ML20151E166
| ML20151E166 | |
| Person / Time | |
|---|---|
| Site: | 05000141 |
| Issue date: | 07/06/1988 |
| From: | Garcia E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Wong J CALIFORNIA, STATE OF |
| References | |
| NUDOCS 8807250393 | |
| Download: ML20151E166 (4) | |
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JULC 1988 Mr. Jeff Wong - -
California Department of Health Services Environmental Radiation Surveillance 2151 Berkeley Way Room 133 Berkeley, CA. 94704
Dear Mr. Wong:
Per your telephone request of July 5, I am enclosing a copy of the NRC order terminating the facility license of the Stanford Pool Reactor and some associated documents. You will note that this license was terminated in 1983.
If I can be of any further assistance, please let me know.
Sincerely, Emilio M. Garcia Emergency Response Coordinator Enclosure as stated cc w/ enclosure:
John Hickman California Department of Health Services Environmental Radiation Surveillance 714-744 P Street,' Room 616 Sacramento, CA. 95814 88072503?3 000706
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NUCLEAR REGULATORY COMMISSION RECCP.O B .l WASHINGTON, D. C. 20555 E80
/ June 21, 1983 l'I3 J.':: 2 7 FM 2 f.F
- Docket No. 50-141 "~
l I;EGiCiiV CE l
Dr. Roland A. Finston, Director Health Physics and Biosafety Stanford University 67 Encina Hall -
Stanford, California 94305
Dear Dr. 'Finston:
The Comission has issued the enclosed Order that terminates Facility License No. R-60 for the Stanford Pool Reactor in accordarce with your application dated September 20, 1973 as supplemented by letters dated November 19, 1973, August 9,1976, December 9,1977, September 13, 1978, June 3,1980, April 28,1982, and January 4,1983. Our related Safety Evaluation is also enclosed.
In connection with this action, we are enclosing two copies of Amendment No.11 to Indemnity Agreement No. E-19. Please sign and return one copy to this office.
A copy of the Order is being filed with the Office of the Federal Register for publication.
Sincerely.
M*
Dennis M. Crutchfield, Chief Operating Reactors Branch f 5 Division of Licensing
Enclosures:
- 1. Order Terminating Facility License
- 2. Safety Evaluation (with Attachments 1 & 2)
- 3. Amendment No.11 to Indemnity Agreement No. E-19 cc w/ enclosures 1 and 2 only:
See next page 1
Dr. Roland A. Finston June 21, 1983 cc w/ enclosures l'and 2_pnly:
. John B. Martin, Regional Administrator Nuclear Regulatory Commission, Region V 1450 Maria Lane Walnut. Creek, California 94596 Attorney General 555 Capitol Mall Sacramento, California 95814 California Department of Health ATTN: Chief, Environmental Radiation Control Unit Radiologic Health Section 714 P' Street, Room 498 Sacramento, California 95814 Sacramento _ County Board of Supervisors 827 7th Street, Room 424 Sacramento, California 95814 H. E. Book, Chief Radiological Saf' .y Branch Nuclear Regulatory Commission, Region V 1450 Maria Lane, Suite 210 Walnut Creek, r.alifornia 94596 t
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7590-01
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UNITED STATES T 'i(, NUCLEAR REGULATORY COMMISS!ON v WASHINGTON, D. C. 20555 k.'..',/ e
.. STANFORD UNIVERSITY DOCKET NO. 50-141 ORDElf TERMINATING FACILITY LICENSE By application dated September 20, 1973, as supplemented by letters dated November 19, 1973, August 9,1976, December 9,1977, September 13, 1978, June 3,1980, Ar-il 28,1982 and January 4,1983, Stanford University (the licensee) requested authorization from the Nuclear Regulatory Comission (the Comission) (NRC) to dismantle the Stanford Pool Reactor (the fccility),
a research reactor. located on the University's campus near Palo Alto, California and to terminate Facility License No. R-60. The authorization would allow the licensee to dismantle the facility, dispose of the component parts in accordance with the plan submitted as part of the application, and tenninate Facility License No. R-60. Authorization to dismantle the facility and dispose of its component parts was given by the Comission's Order dated May 12,1974. A "Notice of Proposed Issuance of Order Authorizing Termination of Facility License,' dated January 18, 1978, was published in the Federal
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P.egister on January 26, 1978 (43 FR 3634). No request for a hearing or petition for leave to intervene was filed following notice of the proposed action.
The Comission has found that the facility has been dismantled and decontaminated, and that satisfactory ciisposition has been made of the components parts and fuel in accordance with the Comission's regulations l
in 10 CFR Chapter 1, and the Comission's Order dated May 12,1974, and in a manner not inimical to the comon defense and security or to the health and safety of the public, i
agn(n26n W h SM w-
7590-01 2
The facility area._ bas been inspected by Nuclear Regulatory Comission (NRC) Region V inspectors and radiation surveys confinn that radiation levels meet the values defined in the dismantling plan, as revised by NRC letters dated March 15, 1981 and April 21, 1982 and the area is available for unrestricted access.
Therefore, pursuant to the application filed by the Stanford University, Facility License No. R-60 is hereby terminated as of the date of this Order.
For further details with respect to this action see (1) the application for authorization to dismantle facility and dispose of components parts and for termination of facility license dated September 20, 1973, as supplemented by letters dated November 19, 1973, August 9,1976 December 9,1977, September 13,1978, June 3,1980, April 28 '
982 and January 4,1983,(2)
- the Comission's Order Authorizing Dismant. ag of Facility and Disposition of Components Parts dated March 12, if/4, and (3) the Comission's Safety l Evaluation related to the termination of license. Each of these items is available for public inspection at the Comission's Public Document Room, l 1717 H Street, N.W. , Washington, D.C. A copy of items (2) and (3) may be l obtained upon request addressed to '5e U. S. Nuclear Regulatory Comission, l
I Washington, D.C. 20555, Attention: Director, Division of Licensing.
l Dated at Bethesda, Maryland, this 21 day of June 1983.
1 FOR THE NUCLE REGULATORY COMMISSION
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4 Darrell LLVV\ %b Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation g
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- g'o g JNITED STATES w NUCLEAR REGULATORY COMMISSION g .E WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING ORDER TERMINATING FACILITY LICENSE NO. R-60 STANFORD P0OL REACTOR DOCKET N0. 50-141 1.0 -INTRODUCTION By application dated August 9,1976 as supplemented by letters dated December 9, 1977, September 13,1978, June 3,1980, April 28,1982, and January 4,1983 Stanford University (the licensee) provided radiation measurements and analysis of the dismantled Stanford Pool Reactor and requested that Facility License No. R-60 be terminated.
2.0 DISCUSSION AND EVALUATION The Stanford Pool Reactor has been dismantled and disposed of in accordance with the licenses's dismantling plan dated November 19, 1973, and our dis-inantling order dated March 12, 1974. Measurements of surface contamination at the dismantled facility are consistent with Regulatory Guide 1.86, "Termination of 4 erating License for Nuciaar Reactors," and gama radiation from activated ceoponents is consistent with our criteria of March 17, 1981, as supplemented by letter dated April 21, 1982. On September 16 and November 5,1982, the NRC Region V inspectors conducted a radiation survey of the reactor facili y (Attachtr.c..t L.1) and determined that the facility met the surface contat.iination limits of Regulatory Guide 1.86. The inspectors also determined that gama radiation from activated components met our gama radiation criteria of 5 micro R/hr above background at one meter, except for inside the concrete reactor shield where the gama level was a maximum of 20 micro R per hour above background. The licensee has shown, however, that the potential occupancy inside the shield is such that the maximum exposure to an individual would be less than 10 cRem per year and would therefore meet ou annual exposure criteria for release of a facility to unrestricted access.
Our review of the licensee's request for license termination focused on consideration of their estimate of potential occupancy of the reactor shield area since the facility meets our surface contamination criteria and gama level criteria in all other facility areas. Our review of the licensees analysis of the shield activity and potential occupancy is discussed below.
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The activity in the concrk_te shield has been analyzed by the licensee thrwJgh use of core samples (letters from Stanford University dated December 9, 1977 and June 3,1980). The core sample results showed that the only significant reactor generated isotopes in the shield were Cobalt 60 and Europium 152 with
' half lives of 5.2 and 13.4 years, respectively. The concentration of these isotopes decreased with depth and distance from the core as would be expected for activation products. The NRC inspectors also identified a small per-centage of Europium-154 during their in-situ gama spectrum. measurements at the facility. The licensee later confirmed (letter dated January 4,1983) that Europium-154 was present in their core samples in the ratio of 1 to 17 with respect to Europium-152. Since Europium-154 has an 8.2 year-half-life and a ratio of 1 to 17 with respect to Europium-152, it ;as no impact on the conclusions of this safety evaluation.
By letter dated April 28, 1982 the licensee provided a review of the potential exposure to individuals from the residual activation in the shield.
The licensee estimated that because of the physical structure of the bio-logical shield that potential occupancy in a location, where exposure could occur, would be only a small fracticn of the time required to accumulate an exposure of 10 mrem per year.
We have evaluated the licensee's analysis of potential exnasure from activi-ation products and have determined that the 31/2 foot craw 1way opening into the inside of the shield structure would make it impractical for anyone to consider using the small space inside for a residence, laboratory, workshop or office. Also, the space inside the shield is only 7 feet, 71/2 inches in diameter, has no ventilation, no lighting, and no electricity. These factors would, also, make the shield ir. ractical for useful yace. Therefore, we estimat that occupancy of this space by individuals would be no more than a few hours per year. If we assume a conservative occupancy of the inside of the reactor shield at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> per year, and a dote rate of 20 micro Rem /hr, the yearly dose equivalent exposure would be about i.,.4 mrem or a small fraction of our 10 mrem per year criteria. The dose equivalent rate and com-mensurate annual axposure will decrease as discussed it. Attachment 2 hereto and in accordance with Figure No.1 of Attachment 2.
The licensee has stated that the reactor shield " in a building on University grounds which are part of a land grant and, therefore, the property can not be sold. Moreover, the licensee has stated, and we agree, that the industrial hazard and the expense of chipping away the slightly activated concrete would greatly exceed the benefit of the small reduction in the exposure rate inside the reactor
All fuel has been removed from the site and transferred to the Department of Energy (DOE) facilities in Idaho for reprocessing. As discussed above, the facility has been dismantled in accordance with the licensee's dis-mantling plan and our Order dated March 12, 1974. Furthennore, ' the residual activity meets our criteria for surface contamination and potential exposure to individuals.
3.0 CONCLUSION
We have concluded that the termination of License No. DPR-60 involves an action, which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR 551.5(d)(4), an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the terminating order.
Based on the reasons discussed above, we conclude that the license termination is in compliance with the Commission's regulations, that there is reasonable assurance that the license can be terminated without endangering the health and safety of the public and that the termination of license will not be inimical to the comon defense and security or to the health and safety of the public.
4.0 ACKNOWLEDGEMENT This evaluation was prepared by. P. Erickson and S. Block.
Attachments:
- 1. Inspection Report (memorandum dated March 11,1983 from H. E. Book NRC Region V, to
. P. Erickson, NRR) l 2. Dose Equivalent Rate and j Comensurate Annual Exposure Date: June 21, 1983 l
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.- Attachment 1
- [ pag}
UNITED STATES y s, g NUCLEAR REGULATORY COMMISSION g ;e REGION V Q 1450 MARIA LANP,SUliE 210 WALNUT CREE K, CAllFORNIA 94596
% * * " * ,e,4 MAR 111903 MEMORANDUM FOR: Peter R. Erickson, Project Manager Standardization and Special Projects Branch, NRR FROM: H. E. Book, Chief Radiological Safety Branch, Region V
SUBJECT:
Close-Out Inspection for the Stanford University Research Reactor, Docket No. 50-141 Pursuant to your telephone request of June 1,1982, we have conducted a close-out inspection of the Stanford University Research Reactor. The inspection report is enclosed.
Based on the findings of the inspection, including the results of independent measurements performed by my staff, we confirm that the facility is not contaminated above the limits of Regulatory Guide 1.86 Furthermore, the radiological conditions are as described in the licensee's coninunications to your office of August 9,1976, December 9,1977, June 3,1980 and April 28, 1982, with the exception that we identified the presence of Europium-154. We do not believe that this finding significani,1y changes the impact on the public's health and safety. It also appears that the licensea's estimate for the biological shield occupancy is reasonable. We concur that the facility may be released for unrestricted use.
. f, G H. E. Book, Chief Radiological Safety Branch
Enclosure:
As stated l'
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U. S. NUCLEAR RECUL\ TORY COMMISSION REGION V Report No. 50-141/82-02 Docket No. 50-141 treens, no, R-60
_ safeguards croup Licensee: Stanford University 67 Encina Hall Stanford, California 94305 racility name: Stanford Research Reactor .
Inspection at: Stanford, California Inspection conducted: September 16 and Novembed,1982 and followup telephone calls con suairyg on marco A, noo -
Inspectors: @/. 3 //O/gj
_E. M. Garcia, Radiation Specialist Date signed
- hh. ; 3 - \C)- M G. WI Temple, Radiat' ion Technician (I.nstrumentation)
, Date signed Approved by-Date Signed Approved by:
F. . Wenslawski, Uniet J //o /73 c date' Signed orgadiatgn Rratection Section s &- G cro-L.
H. t. book, Unlet, Kaolological daT ecy oraric
&ll0llO Date Siped
, , , - Sumary: .
Inspection on September 16 and November 5,1982 and followup telephone calls concluding on March 1,1983_ _(Report No. 50-141/82-02)
- Areas Inspected: Special announced closeout inspection by regionally based l inspectors of the Stanford Research Reactor. This inspection was to verify l that tiie release criteria, as presented in the letter of March 17, 1981 from i
John F. Stolz, Division of Licensing to Dr. Finston, and clarified in the letter of April 21, 1982 from James R. Miller, Division of Licensing had been satisfied. The inspection included review of records, interviews with
! personnel and independent measurements. The inspection involved 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> l onsite by two NRC inspectors.
Results: The inspection confirmed the licensee's findings as reported in l communications dated August 9,1976, December 9,1977, June 3,1980 and April 28, i 1982 with the exception that Europium-154 was identified as being present.
l The release criteria as identified in the paragraph above has been satisfied.
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- l RV Form 219 (2) l 90 iM9 .
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. .. DETAILS -
- 1. Persons Contacted R. A. Finston, Director, Health Physics and Bio-Safety J. A. Holmes, Senior Health Physicist Mr. J. L. Brown, Associate Health Physicist with the State of California, Department of Health Services, Radiological Health Branch, was present during the insnection of November 5,1982.
- 2. Release Criteria The radiation levels for release of a reactor facility to unrestricted access are established in Enclosure 1 to tne letter of March 17, 1981 from John F.
Stolz, Division of Licensing, to Dr. Finston. Two sets of levels are established; surface contamination, and radioactive material other than surface contamination.
Surfaces must be decontaminated to levels consistent with Table 1 of Regulatory Guide 1.'86. Radioactive material, other than surface contamination that may exist in concrete, components, structures and soil, must be removed such that the radiatiu level from those nuclides is less than 5 ur/hr above natural background, as measured at one meter fron: the surface. Natural background is defined as radi_ation from naturally occurring radionuclides measured at a comparable, uncontaminated structure or exterior soil surface.
In a letter dated April 21, 1982 James R. Miller, Division of Licensing, refines the release criteria:
"Since March 17, 1981, we have refined further our position with respect to release criteria and have determined that radiation from gama emitting isotopes is also acceptable if the potential exposure to individuals is less than 10 mRLn per year with reasonable occupancy assumptions. If you wish to justify gamma exposure rates from reactor generated isotopes that are greater than 5 micro Rem per hour, you should show that reasonable occupancy of that area would be sufficiently less than 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year, which would result in exposures of less than 10 mrem per year."
- 3. Surface Contamination In the licensee's letter of August 6,1976, the results of the surface contamination surveys are described. According to this report, the site was surygyed for
- fixed and removable contamination using a thin window GM and 100 cm wipes.
l Fixed contamination was measured with the GM at approximately 1 cm from l the surface. less than 0.03 mr/hr. The wipes l were reported asThe activity less was than 100 recorded dpm/100 cm ag' alpha, beta, and gama. Subsequent l to the site inspection the inspector noted two inconsistencies between the l licensee's survey results and Table 1 of Regulatory Guide 1.86 These inconsistencie involved the reportad lower limit of detection for removable alpha contamination
! and the units used (mr/hr) for average and maximum surface contamination.
l Afterreviewingtheirdatathelicenseedeclaredthatforrgmovablealpha i
contamination the lower limit of detection is 20 dpm/100 cm . In a followup letter to NRC Region V dated March 3,1983, the licensee reported thgt the "less than 0.03 mr/hr at 1 cm." represents less than 2500 dpm/100 cm above I background. The detector efficiency for Co-60 and conservative assumptions l were used in determining this limit of detection. The licensee's clarifications I eliminates the inconsistencies noted by the inspector.
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l The NRC staff collected a total of 67 surf ace smears from areas of the Ryan b Laboratory that had been part of the restricted area. These samples were analyzed on NMC PC-55 windowless proportional counter, serial number 77-2712-05.
The samples were analyzed for gross alpha and beta contamination. The instrument efficiencies for Pu-239 and Tc-99 were used in determining the activity.
At the 95% confidence level g contamination was less than 6 dpm/100 cmg alpha and less than 28 dpm/100 cm beta. These values are not statistically different from background.
The inspector surveyed the restricted area for gamma radiation. Measurements of direct radiation identified no residual activity above background, except for areas inside the biological shield (see paragraphs 4.a and 4.b below).
The background was measured at less than 5 + 1 ur/hr. These measurements were performed using Eberline PRM-7 micro-rTmeter, serial number 247, due for calibration on March 31, 1983. This instrument was calibrated using Cs-137.
- 4. Radioactive Material Other Than Surface Contagination
- a. Radiation Surveys The licensee had identified induced radiation in the concrete floor inside the biological shield. Based upon surveys conducted in 1977, the licensee had estimated a maxiiaum dose equivalent of 30 urems/hr at one meter from the floor of the tank. The NRC measured dose rate is a maximum of 20 ur/hr above background. This value was determined by measuring the dose ^ rate inside the biological shield one meter above the area where the maximum surface dose rate was measured (photograph 4),
and subtracting the background dose rate measured in a similar area of the building (photograph 5), that was not part of the restricted area.
The maximum surface dose rate area was identified using the micro-r-meter described in the paragraph above. The dose rates at a meter were determined ut ing a Reuter-Stokes Environmental Pressurized Ion Chamber, Model RSS-111, Serial Number Z-3999. The ion chamber had been calibrated on July 26, 1982 using a Co-60 standard. Traces from the strip chart for background and biological shield are Charts 1 and 2 respectively. For conservatism the background value was truncated down to 5 ur/hr and the biological shield value rounded up to 25 ur/hr.
- b. Nuclide Identification The release criteria as stated in Enclosure 1 to the NRR letter of March 17,1981, specifies the nuclides to be considered when determining the dose rate. To this end, the licensea had collected and analyzed core samples from the concrete floor. In the licensee's letter of June 3,1980, Potassium 40 (K-40), Cobalt-60 (Co-60), and Europium-152 (Eu-152) are reported as the nuclides found in the core samples. The
NRC, staff gathered an in situ garrrna spectrum to independently and qualitatively determine the nuclides present (photograph 6). Besides the two activation products identified by the licensee, Co-60 and Eu-152, . (K-40 is ~a naturally occurr ng nuclide), the NRC staff identified Europium-154 (Eu-154).
After discussions with the inspector, the licensee reviewed their data and in a. letter to NRR (January 4,1983, Dr. Finston to P. Erickson) confirmed the presence of EU-154. The presence of Europium-154 in a very small quantity will slightly increase the time before the dose equivalent rate is less than or equal to 5 urem/hr. This has no significant increase in the impact on the public's health and safety.
The in situ gamma spectrum was gathered using a portable intrinsic germanium detector and a Nuclear Data ND-6 portable multichannel analyzer. These instruments were on' loan from Nuclear Data. The spectrum was analyzed using NRC's Nuclear Data ND-6620.
- c. Occupancy In the letter of April 28, 1982 the licensee estimated that the annual occupancy of the biological shield would be much less than 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />.
The reactor facility is located in a building called Ryan Laboratory (photograph 1). Although the building is primarily used for the storage of furniture and other equipment, the Art Department uses some of the rooms as artist's studios. These rooms were not part of the restricted area when the reactor was operating. The only remaining reactor component is the biological shield (photograph 2). Access into the shield is either by climbing down from the top or through the thermal column rolling door (photograph 3). The rolling door ~is padlocked closed and the shield top is closed with nailed down plywood. The building has i
an electronic security system to detect unauthorize'd entry. The security I
provisions, condition of the building and the licensee's plans indicate that the licensee's occupancy estimate is reasonable.
- 5. Conclusions The evaluations and surveys conducted by the licensee and NRC indicate that the surface contamination criteria is satisfied. The dose rate due to radioactive l material other than surface contamination is greater than 5 ur/hr but due to the low occupancy rate the potential exposure to individuals is less than 10 mrem per year. The criteria for release of the reactor facility to unrestricted access appears to be satisfied.
l 6. Exit Interview At the conclusion of the inspection the inspector met with the individuals noted in paragraph 1. The scope and preliminary findings of the inspection were discussed.
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j Photograph 1 Outside of Ryan Laboratory Stanford University September 16, 1982 .
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View from Top of Biological Shield Looking into the Tank.
Environmental Pressurized Ion Chamber in the Center Bottom of Tank.
Septaber 16, 1982
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Note: Biological Shield Right Background.
September 16, 1982
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t Photograph 6 View of Portable MCA and Intrinsic Ge Detector Inside the Biological Shield Septaber 16, 1982 1
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The distance from the floor surface to the center of the chamber was 10 0 05 meters. The strip chart scale is 50 ur/hr full scale.
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Strip chart from Reuter Stoke's Environmental Pressurized Ion Chamber Model RSS-lli Serial No. 2-3999 calibrated on July 26, 1982. The measurement was taken on September 16,1982 be9 inning at 11:29 A.M. These measurements were made inside the Stanford reactor biological shield. The chamber was located above the maximum dose rate-area as located with a Micro-R-Meter. The distance from the floor surface to the center of the chamber was 1.! 0.05 meters. The strip chart scal'e is 50 ur/hr full scale.
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Attachment 2
.- [,m g"'q UNITED STATES
[ g NUCLEAR REGULATORY COMMISSION
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- E WASHINGTON, D, C. 20555 l
\., .....f DOSE EQUIVALENT-RATE AND COMMENSURATE ANNUAL EXPOSURE SUPPORTING ORDER TERMINATING FACILITY LICENSE NO. R-60 (DOCKET NO. 50-141)
Calculation of Annual Dose and Dose Rate as a Function of Time By letters dated June 3,1980 and January 4,1983, Stanford University (the licensee) analyzed concrete cores, taken from the floor of the reactor tank structure. A Germanium-Lithium (GeLi) spectrometer was used, and peak energies at Potassium-40, Cobalt-60, Europium-152 and a small amount of Europium-154 were found. Since the latter three radionuclides, Cobalt-60, Europium-152 and Europium-154 are neutron induced, our evaluation will be specifically directed to them. The licensee's specific activity analysis did not include Potassium 40, because it is a naturally occurring radionuclide.
Consequently, we will assume that the dose rate contribution from Potassium 40 is part of the background. We, also, did not include Europium-154 in our analysis because of its 1 to 17 isotopic yield as compared to Europium 152.
The Cobalt-60 and Europium 152 activity, found by the licensee's analysis, was 82 pCi/gm and 154 pCi/gm respectively. Now the gama function, y, in units of R/hr at 1 meter pe. curie, for these radioriuclides is as follows:
Cobalt 60 = 1.32 l
Europium 152 = 0.58 From Radiological Health Handbook, Page 131 l
- ,,' 0 The licensee measured ma.&_ i mum dose equivalent (DE) rate at 1 meter from the floor of the reactor shield was 30 micro R/hr in 1980. A maximum measurement of 20. micro R/hr at one meter from the concrete surface was obtained in November 1982 by the NRC. The dose rate contribution for each component of the radiation can be found from the specific activity of each radionuclide (pCi/gm) .and the value of y, as shown above, as follows
For Cobalt 60 ( A) = 82 X 1.32 = 108.2 For Europium 152 (B) = 154 X 0.58 = 89.2 The ratio of A/B = 108 = 1.21, which shows that the DE contribution of dose rate from Cobalt 60 is 1.21 times greater than Europium 152. Solving for the respective dose rate we get:
A + B = 20 prem/HR l and A = 1.21 B so that
- Cobalt 60 - 11.0 prem/hr Europium 152 = 9.0 prem/hr
.693 Corrections for decay, follow the equation Ic e-At, where A =M2, where il/2 is the half-life of the radionuclide of interest, and t is the time interval l
l under consideration (e.g., 5 through 20 years). Each component of the DE rate was corrected for respective decay, with Cobalt 60 T1/2 = 5.3 years, and Europium 152 T1/2 = 13.4 years, and summed to provide the curve shown in Figure 1. To determine the respective dose to an individual occupying the area for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> per year we get for t = 0:
1 O.020 mrem /hr x 20 hr = 0.40 mrem /yr.
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