assumptions used for characterization and DOT and NRC classification results.
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Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 1-1
2.0 STEAM GENERATOR CHARACTERIZATION 2.1- Input Parameters characteristics, representative independent results, and detailed radiation measurements ot/ained on the steam gen GPUN is enclosed as Appendix A.Each type of input is disc sample data has been obtained from the Saxton steam generator Physical characteristics - A summary of physical characteristics of the Westinghouse Saxton Project Vertical steam generator was obtained Westinghouse Technical Manual 1440-C49 9 (Ref.1). The physical characteristics were used to develop a dose to curie conversion mode use with WMG's MegaShield computer program.
The internal surface areas were used to estimate contamination levels to ensure th Contaminated Object requirements, as per 49 CFR 173.403, were satisfied.
A summary of the physical characteristics used for the analysis is shown Table 2-1 below.
Isotopic Sample Data -
l Representative swipe data was obtained in
! November 1995 (Ref. 2) and two steam generator samples were sent to Babcock and Wilcox for analysis.
! results for samples SX825950111 and SX865950053The independ are enclosed in Appendix A. These results were screened to eliminate LLD values with the exception of the 10 CFR Part 20 Appendix F radionuclides (H-3, C-14, Tc-99 and 1-129),
Table 2-2. and an averaged isotopic distribution was defined as shown in Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization 2-1 Rev.1 3/98
Table 2 - 1 Saxton Steam Generator Physical Characteristics Parameter Description General Information Total Dry Weight (Ib.) 52,000 Overall Height 240" Maximum Diameter 52.25" Nominal Shell Thickness 2.875" Tube Data Tube Bundle Diameter 40.0" Number of Tubes 736 Tube Outer Diameter 0.625" Tube Wall Thickness 0.058" Tube Height (above tube sheet) 110.25" Tube Sheet Thickness 9.5" Tube Sheet Clad Thickness 0.375" Wrapper Outer Diameter 42.5" !
Wrapper Thickness 0.25" l Effective Heat Transfer Area (ft') 2,300 Tube Material 304 Stainless Tube Material Density (g/cc) 8.04 Internal Surface Areas (cm') 1.89E+6 !
Tube Bundle Section 1.69E+6
- Tubes in Tube Sheet Section 1.40E+5 :
- Channel Head Bowl Section 2.74E+4
- Divider Plate Section 1.50E+4
- Tube Sheet 1.01 E+4
- Referred to as Channel Head Area Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 2-2
l Table 2 - 2 Saxton Steam Generator Isotopic Distribution l
j Sample l Nuclide Average SX865950053 SX825950111 Units _
l H-3 <8.37E-03 <8.37E-03 <7.50E-04 uCi/gm C-14 <1.19E-02 <1.19E-02 <2.58E-03 uCi/gm Tc-99 <5.86E-04 <5.86E-04 <2.52E-04 uCi/gm I-129 < 1.41 E-03 <1.41 E-03 <6.37E-05 uCi/gm l Fe-55 2.468 2.5442 2.3921 %
Ni-59 0.497 0.5503 0.4439 %
Ni-63 42.231 44.028 40.4345 %
Sr-90 3.875 0.2853 7.4648 %
Nb-94 0.011 0.0080 0.0146 %
Co-60 19.386 17.2829 21.4897 %
i Cs-137 13.745 0.4007 27.0883 %
Eu-154 0.016 0.0321 0.0000 %
Eu-155 0.004 0.0082 0.0000 %
U-234 0.000 0.0000 0.0000 %
U-238 0.000 0.0000 0.0000 %
Pu-241 1-4.088 27.614 0.5627 %
Pu-238 0.558 1.0959 0.0195 %
Pu-239/240 1.225 2.4090 0.0410 %
Am-241 1.895 3.7414 0.0489 %
Alpha 3.678 7.246 0.109 %
I Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 2-3
. Additional Swipe Data -
Additional swipe data obtained in October 1997 was used to determine the relative contamination levels in the following steam generator areas of interest:
0 Bowl region 0 Divider plate 0 Tube sheet 0 Inside a tube in the tube sheet These swipe samples were counted in-house with a high purity Germanium detector and the gamma isotopics are included in Appendix A and summarized in Table 2-3.
Table 2 - 3 Saxton Steam Generator Additional Swipe Data in uCi/100cm' Survey 891-97-1591 Dated 10/29/97 Location Bowl Tube in Tube Sheet Tube Sheet Divider Plate Sample # SX8-25-97-0181-SM SX-25-97-0182-SM SX-25-97-0183 SM SX-25-97-0179-SM Swipe Data in uCi/100cm2 Nuclide Co-60 1.052E-01 1.832E-02 5.498E-02 1.436E-01 Cs-137 2.945E-01 9.421E-02 2.160E-02 6.781 E-02 .
Eu-152 - - - -
Eu-154 8.562E-04 - - -
Am-241 1.665E-02 6.054E-03 3.284E-03 5.477E-03 Total 4.172E-01 1.186E-01 7.986E-02 2.169E-01 Gross p-y 1.20E+00 2.00E-01 1.60E-01 6.00E-01 l Gross a 6.40E-02 1.30E-02 7.00E-03 2.50E-02 (Gross p-7) Contamination Ratio Relative to Tubes in Tube Sheet
, 6.002 1.000 0.600 3.001 l
l l Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 2-4 l
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- .. External Radiation Surveys -
GPU Nuclear performed detailed radiation surveys on the steam generator at two foot axial increments at four circumferential locations in February 1997. Those survey results are enclosed in Appendix A and summarized in Table 2-4.
2.2 Characterization Calculation Assumptions Several assumptions were made to characterize the steam generator based on the input parameters described above. It was assumed that the secondary side activity is negligible and all the activity is distributed on the primary side surfaces. These surfaces were divided into five sections; tube bundle, tubes within the tube sheet, tube sheet, channel head boy.1, and divider plate.
Historically, steam generators have been characterized by performing dose to curie conversion calculations on the tube bundle sections and then applying a factor to estimate activity in the channel head area. For the Saxton project, non-uniform distribution of crud on tube surfaces was considered in a manner similar to that employed for the Salem Steam Generator Project. A 1983 EPRI report (Ref. 3) was cited as evidence of such non-uniformity and the findings in this report, as well as another EPRI report (Ref. 4), led to a change in the characterization methodology used for steam generators. The approach used for the Saxton steam generator is a similar two step characterization process:
(1) Perform dose to curie conversion using exterior radiation survey results (Table 2-4) to determine activity in the tube bundle area, and i
(2) Perform a series of ratio calculations using channel head survey results (Table 2-3) to determine contamination levels outside the tube bundle area.
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. Channel head swipe results and the EPRI results were used to develop ratios of contamination levels relative to the tube bundle. The ratio of contamination levels and surface areas were combined to quantify the activity in each area and summed with the tube bundle activity to determine the total steam generator activity. Both sets of calculations used the averaged radionuclides content and distribution from the independent lab samples.
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Table 2 - 4 Saxton Steam Generator External Survey Data in mR/hr l
l Axial
' Radial Quadrant Axial Location Location R1 R2 R3 R4 Average A1 0.8 0.5 0.1 0.1 A2 1.5 0.7 0.1 0.1 A3 1.5 0.5 0.5 0.6 0.78 B A4 4 5 2 1.5 3.13 T U A5 5 5 3 3 4.00 U N A6 5 5 4 3 4.25 B D A7 5 5 4 5 4.75 E L A8 5 5 6 7 5.75 E A9 5 5 2 3 A10 15 6 10 10 Tube Bundle Average 3.78 I
Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 2-6
Tube Bundle Section - Three sets of independent laboratory data we analyzed for an average radionuclides distribution which was normalize produca program.
shielding a 1 curie source for input into the MegaShield* point kernel The tube bundle section was then modeled as a homogenized cylinder using fractional density inputs based on the cros sectional areas of metal and air. The wrapper and shell were incorporated as shielding regions with the air gaps inserted between the source region, wrapper and shell.
The total activity calculated for the tube bundle wrapper is assumed to be distributed between the source region and shell. ,
The total activity calculated for the tube bundle is assumed to be distributed between the straight section of the tube bundle and the U-bend section of the tube bundle.
Channel Head Area - Two EPRI studies (Refs. 3 and 4) con the early 1980's provided some insights into the relative distribution of crud within steam generators.
The results from these studies were reviewed and applied to quantify activity and to provide assurance that non-uniform distributions of crud could not result in depositions which exceed the SCO-Il limit of 20 uCi/cm'(p-y/LTA) and 2 uCi/cm 2 (y),
Application of these early study results to the Saxton steam generator is limited because the Saxton steam generator tubes are type 304 stainless steel instead of Inconel. Due to different construction materials a limited Saxton plant operating history as compared to that of commercial plants, the majority of EPRI report relationships between the tube bundle and channel head are not considered reasonable, Non-uniform crud deposition within the tubes is assumed to be due to flow characteristics.
Therefore, the only applicable EPRI relationship is the ratio (1.82) of the contamination in the tube bundle relative to that of the tubes within tubesheet. Another key characterization assumption is that the fixed-to-removable contamination ratio is relatively constant throughout the channel head region.
Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 2-7
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2.3 ' Characterization Results -
The characterization results yielded a total steam generator activity of 1,060
. millicuries as of August 1,1998. .
r 2.3.1 Characterization of Tube Bundle Section i
WMG's MegaShield" point kernel shielding program was used to calculate dose to curie conversion factors for the steam generator tube bundle region from survey data in Table 2-4. The axial survey locations corresponding to the tube bundle are the third through the eighth (inclusive) from the top. The dose to curie conversion factor (2.10E+02 mci /(mR/hr)) was then applied to the average contact dose rate along the tube bundle height (3.78mR/hr) to estimate total activity of the tube bundle section at 791 mci at the time of measurement. The resultant activities by radionuclides were then decay corrected to 8/01/98 yielding 744 mci. A summary output report for the final MegaShield" model is enclosed in Appendix B.
2.3.2 Characterization of Channel Head Area The total activity calculated for the tube bundle in section 2.3.1 was uniformly distributed on the interior surface area of the tube bundle to ,
obtain the tube bundle contamination level of 0.44 uCi/cm2 . As stated previously, application of EPRI study findings regarding channel head area contamination levels relative to the tube bundle is limited to the use of the ratio of 1.82 (conservatively rounded to 2) between the tube bundle and the tubes within the tube sheet. Therefore, the contamination of the 2
' tubes in the tubesheet is estimated as 0.88 uCi/cm . Contamination level relationships, in the form of ratios between the bowl, tubesheet, and divider plate relative to the tubes in the.tubesheet, were determined.
g through a comparison of the gross p-y swipe survey data in Table 2-3.
l The area-specific contamination levels were multiplied by the l corresponding surface areas to obtain localized activities, f-l Saxton Steam Generator and- WMG Report 9803-7025 4 Pressurizer Characterization Rev.1 3/98 2-8 4
These calculations resulted in a total activity of about 315 mci distributed within the channel head area as shown below in Table 2-5:
Table 2-5 Saxton Steam Generator Contamination Tube Tubes in Bowl Tube Divider Bundle Tube Sheet Sheet Plate Activity, mci 744 123 145 7 40 Area, cm2 1.69E+6 1.40E+5 2.74E+4 1.01 E+4 1.50E+4 Contamination, 0.44 0.88 5.28 0.70 2.64 2
uCi/cm EPRI Ratio 2.00 Bowl Ratio 6.00 Tube Sheet Ratio 0.80 Divider Plate Ratio 3.00 .
Thus, the total activity for the Saxton steam generator is 1,060 mci with about 70 percent of the activity (744 mci) in the tube bundle section.
2.3.3 Dose to Curie Characterization Methodoloav and Uncertainty The methodology for characterizing the Saxton steam generator tube bundle section was WMG's MegaShield program which was used to calculate dose to curie conversion factors. The MegaShield program uses a point kernel shielding methodology and employs input libraries based on the following information: _,
. ANS 6.4.3 Geometric Progression Buildup Factors Published in NUREG/CR-5740 dated August 1991.
. ANSI /ANS-6.1.1 Gamma-Ray Fluence to Dose Factors August 1991.
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. DLC-136/PHOTX photon interaction cross section library from the i National Institute of Standards and Technology, ORNL April 1989. l
. The radionuclides library includes photon energies and yields from Kocher's compilation.
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. Saxton Steam Generator and WMG Report 9803-7025 l Pressurizer Characterization Rev.1 3/98 j 2-9
The MegaShield* program was used to develop dose to curie conversion I models based on an assumed homogeneous source distribution throughout the tube bundle section of the steam generator. The point kernel shielding methodology with these same assumptions were employed for the Yankee Rowe, Trojan, and Salem steam generators. {
There are several areas of uncertainty involved with any analysis of this type. The nuclear constants employed in the analysis include American National Standards where available, or industry accepted standards such as ANS 6.4.3 GP Buildup factors which represent the largest source of uncertainty. Additional uncertainty is introduced by the radiation detector ;
efficiency, survey conditions and the effects of backscatter and statistical !
variation in the isotopic distribution used. It is impossible to quantify the percent error associated with an analysis of this type without additional empirical data. The point kernel methodology is utilized throughout the I industry, and provided that sufficient detail is incorporated into the model, the results are generally accepted to be reasonably accurate.
Aside from the methodology employed, there is uncertainty associated with the analysis assumption that the source region is homogeneously ,
distributed within the tube bundle.
Due to the uncertainties discussed above, the background radiation was not subtracted from the radiation survey results to ensure the characterization results are conservative.
2.4 Steam Generator Classification Based on application of the requirements in 49 CFR Part 173 et al., and 10 CFR Part 61, the Saxton steam generator is a Greater Than Type A Quantity of SCO-ll Material containing radionuclides concentrations below NRC waste form Class A limits. The besis for these conclusions is presented below.
During NRC classification, the waste weight and volume used for classification neglect any grout. This results in calculation of the highest possible radionuclides concentrations leading to conservative NRC classification.
l Saxton Steam Generator and WMG Report 9803-7025 i Pressurizer Characterization Rev.1 3/98 i
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2.4.1 DOT Transportation Class Exhibit 1 presents the DOT Classification summary for the Saxton steam generator. As shown. it contains a Greater Than Type' A Qaantity of Radioactive Material with the sum of A2 fractions equal to 8.2.
Exhibit 1 also shows the maximum contamination level from the bowl region of the Saxton steam generator to be approximately 25 percent of the SCO-ll limit.
2.4.2 NRC' Waste Form Class Exhibit 2 presents the NRC Classification summary for the Saxton steam generator. As shown, radionuclides concentrations are approximately 45
_ percent the 10 CFR Part 61 Class A limits.
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Exhibit 1 Saxton Steam Generator DOT Classification Summary SG A2 A2 RQ RQ Nuclide mci Curies Fraction Curies Fraction H-3 LLD --- --- --- ---
l-129 LLD --- --- --- ---
Fe-55 1.93E+01 1.08E+03 1.78E-05 100.00 1.93E-04 Ni-59 5.60E+00 1.08E+03 5.18E-06 100.00 5.60E-05 Ni-63 4.71 E+02 8.11 E+02 5.81 E-04 100.00 4.71 E-03 Sr-90 4.22E+01 2.70E+00 1.56E-02 0.10 4.22E-01 Nb-94 1.27E-01 1.62E+01 7.85E-06 10.00 1.27E-05 Co-60 1.82E+02 1.08E+01 1.68E-02 10.00 1.82E-02 Cs-137 1.50E+02 1.35E+01 1.11 E-02 1.00 1.50E-01 Eu-154 1.62E-01 1,35E+01 1.20E-05 10.00 1.62E-05 Eu-155 3,80E-02 5.41 E+01 7.03E-07 10.00 3.80E-06 Pu-238 6.21 E+00 5.41 E-03 1.15E+00 0.01 6.21 E-01 Pu-239/240 1.38E+01 5.41 E-03 2.55E+00 0.01 1.38E+00 Pu-241 1.48E+02 2.70E-01 5.49E-01 1.00 1.48E-01 Am-241 2.13E+01 5.41 E-03 3.93E+00 0.01 2.13E+00 Totals 1.06E+03 Type A: 8.22E+00 RQ: 4.87E+00 SCO ll Determination SG SCOll S C O II uCi/cm2 Limit Fraction Fixed Beta / Gamma N/A * ---
Fixed Alpha N/A * ---
l Removable Beta / Gamma N/A * ---
Removable Alpha N/A * ---
Inaccessible Surfaces 5.07 20 0.254 Combined Beta / Gamma Combined Alpha 0.21 2 0.103 I Notes: * - All steam generator contaminated surfaces are assumed to be inaccessible. Therefore i only the limits for combined contamination in inaccessible areas are applicable.
- Alpha contamination level estimated based on averaged independent lab data.
- Contamination levels represent worst case estimates from the bowl region.-
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f Exhibit 2 l Saxton Steam Generator NRC Classification Summary Pad 61 Pad 61 Activity Table 1 Table 2 Nuclide mci uCi/cc nCi/gm Limits Fraction Limits Fraction l H-3 LLD
! C-14 LLD Tc-99 LLD l-129 LLD Fe-55 1.93E+01 6.53E-03 700 0.000 Ni-59 5.60E+00 1.90E-03 220 0.000 Ni-63 4.71 E+02 1.60E-01 3.5 0.046 Sr-90 4.22E+01 1.43E-02 0.04 0.357 Nb-94 1.27E-01 4.31 E-05 0.2 0.000 Co-60 1.82E+02 6.16E-02 700 0.000 Cs-137 1.50E+02 5.08E-02 1 0.051 Eu-154 1.62E-01 5.49E-05 Eu-155 3.80E-02 1.29E-05 700 0.000 Pu-238 6.21 E+00 2.63E-01 100 0.003 Pu-239/240 1.38E+01 5.84E-01 100 0.006 Pu-241 1.48E+02 6.28E+00 3500 0.002 Am-241 2.13E+01 9.02E-01 100 0.009 Part 61 Class A Sum of the Fractions 0.020 0.454 I
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3.0 PRESSURIZER CHARACTERIZATION 3.1 Input Parameters 1
The characterization input parameters consisted of the pressurizer physical !
characteristics, representative independent laboratory isotopic sample analysis l
results, and detailed radiation measurements obtained on the pressurizer. Each type of input is discussed separately below and selected data provided by GPUN is enclosed in Appendix A.
Physical Characteristics - A summary of the Westinghouse Saxton Project pressurizer characteristics was obtained from Westinghouse Technical Manual 1440-C57 (Ref 5). The physical characteristics were used to develop dose to curie conversion models for use with Microshield and 4 WMG's MegaShield computer programs. The internal surface areas were used to estimate total contamination levels to ensure that Surface Contaminated Object requirements in 49 CFR 173.403, were satisfied. A summary of the physical characteristics used for the analysis is shown in Table 3-1 below.
Isotopic Sample Data - Representative swipe data was obtained (Ref. 2) and three pressurizer samples were sent to Babcock and Wilcox for analysis.
The independent laboratory analysis results for samples SX25970175-SM, SX825970176-SM, SX825970177-SM are enclosed in Appendix A. These results were screened to eliminate LLD values with the exception of the 10 CFR Part 20 Appendix F radionuclides (H-3, C-14, Tc-99 and 1-129) and an averaged isotopic distribution was defined as shown in Table 3-2.
. External Radiation Surveys - GPU Nuclear performed detailed radiation surveys on the pressurizer at two foot axial increments at four circumferential locations and documented on the survey sheets enclosed in Appendix A.
Surveys were obtained on contact with the pressurizer and are summarized in Table 3-3.
Thermoluminescent Dosimeter (TLD) Survey -
GPU Nuclear also i performed a survey where a string of TLDs were placed along the centerline I of the pressurizer to the heater bundle area. Survey results can be reviewed alongside the extemal survey data in Table 3-3 and r.ce also included in Appendix A.
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Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 3-1
Table 3-1 l
i Pressurizer Physical Characteristics i
Parameter Description General information Total Dry Weight (Ibs.) 25,000 Material of Construction Carbon Steel cladded with Stainless Steel Overall Height 200.75" l Outer Diameter 43" Inner Diameter 36.375" l Free Internal Volume (ft') 94.5 ,
Nominal Shell Thickness 3.3" l Heater Bundle Diameter 12.3" i l Heater Bundle Height 43" Heater Bundle Weight 73lbs.
l l Internal Surface Areas (cm') 1.58E+05 l Pressurizer Shell Section 1.33E+05
- Heater Bundle Section 1.31 E+04
- Surge Ring Section 1.15E+04 Spray Nozzle Section 5.26E+02 i
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, Table 3-2 Saxton Pressurizer Isotopic Distribution Sample Nuclide Average SX25970175-SM SX825970176-SM SX825970177-SM Units H-3 <1.96E-03 * <1.94E-03 <1.75E-03 <1.96E-03 uCi/gm C-14 <1.12E-03 * <1.20E-03 <1.06E-03 <1.12E-03 uCi/gm Tc-99 <8.00E-05 * <7.91 E-05 <7.18E-05 <8.00E-05 uCi/gm I-129 <1.19E-04 * <1.14 E-04 <7.32E-05 <1.19E-04 uCilgm Fe-55 2.194 2.0990 2.3296 2.1524 %
Ni-59 0.805 0.6889 0.8022 0.9230 %
Ni-63 66.825 61.5925 72.3047 66.5766 %
Sr-90 0.364 0.6200 0.1629 0.3083 %
Nb-94 0.013 0.0124 0.0095 0.0174 %
Co-60 16.488 13.8988 16.8784 18.6868 %
Cs-137 0.164 0.1601 0.0611 0.2712 %
Eu-154 0.018 0.0539 0.0000 0.0000 %
Pu-238 0.193 0.2946 0.1180 0.1664 %
l- Pu-239/240 0.464 0.7172 0.2879 0.3858 %
! Pu-241 11.660 18.5588 6.5492 9.8730 %
l Am-241 0.792 1.2683 0.4835 0.6242 %
l Cm-242 0.006 0.0116 0.0030 0.0033 %
Cm-243/244 0.015 0.0239 0.0102 0.0116 %
Alpha 1.470 2.316 0.903 1.191 %
Note: * - The maximum LLD values were used for the 10 CFR 20 Appendix F Radionuclides (H-3, C-14, Tc-99,1-129) l- .
l Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 3-3
Table 3-3 Saxton Pressurizer Survey Results in mR/hr Exterior Interior Axial Radial Quadrant Location TLD mrem /hr Location Location Average Location Average R1 R2 R3 R4 A1 0.5 2 0.5 2 1.25 0 10
-1 13 A2 1 3 0.5 3 1.88 -2 16
-3 18 A3 2 6 3 5 4.00 -4 21
-5 34 A4 2 7 2 8 4.75 -6 44
-7 47 S A5 2 7 2 6 4.25 -8 46 H
-9 45 E A6 1 15 3 15 8.50 -10 42 L
-11 49 45.8 L A7 3 8 4 8 5.75 -12 102 H
-13 108 105 T A8 7 8 4 10 7.25 R Pressurizer Surface Average 4.70 l
l 1
i Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 3-4 L- __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ .
a, 3.2 Characterization Calculation Assumptions Several assumptions were made to characterize the pressurizer based on the input parameters described above. It was assumed that activity is uniformly i distributed on all interior surfaces of the pressurizer. Geometrically the pressurizer can be separated into two major dose contributors, the heater bundle ;
and the shell. I Unlike steam generators, pressurizer can be represented by relatively simple i geometries, lack major internal components, and can be characterized via !
relatively simple dose to curie conversion calculations. The average independent laboratory analysis radionuclides distribution was normalized to produce a 1 curie source strength for input into both the MegaShield* and Microshield* point kernel shielding programs.
Shell Section - The interior surface of the shell was modeled as a hollow cylinder of minimal thickness having the density of stainless steel. The carbon steel shell was in'.:oiporated as the shielding region for the MegaShield* model.
{
For the Microshield model, the TLD survey location band of -7 to -11 was j selected as representative from two perspectives; (1) The localized source, from the heater bundle (locations -12 and -13), 4 does not appreciably influence the survey results in locations -7 to -
11,and .
{
(2) It does not include the less contaminated region of the steam bubble (locations O to -6) as discussed below. '
{
. Heater' Bundle Area - The TLD survey data infers less contamination closer to the spray nozzle and a greater source of activity near the heaters. This distribution is expected from two perspectives. First, high crud depositions are not expected high in the steam bubble, close to and above the spray I nozzle, because the crud containing spray falls to the bottom of the j pressurizer during normal operation. Second, the surface area associated i with the heater bundle represents. a localized source. l l
- l l
Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 3-5 1
l 3.3 . Characterization Results 4
The characterization results yielded a total pressurizer activity of 755 millicuries as of August 1,1998. The results from three characterization models were reviewed to provide assurance that non-uniform distributions of crud could not result in 2
depositions exceeding the SCO-Il limit of 20 uCi/cm 2(p-y/LTA) and 2 uCi/cm (a).
3.3.1 Characterization of the Pressurizer Shell Section Characterization of the pressurizer shell was accomplished via two distinct models.
The first model used WMG's Megashield point kernel shielding program to calculate dose to curie conversion factors for the pressurizer shell region. The dose to curie factor of 1.69E+02 mci /mR/hr was then applied to the average contact dose rate of 4.70 mR/hr along the (exterior)
' pressurizer height to estimate the corresponding activity at 794 mci as of.
the survey date. The 794 mci source was decay corrected to 755 mci as of 8/1/98. A summary output report for this MegaShield* model is enclosed in Appendix B.
To verify that pressurizer characterization results from the MegaShield
-model were reasonable, a second model using the interior TLD string survey data was developed. MegaShield does not support detector placement within the source envelope, so the Microshield point kernel shielding program was used to calculate a dose to curie conversion factor for the pressurizer shell from the TLD survey. The dose to curie factor of 1.78 mci /mR/hr was then applied to the average TLD dose rate of 45.8 mR/hr to obtain 81.4 mci of activity on the shell as of the survey date.
The 81.4 mci source was decay corrected and uniformly distributed on a 6 foot high area'of the shell surface of 5.3E+04 cm2 corresponding to TLD locations -7 to -11 to estimate the shell surface contamination level at 1.5 2 ~
uCi/cm . A summary output report for the Microshield model is enclosed in Appendix B.
The characterization results based on the internal TLD dose rates are considerably lower than those from the external survey results. This is ;
presumably due to the effects of background radiation on the external )
survey results.
Saxton Steam Generator and . WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 3-6 l
)
y s e 3.3.2 Characterization of Heater Bundle Area As stated above, the TLD string survey data infers non-uniform crud deposition within the pressurizer. The heater bundle area was modeled as a homogenized cylindrical source using fractional density inputs based on the cross-sectional areas of metal and air. The source used a normalized curie of activity in WMG's MegaShield point kernel shielding program to calculate a heater bundle dose to curie conversion factor.
The dose to curie factor of 0.60 mci /mR/hr was then applied to the TLD dose rate of 105 mR/hr that corresponds to the " top" of the heater bundle (pressurizer locations -12 and -13) to estimate the heater bundle activity at 62.6 mci. The 62.6 mci source was decay corrected and uniformly distributed on the 1.3E+04cm 2 of the heater bundle to estimate the contamination level at 4.5uCi/cm 2.
3.4 Pressurizer Classification Based on application of the requirements in 49 CFR Part 173 et al., and 10 CFR Part 61, the Saxton pressurizer is a Greater Than Type A Quantity of SCO-Il Material containing radionuclides concentrations below NRC waste form Class A limits. The basis for these conclusions is presented below.
3.4.1 DOT Transportation Class Exhibit 3 presents the DOT Classification summary 'for the Saxton pressurizer. As shown, it contains a Greater Than Type A Quantity of Radioactive Material with the sum of the A2 fractions equal to 2.47.
All of the radioactivity is contained within the pressurizer shell as crud deposition.
As shown, the maximum internal fixed contamination level is well below 2 -
the SCO-Il limit of 20 u,Ci/cm (D-y/LTA) and 2 uCi/cm2 (y),
3.4.2 NRC Waste Form Class Exhibit 4 presents the NRC Classification summary for the Saxton pressurizer. As shown, radionuclides concentrations are about 16 percent of the 10 CFR Part 61 Class A limits.
! - Saxton Steam Generator and WMG Report 9803-7025 l Pressurizer Characterization Rev.1 3/98 3-7
,, sp.
L
Exhibit 3 Saxton Pressurizer DOT Classification Summary
)l PZR A2 A2 RQ RO Nuclide mci Curies Fraction Curies Fraction H-3 LLD --- --- --- ---
I-129 LLD --- --- --- ---
Fe-55 1.21 E+01 1.08E+03 1.12E-05 100.00 1.21 E-04 Ni-59 6.39E+00 1.08E+03 5.91 E-06 100.00 6.39E-05 Ni-63 5.25E+02 8.11 E+02 6.48E-04 100.00 5.25E-03 Sr-90 2.79E+00 2.70E+00 1.03E-03 0.10 2.79E-02 Nb-94 1.04E-01 1.62E+01 6.42E-06 10.00 1.04E-05 Co-60 1.09E+02 1.08E+01 1.01 E-02 10.00 1.09E-02 Cs-137 1.26E+00 1.35E+01 9.35E-05 1.00 1.26E-03 Eu-154 1.28E-01 1.35E+01 9.47E-06 10.00 1.28E-05 Pu-238 1.52E+00 5.41 E-03 2.80E-01 0.01 1.52E-01 Pu-239/240 3.68E+00 5.41 E-03 6.80E-01 0.01 3.68E-01 Pu-241 8.66E+01 2.70E-01 3.21 E-01 1.00 8.66E-02 ;
Am-241 6.27E+00 5.41 E-03 1.16E+00 0.01 . 6.27E-01 !
Cm-242 5.48E-03 2.70E-01 2.03E-05 1.00 5.48E-06 Cm-243/244 - 1.15E-01 8.11 E-03 1.41 E-02 0.01 1.15E-02 Totals 7.55E+02 Type A: 2.47E+00 RQ: 1.29E+00 SCO il Determination l
~
PZR SCOll SCO II
~
uCi/cm2 Limit Fraction Fixed Beta / Gamma N/A * ---
Fixed Alpha N/A * ---
Removable Beta / Gamma N/A * ---
Removable Alpha N/A * --- 1 Inaccessible Surfaces 4.15 20 0.208 Combined Beta / Gamma Combined Alpha 0.62 2 0.310 <
~
Notes: * - All pressurizer contaminated surfaces are assumed to be inaccessible. Therefore only the limits for combined contamination in inaccessible areas are applicable.
- Alpha contamination level estimated based on averaged independent lab data.
- Contamination levels represent worst case estimates from the extemal MegaShield case.
Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 3-8
l Exhibit 4 Saxton Pressurizer NRC Classification Summary PZR Pad 61 Pad 61 Activity i
Nuclide Table 1 Table 2 mci uCi/cc nci/gm Limits Fraction H-3 Limits Fraction LLD C-14 LLD Tc-99 LLD l l-129 LLD Fe-55 1.21 E+01 8.51 E-03 Ni-59 700 0.000 6.39E+00 4.50E-03 220 0.000 l Ni-63 5.25E+02 3.70E-01 Sr-90 3.5 0.106 2.79E+00 1.97E-03 0.04 Nb-94 1.04E-01 0.049 7.33E-05 0.2 0.000 Co-60 1.09E+02 7.68E-02 Cs-137 700 0.000 1.26E+00 8.89E-04 Eu-154 1.28E-01 1 0.001 9.01 E-05 i Pu-238 1.52E+00 1.33E-01 i
100 0.001 Pu-239/240 3.68E+00 3.24E-01 100 0.003 Pu-241 8.66E+01 7.63E+00 3500 0.002 Am-241 6.27E+00 5.53E-01 100 0.006 Cm-242 5.48E-03 4.83E-04 20000 0.000 Cm-243/244 1.15E-01 1.01 E-02 100 0.000 Part 61 Class A Sum of the Fractions 0.013 0.156 t
1 Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98 3-9 i
4.0 REFERENCES
- 1. Westinghouse Electric Corporation, " Technical Manual Vertical Steam Generator," Technical Manual 1440-C49, Essington, PA,1960.
- 2. GPU Nuclear, "Saxton Nuclear Experimental Corporation Facility, Site Characterization Report,"1996
- 3. Westinghouse Electric Corporation, " Gamma-Ray Exposure Rate Distribution in a. Steam Generator," NP-3107, Electric Power Research Institute, Palo Alto, CA, May 1983.
- 4. Westinghouse Electric Corporation, " Primary-Side Deposits on PWR 4 Stesm Generator Tubes," NP-2960, Electric Power Research Institute, Palo Alto, CA, May 1983.
- 5. Westinghouse Electric Corporat_ ion, " Technical Manual Saxton l Pressurizer," Technical Manual 1440-C57, Essington, PA 1961.
- 6. Worku, G and Negin, C.A. "M/CROSH/ ELD", Version 4.21, Grove Engineering Inc., February 1995. I l
- 7. WMG, Inc., "MegaShield", Version 1.2, Peekskill, NY 1996.
- - l l
1 i
l Saxton Steam Generator and WMG Report 9803-7025 i Pressurizer Characterization Rev.1 3/98 4-1
l APPENDIX A GPUN PROVIDED DATA 1
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B & W RESULTS 11/3/95 - Saxton Samples SYSTEM = = = > Safety /niecean piorna - Pipe secton Between RVanet check Valve i Sarton i BAW '
ll 72806.000011 isama/e Numoer Samo/e Numoer Total TRU Content (nCi/g) l SX865950053 9509065-01 '
l 41 44004 Total Uranium Content (nC1/g)
I Sample Type i Pipe Section d i Desenotion 4 SCP, item 6, int. Il Results Normalized Nuclide Notes uCl/c to 100 % Nuclides RATIOS l 1 I H-3 LESS THAN I 8.37E-03 6 0.0040% r/##/###/###/####/###Auwwwwwwwwuuht 2 C- 7 4 i LESS THAN I 1.19E-021 0.0057% f#//#/#/###//#//####///// Ann %gg%gwwwwn) 31 Fe-55 l i 5.27E+0ol 2.5318% I Fs- 55/Co-80 l 0.14721t ,
4 Ni-59 i i 1.14 E +001 0.5477% f###/#/##/#/####//###/Annggnhwgg\ngT1 5 Ni-63 1 9.12 E+oi s 43.813e% r///#/###/#####/#//##/#Awwggnhugwwwwy 6 Sr-89 LESS THAN I N/R 1 0.0000% f#######/##/###//##/#A\wihnggggwngg) 7 Sr-90 l I 5.91 E-011 0.2e3e%Il Sr-90/Cs-137 I o.71205 il )
8 Nb-94 I i 1.65E-021 0.0079% f#/#/###/#/###//#####/Au%ung\nh\nntw) 9 Tc-99 i LESS THAN I 5.86E-04: 0. coo 3% r#/#/###/#/#f/######/#Autung\wn\wwwnt to /-729 i LESS THAN I 1.41 E-031 0. coo 7% r##//###/#/#//###//##/#A%gh\\kuwhw%wn) 11 No-237 i LESS THAN l 3.SoE-011 0.1826% l'//###//#####/###/###/Awwwww\nu\www\\1 j 12 Pu-238 i i 2.27E+0ol 1.0905 %I Pu-238/Am-241 l n.29290il I 13 Pu-239/240 i i 4.99E+001 2.3973% l'/#/###///#!//#####//#///An%gggghhuungt idl Pu-24f I 5.72 E+ 011 27.4796 %I Pu-241/Am-241 1 7.380654 15l Pu-242 i LESS THAN 2.35 E-02 I o.0113% r/#/########//#####'#Anuwwwwunghunj 1 61 Am -24 f i I 7.75E+001 3.7232%8 Am-241/Pu-238 6 3.4141 o j 171 Am-243 i LESS THAN 4 1.52 E-02 f 0.oo73% i'###/#/##////#/###//#//#A%wwwwn%%%%%%\
ist Cm -242 i LESS THAN i 8.11 E-021 0.0390%i'/#//#/##/#/#####/##////A%%%%%%%%%%%w1 19 ! Cm-243/244j LESS THAN i 9.62E-021 0.0482% l'####//#/#/####/#####A%%%%%%%%%%%%t 20 l U-234 I LESS THAN I 2.36E-021 0.0113% l'############/#/#/#//An%%%%%%%%%%%%
21 ! U-235 i LESS THAN i 6.54E-031 o.oo31 % f#//#/#!#####//#/#///N#/Aw%%%%%%%%%u%1 22 U-238 LESS THAN I 1.13E-021 0.0o54 % # U-238/Am-241 I o.oo146il 231 Co-60 l 3.58E+011 17.1988%I Am-241/U-238 i 685.640714 241 Au- 706 i LESS THAN 6 1.34 E-011 0.0644% f f/##//////#########/##A%%%%%%%%%%%%%
2Si Aa-708m I LESS THAN I 1.45E-021 0.co70% f//#####///#////////##/##1A%u\nu\%%%%%%%t '
4 26i Sb-f25 I LESS THAN l 3.17E-021 0.0152% f//##//#####/#/###////#/Aw%%%%%%%%%%%)
271 Cs- 134 ! LESS THAN I 2.33E-02t o.0112%4 Cs-184/Cs-137 0.02807i l 281 Cs-f37 i i 8.30E-011 0.3957% l Co-80/Cs-137 43.132531 i
291 Ce-f44 LESS THAN I 4.34E-02i 0.0208% ( Cs-137/Co-60 0.02318(
301 Eu- f 52 LESS THAN i 1.07E-011 0.0514% F#/#////#//#/####///#/##/Au%\hunuuuuuu\1 31i Eu- f 54 1 6.65E-021 0.0319% %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%w1 32i Eu- f 55 8 1.70E-02 i o.oo82% 1##//##/##/#/####/////#tA%%%%%%%VMVM l Totals = = > t 2.08E+021 100.00 %i1 l N/A - Not Reported i
l i
A-5 l
~~
B & W RESULTS 11/3/95 - Saxton Samples
- SYSTEM = = = > Steam Generator - Cold Leg Side. Bowl Smear Sanon
- B&W s l' 23 9 296r Jample Numberi Sample Numcer ' Total TRU Content (nCl)
' SX825950111 9509067-03 1 e 024932 Sample Type S/G Smear 1 Description SCP, item 1 j Results Normalized Nuclide ,
Notes uCi to 100 % Nuclides RATIOS 1i H-3 i 7.50E-04 '
~
LESS THAN O.0209% ////#/#/#/#/##/####/##Aunwauwmmmm 2l C-14 i GSS THAN 2.58E-03 i 0.0720% F///#//##////#/#//####///Amuunnununnut 31 Fe-55 . - 8.46E-02 ' 2.3599 % l, Fe-55/Co-60 '
O.11132; 44 Ni-59 l 1.57E-02
- 0.4379% r///U////#//#/##//###/////Au%%%%%%%%%%% 4 5i Ni-63 i 1.43E+ 00 , 39 8896% r////#/#/##/#/######///AWWWWmunuunM 61 Sr-89 ' ESS THAN N/R O.0000% f////####/######///##/AnumhuhMwmM '
7' Sr-90 a 2.64E-01 ' 7.3642 % ( St-90/Cs-137 8 0.27557:
S .. ND- 94 5.17E-04 : 0.0144% F///####///#//##/###//#AMuwmmanunng 94 Tc-99 . ESS THAN 2.52E 0.0070% f####/###/#//#/##///#/Anwnum%munnu 103 /-f29 . ESS THAN > 6.37E-05 0.0018% F##!####/#######//#!AnunmnwunngM 11 Np-237 GSS THAN 180E-04 ! 0.0050% F#/#/#/##//#/#//#///###Anunmum\nnuM 12t Pu-238 ,. 6.88E-04 : 0.0192%i Pu-238/Am-241 i 0.39769:
13 F Pu-239/240 1.45E-03 : 0.0404% )/#/#/#////#/##///##/#///AuwnunumnunM 141 Pu-24 f i 1.99E-02 f 0.5551%( Pu-241/Am-241 i 11.502891 15J Pu-242 ! GSS THAN 8.32E-06 ' O.0002% F#/#######/#/#//#/##/A%%%%munuwwM 161 Am-24 f , 1.73E-03 0.0483% d Am-241/Pu-238 ! 2.51453 17* Am-243 4 ESS THAN 2.91 E-06 : 0.0001 % F#/#####//#/#/////#/#/#A%%Bunu%%numt 189 Cm-242 1 ESS THAN 1.37E-05 i 0.0004% (//#####/#////##/#/###AMuun%uuhu%nn 193 Cm-243/244 GSS THAN 2.38E-05 i 0.0007% r///#/##//#/###/###//#/AMWWWunQuuum ,
204 U-234 . WSS THAN
- 1.38E-04 : 0.0038% F#//#///##////##/#//##/#A%%%%%%%%muut 21.: U-235 ESS THAN t 6.52E-05 i 0 0018% r##///#/#####////un/##Aunu%uuuuuwmt 22' U-238 ; ESS THAN 4.61 E-05 : 0.0013%s U-238/Am-241 ' O.02665 231 Co-60 ,
7.60E-01 : 21.2000% d Am-241/U-238 37.52711:
24.i Bu-706 GSS THAN 1.24E 0.3459% (//#/#/#/#///####Il#/////AhwBWWWWWmWM 25, Aa-108m . ESS THAN 1.23E-03 i 0.0343% 7/#/#####/n#/##//###/AhwBWWumwwmW 26i Sb-f25 i GSS THAN ! 3.68E-03 i 0.1027% F////#/#/#////####//#/////AwwwwwwWmu%M 27 Cs-f34 3 GSS THAN 1.74E-03 i 0.0485 % Cs-134/Cs-137 . 0.00182' 28: Cs-f37 9.58E-01 ' 26.7232W Co-60/Cs-137 0.79332' 294 Ce-f44 WSS THAN 4.67E-031 0.1303 % 1 Cs-137/Co-60 1.260534 30: Eu-f52 4 GSS THAN ' 1.22E-02 ! 0.3403% ///##/#//##/#////##/##//A%%%M%%%nmm) 31' Eu-f54 d ESS THAN 7.05E-03 ? 0.1967% 1B%%%%%%%%%%%%%%%%%%%M%%%%B%
32: Eu-f55 .i GSS THAN 1.22E-03 : 0.0340% ///,/#f!!/////#//f/////////////AWmWBW9WW9%4
-L Totals = = > 13 58E+00; 100.00 % 4 N/R - Not Reported l
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8 Nuclide.Idsntificatica Papart 10 .29-97 5:41:50 PM Paga 4 caoooooooooooseeoooooooooooooooossesseessessessesseeeeeeeeeeseseaeo eeeo, eco**
NOCLIDE* I D E N T I'F I'C A T I O N REPQR ooosseeeeeeeeeeeeee .... eeese*e ....eeeeeeeeeeevesseeeeeeeeeee.T
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Sample
Title:
SX8-25-97-0178-SM Nuclide Library Used: C:\GENIEPC\CAMFTLES\150.NLB
. . . . . . ;.. . . . . . . . . . . . . IDENTIFIED NUCLIDES
'Nuclide
~
Id Energy Vield Activity Activity Name Confidence (kev) (%) .(uci/UN .) Uncertainty 00-60 O.983' 1332.49* 100.00 1.351E.-01 1.312E-02 -
CS-137 0.981 661.65* 85.12 6.42SE-01 6.480E-02
~
723.30 '19.70 -
1004.76 17.90 '
1274.45 35.50 -
i AM-241 0.972 59.53' 35.90 1.858E-02 2.052E-03
- = Energy line found in the spectrna.
Energy tolerance used was 2.000 key
.Nuclide confidence index threshold = 0.30 Errors quoted at 2.000 sigma .
ca,sessee** U N I D E N T I F'I E D, P R A K S' *********e l Peak Locate Performed on: 10-29-97' 5:41:31 PM Peak IAM: ate From m=nnels- 80 Peak. Locate To Channel:- 4096 Peak Energy . Peak Size in Peak CPS No. (kev)
Counts per Second 5 Uncertainty .
4 703.08 ' -
1.0071E-01 73.52 M = First peak in a multiplet ~ region m' = Other peak in a multiplet region ,
/
Errors quoted at 2.000 sigma ,
w-wv -
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Nuclide,Idditification 'Rsport 10-29-97 6 53:03 PM P:g*2 4
................................. .................es....... .....e.... .
Co*** NUC.LIDE IDENTIFIC-ATION REP evesessesee.s oooses?*.ee.eeeeee......ee. .se**s.es.o...eeeeeeeeeeeees.sa.O R T-Samp1'e Titlei SX8-25-97-0179-SM '
Nuclide Library Used: C:\GENIEPC\CAMFILES\150.NLB
...........,........ IDEhTIIFIED NUCLIDES ,...........,.........
Nuclide . Id Energy Yield Activity Activity Name Confidence (kev) .(%)' (uCi/UN ) Uncertainty 1.483E-01 1. 439E-02 '
00-60 0.987 1332.49* . 100.00' '
CS-137 0.987 , 661.65* 85.12 6.752E-02 '6.840E-03 344.27 . 26.50 .
778.89 12.74 ,
1112.02 13.30 '
1 1407.95 20.70 19.70 723.30 .
1004.76 17.90
.' 1274.45 35.50 AM-241 0.984 59.53* 3.5.90 5.392E-03 6.6995-04
- = Energy 1ine found in the spectrum. ,
Energy tolerance used.was 2.000 key- '
Nuclide confidence index threshold'= 0.30 '
Errors quoted 'at 2.000' sigma
- U N I.'D E N T I F I E D PBA E5- *********
Peak' Locate Perforne'd on: 10-29-97. 6:52:44 PM Peak Locate From t'hannel: 80 , .
Peak Locate To Channel: 4096 Peak Energy. Peak Size in Peak CPS No .' (kev) Counts per Second-- 5 Unce,tainty r
. 4 *703.06 3.29855-01 62.08 M = Pirat peak in m' multiplet region
~ -
s = Other. peak in a multiplet region .
Errors quote.d sti .2.000 sigma -
^
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Nuclid2 Identification R: port 10-30-97' 2:33:06 PM Pag 3 4 com e s s e s s e s e ooooooo cocoo o o o o o ooo n so oo o ooo oooooo oooo ooooooooococcocooo oo o o ,
00*** N U CT L I D E IDENTIFICATION REPQRT *****
1 aceseeeeeeeeeeeeeeeeeeeeeeeessessessessesseerreeeessenessesesenessesseeee Sample"
Title:
SX8-25-97-0179-SM-Nuclide Library Used: C:\ GEN.IEPC\CAMPILES\150.NLB ,
.....,..,.............. IDENTIFIED NUCLIDES ....................
'Muclide Id Energy Yield Activity Activity Name Confidence '.(kev) (%)' (uCi/UN .) Uncertainty 0.997 1332.498 l'. 4 3'6E 1.397E-02 CO-60 10'O.00 l CS-137 0.987- 661.65* 85.12 -
6.781E-02 6.891E-03 AM-241 0.964 59.53* 35.90 5.477E-03 7.318E 04
- = Energy line found in the spectrum.
Energy tolerance used was 2.000 kev ,
l
. Nuclide confidence index threshold = 0.30 [
i Errors quoted at 2.000 sigma ,
- , UNIDENTIFIED PEAKS . ********** ,
l . Peak Locate Performed on:' 10-30 2.32:49 PM .
I Peak Locate From Channels 80
. Peak Locate To channel:
- 4096 Peak size in
. Peak Energy Peak CPS No. (kev) Counta per Second 5 Uncertainty-All peaks were identified.
l ,
~ j M = First peak in a multiplet region m = Other. peak in a multiplet region ,
Errors quoted at
~
2.000 sigma I
S mem S cvng t.eg., nas n .
- Sfc.( WL) D men- PLes~ S $nNL No. H ee.s SveNA@ No. 8% %*i-\ft. .
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l Se e A ' O TOTAL. P.23 !
' 814 635 2671 - P.03 OCT--35-1997 15:45 i l '
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' Nuclidy Idrntificatica Report 10-29-97 S:06:16 PM- Pcte 4
..............................5.J.............................w........'
.oco** NUC IDENT C'A T I O N R'E P Q R T *****
000....e......L.I D E. s...,.......*.......I.F...I ........t......................s
' Sample
Title:
SX8-25-97-018'l-SM '
Nuclide Library Used: C:\GENIEPC\CAMPILES\150.NLB
.................... IDENTIFIED NUCLIDES ....................
Nuclide . Id Energy Yield Activity Activity Name Confidence (kev) (%) '(uci/UN ) Uncertainty 00-60 0.986 1332.49* 100.00 1.052E-O'1 1.023E-02 CS-137 0.984 661.65* 85.12 2.9.45E-01 2.972E-02 .
723.30 '19.70 .
1004.76 17.90 EU-154 4.939E '
0.763' 1274.45* 35.50' 8.562E-04' AM-241 0.974 59.53* 35.30 1.665E-02 1.817E.-03
- 'e Energy line*found in.the spectramt. .
' Energy' tolerance used was 2.000 key.
'Nuclide confidence index threshold =
- 0.30 '
Errors quoted at 3.000 sigma ,
\
C********* P E A.K'S ******* m -
U N.I D E N T I F'.I E D .
' Peak Locate Perf*ormed'on: 10-29-97 6:05:57 PM Peak Locate From Channel: 80 '
Peak Locate To Channel: 4096 t Peak Energy Peak Size in . Peak CPS No. (kev) Counts per.Secon,d 5 Uncertaint.y ,
All; peaks were identified. l
~
. M = First' peak in a multiplet region . . .
m = Other peak in a.sultiplet region -
Etrors ciuoted at 2.000 sigma.
s e a . S. s ,. scany esi-wr-issj ,.
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,'Nuclide Id2ntifiesticn Report ,
10-29-97 6:31:25 PM Paga 4 ocesseesseeeessessessessesseseessee senseseass'essee seesenesseeeeeeeseesse'
.00c** NUCL1DE IDENTIFICATION RE OR coce ee e e s e s s e ss e e s s e.. esse ee.* *e s se n e s se me es e essesee.
= =es.....P.
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r' Sample _
Title:
.SX8-25-97-0182-SM .
'Nuclide Library Used: C:\GENIEPC\CAMFILES\150.NIS
. . IDENTIFIED NUCLIDES ............'........
.Nuclide. Id Energy Yield Activity Activity Name Confidence' (kev) (E) (uci/UN -) Uncertainty CO-60 0.989 1332.49' 100.00 1.832E-02 Ic.821E-03 s CS-137 0.989 661.65* 85.12 9.421E-02 9.524E-03 723.30 19.70
'1004,76 . *17.90 1274.45 35.50 -
AM-241 0.988' 59.53* 35.90 6.'054E-03 6.991E LO4 1 i*
- Energy line found in.the spectrum. ,
. Energy tolerance used was 2.000 key *
' Nuclide confidence index threshold =
0.30
, Errors quoted at 2.000 sigma -
j J
U N I D E N T I F I E'D -
P'E A E 5 -
Peak Locate Performed.'on: 10-29-97 6:31:07 PM '
Peak Imcate From Channel: 80 Peak Locate To Channel:.
'4096 '
, ' Peak Energy Peak Size in Peak CPS No . , ,,
(kev) ,
Counts per Second E Uncertainty .
,All peaks were identified. -
l M = First' peak in'a. multiplet region .
m = Other peak in a multiplet region-
- .Brrors quoted at .2,.0'00 sigma .
No'., ' 59 dl~4-- I r1 j 6mnAa 'rdo.
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Nucildo Identification Report 10-29-97 5:18:10 PM Pagn. 4 ooses e= = sees s esee s ses ene sseseses e' see eeeme sseme ne ssessesse es'e s ee ssessesse aces * . .N U C L I D S-oce.cocessessessemeseeeeeeeeeeeeeeeeeeeeeeee... IDENTIFICATION..eeee.R . .. E.se...esessee... PORT' Sample
Title:
SX8-25-97-0183-SM -
Nuclide Library Used: C:\GENIEPC\CANFILES\150.NLB
. IDENTIFIED NUCLILAS ......;..... .......
Nuclide
- 2d Energy Yield
'Activity Activity -
"Name, Confidence (kev) . (%) (uci/UN ) Uncertainty .
C0-60. 0.985 1332.49* 100.00 S.488E-02 5.367R-03
- CS-137 0.981- 661.65* 85.12 2.180E-02 '2.211E-03 .
723.30 19.70-1004.78 ' 17.90 .
. 1874.45 35.50 -
AM-241 0.972' 59.53* 35.90 .' 3.284E-03 4.1,31E . .
- = Energy' IIne found in the spectrias. .
Energy tolerance used was 2.000 key . .
Nuclide confidence inder threshold.=. 0.30 . -
. Errors quoted at 2.000 sigma e*o*******' UNIDEN'TIFIED F E A E S- **********' '
Peak Locate Performed on': 10-29-97 5:17:$2 PM Peak Locate From Cha w l: 80
- Peak Locate To Channel: 4096 Peak Energy Peak' Size in. . Peak CPS .
No. (kev) Counts per Second 5 Uncertainty i .
All' peaks were identified.
Me 'First peak in.a multiplet region -
.a = Other peak in a anitiplet region 2.000' sigma' l .
Errors' quoted at '
i- - -
l .
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APPENDIX B MEGASHIELD / MICROSHIELD CASES l
I l
f Saxton Steam Generator and WMG Report 9803-7025 Pressurizer Characterization Rev.1 3/98
MegaShield 1.2 - Summary Output ,, -
Case
Title:
Saxton Steam Generator on Contact Run By: /
File Name: C:\GPU-SAX \ SAX.MS1 Total Run Time:1641 seconds Reviewed By:
Q Run Date: 3/25/961:49:03 PM _ s Case Model Source Data Geometrr Cyhnder-Vtl I
/g . - .
Radius. 5334 cm
- ,; , . g .j -
Heght: lKB.25 cm
. 3J;$. ,.
" '~ Mass: 2.643EG gm
. . W yg g gepip;. .
{d<,'g
.;g pgg p y
-l$gg Qdg 5; Volume:
Matenal/ Density-2.764EG cm*3 Iron 096 6;< . .;;. d4 .
p w.- m c x.:: e 1.
Shield Data IIJ; ?$@A#
- 4. ~
$ Shield #: 1
{bNEM[ [ Geometry inner Rad:
Cyhndncal S3.34 cm IJ y hk.[.lsl e D Outer Rad: 53.98 cm 064 cm Y: Thickness:
d I .'1hijiN );
7
- Matenal/ Density' fron 7.87
^
Shieki #: 2 B Up
$14 ~ ,
- if 4 Geometrr Cyhndrical
): . .-
h, ]t inner Rad:
Outer Rad:
58.74 cm 66.04 cm
.$ ff""'
. f f; W. -
. (- Thickness: 7.30 cm y "* Vfp _ }[ Matenal/Denstr Iron 7.87 Integration Pararneters: Radial:100 Circumferential:100 Axial:100 Source Activity - Std. Grouping - Lower Energy Cutoff = 1.000E-01 Curies Sq l l Nuclide 1.300E 01 4810E@
l Ba-137m jCo'30 1.9 E E 01 7.173E+09
)
l Case Results !
i Exposure Rates m Ar Dose Equrvalert Rate l
Detector Coordinates l in cm mR/hr mrem /hr for AP* Exposure l
Detector # ; X !
Y j Z l w/o Buildup j wth Buildup j w/o Buildup j wth Bulk 2Vp 7.209E41 4.2S4E+00 1 1 67.04 l 154 62 j 0.00 j 8.032E41 l 4 773E+00 l } {
- AP refers to Antenor-Postenor Exposure Geometry as per ANSl/ANS-6.1.1-1991 I
/c 00 mc: # c. .
D CP h 9. T 5/a73*%[ m e/,
f f-l I
2 Assuming the rated heat transfer area (2300 ft ) pertain's to the secondary side area of the tubes above the tube sheets, and that the tubes are essentially right circular
, cylinders, the primary side area of the tubes can be calculated from a ratio of tube outer and inner diameters by.
Outer Tube Area = Outer Tube Circumference x length
= x x Tube OD x length inner Tube Area = Inner Tube Circumference x length
= x x Tube JD x length '
Assuming the primary versus secondary side tube length disparity associated with the tube bend is negligible and therefore the outer and inner length terms are identical, then the preceding two relationships can be combined to yield:
Inner Tube Area = Outer Tube Area x Tube ID + Tube OD
= 1,692,322 cm2 The tube sheet is 9.5 inches thick. The primary side area of 1472, 0.495 inch diameter, 9.5 inch long tubes is 140,299 cm2,
- The outer channel bowl spherical radius is 27.125 inches. If the wall thickness is 2.87 inches, the inner spherical radius is 24.25 inches. A hemisphere of 24.25 inch radius 2
has an area of 23,838 cm [ Area = 2r.(R')]. The channel bowl area is not a perfect hemisphere and has an approximately 5 Mch tall right cylinder of identical radius .
l adjacent to the tube sheet. The wette,d periphery that cylinder has an area of 4915 2
l cm . The areas associated with the 10.834 inch diameter inlet and 12.084 inch l - diameter outlet nozzles were approximated as circles having a total area of 1335 cm*;
The combined geometry described above leads to a net channel bowl area of 27,418 2
cm .
l The. circular tube sheet shares the inner channel bowl radius of 24.25 inches but is
' perforated with the 1472 tube holes. The tube sheet area is therefore x(24.25, 2 (368(0.4952))) or 10,091 cm2, Each side of the divider plate is a combination of a hemicircle of 24.25 inch radius and a mating rectangle 5 inches high. The two sides of the divider plate have an area of
. 15,048 cm*.
The total primary coolant wetted area of the Saxton steam generator is thus 1,885,178 cm 2,
(
v 2/n/w
. g/
b - 2.
C ._____m-___ - - - - - - - - - - - - - - - - - - - - - - --
Pressurizer Area Calculation Ref: Tech Manual 1440-C57. Fig.1-3 ftem in / inn em2
- 1. Shell Area ID 36.375 Length of Cylindrical Section 144 Hemispherical Head (s) Area 4156.769 Cylindrical Area 16455.66 Interior Area of Shell 20612.43 132983.2
- 2. Spray Nozzle Spray Nozzle OD (1in Sch 80S) 1.315 Spray Nozzle Length 19.75 Spray Nozzle Area 81.59109 526.3931
- 3. Heaters Hester OD 0.573842 Heater Length 43 Heater Area 1395.35 9002.24
- 4. Hester Bundle Plate Bundle Plate OD 14 Bundle Plate Thickness 0.5 Non Heater Hole Diameter 1.75 Hester Hole Diameter 0.671875 Center Hole 1.6875 Bendle Plate Area 297.7077 1920.691
- 5. Hester Support Pipe Support Pipe OD (1.25 Sch 80) 1.06 Support Pipe ID 1.278 Support Pipe Length 37 Support Pipe Area 341.51 2203.286 6.' Surge Ring Surge Ring OD (3in Sch 80S) 3.5 Surge Ring 10 (3in Sch 80S) 2.9 Approx Surge Ring Diameter 28.25 Surge Ring inner and Outer Area 1784.424 11512.39
- 7. Total Pressurizer Area 24513.01 188148.2 r
1
. ,, . (
1 MegaShield 1.2 - Summary Output Case
Title:
Saxton Pressurizer Surface Source with New data Run By:
]s File Name: C:\ TEMP \SAXTON\PZR-NEW.MS1 Run Date: 12/17/9710:54:22 AM Total Run Time:14 seconds Reviewed By:
Case Model Source Data 5 l Geometry. Cytinder-Sfc ,
- - j j Radius. 45 72 cm
- , .,, Heght
- 472.44 cm
[
g Surface Area: 1357E+05 cm^2 l ,
-[ - [
J';.
Mass:
Volume:
N/A N/A
",p. j[ Matenal/ Density: N/A I
- v - \
-7 f . Shield Data
- I'; Shield #: 1 B Up 4 I' . Geometry: Cylindncal
( .f/; .,1 Inner Rad: 46.20 cm
? f;j.'- . Outer Rad: 5493 cm j{/ j Thickness: 8.73 cm l
- 'j . , q Matenal/ Density- Iron 7.87 l gp.
h
$,,5..
i Y .N y
n . *^a .
_ '~e integration Parameters: Circumferential:100 Axial:100 Source Activity - Std. Grouping - Lower Energy Cutoff = 1.000E-01 Nuclide Curies Sq l Ba-137m 1.500EG 5.550E+07 Co-80 1.649E-01 6.101E+m ;
ICCO M i hPAL\ '
DC F = g,g,me =M 4 mP/'h--
h .-
Case Results i Detector Coordinates Exposure Rates in Air Dose Equrvalent Rate in em mR/hr mrem /hr for AP* Exposure l
X j Y j Z j w/o BuddUp with Buildup w/o Buildup with ButidUp l Detector # l !
l 1 i 55.93 j 236.22 j 0.00 l 1.115E+00 ; 5.926E+00 1.001 E+00 l 5.316E+00 j
'AP refers to Antenor-Postenor Exposure Geometry as per ANSI /ANS41.1-1991
~
w . . -
MicroShield 4.00 -
Serial #4.00-00112 WMG Inc.
Page : 1 File Ref: (JPu so7s n DOS File: SAX-PZR.MS4 .
Date: / /1 Run Date: January 7, 1998 By: _
Run Time: 10:33 a.m. Wednesday Checked: /E'f Duration: 0:00:29 Case
Title:
Saxton Pressurizer Internal Dose Point for PZR Shell GEOMETRY 9'- Cylinder Surface - Internal Dose Point centimeters feet and inches Dose point coordinate X: 0.0 0.0 .0 Dose point coordinate Y: 91.44 3.0 .0 Dose point' coordinate Z: 0.0 0.0 .0 Cylinder surface height: 182.88 6.0 .0 Cylinder surface radius: 45.72 1.0 6. 0 Source Area: 52535.4 sq cm 56.5487 sq ft. 8143.01 sq in.
MATERIAL DENSITIES (g/cm*3)
Material- Cylinder Shield Air- 0.00122 BUILDUP Method: Buildup Factor Tables The material reference is Cylinder INTEGRATION PARAMETERS Quadrature Order Axiar (along Z) 98 Circumferential 98 SOURCE NUCLIDES Nuclide curies gCi/cm8 Nuclide curies pCi/cm2 Ba-137m 1.5000e-003 2.8552e-002 Co-60 1.6490e-001 3.1388e+000
========================= RESULTS =========================
Energy Activity Energy Fluence Rate Exposure Rate In Air (MeV) (photons /sec ) (MeV/sq cm/sec) (mR/hr)
No Buildup With Buildup No Buildup With Buildup 0.6 5.093e+007 6.403e+002 6.436e+002 1.250e+000 1.256e+000 1.0 6.101e+009 1.280e+005 1.284e+005 2.359e+002 2.368e+002 1.5 6.101e+009 l'.921e+005 1.927e+005 3.233e+002 3.241e+002 l- TOTAL: 1.225e+010 3.208e+005 3.217e+005 5.605e+002 5.622e+002 gp . /0* M'i : /. 779 562.1 Cf. 91Eht" f./
B -5
Pressurizer MegaShield input Ref: Dwg 6490772 Item Value Heater Bundle Weight, (excluding Diaphram Plate &
Connectors)Ibs 42 12 Heater Circle Diam, in 12 Approx Heater OD, in 0.671875 Heater Length (above Diaphram Plate), in 43 Outer Element Diam, in 12.33594 Volume in' 5139.284 Ref: Dwg 6490774 Surge Ring Weight, Ibs 31 Combined Weight,lbs 73 Volume in' 5139.284 Eq Source Density, g/cc 0.393174 l
I l
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i 43.0 i
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. .' w MegaShield 1.2 - Summary Output Case
Title:
Saxton Pressurizer Heater Bundle Run By:
V File Name: C:\ TEMP \ HEATER 2.MS1 Run Date: 12/17/97 3:37:51 PM Total Run Time:600 seconds Reviewed By: [
Case Model . Source Data Geometry- Cyhnder-Hrz B Up
, H 1 ,
Mass: 3.300E+04 gm Volume: 8.415E+04 cma 3 Matenal/ Density- tron 0.39 Shield Data
<None>
l
,f Integration Parameters: Radial:15 Circumferential: 98 Axial:100 Source Activity - Std. Grouping - Lower Energy Cutoff = 1.000E-01 Nuclide Curies Sq ,
Ba-137m 1.500E-G3 5.550E+07 Co40 1.649E41 6.101 E+09 l
tccomC, w mO ec== M 6 rJ' h r- =-
% mR/'hr
/
Case Results Detector Coordinates Exposure Rates in Air Done Equrvalent Rate l
in cm mRihr mrem /hr for AP* Exposure Y Z w/o Buddup witn Buddup w/o Bundup with BuddUp Detector # X 1 85.09 15.66 0.00 2.550E+02 3.372E+02 2.289E+02 3.027E+02 115.57 15.66 0.00 1.CD7E+02 1.435E+02 9.309E+01 1.289E+02 2
15.66 0.00 5.632E+01 8.027E+01 5.056E+01 7.207E+01 3 146.05 15.66 0.00 3.545E+01 5.155E+01 3.183E+01 4 628E+01 4 176.53 ,
5 55 61 15.66 0.00 1.341 E+G3 1.678E+03 1.204E +03 ifD7E+03 j
- AP refers to Antenor-Postenor Exposure Geometry as per ANSt/ANS-6.1.1 1991 l
b7 L
i Attachment 1 s
l Saxton Nuclear Experimental Corporation Facility Reactor Vessel Shipment Exemption 1 Request -
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ATTACHMENT 1 EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPNENT OF THE SAXTON REACTOR VESSEL 49 CFR 107.105 Application for exemption.
49 CFR 107.105(a) General
- 49 CFR 107.105(a)(1): The requested ne' e d date for this exemption is September 9,1998.
Two copies of this exemption have been delivered to:
Associate Administrator for Hazardous Materials Safety Research and Special Programs Administration U.S. Department of Transportation 400 7th Street, SW Washington, D.C. 20590-0001 Attention: Exemptions, DHM-31 49 CFR 107.105(a)(2): The correct applicant name, address and responsible agent for this exemption is:
Applicant Saxton Nuclear Experimental Corporation (SNEC) and GPU Nuclear
' Inc.
Mr. G.A. Kuehn Jr.
Program Director SNEC Facility
, GPU Nuclear Inc.
2574 Interstate Drive Harrisburg,PA 17110 49 CFR 107.105(a)(3): Saxton Nuclear Experimental Corporation and GPU Nuclear Inc. are i United States corporations.
49 CFR 107.105(a)(4)i This is not a request for a Manufacturing Exemption.
49 CFR 107.105(b):' Confidentialtreatment
. Confidential treatment of this exemption is not requested.
j 49CFR l07.l05(c) Description ofexemptionproposal L
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l RVAttach. doc .10.07 AM 6/24/98 - Page1 '
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ATTACHMENT 1
' ' EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPMENT OF THE SAXTON REACTOR VESSEL 49 CFR 107.10S(c)(1): With regard to the transportation of one reactor vessel from the Saxton Nuclear Experimental Corporation Facility site in Saxton, PA to the burial site at Barnwell, S.C., the Applicant seeks relief from the requirements of 49CFR173 as follows; P ACKAGING REQUIREMENT The requirement of 49 CFR 173.427(a) that low specific activity (LSA) material must be packaged in accordance with 49 CFR 173.427(b) or (c).
DOSE RATE AT 3 METERS The requirements of 49CFR 173.427(a)(1) regarding the 10 mSv/hr (1 Rem /hr) radiation dose limitation at 3 meters from the unshielded material.
LS A III DEFINITION l
The requirements of 49CFR173.403 regarding the definition of LSA-Ill r
which does not provide for surface contaminated LSA material s.
I LSA Ill MATERIAL LEACH TESTING The leach testing required by 49CFRl73.468 for LSA III material which is also included in the definition of LSA Ill in 49 CFR 173.403.
Sec 173.403 Definitions requires that Low Specific Activity (LSA) i materials consist of Class 7 (radioactive) material with limited specific activity and the determination of such specific activities may not consider
- the shielding materials nrrounding the LSA material. For LSA-III solids, i
this Section further proddes:
- 1. That such materials meet the requirements of 49CFR 173 468, which provides detailed requirements for the LS A III leach testing and ,
i
- 2. Have the Class 7 (radioactive) material " distributed throughout a solid or a colle: tion of solid objects", and
- 3. Have an average specific activity not to exceed 2x10-3 A2/g, and j i
RVAttach. doc 10:07 AM 6/24/98 Page 2
ATTACHMENT 1 EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPMENT OF THE SAXTON REACTOR VESSEL 4 Consist of Class 7 (radioactive) material which is relatively insoluble so that even under loss of packaging, the loss of material by leaching in water for 7 days shall not exceed a 0.1 A2 quantity.
49 CFR 107105(c)(2): The specific modes of transponation for this exemption request are
- 1) Motor Vehicle Transportation
- 2) Rail Transportation The reactor vessel will be transponed from the SNEC facility site propeny by land transporter to a local railroad siding for transport by rail car to a rail siding at the burial facility site.
49CFR107.105(c)(3): A detailed description of the proposed exemptions follows as well as the 49CFR107.105(c)(5): basis for the exemption requests.
The Saxton reactor vessel! internals consist of the activated internal components and the activated reactor vessel which surrounds the intact immovable (grouted) internal components. A detailed description of the reactor vessel package is provided in Attachment 2. These materials comprise the Class 7 (radioactive) material. These materials are packaged within a transponation system which provides equivalent safety to an Industrial Package Type 2 (IP-2) as described below. The determination of specific activities did not include the container shielding or the grout within the reactor vessel.
The Saxton reactor vessel with its intact grouted internals, when packaged in a transportation system as described below, meets the above requirements whether the activated metals which comprise the material are considered "a solid" or "a collection of solid objects" since each component within the vessel and the vessel itself have specific activities l below the LSA-Ill limit of 2x10-3 A2/g. On average, the vessel plus the l internals, excluding the grout, (i e., the Class 7 material) has a specific activity of 1.7E-6 A2/g and the most radioactive individual component within the internals has a specific activity of 7.lE-5 A2/g. These specific activities correspond to 0.1 and 3.6 percent of the LSA-Ill limits, respectively.
9 j RVAttach. doc 10.07 AM 6/24/98 Page 3
9 ATTAClufENT I l
EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPMENT OF THE SAXTON REACTOR VESSEL l
The bases for the exemption requests are due to the unique characteristics of the radioactive material to be transponed and the t
administrative controls that will be implemented during transportation.
The basis for each exemption requested is discussed below.
I PACKAGING REOUIREhfENT 49 CFR 173.427(a) requires LSA material to be packaged in accordance with paragraph (b) or (c) of this section. For LSA 111 material transported as an exclusive use shipment,49 CFR 173.427(b) and Table l 8 would require that the vessel be packaged in an Industrial Package type
_ (IP-2). IP-2 package design and certification requirements are stipulated in 49 CFR 173.411. Under the requirements of 49 CFR 173.411(b)(2) each IP-2 must meet the general design requirements of 49 i CFR 173.410 and prevent the loss or dispersion of radioactive material
! and significant increases in the radiation levels under the testing requirements of 49 CFR 173.465(c) and (d) or evaluated in accordance with 49 CFR 173 461(a).
The applicant proposes to transport the reactor vessel in a transportation system (non-specification packaging) which provides equivalent safety to an IP-2 package when transported in accordance with the transportation plan and emergency response plan. A description of the Saxton reactor vessel transportation system is enclosed as Attachment 2, the transportation plan is included as Attachment 3 and the emergency response plan is included as Attachment 4.
i The proposed transportation system provides containment of the radioactive material by the following means:
- All reactor vessel penetrations will be seal welded to provide closure.
- The reactor vessel interior will be filled with low density giout (25 lb/ff) to fix the surface contaminants and components in place.
- The vessel exterior surfaces will be painted with an epoxy coating to l fix external contamination.
i a The reactor vessel will be placed in an outer two (2) inch thick steel container for double containment.
- The space between the reactor vessel and outer container will also be filled with low density grout (65 lb/f13).
{ RVAttach doc 10.07 AM 6/24N8 Page 4 )
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ATTACHMENTS EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPMENT OF THE SAXTON REACTOR VESSEL This transportation system is a robust package which includes multiple barriers to prevent release of any radioactive material under normal conditions of transport.
The transportation system provides equivalent safety to an IP-2 package by ensuring that it is designed in accordance with all the general design requirements specified in 49 CFR 173.410. The transportation system is also designed in accordance with the additional design requirements for IP-2 packages of 49 CFR 173.465(c) and (d) within the limitations of the transportation plan.
All Saxton reactor vessel activities will be controlled by the Transportation Plan and the Emergency Response Plan which will in part require that the Saxton reactor vessel remain in a horizontal position during all transportation evolutions. In addition, all external lifting devices will be removed or disabled. Therefore, reactor vessel qualification for a free drop from any orientation up to 30 from horizontal was analyzed as opposed to the orientation which would cause
" maximum damage to the safety features being tested" Analyzing horizontal drop scenarios up to 30* is conservative for the conditions of transport as regulated by the Transportation Plan.
It should be noted that the transportation requirements for LSA material presented in 49 CFR 173.427(a) will be met with the exception of the packaging requirements discussed above and the dose rate at 3 meters from the unshielded material specified in 49 CFR 173.427(a)(1) presented below.
DOSE RATE AT 3 METERS Regarding the dose rate limitation of 10 mSv/hr in 49CFR 173.427(a)
(1), the worst case dose rate at 3 meters from the vessel exterior is calculated to be 6.6 mSv/hr (660 mrem /hr). This is an estimated dose rate based on historic radiation survey data taken on the exterior of the reactor vessel. The irradiated reactor vessel and internal components are considered as a collection of solid objects and it has been shown above that the worst case component concentrations are well within the limitations of 49CFR 173.403. The unshielded dose rate at 3 meters
} from some components, if considered separately, will exceed 10 mSv/hr.
However, these internal components are an integral part of the reactor vessel.
RVAttach doc 10:07 AM 6/24/98 Page5 l
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9 ATTACIBENT 1 EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPNENT OF THE SAXTON REACTOR VESSEL The 3 meter radiation level requirements of Sec 173.427 (a)(1):
The basis for this requirement is loss of package shielding under normal conditions of transport and the resultant dose if the package surface radiation level exceeds 10 mSv/hr at 3 meters. The worst case dose rate at 3 meters from the vessel exterior is calculated to be 6.6 mSv/hr (660 mrem /hr).
Some components within the reactor vessel, by themselves, will lead to 3 meter dose rates greater than 10 mSv/hr. However, these internals components are an integral part of the reactor vessel. They are mechanically fixed to the reactor vessel itself and surrounded by grout j within the vessel. Thus, even if the integrity of the package is breached I in its entirety under normal conditions of transport, the dose rate at 3 meters from any of these components could not exceed the maximum dose rate of 6.6 mSv/hr at 3 meters from the exterior surface of the reactor vessel.
LS A III DEFINITION The defmition of LSA III material includes provisions for consideration of activated materials as LSA Ill but does not specifically address activated materials which are also surface contaminated. Although the definition does not include surface contaminated LSA material, the applicant does not believe it was intended to exclude activated metals with surface contamination. As a practical matter, any activated metals generated in a commercial reactor will have some level of surface contamination.
LS A HI LEACH TESTING REQUIREMENTS The internals components and the reactor vessel interior have l contaminants on their surfaces which are conservatively estimated to consist of 3.4 curies of activity which corresponds to 48 A2 values.
Uniform distribution of this activity results in contamination levels of 3 l 2 uCi/cm . These contaminants will be grouted onto their surfaces and j enclosed within the reactor vessel. l 1
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l RVAttach doc 10:07 AM 6/24N8 Page 6
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ a
ATTACHhfENT I EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR !
THE SHIPMENT OF THE SAXTON REACTOR VESSEL
(
The exterior surfaces of the reactor vessel contain relatively low levels of surface contaminants. It is conservatively assumed that both the removable and fixed contamination would be available for release under the leach testing requirements. Based on swipe sample data, the level of exterior remevable surface contaminants is estimated at 21.2 millicuries.
Based on the radionuclides distribution of these contaminants, the total amount corresponds to 7.7E-03 of an A2 quantity. This estimate is conservative since it assumed the worst case swipe sample data obtained 2
to date, 120,000 DPM/100 cm , is distributed over the entire exterior surface of the vessel and insulation can. These exterior surfaces of the vessel will be encapsulated and surrounded by cement grout within the container. Therefore, it is unlikely that all the exterior surface cont.aminants could be available for release. However, even if all surface contaminants are released, they are well below the 0.1 A2 limit prescribed in 49 CFR 173.403 and 49 CFR 173.468.
The leaching requirements for the reactor vessel internal surfaces as required by Sec 49CFR 173.468:
The basis for this request is the shipping configuration which provides that the vessel interior surfaces will be grouted with all vessel penetrations completely enclosed. Thus, under normal conditions of transport, there is no credible pathway which could lead to contact i between internals surfaces and water leading to leaching of these contaminants.
l' 49 CFR 107.105(c)(4): The current project schedule requires shipment of the reactor vessel from the Saxton facility site on or about October 1,1998. Land transport to the railroad siding and rail transport to the ultimate disposal facility should be accomplished within 120 to 180 days. In order to accommodate unforeseen delays in our schedule, we ask that the exemption for the reactor vessel be applicable for 2 years from the date of approval .
'49 CFR 107.105(c)(6): The Applicant is not requesting emergency processing under Sec.
% ,, 107.117.
49CFR10'/.105(c)(7): Identification and description of hazardous material:
RVAttach. doc 10.o7 AM 6/24/98 Page 7
ATTACIBfENT 1 EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPMENT OF THE SAXTON REACTOR VESSEL A detailed description of the characteristics of the reactor vessel and activated internal components is provided in report WMG- 9801-7025 entitled "Saxton Reactor Pressure Vessel and Internals Final Characterization" which is included as Attachment 5 to this exemption request.
49 CFR 107.105(c)(8): An exemption is requested for the following shipment:
The Saxton reactor vessel approved as LSA-III material within a transportation system which is a non-specification package transported in accordance with a transportation plan and emergency response plan A detailed description of the transportation system, including packaging, is provided in Attachment 2.
49 CFR 107.105(c)(9): Documentation for quality assurance controls, package design, manufacture, performance test criteria, in-service performance and service life limitations for the proposed non-specification package is provided in Attachment 2.
49 CFR l07.l05(d) Just$ cation ofexemptionproposal 49 CFR 107.105(d)(1): A description of relevant shipping and incident experience follows:
l The Shippingport reactor vessel was successfully shipped to Hanford, l Washington under DOE regulations. The Yankee Rowe reactor vessel was successfully shipped to Barnwell, South Carolina under h1C regulations (10CFR71). These vessels are both larger and heavier than the Applicant's reactor vessel.
Steam generators have also been successfully transported by land and water in the past from Salem Trojan, Millstone and Yankee Rowe. The Salem, Portland and Millstone steam generators are larger in size and weight than the Applicant's reactor vessel.
49 CFR 107.105(d)(2): The Applicant is not aware of any increase in risk to safety or property that would result from issuing the requested exemptions.
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RVAttach doc 10.07 AM 6/24/98 Page 8
ATTACHMENT 1 lu f j
EXEMPTION REQUEST FROM THE PACKAGING REQUIREMENTS OF 49 CFR 173 FOR THE SHIPMENT OF THE SAXTON REACTOR VESSEL 49 CFR 107105(d)(3): Either - (Note: 49 CFR 107.105(d) allows justification by complying i
with section 105(d)(3)(i) 9.r 105(d)(3)(ii). GPU Nuclear will use 105(d)(3)(i) to demonstrate compliance.
49 CFR 107.105(d)(3)(i): The applicant has designed the transportation system in accordance with all the general design requirements specified in 49 CFR 173.410 as well as the testing requirements 49 CFR 173.465(c) and (d) within the
! limitations of the transportation plan.
The reactor vessel is fully enclosed and grouted inside a cylindrical container. This transportation system was evaluated to confirm the capability to safely withstand a free horizontal drop of the package from a height of one foot in accordance with Table 12 of Sec.173.465 onto a flat non-yielding surface without loss of containment. Horizontal drop scenarios were considered at angles up to 30* from horizontal in these evaluations which are conservative relative to the limitations of the transportation plan. The lifting trunnions and lug attached to the container will be removed prior to transport. There are no other attachments or protrusions on the container except those container stmetural components such as top and bottom cover plates. and the flange / stiffeners assembly nt the mid-plane of the container.
Compliance with the testing requirements specified in 49 CFR 173.465 for the one foot horizontal drop calculations is demonstrated in accordance with 49 CFR 173.461(a)(4). A copy of these calculations are provided as pan of the transportation system description in Attachment l 6.
l Due to the physical configuration of the Saxton reactor pressure vessel and the intact vessel internal components within the transportation l system, this exemption request does not pose increased risk to the public
- health and safety since there is no credible scenario under normal i transport conditions resulting in direct exposure to the internal components. ;
t 49 CFR 107.105(d)(3)(ii): This section is not applicable to this exemption request.
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l RVAttach. doc 10:07 AM 6/24/98 Page 9
I Attachment 2 1
Saxton Nuclear Experimental Corporation
- Facility Reactor Vessel Shipment Transportation System Description J .
I
l REACTOR VESSEL l
TRANSPORTATION SYSTEM I
i
ATTACHMENT 2 REACTOR VESSEL TRANSPORTATION SYSTEM The Saxton reactor vessel shipment will be transported from the Saxton Nuclear Facility in accordance with the following transportation system:
REACTOR VESSEL CONTAINER The Saxton reactor vessel will be placed into a shipping container meeting the requirements of an IP-2 package. The container will be constructed of ASTM-A-36 steel with a minimum thickness of two inches and a maximum thickness of four inches. The reactor vessel shipping container is shown in figures 1,2 and 3.
The reactor vessel shipping container is considered an Industrial Package Type 2 (IP-2)
The reactor vessel container will be designed to meet the general design requirements in 49 CFR 173.410, and IP-2 specific requirements in 49 CFR 173.465 including design to sustain a one foot drop in the horizontal position.
The container will be fabricated in accordance with an owner approved Quality Control Inspection Plan. The welds of the shipping container will be made in accordance with AWS.DI.1, and the welding will be performed by ASME/AWS qualified welders. All welds will be visually inspected, and magnetic panicle (MP) or liquid penetrant (PT) tests will be performed on designated container welds. Examinations will be performed by certified inspectors.
The reactor vessel shipping container is a single use container, and will not be reused for any other shipments. Inspections of the container will be performed in accordance with the quality assurance plan, and as required to meet IP-2 requirements. Reinspection will not be required because the container will not be reused.
The assembled reactor vessel shipping container package will measure 21 feet 6 inches long and have a maximum diameter of 9 feet 4 inches. The package will have a total approximate weight of 213,000 pounds.
All reactor vessel nozzle openings will be seal-welded closed, and the reactor vessel internal volume will be grouted before placement into the shipping container.
The void area between the container and the exterior of the reactor vessel will be grouted, thus providing a solid monolithic stnacture.
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ATTACIBfENT 2 l~ REACTOR VESSEL TRANSPORTATION SYSTEM SHIPPING CRADLE '
The reactor vessel container will be down-ended into a shipping cradie. The cradle will consist of two . saddles 24 inches in width and be connected by two 14 X 120 wide flange beams Four wire ropes on each saddle of 1.25 inches diameter will hold the container to the shipping skid. End blocking beams will be attached to the shipping skid to restrain the container. The sh;pping cradle is a single use design and will not be used for other l
shipments The shipping cradle is shown in figures 4, 5. 6. 7, 8. and 9.
l TIE DOWN The reactor vessel shipping container and skid assembly will be attached to the trailer transporter by means of turnbuckles and anchor brackets that will be attached to the designated trailer transporter load points. The tie down system will be designed to meet the requirements of ANSI-N-14.2 (draft). The reactor vessel container tie down system for the trailer transporter is shown on figures 4, 5, and 6.
The railcar tie down system will consist of hold down clips that will be welded to the railcar deck and that will attach the container and cradle to the railcar. End blocking clips will be welded to the railcar deck at each end of the cradle to restrain the container, and l cradle assembly. The railcar tie down system will be designed to meet the requirements of j AAR-Section 1 open top rules. The rail car tie down system is shown on figures 7. 8. and
! 9 l
The tie down system for the 1/2 mile transport from the offloading facility to the Barnwell Disposal facility will be the same configuration that will be used for the tie down system shown in figures 4, and 5. The design of this tie down system will be based on the 49 CFR 393.102(b) requirement of 1/2 times the weight of the components using the aggregate l working load limit of the tiedown assemblies OVER THE ROAD TRANSPORI The reactor vessel shipping container and skid assembly will be transported from the Saxton site to the rail siding at Huntingdon, PA. by a heavy-duty transporter with 16 axles and eight tires per axle. The reactor vessel shipping container and skid assembly will l weigh approximately 240,000 pounds, and the trailer about 130,000 pounds for a total l weight of approximately 370,000 pounds. The over the road transporter is shown on figures 2 and 3.
The route from Saxton to the Huntingdon, PA rail-siding is approximately 28 miles. The prime mover will be a three-axle heavy-duty tractor. An additional heavy-duty tractor will ;
accompany the transporter as a back-up prime mover. !
t f
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ATTACHhENT 2 REACTOR VESSEL TRANSPORTATION SYSTEM O
_I_R,ANSFER TO RAILCARS The Rail-siding will be prepared to receive the railcar and transponer. The railcar will be cribbed to provide stability for load transfer. The transfer of the Reactor Vessel Package will be accomplished by one of the following methods.
The first method will require that timber cribbing be placed between the transporter and the railcar. Rolling-plates will be placed on the railcar deck and timber cribbing. The skid and Reactor Vessel Container will be jacked and rolled onto the railcar deck.
The second method for transfer of the Reactor Vessel Package to the railcar will reduce the transfer time and assist in keeping exposures ALARA. In this method, the rail car will be positioned next to a suitable (-275 ton) mobile crane and the Reactor Vessel Package transporter will be positioned on the opposite side from the railcar. The crane will then pick the Reactor Vessel Package from the transporter using slings and a lifting beam and place Reactor Vessel Package on its railcar. The lift will be made so as to comply with drop analysis criteria.
RAll TRANSPORT The reactor vessel package and skid assembly will be transported from the Huntingdon, PA Rail-siding to the Disposal Facility at Barnwell, SC by Conrail and CSX. The proposed rail route will be from Huntingdon, PA to Hagerstown, MD; Washington.DC; Richmond,VA; Rocky Mount,NC; Florence, SC; Charleston,SC; Yemassee,SC; Robbins,SC; Dunbarton,SC; ( Barnwell, SC).
The reactor vessel package will be transponed on a heavy-duty QTTX flat railcar with a capacity of 390,000 lbs. The total railcar and package height will be 15 feet 2 inches and will have a total width of 10 feet 8 inches The rail car and reactor vessel container is shown on figures 7,8, and 9.
A dedicated train with escorts / riders will be provided for the transport of the reactor vessel package. Travel times and speed will be established by the governing railroad, and DOT regulations. Scheduled stop-overs, and shift change locations will be identified in the J transponation plan and supporting procedures.
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ATTACHMENT 2 REACTOR VESSEL TRANSPORTATION SYSTEM
)
RADIATION PROTECTION The radiation protection program for the transport of the reactor vessel package will mee the requirements of 49CFR174. The radiation protection program will be controlled and t implemented by the owner and licensee of the shipment, and will meet all Railroad and DOT regulations. Two radiation protection personnel will accompany the shipment from the Saxton Site along the over the road, and rail transpon routes to the final destination at the Disposal Site in Barnwell, SC.
The radiation protection personnel will conduct all required surveys and inspections alo the route, and provide direction on all radiologicalissues.
SECLRITY The security program for the transport of the reactor vessel package will be covered under the owner's security program. This includes the transport of the reactor vessel package from the Saxton project site to/and including the work at the rail siding.
Security will be provided by the responsible Railroad Company as needed along the rail route, and will be in accordance with the railroad security program. Security at the Disposal Facility will be in accordance with the Disposal Site Procedures and security coverage will be provided by the Disposal Site contractor.
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f Attachment 3 Saxton Nuclear Experimental Corporation
(
Facility Reactor Vessel Shipment Transportation Plan i
i l
i
(
t The Transportation Plan (TP) for the SNEC facility Large Component Removal Project l (SNEC LCRP) is still in the development stage. An overview of the Plan is provided at this time with the full Plan to follow as soon as it is available.
' The following parties will approve the SNEC LCRP TP:
. GPU Nuclear- the SNEC Facility co-licensee
. Raytheon Nuclear- the SNEC LCRP prime contractor F. W. Hake, Inc. - the SNEC LCRP heavy lift and road transport contractor
. Conrail and CSX - the rail shipment companies
. Chem-Nuclear Systems Inc. - the disposal contractor The table of contents and a brief discussion of each section follows.
, t
l' l
l 1-i t
l SAXTON LARGE COMPONENT REMOVAL PROJECT
! TRANSPORTATION PLAN l
l t
l I
1 i
[
Prepared By: Signature:
Date:
Reviewed By: Signature:
Date:
j Approved By: Signature:
Date:
l GPUN Approval: Signcture:
- Date
1 Document No: Rev.
Date:
l l
l l'
i 1
r 9
l L____-_-_-___--_-___--_--- --
t
'l .
i TABLE OF CONTENTS 1.0 SCOPE.. .
'2.0 ' REFERENCES.... . . . . . . . . . .
I j 3.0 '
' ORGANIZATION & RESPONSIBILITIES .. . .. . , ,
4.0 SPECIAL INSTRUCTIONS / REQUIREMENTS . . . . , . ,.. . .
5.0 PACKAGE DESCRIPTION . .. .
6.0 ROUTE DESCRIPTION. . .. . .. . . . . .. . .. .. . . . ,
7.0 PREPARATION FOR TRANSPORT .. . . . . . .. .. . .
8.0 EQUIPMENT DESCRIPTION.. .. . .. .. . . . . . . . .. . ..
9.0- ' IN TRANSIT PROCEDURES. ... . . . .. . . . . .. . . . . . . . . . . . . . . . .
10.0 SAFETY........... ., , , , . , , , , , , , , , , , , , , , , , , , , ,
11.0 . RADIATION PROTECTION ... . . ... . . .. . .. . . . . . . . . . . . . . . . . ... . . . .
12.0 SHIPMENT HAZARDS . . . . . . .. . . . . . . .... .. . . . . . . . . . . . . .
"13.0 MARKING / PLACARDING .. . ... . . . .. . .. . . . . . . . . . . . . . . . . . . . . .
14.0- TRAINING . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .
15.0 ATTACHMENTS.... . . . . . . . . . . . .. . . . . . . . . . . . . ... .. . . . . . . . . . . .
Attachment 1 - Prerequisites Checklist .. .. . . . . . . . . . . . . . . . . . . .
Attachment 2 - Pre-transport Checklist Road Transport . ... . . , . . . . ..
Attachment 3 - Pre-transport Checklist Rail Transport.. . .. . . ... . . . ..
' Attachment 4 - Procedure For Transfer To Rail Car. . . . . . . .. . . .. .
Attachment 5 - Route Maps
L The scope of each section is as follows:
i 1.0 SCOPE The " SCOPE" section will dictate the applicability of the TP and bound its coverage. It will ensure coverage of all applicable phases of the SNEC LCRP including package preparation, loading and tiedown, transportation requirements, precautions and limitations, radiation protection j
requirements, safety, training, routing and provide checklists.
p
2.0 REFERENCES
Will provide a list of applicable references for the SNEC LCRP.
3.0 l ORGANIZATION & RESPONSIBILITIES Will provide an organization chart and will dictate the lines of authority and responsibility for all major phases of the SNEC LCRP.
4.0 SPECIAL INSTRUCTIONS / REQUIREMENTS This section will spell out the particular operating and performance criteria for s~pecialized equipment and practices for the SNEC LCRP. Examples include the operating restrictions on the heavy haul truck trailers, surveillance requirements on the package tiedowns, etc. In addition, i
special instructions, precautions and prerequisites will be covered in this section.
5.0 PACKAGE DESCRIPTION This section will contain a description of the pressurizer, steam generator
, and reactor vessel shipping container packages similar to that in the L
" Transportation System" descriptions. This section will clearly delineate
. what the packages consist of.
I l
6.0 ROUTE DESCRIPTION This section will describe in detail the route, including highway and rail. It will not describe the route features but serve as a verbal
- road map" 7.0 PREPARATION FOR TRANSPORT This section will describe the preparation of the three large components such as the applied coatings, the covering of openings, radiological surveys, removal of lifting attachments, etc. The orientation of the components and the tiedown requirements will also be covered.
8.0 EQUIPMENT DESCRITION.
This section will provide a detailed description of the tiedown, rigging, transfer and hauling equipment including highway and rail.
9.0 IN-TRANSIT PROCEDURES in this section, the procedures required for safe transport of the packages will be covered. This willinclude routine checks of the tiedown system, orientation of the packages to verify no shifting has occurred, radiation surveys, ins'pection of protective coatings, etc.
10.0 SAFETY Industrial safety precautions unique to this shipment will be covered. The radiological protection requirements are covered by the SNEC Radiation Protection Plan and the LCRP Health and Safety Plan.
11.0 RADIATION PROTECTION i j
! 1 l This section will specify the radiological surveys, exposure monitoring of
{
the environment and personnel, radiological escort requirements and data reporting.
i I j
i t
12.0 SHIPMENT HA2ARDS In this section, the specific " hazards" of the shipment will be described including direct radiation exposure, radiological releases in the event of a package breach, and those associated with any heavy load.
13.0 MARKING / PLACARDING The specific marking and placarding requirements of all applicable regulations will be listed for the packages and transport vehicles.
14.0 TRAINING Task or position specific training requirements will be specified. Examples
-include transport drivers, welders, radiological protection personnel, escorts, and quality assurance / verification personnel.
15.0 ATTACHMENTS Attachments will be provided such as route maps, pre-transport checklists, conveyance transfer procedure, etc.
l L
P
g Attachment 4 i
Saxton Nuclear Experimental Corporation Facility Reactor Vessel Shipment Emergency Response Plan
)
u__________-__---___-- - - - - - - - - - - - - - -
f 4
l
, Hea th and Safety Plan :for the Transoort of Saxton Nuc. ear Experimental l
Corporation's Reactor Vesse.., Steam l
Generator & Pressurizer Prepared by: Date:
Reviewed by: Date:
Prepared by: Date: l l
s
~
Table ofContents
- 1. OBJECTIVE. INTRODUCTION AND BACKGROUND.. .1 Description of Packages.. .1 Route Description .2 Work Tasks .7 Pre-Transit Preparation . .7 Tie Down Inspection and Approval. .7 Shipment Prerequisites.. .7 Interchange of Load. .7 Scheduled Stops . .8
- 2. SHIPMENT HAZARDS. .9
- 3. STAFF. ORGANIZATION AND RESPONSIBILITIES.. .12
- 4. REGULATORY REQUIREMENTS.. .15
- 5. GENERAL HEALTH AND SAFETY WORK PRECAUTIONS . .16 Pre-Shipment Briefing. .16 Health and Safety Briefing.. .16 Personal Protective Clothing. 16 Respiratory Protection. .17 External Radiation Protection . 17
- 6. TASK SPECIFIC HEALTH AND SAFETY PROCEDURES. .18
- 7. EMERGENCY RESPONSE. .20 Responsibilities .20 Radiation SurveyAVipe Test inconsistency . .20 Unscheduled Stop in an Urban or Sensitive Area . .21 Transporter or Train Accident. .21 Medical Emergency .21 LIST OF TABLES.. . . . ii LIST OF FIGURES. . iii APPENDIX A (to follow once nackare dose rates are known).. . A-1 LIST OF TABLES Table Tale
. Pace 5-1 Measured Dose Rates from the Packages 17 7-1 Route Segment. Rail Company and Emergency Response Contacts 22 l I'l i
]
l, -
i LIST OF FIGURES Ficure Title Pace j I-l Reactor Pressure Vessel and Shipping Cask 4 l l-2 Pressurizer 3 1-3 Steam Generator 6 21 Dose Rates at Various Distances from the Reactor Package 10 l
2-2 Dose Pates at Various Distances from the Reactor Package 11 l 3-1 Transport Lmes of Communication
! 14 l
l f
1 ia
l..
l 1
Objective, Introduction and Background The Saxton Nu: lear Experimental Corporation (SNEC) Facility, located in Liberty Township.
near the Borough of Saxton. Pennsylvania. is currently being decommissioned. GPU Nuclear.
Inc.. (GPUN) a SNEC Facility co-licensee. is conducting the decommissioning on behalf of SNEC. Part of this process includes the packaging. transport. and disposal of radioactis ely contaminated items. equipment, soil. debris. etc. Central to the successful decommissioning of the SNEC Facility is the removal. preparation. shipment and disposal of the three large components from the nuclear steam supply system. At the SNEC Facility, these components are the pressurizer. steam generator and reactor pressure vessel. This project is referred to as the SNEC Large Component Removal Project (SNEC LCRP). The large components will be packaged and shipped via truck and rail to the low-level radioactis e waste disposal site in Barnwell. South Carolina. The shipment process is the focus of this health and safety plan (HASP).
This HASP provides information necessary to assure the safety of employees and the public during activities associated with transportation of the large component packages. The objective of this plan is to assure that all personnel and the public are exposed to minimal radiological risk in accordance with the principle of maintaining radiation exposures As Low As Reasonably Achievable (ALARA) and in conformance with Federal Regulations and guidelines. Written and appros ed procedures address activities dunng normal transport. as well as notifications and activities that need to occur during accident situations. The accident situations addressed in these procedures will represent credible accident scenarios.
Transpon activities cosered by this HASP include the shipment by truck from the SNEC Facility site to the rail siding in Huntingdon. PA. and transport by rail to the Chem-Nuclear Consolidation Facility near Barnwell. South Carolina. Contracts between GPUN. Raytheon. Inc. the LCRP prime contractor and the carriers Conrail (CR) and CSX Transportation (CSX) describe l
responsibilities of GPUN. Raytheon and the carriers for activities that must take place prior to and during mos ement of the large component packages. However. most details of these contracts are not appropnate for discussion in this HASP except where they relate to issues of health and safety. Therefore. contractual obligations that address health and safety issues has e been incorporated into this HASP.
All cognizant railroad. truck transport and other personnel im oh ed in the shipping operation will i
' receive a copy of this document. supplemented by training in its contents. All such personnel are required to return a signed copy of the HASP review signature sheet to the SNEC LCRP Project Manager before transport occurs.
Description ofPackages Figure 1 1 presents a diagram of the reactor pressure s essel in its shipping caimister. The reactor pressure s essel is a 5-inch thick cylinder with a 5.T outside diameter and is composed of multilayered sheets of carbon steel. It has a bottom hemispherical head and an upper section Pagei Heahh & safen nan ror The Rail Trangort of SNECi Resetor venet. henunzer. A Steam Generator I
w_ _ _ _ _ _ _ _ _ _
consisting of a thickened flange ring nozzle ring. and 2 lifting lugs. The reactor pressure vessel is filled with low density cellular concrete (grout). The grout also serves to fix any radioactive contamination on the intemal surfaces. to ensure that there is no free standing water in its interior. '
and to stabilize intemal components. Grout was also placed in the annulus between the exterior of the reactor pressure vessel and the shipping cannister. Without grout. the un-packaged reactor pressure vessel weighs approximately 67 tons.
The reactor pressure vessel single-use shipping cannister is a two-inch thick steel cylinder with an integral flat circular bottom and flat cover. The maximum dimensions of the shipping package are 21.5 feet long by 9 feet 4 inches wide. The entire cannister. except for the top cover and miscellaneous penetration covers. is a single. welded unit. He top cover and miscellaneous penetration covers were welded on after the reactor pressure vessel was placed inside. The package will be shipped as IP-2 Package under USDOT 49 CFR 173. The reactor vessel shipping cannister gross weight is approximately 213.000 pounds or 106.5 tons. The total weight of the reactor vessel shipping cannister and cradle is approximately 120 tons.
Figure 1-2 presents a diagram of the pressurizer. The pressurizer is an elongated cylinder measuring 3 feet in diameter by 17 feet long. It is composed of 3-inch thick carbon steel and weighs approximately 12.5 tons. He nozzles on the pressurizer will be removed and the inside will be filled with grout. The penetrations will be welded. and the extemal surfaces will be painted to fix any external contamination that may be present. The pressurizer will sene as its own package and will be shipped under USDOT 49 CFR 173. per NRC Generic Letter 96-07.
The final weight of the pressurizer with grout is 28.000 pounds.
Figure 1-3 presents a diagram of the steam generator. The steam generator is composed of a carbon steel shell approximately 3-inches thick and is 20 feet in length and 4 feet in diameter.
The empty steam generator weighs approximately 27 tons, and will be prepared for shipment in the same manner. i.e.. grouting and welding of penetrations. The steam generator is approximately 57.000 pounds when grouted and will also serve as its own shipping package. The steam generator will also be shipped under USDOT 49 CFR 173. per Generic Letter 96-07.
Infonnauon regardmg the reactor pressure vessel shippmg cask and the contents w ere obtained from the Reactor Vessel Contamer Design Details Dwg No 775Rl-021-E-13-ool. Feb 26.1998 Route Description The movement of the packages on the roadway will be performed using a dedicated heavy hauler.
The heavy-duty transporter for the reactor pressure vessel will possess 16 axles and will has e a total length of 130 feet. The heavy-duty tractor-trailer for the pressurizer and steam generator will have 8 axles and have a total length of approximately 60 feet. >
The mosement of the reactor sessel package will proceed at a speed ofless than or equal to 5 !
miles per hour, and will be escorted by public safety officials. The transponders will proceed from )
the SNEC Facility site along Pa. State Route 913 east through the Borough of Saxton. The j transporters will then travel on Pa. State Route 26 north to Huntingdon. PA to Ridge Road where the rail siding is located. The total length of the road transport is approximately 28 miles. There is only one long-span bridge along the route. there are no steep grades. and no sharp comers.
l There will be one stop overnight along the route at a location yet to be determined.
)
I Rail movement of the packages will be performed using a special dedicated train along a predetermined route. Maximum speed will be limited to that determined by each of the rail ,
companies. The route will be as follows: from Huntingdon. PA to Hagerstown. MD. I Page 2 Heahh k Safety Plan For The Rail Trarnport Of sNEC'n Reactor \ euel. Preuunzer. & Steam Generator E _ _ _ _ _ _ - _ _ _ - - - _ - - - - - - - - - - - - 1
Washington. DC: Richmond. VA: Rockey Mount. NC: Florence. SC: Charleston. SC; Yemassee.
SC: Robbins. SC: Dunbarton. SC (Barnwell. SC).
The rail carriers will be Conrail from the Huntingdon. PA to Hagerstown MD: and CSX from Hagerstown. MD to Dunbarton. SC. The route the packages will follow needs to be known for emergency response purposes as discussed in Chapter 7.
l Page 3 Heahh & safety Plan For The P61 Transport Of SNECi Reactor Veuel. Preuurner, k Steam (ienerator j
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t Work Tasks '
Pre-transit Preparation l.
- An esaluation of the route between the SNEC Facility and the rail siding in Huntingdon. PA was conducted and included structures such as culverts and bridges, and also reviewed slope stability along the public roadway. Any repairs or upgrades on the road or roadway structures will be performed prior to package transport. All evaluations pertaining to the roadway haul route will be reviewed by the appropriate regulatory agencies prior to transport.
Tie Down inspection andApproval The tie down systems used for roadway transport of the packages conforms to ANSI N14.2 (draft). Tie down systems used for rail transport of the packages conforms to the requirements of AAR - Section 1 Open Top Rules. Tie down of the packages will be inspected by qualified l GPUN personnel prior to transport.
l-Shipment Prerequisites i Besides other contractual obligations prior to shipment. GPUN will ensure that copies of the following documents are delivered to the Corporate Attorney of the rail carriers prior to the commencement of the operation:
A copy of t he USDOT Exemption Approval for the LCRP.
A copy of the Bill of Lading and Shipping Manifest.
. A copy of this " Health and Safety Plan" l A copy of the SNEC LCRP offsite transportation procedure A copy of all necessary insurance certificates.
Notifications will be made prior to shipment of the package. A notification package will be sent to the Governor's Office for each state the package will pass through. Additional notifications will be made to the appropriate state emergency response agencies. Refer to the Public Information Plan for further details regarding notifications.
Stopover along RoadRoute --
' Due to the length of the transport over road from the SNEC facility to the rail siding and the speed of travel, the reactor vessel package transporter will be required to stop for the night. It will stop at a location yet to be determined. When the location is known, a 24-hour watchman will be posted. and the area cordoned off with caution tape such that unauthorized individuals cannot come within thiny feet of the transporters. A GPUN radiological control specialist will also be posted at the location until the transport resumes.
Interchange ofLoads The packages will arrive at the Huntingdon rail siding on hydraulic platform trailers. At this location. the packages will be transferred to the rail cars in accordance with an approved
- l. procedure.
1' I
Page 7 - . Heahh a sarety Plan For The Rail Tramport of SNEc's Reactor venet. Pressunzer k steam Generator
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- - - _ _ - _ - - - - - - _ ---- d
I The GPUN radiological controls specialsts escorting the load will perform radiation ~
measurements prior to. and after transfer of the packages from the trucks to the rail cars to ensure that dose levels are in compliance with 49 CFR 173 399. These dose limits are. 1 the dose rate at any accessible surface of the container is less than 200 mrem /hr.
the dose rate at two meters from the accessible outer edges of the package is less than 10 mrem /hr, and -
removable surface contamination is less than 1.000 dpm/100 sq.cm. for beta / gamma emitters.
and less than 20 dpm/100 sq.cm. for alpha eTitters (GPUN limit).
The SNEC LCRP Project Manager and Conrail Project Manager will meet to agree that:
loading occurred as specified in the procedure.
' all required health and safety training and paperwork has been completed.
notifications have been made.
the tie-down system has been inspected and approved and the clearance dimensions have been inspected and approved, if required.
After both agree that these issues have been successfully completed. the shipment may proceed.
During the railjourney from Huntingdon. PA to Dunbarton. SC the train crew (s) shall have full responsibility for operating the train in accordance with the established rail company procedures.
Communication consistent with safe operation of the train shall be maintained with the respective Dispatch Control. The radiological content of the packages will not impose any changes in the normal routine procedures for operating the train: however, the weight and size of the load may require the use of special procedures.
During the railjoumey the load will be accompanied by two GPUN radiological controls specialists. In the event of any incident involving delay or damage to the train. or of failure'or suspect failure of the containment of the packages, then the directions of the GPUN radiological controls specialists shall be followed in all actions. In circumstances following an incident when the guidance of GPUN radiological controls specialist is not available. any uninjured rail employee at the scene should follow the emergency response procedure that applies to the event as described in Chapter 7.0. If possible in emergency conditions. r#.1 staff should shut down the train and evacuate the cab to an upwind location.
Scheduled Stops The first scheduled stop will be in Hagerstown. MD. At this point. the GPUN radiological controls specialist will conduct the radiological measurements specified in the procedure. Results of the survey will be recorded. and if they are consistent with the measurements conducted at the s
l tart of the shipment. a completed survey form will be made available to the Conrail Project Manager. Ifmeasurements indicate a significantly greater exposure rate than obsen ed at the start of the shipment. or the presence of radioactive contamination. then the procedure in Chapter 7 of this HASP must be followed.
Other stops will occur along the way. Whenever stops are made. the GPUN radiological controls specialist may perform radiation surveys and wipe tests to confirm the radiological status of the load as deemed necessary, and the results of any suneys shall be documented. All such sun e.s activities shall conform to the requirements of the SNEC Facility Radiation protection Plan.
Page8 Heahh & Safety Plan For The Rad Transport Of SNEC's Reactor venet. Prenurner, & Sicam Generator
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1
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2 Shipment Hazards There are a variety of hazards involved with the shipment of the packages. For the pu this health and safety plan. only radiological hazards are discussed. During normal tran conditions. the radiation levels in the vicinity of the packages are the primary hazard to resi along the road and rail route, and to rail company employ ees and the train passengers.
the level of radiation is within regulatory limits and will not pose a hazard to the health of residents. or rail company employees greater than other day- to-day hazards typically encountered.
The dose rate from each package is designed to be less than 200 mrem per hour on contac less than 10 mrem per hour at a distance of two meters from any side. Smce the level of r reduces as one moves further from the package, all personnel should maintain a reasonable distance i.e., at least 20 feet when not performing work related activities. Personnel w closer than 20 feet in order to perform tasks should follow the As Low As Reasonably Ach (ALARA) principle. The ALARA principle is used throughout the nuclear industry and other organizations that use radioactive materials and sirnply means that one should minimize his or her exposure to radiation as much as practical. This can be accomplished by:
minimizing the time spent close to the packages, maximizing the distance between the individual and the packages. and taking advantage of physical barriers betw een the individual and the packages that will act as shielding.
All project and rail personnel involved with the actual shipment will receive training and will be issued appropriate dosimetry as required by the SNEC Facility Radiation Safety Officer (RSO record any radiation dose received. Training. and issuance and collection of dosimetiv will be performed by the GPUN radiological controls specialists aboard the train.
In the es ent oflengthy (>24 hrs) unscheduled stops. the GPUN radiological controls speciali and rail employ ees must ensure that all unauthorized individuals maintain a distance of at least 30 feet from the rail cars carrying the packages. It may be necessary to obtain security from the railroad or local police department depending upon location and length of the delay.
Diagrams of the calculated dose rates at various distances from the reactor sessel package a presented in Figures 2-1 and 2-2. The purpose of the diagrams is to enable the GPUN radiological controls specialists to make decisions regarding potential doses to individuals under normal transport conditions. The steam generator and pressurizer package dose rates are much lower than the reactor vessel package and are not presented in this plan.
Page 9 Heahh k Sarety Plan for 1 le Rail Trampon Or sNECi Reactor veuel Prenuracr. k Steam Generator
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3 Staff; Organization andResponsibilities Figure 3-1 lists SNEC LCRP organizational structure for shipment of the packages and the lines of authority and responsibilities between GPUN and the rail companies. Health and safety support for the shipment of the packages will be coordinated and administered by GPUN. GPUN will provide administrative and technical support activities and services such as obtaining the proper permits and certificates making the necessary notifications of the shipment. performing health and safety briefings, conducting radiological measurements, providing emergency response resources. providing radiation dosimetry. and providing GPUN radiological controls specialists to accompany the packages on theirjourney to Dunbarton. SC.
Overall responsibility for health and safety activities ultimately lies with each individual.
However it will be the responsibility cf tN SNEC LCRP and Rail Company Project Managers t ensure that all activities are conducted.i:. accordance with this Health and Safety Plan. The GPUN radiological controls specialists Accompanying the shipment will be responsible for implementation of this health and safety plan and act as the GPUN representative solely respons ble for all radiological activities. The Rail Company Project Managers will be responsible for ensuring that their respective personnel involved with the shipment are aware of the contents of this Health and Safety Plan. and that they follow its contents.
A copy of this HASP will be made available to all cognizant personnel invohed in the shipment.
They will also be informed ofits contents before work is initiated through a pre-shipment briefing. The roles and responsibilities of key individuals include the following:
.SNEC LCRP Project Manager / Representative ensure required paperwork is completed and given to the Rail Company ensure appropriate notifications of state and local government agencies were made oversee truck loading operations at the SNEC Facility oversee transfer operations in Huntingdon. PA ccnduct a pre shipment briefing be available during the shipment to address emergencies or events attend the train in Dunbarton. South Carolina to oversee off-loading operations GPUN Radiological Control Specialists /RSO provide health and safety briefings to train crews -
issue and collect TLDs from the train crew perform radiation surveys / wipe tests of shipment and sign survey forms accompany the shipment to final destination provide emergency response guidance to train crew and response teams Page 12 Heahh & sarety Plan For The Rad Transport Of sNEC's Reactor venel Prenurizer, k Steam Generator o
1 SNEC Emergency Response Team '
be available to respond to any emergency during transpon Raytheon LCRP Team i
coordinates the LCRP subcontractors and executes the LCRP Plan Rail Company Chief Engineer review design calculations and drawings to ensure that the rail cars with the packages meet the requirements of the Rai: Companies for safe transpon.
Rail Company Project Representative observe loading at Huntingdon. PA observe SNEC health physics specialist conducting radiat;on measurements obtain copies of all radiation measurements ensure rail company personnel wear dosimetry inspect placarding test radio communication with dispatch control ensure that the known conditions of the track and signal / control equipment between destinations have no known faults that could compromise the safe completion of thejoumey.
check on current and anticipated weather conditions with dispatch control to determine whether deferring or delaying departure is warranted.
ensure that rail company personnel understand their responsibilities in routine operation or in the event of an emergency.
meet the train at transfer points. and obtain copies of radiation measurements conducted while being pulled by their train and crew.
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l Page 13 Heanh a safety Plan For The Rail Transport Of sNEC3 Reactor Venet. Pressumer. A Steam Generator J
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REGULATORYREQUIREMENTS Chapter 5 and 6 of this Health and Safety Plan address the requirements of 10 CFR 20 wit respect to protecting the health and safety of the public and the workers involved in the sh Chapter 7 addresses the appropriate notifications to the various states through w will pass. and what steps personnel should take in the event of a transportation incident. All placarding and manifesting of the shipment will be performed according to Department of Transportation regulations.
An information package has been prepared and was submitted to the Office of the Govern states through which the package will pass. The shipment will pass through the followi Pennsylvania Maryland
. Virginia North Carolina South Carolina in addition. NRC Regions I and 11 will be notified of the shipment. In the event of an incid the NRC. other Federal agencies and appropriate State agencies would be notified For furth information on the requirements to respond to a transportation accident involving radioactiv materials. refer to Chapter 7. Emergency Response.
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! Page 15 Heahh k safety Plan ror The Rail Transport Of sNECs Reactor venei. Preuurner. k Steam (i
r-5 GENERAL HEALTHAND SAFETY WORK PRECA UTIONS Under normal transport conditions, the only hazard presented by the packages are from radiation as described in Chapter 2. General health and safety work precautions ar ensure that workers involved in the road transport. and to rail employees' exposure to radia maintained As Low As Reasonably Achievable (ALARA). It is the Rail Company Proje Representative's responsibility to ensure that all of their employees are aware of and fo suggested work precautions. The work precautions are described below.
Pre-Shipment Briefing Prior to initiating the loading and transport of the packages to Huntingdon. PA. a pre bnefing will be held by the SNEC LCRP Project Manager to allinvolved personne will include a description of the operation that will take place, and the communication and cooperation that will be necessary among all individuals involved. At this time. the SNEC LCRP Project Manager will ensure that all individuals are aware of and understand their roles and responsibilities, and the radiological conditions associated with the shipment.
Health andSafety Briefings The GPUN radiologiul controls specialists will provide health and safety briefings to the tru transporter and train crew prior to each shipment. Appendix A will incorporate the outline of the briefmg that will be provided. At each switch over point when the package is transferred from one rail carrier to another. the GPUN radiological controls specialists will provide the same briefing to the new rail carrier personnel.
Personal Protective Clothing I No personal protective clothing will be necessary for personnel involved with this shipmen purpose of protective clothing is to prevent radioactive contamination from coming into contact with an individual's skin. The packages are designed such that no removable radioactis e {
contamination is present. In addition. the GPUN radiological controls specialists will perio monitor the external surfaces of the package to ensure the absence of radioactis e contamination Page 16 Hohh a samy nan ror The kad Tramport of SNECN e Rea.: tor v ud hewunzer. A Nieam Geaerator
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Respiratory Protection .
Respiratory protection will not be necessary. Respiratory protection is used when there is a l
potential for individuals to inhale radioactive particulate. Under normal transport conditions.
there is no potential to inhale radioactive particulate.
i External Radiation Protection Under normal transport conditions the packages will emit measurable radiation levels, The highest measured dose rates from the sides of the packages are presented in Table 5-1.
Table 5-1: Measured Dose Rates from the RPV Package Measured Surface Dose Measured Dose Rate at 2
' Rate Meters (mR/hr) (mR/hr)
Right Side of RPV Package
- Left Side of RPV Package *-
Top of RPV Package Bottom of RPV Package *
- Note: these dose rates will not be known until the reactor vessel package is prepared, this information will be completed following radiological surveys after package preparation.
Similar tables for the steam generator and pressurizer wil1 be added.
The maximum allowable dose rate on contact with the surface of the package is 200 mrem /hr and
/
10 mrent hr at two meters. Actual dose rates will be measured prior to initiating transportation and at switch our points along the transportation route.
All personnel involved with loading and transportation will be issued appropriate dosimetry as directed by the SNEC Facility RSO. Dosimetry will be processed after the transportation event is complete. and will give an account of the individual's radiation exposure. if any. Dosimetry results will be forwarded to the respective companies for disclosure to their employees.
L When the stopover along the road route occurs, a watchman will be posted and the area cordoned i
off with caution tape such that unauthorized individuals cannot come within thirty feet of the transporters. A GPUN radiological controls specialist will also be posted at the location until the transport resumes.
i Page 17 Heahh & Safety Plan For The Rail Transpon Of sNEC's Reactor venet. Preuunzer, & steam Generator f
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l TASKSPECIFICHEALTH AND SAFETY PROCEDURES SNEC Facility procedures has been developed for the conduct of radiological activ among other activities. includes the following:
Radiation Survey Equipment Use Wipe Testing Radiation Surveys These procedures apply to all radioactive material shipping activities and will be in p following: loading of the packages at SNEC facility onto the transporters; transferrin packages to the rail cars at the rail siding in Huntingdon. PA: switch over between rail and, unloading at Dunbarton. SC.
After loading the packages onto the transporters at the SNEC Facility site. GPUN controls specialists will conduct external radiation suneys in the transporter cabs and of the packages. perform wipe testing of the packages. and ensure the necessary radiological papenvork is completed. The will be made available to the transponer personnel and other mterested panies. The use of the survey equipment. radiation surveys. and wipe testing w conducted in accordance with approved procedures.
The GPUN radiological control specialists will provide the transporter drivers and the Conra Representative with a briefmg and appropriate dosimetry for those individuals involved with inspecting and transporting the packages. The GPUN radiological controls specialists will perform a health and safety briefmg for the rail carrier individuals.
Upon reaching Hagerstown. MD the train crew will switch over to CSX. and the GPUN radiological controls specialists will collect the dosimetry. The GPUN radiological controls specialists will also conduct radiation sun eys and wipe tMing at this point to ensure that the integrity of the package has not changed during transpon. A copy of the results of th wipe testing will be made available to the Conrail representative for their records.
The CSX train crew will be provided appropriate dosimetry by the GPUN radiological co specialists. The GPUN radiological controls specialists will conduct the health and safety briefmg. A copy of the survey and wipe test information will be provided to the CSX representative to identify the status of the package at the time that they take responsibility .
Constant communication between the GPUN escorts / riders accompanying the package passenger car and the engineers in the locomotive must be mamtained at all times sia ponable radio or other means. In the es ent of an emergency. the passengers must be capable of communicating with the engineers to stop. and have access to local emergency response Page 18 Heahh a sareiy Plan For " Die Rail Transport Of sNEC's Reactor vessel. Prenurizer. & Steam
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providers. such as medical EMTs. - In addition. the passengers on the train will have a cellular and satellite phone to contact emergency responders and others as necessary.
' As the train progresses along the route. the GPUN escorts / riders must maintain cognizance of their approximate location for emergency response contact purposes. In addition, if the train must stop for any extended period of time. the SNEC LCRP Project Manager or representative must identify their location. and make all passengers aware of their location in the event an emerg contact must be made. The GPUN escorts / riders will utilize Global Positioning Systems or
- l. similar means to locate their position.
Upon completion of thejourney in Dunbarton. SC the GPUN radiological controls specialists will:
perform a final radiation survey and wipe test of the packages.
collect the dosimetry issued to the CSX employees, and make the data available to disposal site personnel and to the CSX representative.
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7 i EMERGENCYRESPONSE The response to an incident will depend upon the nature of the incident. No incident time the incident hastook occurred place. will depend upon the geographical location e of the tran The types ofincidents that could occur include:
radiation surveys and/or wipe tests that significantly deviate from initial sur unscheduled stop in an urban or otheruise sensitive area, and transporter or train accident.
In any event. all actions shall be made with the following priorities:
Pubbe and Environmental Safety Personnel Safety Transport Protection follow during an incident. This section was developed N14.27-1986. "Canier and Shipper Responsibilities and Emergency Response Highway Transportation Accidents '
Responsibilities i During any event, the GPUN radiological controls specialists or their designee wil SNEC LCRP emergency coordinator and will be responsible for all assign actions until the local emergency response organization takes os er authority SNEC LCRP emergency coordinator will be responsible for making the necessa detailed in Table 7-1. and will work with the local emergency responders as appr If the SNEC LCRP emergency coordinator is not capable of performing their incident. the next available GPUN escort / rider will make the necessary notifica Table 7-1. and assist the local emergency responders. The GPUN/SNEC public representative will iss6e any public notices.
\
Radiation Snrvey/H'ipe Test Inconsistency if the radiation surT ey or wipe testing identifies radiation levels that sienificanth excI measurements must: taken at the last transfer point. then the GPUN radiological controls specia
("significantly"is defined as exceeding an applicable GPUN or regulatory limit) !
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! Ensure the immediate health and safety of train crew by requiring all personnel to Page 20 Heahh k Safety Plan For The kad Transport Of sNEC's Reactor Venel Precurner, k j
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distance of at least 30 feet from the packages.
2.
Contact the SNEC LCRP Project Manager for guidance.
3.
Contact Project Manager.
the appropriate State and NRC emergency response center as directed by the Note: The SNEC V.P/ Director may act in place of the SNEC LCRP Project Manager in all cases.
Unscheduled Stop in an Urban or Sensitive Area If an unscheduled stop occurs in an urban or sensitive area. the responsibility of the GPUN radiological The specialist must: controls specialists is to ensure that the health and safety of the public is preser 1.
- 2. Decide whether surveys are necessary based on the length of the delay (greater than I ho Conduct radiation surveys and ensure the integrity of the package.
3.
Notify the SNEC LCRP Project Manager of the delay.
4.
Ensure that members of the public maintain a distance of at least 30 feet from the train if stopped >24 hrs. (Local law enforcement officials will be contacted to provide security around the train. See Table 7-1 for a listing of Law Enforcement Agencies that provide radiological emergency response in the various states the train will be passing through.)
Note: The SNEC V.P/ Director may act in place of the SNEC LCRP Project Manager in all cases.
Transporter or Train Accident in the unlikely event of a transporter or train accident. the immediate action as recomme the NRC. DOE and professional radiation protection associations is to:
- 1. Rescue victims.
- 2. Provide medical attention.
- 3. Extmguish fires.
4.
Notify the organization located in the area where the incident occurred. Table 7-1 presents the list of organizations requiring notification of a transportation incident involving radioactive materials.
5 Notify the SNEC LCRP Project Manager. The SNEC LCRP Project Manager will activate GPUN's emergency response resources.
Note:
cases.
The SNEC V.P/ Director may act in place of the SNEC LCRP Project Manager in all MedicalEmergency At least one member of the GPUN escort / rider crew must be CPR/First Aid trained first-aid kit available. To facilitate communications with local emergency medical responders, the SNEC LCRP Project Manager must:
1.
2.
Ensure that two-way communication with the locomotive engineers is maintained at all times.
Know the location of the train and the nearest town. municipality. etc. as the train mos es along the rail route.
3.
If the train is stationary for any length of time (i.e.. more than three hours) ensure that the rail police are aware of the train's location. Also. ensure that two-way communication with outside responders is in w orking order at all times. (Note: Cellular and satellite phones ma be in a blacked-cut location so they should be tested when the train is not in motion for the ]
period of time indicated.) j Page 21 Heahh & safety Plan f or The Rail Tramport OrSNEc's Reactor vessel. Pre **urizer. & Steam Generat
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