ML24358A230
| ML24358A230 | |
| Person / Time | |
|---|---|
| Site: | NS Savannah |
| Issue date: | 12/19/2024 |
| From: | Reese J Tidewater |
| To: | Office of Nuclear Material Safety and Safeguards |
| References | |
| CR-176 | |
| Download: ML24358A230 (1) | |
Text
U.S. Department of Transportation Maritime Administration Office of Ship Operations 1200 New Jersey Ave., SE Washington, DC 20590 Docket No. 50-238; License No. NS-1; N.S. SAVANNAH ENCLOSURE 1 CR-176, METALLURGICAL SAMPLES EVALUATION
U.S. Department of Transportation Maritime Administration SAVANNAH Technical Staff N.S. SAVANNAH Metallurgical Sampling Results CR-176 Prepared by:
Tidewater, Inc.
Revision 0 MARAD Accepted:
I Title Date
Prepared by:
December 19, 2024 James Reese, CHP, RRPT Date Reviewed by:
December 19, 2024 Osborne, License Compliance Manager Date Reviewed by:
December 19, 2024 oeun, Decommissioning Program Manager Date Page 2 of 17
INTRODUCTION By letter dated May 30, 2024, the Nuclear Regulatory Commission (NRC) requested additional information (RAI} to support its review of the N.S. SAVANNAH (NSS) License Termination Plan (LTP). Request 2-4(b) requested an evaluation by MARAD for neutron activation of any impacted system components that will remain in place after license termination. The evaluation is to include activation products such as carbon-14 (C-14), nickel-59 and 63 (Ni-59, Ni-63), and cobalt-60 (Co-60). The evaluation must include data from a statistically significant number of samples to supplement the characterization data provided in the License Termination Plan (LTP). Request 2-4(c) further asked MARAD to discuss how the new sample data affects radionuclides of concern (ROC), insignificant contributors, and/or surrogate ratios in its survey plans. MARAD responded to the RAls by letter dated June 27, 2024. In response to RAI 2-4(b), MARAD provided a recap of previously submitted information and stated that a work package was prepared to draw metallurgical samples from the affected components, and that the evaluation would be provided to the NRC when complete. The response to RAI 2-4(c) stated that it was too early to discuss because the sample data was not yet available. This report documents the evaluation of the sample data and draws conclusions regarding neutron activation and the ROCs.
The sampling effort was performed beginning in July 2024, during the period when NRC confirmatory surveys took place. It continued after the confirmatory survey team completed its work and left the site, until the required number of samples was drawn. Three (3) samples were drawn from each location, with one "split" provided to NRC, one sent for analysis by MARAD, and the third retained as a control.
DISCUSSION The sample plan was developed to identify a statistically significant number of locations where neutron activation might reasonably be expected considering the design and arrangement of the components in relation to the core. Previous studies of this issue were summarized in the LTP and recapped in MARAD's June 27 RAI response. The RAI response is repeated herein for context.
The results of a sampling and analysis campaign of the Reactor Pressure Vessel and Neutron Shield Tank in 2005, documented in CR-056, shows no positive activation activity in the NST outer wall. An additional sampling campaign was conducted in 2005 and documented in CR-038, in which 14 metal samples were taken from structural components 1) in close proximity to, 2) outside of the reactor pressure vessel (RPV) and 3) in direct line-of-sight of the core mid-plane. None of these metal samples showed activity above background. Six secondary containment concrete core bores were also collected. Four metal samples and two concrete cores along with steel plugs were sent to GEL labs for confirmatory analysis. The results of these samples showed no positive gamma emitting radionuclides in any metal samples, with the exception of naturally occurring potassium-40 (K-40) being positively identified in one sample.
No Tritium was detected in the concrete cores and only naturally-occurring gamma emitting radionuclides in the concrete.
In 2018, the Containment Vessel (CV) Portal was created to provide horizontal access to the CV.
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Creating the portal required removing the layers of the CV shell: the concrete biological shielding and the laminated redwood and steel collision mat. One wood, one concrete and two metal samples were collected and sent for gamma spectroscopy and Hard-To-Detect (HTD) radionuclide analysis. The results of these samples showed no positive gamma emitting radionuclides in any metal samples or the wood sample, no positive HTD radionuclides in any sample and only naturally-occurring gamma emitting radionuclides in the concrete.
Given that no activation has been identified in any of the samples noted above, no activation of the secondary side portions of the Steam Generators and Pressurizer shell was expected, and none was discovered.
Neutron activation of these components secondary side and the inside of the NST would be limited to the neutrons emitted from the reactor core. As shown in Figures 1 and 3, the Steam Generators, Pressurizer and the NST are all located adjacent to the reactor pressure vessel and thus subject to neutron bombardment. The sampling effort focused on the outer shells of the Steam Generator and Pressurizer on the sides facing the reactor vessel. For the Pressurizer, an additional sample was collected from the side of the shell opposite the reactor core. In addition, MARAD chose to sample the Fuel Transfer Tank and also the support structure (foundation) where the pressure vessel once stood.
Figure 1, Savannah Reactor Layout Neutron Shield Tank The primary (neutron) shield tank surrounded the reactor vessel up to the hot legs.
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When filled, the shield tank formed a 33-inch-thick water annulus that provided the required neutron shielding to prevent excessive neutron activation of material inside the containment vessel and to reduce the neutron doses outside the secondary shield.
Fuel Transfer Tank The fuel transfer tank is similar in dimensions to the primary shield tank. It surrounded the reactor vessel head above the hot legs and was normally empty. It has no lead shielding. Figure 2, Steam Drum and FuefTransfer Tank in Red During refueling operations before the reactor vessel head was lifted, the fuel transfer tank (also called the upper shield tank in the 1968 refueling log) was filled with water to 36 inches above the lower reactor vessel flange to reduce dose during refueling when the head was removed from the vessel.
Figure 4 shows the NST and the Fuel Transfer Tank.
1 Heat Exchanger (Steam Generator U-tube bundle) 2 Let Down Cooler 3 Steam Drum (Steam Generator) 4 Effluent Condensing Taruc 5 Containment Drain Tanlc 6 Pressurizer 7 Check Valve 8 Primary Pump 9 Gate Valve Figure 3, Savannah Reactor Components Page 5 of 17
Reactor Pressure Vessel Foundation The reactor pressure vessel (RPV) has been removed but the foundation for the RPV remains and was potentially subjected to neutron flux from the core. Figure 4 provides a view of the RPV and the foundation.
Steam Generators The heat exchanger tube sheet and U-tubes were removed from both steam generators. The interior of each SG/Heat Exchanger shell has been decontaminated. They will be partially reassembled after license termination and left in place. The interior surface of both steam generator shells has been surveyed with swipes, static and scans. Figures 5 and 6 show the steam generators, risers and the steam drum.
Fuel Trllllfa-T,nk Neulron Shield Tank Figure 4, Pressure Vessel Showing Primary (Neutron) Shield Tank and Foundation Page 6 of 17
Figure 5, Steam Generator and Steam Drum Figure 6, Current Photo of Steam Generator, Risers, and Steam Drum Pressurizer Shell The interior of the upper section of the Pressurizer has been decontaminated to allow leaving it in place on completion of the decommissioning. The lower section containing the heater sleeves and other penetrations was disposed as Low Level Radioactive Waste (LLRW). The remaining Page 7 of 17
interior surface of the Pressurizer was surveyed during Final Status Survey. Figure 7 shows the Pressurizer.
- ' t:
Figure 7, Components to be Sampled (PD-T4 is a previous sample location)
Figure 8 prQvides an overview of where the samples were collected.
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PORT 0
- STARBOARD Figure 8, Sample Locations ANALYSIS AND CONCLU_SION BOW The number of samples collected was reduced from twenty-six (26) as originally planned to ten (10) in order to match sample locations specifically requested by an CRISE (Oak Ridge Institute for Science and Education) site visit the week of July 29th, 2024. The physical effort for setting up and drilling the samples proved to be difficult and was risky from a personnel safety standpoint so the decision was made to minimize any unnecessary sample collection. The NRC and MARAD agreed to this reduction in scope.
Samples collected were analyzed for activation products as requested, C-14, Ni-59 and 63 and Co-60 by GEL laboratories1 an ELAP certified laboratory by gamma spectroscopy using DOE GA-01-R Mod Co-60, and for Ni-63 method RESL Ni-01 M. Table 1 provides the results of the sample analysis.
1 GEL Laboratories 2040 Savage Road Charleston, SC 29047 Page 9 of 17
Only three (3) samples resulted in positive values more than the MDCs; two indicated Ni-63 and one (1) tested positive for tritium. Both radionuclides were used in the development of the current DCGLs in the L TP.
The sampling effort was designed to determine whether any impacted system components that will remain in place after license termination are neutron activated, including activation products such as carbon-14 (C-14 ), nickel-59 and 63 (Ni-59, Ni-63), and cobalt-60 (Co-60). The sampling effort collected ten ( 10) samples from the various components with only three (3) positive results. These results indicate that the systems to remain in place do not indicate levels of neutron activation and the current DCGLs and ROCs listed in the LTP remain valid.
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Table 1, Sample Results Location ROC Result Units MDC Units Co-60 2.29E-07 uCi/g 3.37E-07 uCi/Q Sr-90
-4.30E-07 uCi/g 4.SSE-06 uCi/Q Tritium 2.02 pCi/g 4.99 pCi/g
- 1-PORT Steam Generator Riser Tube 0005 C-14 8.79E-06 uCi/g 3.96E-05 uCi/g Tc-99
-3.39E-06 uCi/g 1.64E-05 uCi/Q Fe-55
-8.43E-05 uCi/g 3.72E-04 uCi/Q Ni-63 1.40E-06 uCi/g 8.14E-06 uCi/g Co-60 1.59E-07 uCi/g 3.76E-07 uCi/g Sr-90
-5.19E-07 uCi/g 4.30E-06 uCi/g Tritium 5.12E+OO pCi/g 4.96E+00 pCi/g
-2.69E-06 uCi/g 7.78E-06 uCi/Q Co-60 1.21E-09 uCi/g 1.56E-07 uCi/g Sr-90
-5.S0E-07 uCi/g 3.77E-06 uCi/Q Tritium 6.14E-01 pCi/g 4.92E+00 pCi/g
- 3-Fuel Transfer Tank Fwd PORT 0008 C-14
-6.03E-06 uCi/g 3.67E-05 uCi/Q Tc-99
-8.85E-06 uCi/g 1.61 E-05 uCi/Q Fe-55 1.35E-04 uCi/g 5.98E-04 uCi/g Ni-63 1.35E-06 uCi/g 6.83E-06 uCi/a Co-60 1.79E-07 uCi/g 5.98E-07 uCi/g Sr-90
-1.33E-06 uCi/g 6.72E-06 uCi/Q Tritium 1.58E+00 pCi/g 5.11E+00 pCi/Q
- 4-STBD Steam Generator Riser Tube 0010 C-14 1.45E-05 uCi/g 3.95E-05 uCi/g Tc-99 3.86E-06 uCi/g 1.60E-05 uCi/g Fe-55 2.45E-04 uCi/g 3.60E-04 uCi/g Ni-63 4.03E-06 uCi/g 8.09E-06 uCi/a Co-60 1.72E-07 uCi/g 7.15E-07 uCi/g Sr-90
-1.40E-06 uCi/g 3.06E-06 uCi/g Tritium 1.66E+00 pCi/g 4.94E+00 pCi/g
-2.93E-06 uCi/g 1.66E-05 uCi/g Fe-55 1.38E-04 uCi/g 3.38E-04 uCi/g Ni-63 8.26E-06 uCi/g 8.0SE-06 uCi/Q Co-60 1.90E-07 uCi/g 3.0SE-07 uCi/Q
- 6-NST Inside Aft 0004 Sr-90 1.78E-06 uCi/g 4.0SE-06 uCi/g Tritium 1.83E+00 pCi/g 4.94E+00 pCi/g C-14 1.09E-05 uCi/g 3.76E-05 uCi/a Page 11 of 17
Location ROC Reault unn.
-2.55E-06 uCl/g 1.59E-05 uCi/g Fe-55 2.83E-04 uCl/g 3.66E-04 uCi/g Ni-63 1.72E-06 uCl/g 7.22E-06 uCi/g Co-60 1.82E-07 uCl/g 4.64E-07 uCi/g Sr-90
-3.82E-06 uCl/g 7.03E-06 uCi/g Tritium 2.35E+OO pCl/g 4.88E+OO pCi/g
- 7-NST Exterior STBD 0001 C-14 1.85E-05 uCl/g 3.73E-05 uCi/g Tc-99 2.S0E-06 uCi/g 1.57E-05 uCi/g Fe-55 1.51 E-04 uCi/g 3.84E-04 uCi/g Ni-63 2.21E-06 uCi/g 7.81E-06 uCi/a Co-60 1.01E-07 uCi/g 3.16E-07 uCi/g Sr-90
-2.07E-06 uCl/g 5.23E-06 uCi/g Tritium 1.61E+O0 pCi/g 4.85E+OO pCi/g
-5.31E-06 uCi/g 1.58E-05 uCi/g Fe-55
-1.39E-05 uCi/g 3.29E-04 uCi/g Ni-63 1.71E-05 uCi/g 7.38E-06 uCi/a Co-60 1.44E-07 uCi/g 3.40E-07 uCi/g Sr-90 3.71E-06 uCi/g 4.21E-06 uCi/a Tritium 5.S0E-01 pCi/g 4.92E+00 pCi/g
-1.54E-07 uCi/g 1.S0E-05 uCi/g Fe-55
-1.55E-05 uCi/g 2.88E-04 uCi/g Ni-63 6.60E-06 uCl/g 7.35E-06 uCi/g Co-60 4.16E-08 uCi/g 2.14E-07 uCi/g Sr-90 2.S0E-06 uCi/g 3.32E-06 uCi/g Tritium 1.05E+OO pCi/g 4.88E+00 pCi/g
-1.98E-06 uCi/g 1.55E-05 uCi/g Fe-55 9.52E-05 uCi/g 4.40E-04 uCi/g Ni-63
-3.83E-07 uCi/g 7.62E-06 uCi/o Bolded values show positive results.
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To provide a relation of the results and the MDC they are plotted in Figures 8 - 15. The sample numbers are given in Table 2.
Table 2, Sample Identification Sample Identification Number
- 1-PORT Steam Generator Riser Tube 0005
- 2-Fuel Transfer Tank Aft STBD 0009
- 3-Fuel Transfer Tank Fwd PORT 0008
- 4-STBD Steam Generator Riser Tube 0010
- 5-Pressurizer STBD 0006
- 6-NST Inside Aft 0004
- 7-NST Exterior STBD 0001
- 8-NST Inside Aft 0002
- 9-NST RPV Foundation Aft 0003
- 10-Fuel Transfer Tank Fwd 0007 Page 13 of 17
4.51 e-05 4.0 I e-05 3.Sle-05 3.0le-05 2.5 le-05 2.0 le-05 I.Sle-05 1.0 I e-05 5.1 0e-06 I.00e-07 0
2 Figure 9, Carbon 14 Results I.00e+00 I.00e-09 0
0 I
2 Figure 10, Cobalt Results Carbon 14 Results*
~
~ample-Numb&
7 a;ample 3 result was -6.03E-06
- t 0
3 Cobalt 60 Results 4
5
- 6 Sample Number Page 14 of 17 7
8 8
9 9
Iron 55 Results*
~ 4.1 0e-04 u
e Samph: Numbei 2
- sampld 1 resalt was ~.43E-05 9
1 O Sample 8 result was -1.39E-05 Sample 9 result was -1.SSE-05 Figure 11, Iron 55 Results 1.53e-05 1.03e-05 5.30e-06 3.00e-07 0
g u
- L Nickle 63 Results*
2 3
8ample!Numbe, 7
8 9
- Sample 2 result was -2.69 E-06 Sample 10 result was -3.88E-07 Figure 12, Nickle 63 Results Page 15 of 17 10 Result e MDC Result e MDC
7~06 1 6.00e-06 5.00e-06 4.00e-06 3.00e-06 2.00e-06 Strontium 90 Results*
Sample Number 1.00e-06
- Sample I =.. 4_30 E-07 Sample 2 ;: -5.iS E-07 Sample 3 = -5.803 E-06 Sample 4 = -1.33 E-06 Sample 5 = -1.40 E-06
~ 1 Sam~le 7 :3-3.83 f *06 S?mp1es9 = w~s -2.01 E-06 9 10 u
- i.
Figure 12, Strontium 90 Results 1.90e-05 1.?0e-05 1.50e-05 1.30e-05 1.10e-05 9.00e-06 7.00e-06 5.00e-06 3.0oe-06 Tecnetium 99 Results*
Sample Number
- 1.00e-06
- Sample 1 = -3.39 E-06 Sample 2 = -1.43 E-06 Result e MDC Result e MDC Sample3 = -8.85 5-06 s,mple t = *2,i13 E-0~ Sa"'E)le 6 E -2.55;E-06 ~mpleg'l = -51;1 E-06 Sample 9 =
-1.5'4 E-07 Sam pre 1 o = -1.98 E-06 g,
u
- i.
Figure 14, Technetium 99 Results Page 16 of 17
5
- 4 3
Result 2
- MDC 0
0 2
3 4
5 6
7 8
9 10 Sample Number Cl
(.)
c..
Figure 15, Tritium Results Page 17 of 17