ML20010J189
| ML20010J189 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/03/1981 |
| From: | STONE & WEBSTER, INC. |
| To: | |
| Shared Package | |
| ML20010J181 | List: |
| References | |
| NA-1574, NUDOCS 8109290684 | |
| Download: ML20010J189 (50) | |
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TEST REPORT ON THE EFFECTS OF COMBINED GAMMA AND NEUTRON RADIATION ON BORATED SILICONE MATRIX CT-40-NS TYPE "B" INSTALLED WITHIN THE REACTOR CAVITY OF THE NORTH ANNA NUCLEAR STATION UNITS 1 AND 2 STONE AND WEBSTER P.O. NO: NA 1574 DATE ISSUED: February 23, 1981 :
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8109290684 810813 3 67 PDR ADOCK 05000338
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INDEX PAGE 1.0 0bjective...................... 2-3 2.0 Testing Requirements........... 4-6 3.0 Samples Preparati 7............ 7-9 4.0 Samples Irradiation........... 10-15 5.0 Physical Analysis............. 16-17 6.0 Chemical Analysis............. 18-19 7.0 Pre and Post Irradiation LOCA Exposure Testinc.......... 20-23 Attchments
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A). Specification for Supplemental Neutron Shielding B). University of Michigan (Irradiation and Physical Analysis)
C). Dow Corning Corporation (Chemical Analysis)
D). Southwest Research (Impact Test)
E). Wyle Laboratories (LOCA Exposure Teuting)
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1.0 OBJECTIVE The objective of this test program is to simulate the shield material's specified operating conditions and possible one-time abnormal condition. Also to demonstrate, under these one-time conditions, the neutron shielding material will not decompose, crumble, dissolve, or melt in any sienificans egree and will not evolve quantities of combustible gases or become a fire hazr" . The shield material need not remain functional as a shield after this one-time event.
I The specified normal operational requirements are as follows:
- 1. NORMAL OPERATION 1.1). Ambient temperatures between 86* F to 105* F.
1.2). Relative humidity form 0 to 100 percent.
1.3). containment pressures of 9 PS.~A to 11 PSIA.
1.4) . The neutron shield material shall be cap:' ale of withstanding an integrated gamma radiation dose 10 of 3 3 x 10 rads over its design life of 40 years.
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1.5). The shield material shall be capable of withstanding neutren fluence over its design life of 40 years of:
1.5.1) . E <.11 Mev - 6.0 x 10 l0 neutrons /sq.cm.
18 1.5.2). 11 s E s 1.1 Mev - 3.0 x 10 neutrons /sq.cm.
1.5.3) . E > 1.1 Mev - 1.5 x 10 17 neutrons /sq.cm.
1.6). The neutron shielding material will be in direct contact with sections of the reactor vessel and nozzle insulation.
During nonnal reactor operations, the continuous contact surface temperature of the shield at the insulation-to-shield interface will be 247*F to 393 F.
- 2. ABNORMAL ONE-TIME CONDITIONS b As specified, the shielding material may be subjected to the following abnormal one-time condition:
2.1). IIE TEMPERf E PRESSURE fH 0 to 2 min. 440*F 45 PSIG 5-11 2 to 60 min. 280*F 45 PSIG 5-11 60 min to 30 days 150*F 0 PSIG 7-9 2.2). 100 percent relative humidity.
2.3). Steam and water jet at 600 F in a localized area for a one minute period.
2.4). A boric acid solution soak (2.000 to 2,500 PPM) with sodium hydro /ide added, ph 5.0 to 11.0 at 150 F for 30 days.
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2.0 TESTING CRITERIA In order to properly accomodate the above normal and abnormal operational requirements, some test criteria integration was required. This integration resulted in the following test requirements.
1.0 The Shield Material-In Pre-0Derational Mode a). Quantify chemical values.
a.1). Hydrogen a.2). Carbon a.3). Oxygen a.4). Silicone b). Quantify physical values, b.*). Specific Gravity b.2). Durometer b.3). Stress / Strain c). Quantify LOCA Resistance c.1). Steam Impingement c.2). Borated Water Soaking 2.0 The Shield Material In Operational Mode a) . G amma and Neutron radiation exposure to 2 levels of integrated doses. ,
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a.1). Sample 1 - 8.25 x 10 9 Rads / Gamma 18 1.5 x 10 N/sq.cm. E <C .11 Mev lI 7.5 x 10 N/sq.cm. .115 E5s1.1 Mev 16 3.5 x 10 N/sq.cm. E > 1.1 Mev Above dose equivelant to 10 operational reactor years.
a.2). Sample 2 - 3 3 x 10 10 Rads / Gamma 18 6.0 x 10 N/sq.cm. E<.11 MeV 18 3.0 x 10 N/sq.cm. .11 SE s:1.1 MeV lI 1.5 x 10 N/ sq . cm. E :> 1.1 Mev Above dose equivelant to 40 operational reactor years.
j Since the shield material is to be in contact with reactor vessel and nozzle insulation, the shield material shall experience normal operating temperatures of 247'F to 393 F and therefore, shall undergo radiati.on exposure testing under equivalent conditions of 3 80* F 20 *F.
b). Quantify Chemical values (See Item 1.0 above) c). Quantify Physical values (See Item 1.0 above) d). Quantify LOCA Resistance d.1). Steam Impingement (Sample 1 and 2)
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e). Quantify Impact Resistance e.1). Accelerate Sample 1 and 2 above to 1500 F.F.S. to simulate Loss of Coolant piping rupture. Recover accelerated material and
. submit to Borated Water Soaking.
f). Quantify LOCA Resistance f.1). Borated Water Soaking.
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3.0 Chemtrol CT-40-MS Tvoe "B" Samoles Preoaration The preparation of sample specimens was carried out at the Chemtrol facilities April 11, 1979. No elaborate procedure were involved in the proper compounding of the matrix other than those approved by the Chemtrol QA/QC. These procedures basically reflect compounding normally associated with dual 4
component polymers.
i For the sake of traceability (Quality Control), all samples produced were labeled to provide continuity of samples. A series of samples were labled A1 thru A12 with another series B1 thru B12. The A series of samples were submitted to the University of Michigan (Phoenix Memorial Laboratory) for experimental purposes; i.e., identification of a reliable irradiation mode within the Ford Nuclear Reactor.
As soon as reliability was estabilished the B series of samples were submitted to the University of Michigan for the required irradiation process.
Due to limitations within the reactor irradiation area, the above A and B series of samples were molded into 1"x7"x3/8"
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long specimens. This geometry pr ovided symmetry of testing since the 380* F environmental heating required during the irradiation process did not allow sample specimens of greater size. Moreover, the sample geometry allowed greater ease of pr"sical propert.ies testing.
In conujunction with the above sample preparation, four additional sample specimens were prepared with an approximate rize of 4"x4"x4". These unirradiated samples were submitteo to hyle Laboratories for the purpose of undergoing exposure testing per Loss of Coolant Accident conditions.
o llote should be taken of the fact that the compounding of Chemtrol CT-40-flS Type "B" Silicone matrix involves the aMitional compounding of Baron Carbide (B4C) in fine powder form within the basic silicone polymer. This compounding is
( the 2nd step in the 2 step process. In the first step, the basic 2 part silicone components A and B are mixed in equal weights or volumne. Immediately thereafter, the boron i carbide (B4C) is added to mix as the second step in mixing process. The amount of B4C to be added to the basic polymer mix is sufficient to result in a net 1.5% minimum weight of boron of the matrix weight.
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An additional note to be made is that ancillary samples were 1
! prepared for the purpose of assuring reliability of boron homogenity within the silicone matrix. Said sample testing resulted in such assurance.
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. o 4.0 Chemtrol CT-40-NS Tvoe "B" Samoles-Irradiation The purpose of the neutron shielding material irradiation study is to determine the changes to physical and chemical propertiem of the shielding material as a function of irradiation exposure under conditions similar to those present within the reactor cavity of the North Anna Nuclear Station.
- In a radiation field, elastomeric materials are the most sensitive to radiolysis of all construction materials.
Likewise, radiation progressively deteriorates the initially optimized properties of elastomers. The general effect is to increase the modulus of elasticity and hardness while decreasing ultimate elongation and tensile strength.
The general process whereby such changes occur is l ionization. The transfer of energy from the radiation beam l
to the atoms of the polymer directly or indirectly causes the ejection of orbital electrons. The energetic electrons produce secondary ionizations. The ions formed by the ejection of orbital electrons recombine with free electrons to form enerjectic, unstable molecules. The excited l -
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. o molecules quickly dissipate their excess energies, largely by bond scission, physical transfer, and possibly molecular rearrangement. The bond scissions produce free radicals and unsaturation, and most of e.h e subsequent overt effects result from these. Cross linking, chain scission, molecular rearrangement, and chemical reaction with environmental agents, especially oxidation and czonization, occur and constitute the preponderant changes. Essentially all the changes in physical properties of the elast 3r material ensue from these basic processes. Although the primary and secondary processes are obviously not temperature sensitive, the resultant chemical reactions are temperature dependant.
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A secondary point of importance, is the effect of dose rate on the clastomer compound. Prior experimentation has shown that there is no dose-rate effect for elastomers within the broad range of 10 through 10 7 rads /hr. i.e.; at least over this range, any variation in dose rate does not significantly effect the change in physical properties of the elastomer compound. L' hat does matter is the total amount of radiation energy absorted and the uniformity of absorption.
A final point to be made is that prior experimgntation has revealed no significant post-irradiation effects in the .
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compound. Measurements of hardness and elasticity has shown no change over a period of three months after irradiation.
The irradiation test program was comprised of two parts.
The_first part involved irradiation of test specimens for a period equ,1 valent to 10 years of reactor operation i.e. , to the same level of accumulative gamma and neutron doses. The second part of the test program involved the irradiation of test specimens to 40 years equivalent reactor operation.
The above irradiation process was to be conducted with material samples in a 380 F 20 F environment.
All of the above irradiations were performed at ,he t
University of Michigan Ford Nuclear Reactor operated by the Michigan Memorial Laboratory. These facilities were chosen for these test experiments simply because the Ford Reactor flux profiles accomidated the majority of incident flux profiles required for the material testing. In addition, the University of Michigan reactor material testing personnel had considerable experimental experience with similar material testing requirements. o ,
v The respective irradiation exposure levels required along with the actual accumulated doses are a.s follows:
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10 Year Exposure Level Irradiation ,
Desired Actual Integrated Gama Exrosure 8.25 x 109Rads 8.4 x 109 Rads 10 Neutrons E <.11 MeV 1.5 x 10 N/sq.cm. 2.0 x 10 10 tVsq.cm.
Neutrons .115 En1.1 Mev 7.5 x 10 17 fVsq.cm. 8.7 x 10 1I tysq.cm.
l0 Neutrons E >1.1 MeV 3.5 x 10 N/sq.cm. 7.1 x 10 lI N/sa.cm.
to Year Excosure Level Type Desired Actual 10 Integrated Ga:::ma Exposure 3.3 x 10 Rads 4.1 x 10 10 Rads Neutrons E 4.11 Mev 6.0 x 10 lb FVsq.cm. 9.8 x 10 l0 fVsq . cm.
Neutrons .116 EnMev 3 0 x 10 l0 fusq.cm. 18 fy3q,cm, 4.3 x 10 0
Neutrons E >1.1 Mev 1.5 x 10 lI N/sq.cm. 3.5 x 10 N/sq.cm.
The above exposure levels were administred at the following incident flux levels.
10 Year and 30 Year Excesure Level Irradiation Measured Dose Rate Dose Rate -10'.
Gama 8.8 x 107 rad /hr. 7.6 x 10I rad /hr.
E >.11 MeV 5.6 x 10 12 n/cm2 /sec. 5.0 x 10 12 n/cm2 f3,'c.
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.11 Mev $E $.1.1 MeV 2.4 x 10 12 n/cm /sec. 2.2 x 10 12 n/cm2 /sec.
E > 1.1 Mev, 2.0 x 10 12 n/cm2 /sec. 1.8 x 10 12 n/cm2 /sec.
The 10 year exposure level required 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> of exposure to reach its desired accumulated dose while the 40 year level required 545.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of exposure. The actual exposure levels were in excess of the desired levels, i.e.; the 40 t
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year specimen received 20% more gamma and over 50% more neutron exposure than specified.
An additional deviation from the norm is the above referenced measured dose rates. A 10% reduction was applied to all gamma and neutron incident flux rates to allow for shielding of the samples 1/4" wall aluminum container. This value reduction was applied per estimates generated by University of Michigan Ford Reactor Engineering personnel.
These estimates are highly conservative.
From the. standpoint of irradiation exposure and resulting I physical and chemical data, the 40 year material sample specimens are not representative to conditions within the reactor cavity of the North Anna Nuclear Station. The material's physical and chemical properties may have undergone radical change beyond irradiation to 3.3 x 10 10 Rads. A more than 50% over exposure to neutrons in the thermal energy range surely had a prominent role in end data. Each thermal neutron to hydrogen reaction (capture) results in the release of a single 2.2 mev gamma photon.
The magnitude of this problem was somewhat reduced since B4C compounding reduced the potential for the H +N event.
Instead, a neutron + Boron 10 event results in gamma emissions of .5 mev. However, the accompanying release of a t,
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highly eaergetic alpha particle does provide a mechanism for localizec ionization. The magnitude of this event is appreciated when one equates the thermal flux rate, and atoms of Boron 10 as a function of cross section.
I i The temperature environment sustained by the material samples during the irradiation process most assuredly contributed to Loss of Chemical and Physical properties.
Gross radiation damage is accelerated by heating or stressing the elastomeric siterial.
On the matter of material activation, the 10 year sample rqrlected a gamma dose rate on contact of approximately 50 Mrem /hr. 16 days af ter removal from the reactor core area.
The identified activation species were CO-60, ZN-65, SB-124, SC-46 and CS-134. No data was obtained for the 40 year
, exposure sample.
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t 5.0 Chemtrol CT no-NS Tvoe "B" Physical Analysis In its cured state, Chemtrol CT-40-NS Type "B" has the following physical properties:
Specific Gravity-------------------------------1.36 Durometer (Shore A)----------------------------55 t
(Lb) (In) (In) (PSI) (In/In)
Tensile Force Sample Length Sample _ Elongation Stress Strain 0 Lo:2.50 0 0 0 1 2.58 .08 8.40 .032 2 2.68 .18 16.81 .072 3 2.75 .25 25.21 .100 4 2.85 35 33.61 ,140
, 9.5 UTS 79.83 .250 Subsequent to irradiation exposure (10 year exposure level) the shield material manifested charges in all areas of physical properties tested. These changes adhered to the general effect of increased hardness and modules of elasticity while decreasing ultimate elongation and tensile strength.
Specific Gravity------------------------1.57 g/cc D u ro m e t e r ( S ho r e A ) --------------------- 95 (Lb) (In) (In) (PSI) (In/In)
Tensile Force Sample Length Sample Elongaticn Stress Strain 0 1.80 0 0 0 20 1.80 0 204.08 0 .
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- No physical data was obtained from the 40 year exposure level sample other than specific gravity due to difficulty in handling sample. Generally speaking said samole was fragile upon handling. As discussed in Section 4.0, the over-irradiation most likely resulted in an exagerated representation of expected physical affects to shield material.
The impact resistance testing simulating accelerations as a result of loss of coolant accident was conducted by the Southwest Research Institute of San Antonio, Texas. The institute received the 10 year and 40 year exposure samples from Wyle Laboratories who had the initial test responsibility of subjecting the samples to steam impingement exposure.
The general results of the impact testing, i.e.,
accelerating samples to 1500 FPS, was to reduce the sample specimens both (10 year ar.d 40 year) to a. fine powder with a density much greater than water.
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6.0 Chemtrol CT 40-flS ivne "B" Chemical Analvsis Since the primary function of the above referenced shield material system is to act as a neutron shield, careful elemental analysis is required. With this fact in mind, Chemtrol Corporation was able to retain the services of Dow Corning Corporation to perfom the chemical analysis. Dow Corning Corporation is the main supplier of the base silicone polymer utilized in the formulation of the shield material and therefore, is well familiar with silicone analysis techniques.
t Initially, one control sample (non-irradiated) and one of the 10 year exposure 1.evel specimens were submitted to them for analysis. Soon thereafter, the 40 year exposure level specimen was also submitted to them for the required analysis.
l The following is the elemental data obtained from the 1
! control, 10 year and 40 year exposed samples:
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~5lement/ Property Control 10 Year Sample 40 Year Sample Carbon (C) 17.77 + .17% 14.22 + .14% 5.96 + .06%
Hydrogen (H) 4.38 + 12% 3.00 + .09% 1.07 + .03%
Silicon (S1) 40.5 + .2% '41.2 + 0.2% 40.2 + .5%
0xygen (0) 37.35 + .2% 41.58 + .25 58.42 + .55
( Specific Gravity 1.359 + .005% 1.571 +.005% 1.68 + .005.
l l 0xygen Values by Difference.
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A review of the elemental data reflects a significant loss in hydrogen and carbon.
It should be noted that the irradiated samples were thin strips which were given a near uniform irradiation under uniform temperature. The insthlled shield material's thickness and orientation allow for self shielding and a temperature gradient throughout its thicknes.
Accumulated dose and resultant det;radation of material properties would, therefore, be attenuated as a function of shield thickness. In addition, the test specimens were given an approximate 50% excess irradiation, as discussed in Section 4.0.
In conclusion, the test specimens demonstrated a definite degradation in f
shielding properties. The degree of degradation or efficiency which will be experienced by the' installed shield material, however, can only be accurately determined by a plant survey pregram of neutron dose vs.
life of shield material.
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7.0 Chemteci CT-40-NS Tvoe "B" Pre and Post Irradiation LOCA Exoosure Testing Qualification testing was conducted on CT-40-t!S Type "B" to simulate Loss of Coolant Accident (LOCA) conditions. The test program was performed by Wy.e Laboratories of Huntsville, Alabama.
The test program was differentiated into two basic programs.
One program element reflected qualification testing of unirradiated shield material, while the other element involved qualification tetting of the 10 yesr and 40 year irradiated material samples. In conjunction with the second program ar.d subsequent to the initial steam impingement portion of the LOCA testing, the material samples were submitted to Southwest Research Institute for impact testing (refer to Section 5.0 and Exhibit E). Thereafter, Southwest Research Institute returned the impact test specimens back to Wyle Laboratories for continuation of LOCA Testing Requirements (Borated Water Submersion Test).
The initial LOCA testing involved unirradiated material samples (4 ea. - 4"x4"x4") comprising of steam impingement (600 F steam - 60 sec.) and submersion in borated water at 150 F for 30 days.
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J The effects of steam impingement on the sample specimens were non-existant. However, the results from the Borated Water Submeraion Test merits some discussion. The area of discussion only relates to the generation hydrogen off-gassing since all four material samples appeared to be unchanged in all other physical respects after removal from the borated water test cylinder.
The LOCA test cylinder had a free volume of approximately .7 cubic foot (20 liters). Of this free volume, approximately 75% (15 liters) was occupied by borated water and 4 unirradiated material sample specimen'. s Gas that evolved collected in the remaining 5 liters of the cylinder at 150 F (66 C) and 1 psi (1.07 atmospheres). Gas analysis, performed at room temperature and atmospheric pressure, resulted in 32 volume percent. hydrogen. Five lit es of gas at 1.07 atmospheres and 150 F is equal to approximately 4.31 liters at room temperature and standard pressure (STP). The
- hydrogen component under the same conditions would be 32% of 4.31 liters, or approximately 1.38 liters. The evolution of gas is a function ofothe surface area of 2.7 sq.ft.
Therefore, the evolution of hydrogen per square foot of a
material is .62 liters /sq.ft.
Chemtrol's CT-40-NS Type "B" Neutron Shield 1nd Material has
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been installed in the North Anna Unit 1 and 2 Nuclear Facility with an exposed surface of approximately 10,000 sq.ft. This materici surface area, in an unirradiatev form, would generate 6,20G liters of hydrogen gas in the event of LOCA. The area of installation (containment) has an 6
approximate volume of 1.8 x 10 cubic feet (5.1 x 10 7 liters). Therefore, the hydrogen cas evolved would account for less than 1.24 x 10-N percent of the containment volume.
This volume percent does not even remotely approach the LEL for hydrogen in air (4%).
The introductica of th'e methane (1.1%) and ethane (.01%)
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components of the off-gassing to the above equation does not alter the basic conclusions.
The LOCA testing on the 10 year and 40 year irradiated specimens was somewhat different in that the test specimens for the Borated Water Submersion were in potider form as a result of impact testing. Nevertheless, the basic results with the exception of increased ethane off-gassing and much reduced hydrogen off-gassing. The overall off-gassing was minute in quantity and therefore, does not p.~ ovide a potential for concern.
A comparative evaluation of CT-40-NS Type "B" (irradiated l
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s vs. unirradiated) shield material system as a function of LOCA exposure testing reveals that as the shield system undergoes irradiation e x p o s 'J r e , its potential for off-gassing is greatly reduced.
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J.O. No. 13075.20 June 8,19 79 NAS 1007
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Specification for l
SUPPLEMENTAL NEUTRON SHIEL,D_ING North Anna Power Station - Unit 1 Virginia Electric and Power Company Richmond, Virginia Rev.0 Rev.I *Rev.2
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S&W Originator
'h>W l 6-9 *4 S&W Lead Engineer S&W Specialist f.it.W
//78T S&W Project Engineer _
VEPC0 Project Engineer VEPCO QA Supervisor I
YEPC0 Approved t
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I. GENERAL 1.9 The reactor neutron shielding will be used to attenuate 1.12 neutrons escaping from the reactor vessel during normal 1.13 plant operations. _The shielding will be located between 1.14 and around the reactor nozzles and will be supported on top of the neutron shield tank. _The shielding will remain 1.16 in place during all phases of operation except when portions of the nozzle shields must be removed for 1.17 periodic inspection of welds. The shield is designed to 1.18' contain both hydrogen and boron to attenuate and _ absorb 1.19 neutrons, thus reducing exposure to operating personnel.
II. SCOPE 1.22
- 1. The work includes fabrication, quality control, and 1.25 delivery of silicone-based neutron shielding material 1.26 to the North Anna jobsite.
- _The supplier will be provided with previously 1.29 fabricated molds (50* and 70* shields) as shown o_n the 1.30 attached sketches, 13075-MKS-13, sheets 1 and 2.
These prefabricated molds are to be filled with the 1.31 silicone neutron shielding material.
l i The supplier shall also supply molds and mold the 1.32 silicone material into the various shapes shown on the a_ttached sketches, 13075 -MKS-13, sheets 3 through 10.
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1.33 quantities shall be as specified. 1.34 2_ . The prefabricated molds require that the 10 gage, 1.36 inside diameter be supported along its vertical length 1.37 such that it is not distorted during injection of the liquid silicone neutron shield material by the liquid 1.38 hydrostatic force. _The purchaser will provide this 1.39 support. The supplier shall inform the purchaser of 1.40 any restrictions on the support material. Items to be 1.41 considered are cure temperature that the support may l experience and other supplier identified
. considerations. Ihis information is required as part 1.42 of the supplier's bid.
III. OPERATING CONDITIONS 1.45 l
The neutron shielding will be located within the 1.48 containment building in close proximity to the reactor 1.49 vessel. It will be subjected to mntainment atmospheric 1.51 conditions as follows:
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- 1. Normal Operation 1.54
- a. Ambient tercperatures between 86*F to 105*F 1.56
- b. Relative humidity,from 0 to 100 percent 1.57'
- c. Containment pressures of 9 psia to 11 psia 2.1
,d_ . The neutron shield material shall be capable of 2.2 withstanding an integrated gamma radiation dose of 2.3 3.3 x 1010 rads over its design life of 40 yr.
- e. The materials shall 'be capable of withstanding 2.4 neutron fluence over its design life of 40 yr of:
- 1. E<.11 Mev - 6.0 x 10 *
- nutrons/sq cm 2.6
- 2. .115ES1.1 Mev - 3.0 x 1018 neutrons /sq cm 2.8
- 3. E>1.1 Mev - 1.5 x 10:7 neutrons /sq cm 2.10
- f. The neutron shielding material will be in direct .:.13 contact with sections of the reactor vessel and 2.14
' nozzle insulation. _During normal reactor 2.15 operations, the continuous contact surface temperature of the shield at the insulation-to- 2.16 shield interface will be 247*F to 393*F.
i 2_ . The shielding material may be subjected to the 2.18.
! following abnormal one-time condition in addition _to 2.19-those listed in III.1.
- a. Time Temperature Pressure p_H_ 2.22 0 to 2 min 4400F 45 psig 5-11 2.25 2 to 60 min 280*F 45 psig 5-11 2.26 60 min to 30 days 150*F 0 psig 7-9 2.27
- b. 100 percent relative humidity 2.32
- c. Steam and water jet at 6000F in a localized area 2.33 for a one-minute period d_ . A boric acid solution soak (2,000 to 2,500 ppm) 2.34 with _ sodium hydroxide added, pH 5.0 to 11.0 at 2.35 150*F for 30 days.
Ender these one-time conditions, the neutron shielding 2.37 material must not decompose, crumble, dissolve, or melt in any ,s_ignificant degree and must not evolve 2.38 quanitites of combustible gases or become a fire c-13075.20-1b 06/07/79 080
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hazard. The degree and type of degradation expected 2.39 under the one-time conditions shall be indicated in the supplier's bid. The shield material need not 2.40 remain functional as a shield after this one-time event. .
IV. DESIGN REQUIREMENTS 2.43 1 The shielding material shall be of a Dow Corning 2.46 "Sylgard 170* base with a boron loading of 2 percent 2.47 by weigLt. If there are any exceptions or changes in 2.49 the Seller *s proposed composition, list under. the
" Exception to Requirements" section of the proposal. 2.50
_The following information shall be provided in the bid 2.51 package:
Specific gravity 2.54 (Minimum 1.3 g/cc) 2.55-Thermal conductivity 2.57 Bydrogen density after 32EFPY 3.1 (Minimum 0.055 g/cc) 3.2 Carbon (percent) 3.4 i
Silicone (percent) 3.7 Oxygen (percent) 3.9 Bydrogen (percent) 3.11 Boron (percent) 3.13.
- 2. Toxicity: The material shall exhibit no toxic effects 3.18 to personnel handling it.
- 3. Fire resistance: The following requirements are set 3.20 forth in the U.S. Regulatory Guide 1.120, Revision 1, dated November 1977. The material should be 3.22 noncochustible,or listed by a nationally recognized testing laboratory such as Factory Mutual Underwriters 3.23 Laboratory, Inc., for:
2 Surface flamespread rating of 50 or less when 3.25 tes ced under AS'IM E-84.
c-13075.20-1b 06/07/79 080
4.
_The supplier shall provide the test data necessary to 3.2L support ANI and NRC acceptance. _Any test data whiche 3.29 is not available shall be identified in the supplier's bid.
,4. The cured shielding material shall be sufficient 3.3-strength to be self-supporting when subjected to the 3.32 normal and one-time operating _ conditions specified in 3 . 3.*
Section III above. The supplier shall certify that 3.3L i the neutron shield material will exhibit the above
! physical and chemical properties for 40 yr under the 3.34 operating condition (Section III) specified above.
- 5. . The silicone neutron shield material shall be cured by 3.36' the Seller to ensure that it will exhibit the physical 3.37 and chemical properties listed above in Section IV when subjected to the normal _and one-time operating 3. 3 t '
conditions of Section III. The supplier shall ensure 3.39' that void spaces in the silicone material are minimized during mold injection. The supplier shall 3.4C provide his curing procedure as part of his bid.
V. INSTRUCTIONS AND DOCUMENTATION 3.43 1 The Seller shall furnish the buyer with any available 3.46 documentation required for support of licensing 3.47 applications. _This would include such things as fire 3.49 protection, radiation resistance, quantity of hydrogen offgassing ,during the one-time conditions of 3.50 Section III, chemical properties, physical properties, etc.
- 2. The Seller shall furnish with his bid the proposed 3.51 que'.ity control procedures to ensure that the s_upplied 3.52!
neutron shield material meets the above-listed (Section IV) physical and chemical properties and the method for which void spaces are minimized during 3.53 manufacturing.
' -3. The Seller shall furnish with this bid a schedule of 3.54 fabrication and shipping milestones for _the material 3.55' identified in Section II.
O e
k c-13075.20-1b 06/06/79 082
5.
i VI. IDENTIFICATION AND PREPARATION FOk SHIPMENTS 3.5E J,. Packing slips or other identification papers shall 4.3 I accompany all shipments and must bear the following: 4.4 Supplementary Neutron Shielding 4.6 )
North Anna Power Station - Unit 1 4.7 Purchase Order No. 4.8 Item No. _ 4.9 1
- 2. Each item shall be marked with the appropriate item 4.12l number.
3_ . The Seller or subcontractor shall prepare and 4.14 weatherize all articles for shipment in a good 4.15 consnercial manner so as to procect the material from
. damage to which it might reasonably be subjected, both 4.16 in transit. and handling. P,ackaging shall allow for 4.17-temporary outdoor storage.
VII. QUALITY ASSURANCE PROGRAM 4.2C Each bidder shall submit with his original proposal one 4.23' copy of his quality assurance procedure, c,overing the 4 . 2 t.
, quality control and quality assurance measures (1) imposed i
by him on his work and 12) imposed by him upon 4.25 subsuppliers or subcontractors.
- A procedure acceptable to the Engineers shall be a 4.27-prerequisite for a bidder being chosen as Seller.
The Seller and his subsuppliers shall have in effect in 4.28.
their shops at all times _an inspection, testing, and. 4.29-documentation program that will ensure that the equipment furnished under this technical document meets in all 4.30' l respects with the requirements of this technical document.
The Seller shall' implement and maintain this procuedure 4.31 carrying out the requirement of this technical dccument, 4.32 and all proposed major changes shall be sulxnitted to and approved by _the Engineers prior to implementation. 4.33i In order to attain assurance that the neutron shielding 4.34 material will.give satisfactory, dependable, trouble-free 4.35.
service, this technical document requires that a "Ceritificate of Conformance" be submitted by the Seller, 4.36 verifying that the neutron shielding material is in conformance with the set-forth requirements.
The Seller shall specifically ensure that a copy of this 4.37-i technical document with all addenda thereto, or c-13075.20-1b 06/06/79 082 l
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appropriate work instructions which include the technical 4.38 document requirements are readily available at each p_f his, 4.39 fabricating or production locations where work covered by this document is in progress.
VIII. INSPECTION 4.42 The Seller shall give the Purchaser or Engineers 4.45 notification 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to initial pouring of molds for j;tossible inspection. 4.46 Authorized shop inspectors or other representatives of the 4.48 Purchaser or the Engineers shall be allowed access to the 4.49 engineering offices, shops, and working area of the Seller and h_is subsuppliers at all reasonable times. They shall 4.57 have the right to such information as is necessary to demonstrate that engineering, procurement, and production 4.52 are proceeding in accordance with the established schedules. They shall also have the right to inspect the 4.52 materials or equipment, or the Seller's or subsupplier's production and inspection procedures, to confirm that the 4.54 requirements of this document are being complied with; the Seller or subsupplier shall provide all tools, 4.55 l
instruments, etc, necessary to facilitate these
. inspections. A final inspection will be made prior to 4.56 shipment of the neutron shielding material.
IX. GUARANTEE 5.1 The materials furnished shall be new and guaranteed to 5.4 perform as required herein. tiaterials and workmanship 5.6 shall be guaranteed to meet the requirements set forth, above. Should any defect in material, workmanship, or 5.7 performance develop during the first year following 5.8 initial commercial operation of the plant, the Seller shall agree to make repairs or necessary replacements .f_ree 5.9 of charge as long as the materials were not subjected to unusual operating conditions or other hazardous requirements. 5.10 The Seller shall repair or replace any item which is torn, 5.11 gouged, or broken upon receipt. The proposed repair 5.12 procedure shall be submitted with the supplier's bid.
X. PROPOSAL DATA 5.15
,1. . The bidder shall include in his proposal sufficient 5.18 l data and other information to completely 5}.escribe the 5.19' material to facilitate a thorough evaluation. The 5.29 proposal information shall include, but not.be limited I to, the following:
c-13075.20-1b 06/06/79 082
~. ~.
- 7.
- a. Completed' summary of proposal 5.23
- b. Warranty statement 5.24
- c. Material specific.ation and data sheets 5.25
- d. Performance test reports 5.26
- e. Quality Assurance and Quality Control procedures 5.27 to be followed during the udxing _and injecting of 5.28 the neutron shielding mate-ial.
- f. Any other data or information requested'elsewhere 5.29 in this document.
- g. Any other information that may assist the 5.30 Purchaser or Engineers in evaluating the bidder's proposal.
- 2. If the bidder takes exception to any of these 5.32 requirements, these exceptions should be listed separately _under the heading " Exceptions to 5.33 Requirements.* _Any exceptions r'tould reference the 5.3a paragraph to which exception is taken. I,f no 5.35 exceptions are taken, the bidder should so state.
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VIRGINIA ELECTRIC AND POWER CCMPANY CERTIFICATE OF CDNFORMANCE Project Name i
Seller Address Item or Service Mark So.
Specification No. and Title Purchase Order No. J .O .No .
Seller Identifying No. Drawing No.
Deviations from Specification Requirements: (If I:one, so state) attach Cbpies of Deviation Approval Documents
- 1. 4.
- 2. 5.
- 3. 6.
The Seller, including his subsuppliers, hereby certifies that the item or service, s_upplied on this order, complies with the above-listed specifications, drawings, applicable codes,,s_tandards, and procedures. _The Seller certifies that all deviations from specification requirements are listed a_bove and that deviation approval docu:nents are attached.
Signature Quality Assurance Manager or Equivalent O
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INSTRUCT ONS FOR COMPLETING VEPCO'S .
CERTIFICATE OF CONFORMANCE The vendor shal. complete the lines numbered below, as applicable, and return one copy with the shipment, and another copy mailed to the
. project location - ATTN: Resident QC, Engineer:
LINE NO. INSTRUCTIONS ,
'l . . . PROJECT NAME (Surry, North Anna) ,
- 2. VENDORS NAME .
'3 . VENDORS ADDRESS .
- 4. NAME OF COMPONENT OR SERVICE PERFORMED i
- 5. MARK OR PART UUMBER OF COMPONENT .
- 6. THE SPECIFICATI0M TITLE, NUMBER, REVISION AND DATE, SPFf'.TFTC TMDUSTD.Y STANDAD.DS
~
~~- OR CODES ATTESTED TO AS REQUIRED BY THE
' - PURCHASE ORDER.
- 8. JOB ORDER NUMBER
- 9. - VENDORS JOB NUMBER OR SHOP NUMBER
- 10. APPROVED FALRICATIO!! OR ENGINEERING DRAUINGS AND LATEST REVISION
- 11. ALL VENDOR DEVIATIONS FROM Tile PURCHASE ORDER OR SPECIFICATION UIT11 APPROVAL LETTERS ETC. .
TO VERIFY ACCEPTANCE OF DEVIATION
- 12. / Q.A. MANAGER OR EQUIVALENT RESPONSIBLE VENDOR REPRESENTATIVE 9
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ATTACHMENT B .
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- THE UNIVERSITY OF MICHIGAN hl PHOENIX MEMORIAL LABORATORY FORD NUCLEAR REACTOR ANN ARBOR MICHIGAN 48109
% (313) 764-6220 September 17, 1979 Mr. Vincent Cataldo Chemtrol Corporation 8805 Solon Building G5 Houston, Texas 77064 dear Mr. Cataldo:
Enclosed are tabulations of the total gamma and neutron doses administered to samples of Chemtrol material CT-40-NS Type B. Samples were irradiated at 380 20 F for 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> and 545.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at two megawatts. The specific gravities of three separate pieces from the 110 hour0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> samples are also enclosed.
Pieces of the i M hour samples have been sent to Southwest Research for analysis.
The 545.6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> samples are still too radioactive to be handled. Present plans are to prepare them for analysis on September 24,1979.
Sincerely, Reed R. Burn k bLu
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Reactor Manager enclosure .
I RRB/sv
- . . . . - . . = = . * = * - .
CHEMTROL MATERIAL CT-40-NS TYPE B Short Term Irradiation irradiation Temperature: 380!20 F
+
2 MW Measured Dose Rote Total Dose Irrodiotion Time . Dose Rote -10% Actual Desired 7
Gamma 110 hr 8.4 x 10 rod /hr 7.6 x 10 8.4 x 10 8.25 x 10' I 12 12 18 18 nc .11 MeV 110 hr 5.6 x 10 cm 2/5 5.0 x 10 2.0 x 10 1.50 x 10 12 12 I7 37
.11 WV < n< 1.1 MeV 110 hr 2.4 x 10 2.2 x 10 8.7 x 10 7.5 x 10 12 12 I7 16 n > 1.1 MeV .
110 hr 2.0 x 10 1.8 x 10 7.1 10 3.5 x 10 Long Term Irradiation I 10 Gamma 545.6 hr 8.4 x 10 rod /hr 7.6 x 10 4.1 x 10 3.3 x 10 12 12 18 18 nc .11 MeV 545.6 hr 5.6 x 10 n/cm 2/5 5.0 x 10 9.8 x 10 6.0 x 10 12 12 18 18
.11 MeV< n < 1.1 MeV 545.6 hr 2.4 x 10 2.2 x 10 4.3 x 10 3.0 x 10 n > 1.1 MeV 545.6 hr 12 12 18 I7 2.0 x 10 1.8 x 10 3.5 x 10 1.4 10 l
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7 THE UNIVERSITY OF MICHIGAN PHOENIX MEMORIAL LABORATORY FORD NUCLEAR REACTOR ANN ARBOR, MICHIG AN 48105 August 13, 1979 Mr. Vince Cataldo Chemtrol Corporation 8805 Solon Building G5 Houston, TE 77064
Dear Vince:
Enclosed is sample B6 of your CT-40NS, Type B material which has been irradiated as follows:
Gamma 8.4 x 10' rad 0
Neutron E < .11 MeV 2.0 x 10 ry'em I7 *
.11 MeV < E<l .1 MeV 8.7 x 10 I
E > 1.1 MeV 7.1 x 10 The gamma dose rate on contact from this sample was approximately 50 mrem /hr.16 days after removal from the reactor core.
Sincerely, kr k L t_ _
Reed R. Burn Reactor Meager Ford Nuclear Reactor Enclosure
(.
. p THE UhlVERSITY OF MICHIGAN
'e-
}Mg!,g tv, g q ,!j PHOENIX MEMORIAL LACORATORY FORD NUCLEAR REACTOR
- ANN ACBOR, MICHIGAN 48109 (31 ) 7s4 s220 Qe.dd July 29,1979 Dear Vince; Enclosed are the residual radioactivity results from the short term sample which was irradicted at 350-400 degrees.
The long term sample completed its first cycle at temperature and is doing well.
Sincerely, f
(
DiSIEUBUTlG l U n* Cetol;--
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4B '6C 4. * * $ * * * * * * * * *
- 4. * * * $ * * *
- t * * * * * * * * * $ 4. * @ * * * * * $ $ *
- p p * * * * * * * * ,4 * * *
..*&c.w***.+*****+** 20 JUL 1979 2:29:28 PM *******++********
- .***********b***************t***********t:****k+**+.F**4**
CHEMTROL EXPERIMEFJT i t
SAMPLE DATE: 191UL79 1446:00 SAMPLE IDEt3TIFICATIOft: 110 '
TYPE OF SAMPLE: SOLID SAMPLE QUAR 4TITY: 0.1510000 UNITS: GRAMS SAMPLE GEONETRY: SAMPLE GEOMETRY '
EFFICIENCY FILE NAME: EFF.TABN
- ++***************++**************+*****+*************e.**
ACQUISITIOrt DATE: 19JUL79 1446:00
- FWHM(1332) 2.719 PRESET TIME < LIVE): 1800. SEC
- SEtJSITIVITY: .
16.000 ELAPSED REAL TIME: 1840. SEC
- SHAPE PARAMETER : 10.O %
ELAPSED LIVE TIME: 1800. SEC
- NBR ITERATIONS: 5.
- gr*******************************+.*:+e*******++*****************
- (
DETECTOR: ADC DETECTOR
- LISRARY:NUCL.LIBM DATE CALISPRTED: 19JUL79 744:23
- ENERGY TOLERANCE: 2.500KV KEV /CHNL: 1.0012299
- HALF LIFE RATIO: 8. 00 OFFSET: -0.7146829 KEV
- ASUNDAtlCE LIMIT: 50.00%
Q. COEFF. : -2.733E-07 KEV /C**2
- r
( %*******************************************:k*********.********
i I
EtlERGY HINDOM 99.406 TO 2000.652 PK IT ENERGY AREA BKGND FWHM CHANNEL LEFT PM CTS /SEC % ERR FIT t-t 1 1 510.85 20623. 13157. 3. 20 511.01 504 14 1.146E 01 1.1 4.51E C 2 1 602.67 1422. 8628. 2. 73 602.74 599 9 7.903E-01 9.6 1.29E _~
3 1 889.01 3427. 16571. 2. 52 888.84 885 9 1.904E 00 5.6 4.13E-4 1 1115.21 182305. 4137. 2.58 1114.89 1107 19 1.012E 02 0.2 1.79E C 5 1 1172.95 361. 236. 3.25 1172.59 1167 16 2.006E-01 8.0 3.87E r 6 1 1291.44 436. 257, 3. 09 1291.02 1284 19 2.424E-01 7.1 1.78E 7 1 1322.17 312. 117. 2. 63 1331.73 1326 13 1.732E-01 7.5 1.09E i 8 1 1407.33 141. 153. 3.31 1406.86 1401 18 7.819E-02 15.0 2.40E E 9 1 1460.48 137. 75. 3.04 1459.98 1456 10 7.626E-02 12.3 1.71ti .
10 1 1690.46 239. 56. 2. 63 1689.83 1683 16 1.226E-01 7.8 2.13E i PEAK SEARCH COMPLETED
(
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, 60-01 tiUCLIDE IDENTIFICATION SYSTEM
SUMMARY
OF flVCLIDE ACTIVITY PAGE I
. (
TOTAL LIf1ES IT4 SPECTRUM 10 LIr4ES NOT LISTED If4 LIBRARY 0 IDENTIFIED Iti SUf1 MARY REPORT 10 100.00% (
t ACTIVATIOtt GAS tJUCLIDE HLIFE HLSEC DECAY UC/UT ERROR % ERR 5 AR -41 1.82E 00H 6.588E 02 6.111E -4 7.646E -2 5.400E -4 7.06 i
ACTIVATIOff PRODUCT yg g 11UCLIDE HLIFE HLSEC DECAY UC/UT ERROR % ERR
'CO-60 5.26E 00Y 1.660E 08 5.556E -4 5.veca -2 2.792E -4 7.49 ZN-65 2.44E 02D 2.10SE 07 5.556E -4 4.971E O 1.190E -2 0.24 58-124 6.02E 010 5.201E 06 S.556E -4 1.061E -2 8.229E -4 7.85 SC-46 8.29E 01D 7.249E 06 5.556E -4 3.858E -2 2.153E -2 5.58
{C5-124 2.05Y 6.457E 07 5.556E -4 ** KEY LINE tiOT PPESErlT**
tJATURAL PRODUCT
( '40CLIDE HLIFE HLSEC DECAY UC/UT ERROR % ERR K-40 1.00E O2Y 2.156E 10 5.556E -4 2.210E -2 2.728E -212.25 k
t e
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THE UNIVERSITY OF MICHICAN
. PHOENIX MEMORIAL LABORATORY FORD NUCLEAR REACTOR ANN ARBOR, MICHIG AN 48109-(313) 764 6220 D T C,Nr.. (;k . r:=-tpb d> ~a OCT 10 13sb
.Chemtrol October 5, i979 Mr. Vincent Cataldo Chemtrol Corporation 8805 Solon Building G5 Houston, Texas 77064
Dear Vince:
I sent both short term and long term samples of CT-40-NS-Type B to Ron Estophon, Southwest Research Institute and G.L. Elam, Wyle Laboratories. G. L. ciam is Wyle's Radiction Control Officer. You might contact the octual user of the material to worn him to expect it. Enclosed are letten sent prece' ding the packages.
Sincerely, T LLW Reed R. Burn Reactor Manager Enclosures .
RRB/sv i
$f-[ . THE UNIVERSITY OF MICHIGAN Ih ,
PHOENIX t.82MORIAL LABORATORY FORD NUCLEAR REACTOR
/ ANN ARBOR, MICHIGAN 48109 d[ @,h (313) 764-6220 October 5,1979 Mr. G. L. Elam Radiation Control Officer Wyle Laboratories 7800 Governor's Drive, West Huntsville, Alabama 35807
Dear Mr. Elam:
( l have packaged and sent to you two irradicted samples of Chemtrol silicone polymer base neutron shielding material CT-40-NS-Type B. The samples are in lead pigs within the box. Each sample is icbeled on its lead pig. An irradiation data sheet is enclosed which provides irradiation details.
The samples are radioactive. Jose rates are provided on each lead pig.
Sincerely, i
h-Reed R. Burn IA, W Reactor Wanager i Enclosure RRB/sv
) xc: Vincent Cataldo Chemtrol Corporation I
l l
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ATTACHMENT C k.
o .
- _ ___ _~
April 16, 1980 Mr. Vincent M. Cataldo Chemtrol Corporation hQq D
P.O. Box 38556 )T Houston, TX 77088
Dear Mr. Cataldo:
In response to your recent telephone inquiry about the labeling of Sample No. 16705 (CT-40NS Type B), designated as irradiated with 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> exposure per R. G. Niemi's letter of March 3, 1980, I have talked further with Mr. R. R. Burn at the University of Michigan and also rechecked our records.
Mr. Burn sent us two samples on December 7, 1979; one irra-diated 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br />, the other SAS.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. In recording these samples, our sample clerk assigned both of them the same sample number (16705). Thus, we do not know wnich of these two samples was actually analyzed. Since the identification
( of the material labeled 16705 is ambiguous, I suggest that you simply delete that data from your consideration. The sequence of samples is therefore as fellows:
Sample 12751 - Non-irradiated 12750 - Irradiated 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> 17296 - Irradiated 545.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> I regret the inconvenience that this confusion has caused, and hope the data will now be useful to you.
Yours very truly, 3
/ -
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.M' Q)2 A. Lee Smith Manager Analytical Services Dept.
ALS/jlm .
xe: Mr. Reed R. Burn, Reactor Mngr.
Ford Nuclear Reactor Phoenix Memorial Laboratory Ann Arbor, MI 48109 R. G. Niemi (Dow Corning Corp.)
DOW CORNING CORPORATION, MIDLAND, MICHIGAN 48640 TELEPHONE 517 496-4000
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I i OCT ia cl3 October 8, 1979 Chemtrol 6
Mr. Vincent M. Cataldo Chemtrol Corporation P.O. Box 38556 Houston, TX 77088
Dear Mr. Cataldo:
We received the two cured samples of SYLGARDe 170 Elastomer which you submitted specially formulated with baron carbide filler and designed for neutron shielding applications. As per your comunication and designation, sample B-7 has undergone neutron and gama irradiated exposure at the University of Michigan. For comparison, your sample designated B-12 represents a non-irradiation control. Both samples were submitted to our Analytical Department (Dr. A. '.. Smith) for specific element and specific gravity analysis. These results, along with our Analytical Department reference numbern and Corporate Test Methods (CTM) are listed below:
B-7 (12750) B-12 (12751)
CTM Element / Property Irradiated Non-Irradiated 0030 Carbon (C) 14.22 .14 17.77 .17 0030 Hydrogen (H) 3.00 .09 4.38 .12 j
0522 Silicon (Si) 41.2 0.2 40.5 0.2 0540 Specific Gravity 1.571 .005 1.359 .005 I' hope that this analytical data will satisfy your requirements. Let me know if you have any questions.
l Very truly yours, s.
R. G. Niemi Sr. Specialist Elastomers, Technical Service and Development Phone: (517)-496-5380 sdk l
l DOW CORNING CORPORATION, MIDLAND, MICHIGAN 4864o TELEPHONE 517 496-4000 L
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FAWM hhPR'rno March 3, 1980 'Q :"::: W e u iy p.
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Mr. Vincent M. Cataldo Chemtrol Corporation ,
Cheratiu/
P. O. Box 38556 Houston, TX 77088 ,
Dear Mr. Cataldo:
We received the two additional samples of cured SYLGARDe 170 Elastomer that was specially fonnulated with bomn carbide by Chemtrol (earlier samples were designated B-7 and B-12). These latest samples were designated by Chemtrol as CT-40 NS Type B and were respectively subjected to 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> and 545.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of neutron and gamma irradiation exposure at the University of Michigan. These samples were submitted to our Analytical Department (Dr. A. L. Smith) for specific elemental and specific gravity analysis. These results, along with our Analytical Department reference numbers and Corporate Test Methods (CTM) are listed below:
(17296) (16705)
CT-40 NS Type B CT-40 NS Type CTM Element / Property 545.6 Hours Exposure 110 Hours Exo ure hr 0030 Carbon (C) 5.96 0.0M 4.97 0@5%
0030 Hydrogen (H) 1.07 0.03% 0.83 0.02%
0522 Silicon (Si) 40.2 0.5% 38.(# 1.0%
0540 Specific Gravity 1.681 0.005 fhsufficient Sample Again, I hope that this analytical data will satisfy your requirements.
Let me know if you have any questions.
Very truly yours, fkf. OH R. G. Niemi Sr. Technical Specialist Elastomers, Technical Service '
and Development Phone: (517)-496-5380 ,
sdk i
DOW CORNING CORPORATION, MIDLAND, MICHIG AN 48640 TELEPHONE 517 496-4000 I
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