ML20207J467
| ML20207J467 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 07/23/1986 |
| From: | Agosti F DETROIT EDISON CO. |
| To: | Berkow H Office of Nuclear Reactor Regulation |
| References | |
| VP-86-0092, VP-86-92, NUDOCS 8607290139 | |
| Download: ML20207J467 (57) | |
Text
-
ce Pre's ent i
Nuckar Operations OR rer-i n Edison 55BEF" W rat.L.
July 23, 1986 VP-86-0092 Mr. Herbert N. Berkow, Director Standardization & Special Projects Directorate Division of PWR Licensing-B U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Berkow:
i i
Reference:
- 1) Fermi 1,
NRC Docket No. 50-16 i
NRC License No. DPR-9 i
l
- 2) Detroit Edison to NRC Letter, " Amendment Request l
for Extension of the ' Possession Only' License for Fermi 1", NE-85-0714, dated May 17, 1985 1
- 3) NRC to Detroit Edison Letter, " Request for Additional Information", dated February 27, 1986 i
Subject:
Supplemental Information on Fermi 1 I
Reference 2 transmitted a request to amend the Fermi 1
" Possession Only" license to reflect an extension beyond the l
current expiration date. Reference 3 identified six questions l
which the NRC needed resolved prior to completing their review i
of the Detroit Edison amendment request. Detailed responses to l
the subject questions are provided in Enclosure 1.
A suasary of l
the responses is also provided as a forward to the detailed responses.
The enclosed information reflects procedures and practices currently employed to meet the requirements of the Enrico Fermi 1 " Possession Only" license and its Technical Specifications. While Detroit Edison is committed to l
implementing the requirements of this license, the tools (e.g.,
procedures) it uses to meet the requirements may change throughout the life of the license.
l The enclosed information should support both closure of the subject questions and issuance of the requested amendment to Fermi I License No. DPR-9.
6 0
72hh 8
p P
Mr. Herbert N. Berkow July 23, 1986 VP-86-0092 Page 2 Please direct any questions on the enclosed information to Mr.
R. L. Woolley at (313) 586-4211.
Sincerely, n
~.s.---.k. [
i cc:
Mr. J. Eckert (Monroe County OCP)
Mr. P. B. Erickson Mr. D. Hahn (State Health Department)
Mr. W. G. Rogers USNRC Document Control Desk Washington, D.C. 20555 t
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___ to VP-86-0092 DETROIT EDISON'S RESPONSE TO FERMI 2 LICENSE APPLICATION QUESTIONS TRANSMITTED IN A FEBRUARY 27, 1986 NRC LETTER a.
Summary of Responses b.
Detailed Responses l
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Enclorura 1 Pag 3 1
SUMMARY
1.
ALARA MEASURES o
All of the highly radioactive material (fuel, blanket material, cold trap components, equipment used to package the material and waste produced in the operations) were removed from the site in the 1973 to 1975 period, o
All access possibilities to the interior of the reactor primary shield tank and its surrounding concrete biological shield (nominal 3 foot wall) have been welded shut.
o The primary sodium (1344 drums representing 99 3% of the total inventory) was shipped to Argonne National Laboratory-West during the final quarter of 1984.
An ongoing ALARA program is accomplished by the application of Fermi-1 o
" Administrative and Surveillance Procedures" and Fermi-2 relevant Health Physics Procedures.
o If all of the liquid waste water contaminated by residual activity on the interior sides of the drain lines, pumps and tanks) were to be released to the environment, the maximum calculated whole body dose to an individual would be 0.031 mrem which compares with a whole body dose of 0.24 mrem received from natural background radiation over the same time period.
o If all of the residual sodium radionuclide inventory were released to the environment in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the maximum calculated organ dose to an individual at the site boundary would be.0004 mrem. This compares with 0.24 mrem received in the same period from natural background radiation.
(
2.
DOSE ASSESSMENT o
The total cumulative dose received at Fermi 1 for the period 1973 through 1985 was 21 3 man-rem.
o Of the total cumulative dose, 20.5 man-rem (96%) was recorded during the three year period (1973-1975) in which all fuel, blanket equipment, cold trap components, and equipment contaminated in the cut-up shipment, and clean-up operations were removed from the site.
o 0.2 man-rem (1%) was recorded during the interim six year period of surveillance and maintenance activities.
(1976-1981) o 0.6 man-rem (3%) was recorded during the last four year period (1982-1985) which includes the sodium drumming and shipment operations. Much of the dose received in 1985 (0.1 man-rem) was from I
operation of a TLD calibration facility located in the Fuel and Repair Building (FARB), a Fermi 2 activity.
Enclo;urs 1 Pags 2 o
With the sodium removed, an annual dose of 0.04 man-rem is estimated for maintenance, repair and surveillance activities, for a total of 1.6 man-rem during the next forty years.
Decommissioning at this time would be severely impacted by the lack of o
disposal site space in general and the reluctance of sites to accept equipment that has contained sodium.
Natural decay of the radioactive material (approximately 90% of the o
remaining inventory over the next forty years) and the development of improved radwaste disposition alternatives, adoption of de minimus levels, and the development of acceptable methods for the neutralization of sodium contaminated surfaces would provide a significant reduction in the work space, man-rem dose, and volume of radioactive waste resulting from decommissioning activities.
3 PROCEDURES Health Physics protective actions (personnel monitoring, bioassay and o
potential airborne contamination control and personnel protection) are provided through the Fermi-2 Radiological Control Program documents (Nuclear Operations Directives, Enrico Fermi-2 Plant Orders and Procedures).
Fermi-2 procedures are also used for handling and storage of sealed o
and unsealed byproduct materials.
o The fuel and blanket material, have been removed.
4.
REVISED AMENDMENT REQUEST provides a revision to Detroit Edison letter, NE-85-0714, o
dated March 17, 1985 I
l 5.
INVENTORY OF RADIONUCLIDES The inventory of radionuclides for the reactor components and primary o
biological shield was obtained by calculations.
Direct measurements or sample analyses could not be done without disruptive torch and jackhammer operations, which are not compatible with our ALARA
- program, i
o The following total activities were calculated:
i l
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,--..n-.,
Encle;ura 1 P g2 3 Reactor Primary Shield Reactor Vessel Tank Internals Radionuclide (Curies)
(Curies)
(Curies)
Nb-94 1.9E-06 4.1E-02 2.7E+02 Co-60 1 3E+00 3 9E+00 Ni-59 2.1E-03 8.7E+01 Ni-63 2.5E-01 C-14 9.5E-10 3 9E-10 Fe-55 5.1E-01 8.1E-02 1.1E+02 An assumption was made that the concrete contained 0.01% europium and a value of 2.8E-03 curies was obtained for Eu-152.
In forty years the activities will have reduced to the following curie o
values:
Concrete Reactor Primary Shield Reactor Radionuclide Shield Vessel Tank Internals 4.1E-02 Nb-94 1.9E-06 1.41E+00 Co-60 6.7E-03 3 9E+00 2.1E-03 Ni-59 6.4E+01 1.9E-01 Ni-63 C-14 9.5E-10 3 9E-10 F-55 1.8E-05 2.8E-06 2.5E-03 Eu-152 3 4E-04 This constitutes about a 80% reduction in the radionuclide inventory for those parts that are inaccessible.
The inventory of radionuclides in the waste water was based on o
measurements of samples from the radwaste sump, tank dose rates and activity measurements for the final total discharge after the 1973 to i
1975 decommissioning work. The present total activity for Co-60 and j
Cs-137 was estimated at 6.0E+3uCi each. This activity will decrease to 3 1E+1uCi for Co-60 and 2.4E+3uci for Cs-137 in forty years, a j
reduction of 80%.
o The residual sodium radionuclide inventory was measured (random sampling of drums) and found to be 9.8E+02uCi for Na-22 and 4.8E+3uCi for the Cs-137 After 40 years the activity reduces to 2.5E-02uci and 2.0E+03uci respectively. This amounts to a 65% reduction in activity.
Significant inventory reductions will be obtained by the Safstor o
license extension of forty years.
Enclorura 1 Paga 4 6.
RELEASE CRITERIA APPLICATIONS o
A radiological survey was performed of all accessible areas within the protected area.
o The environmental background radiation dose rates as measured in this region of Michigan range from 10-11 micro R/hr.
Presently the first floor and mezzanine of the FARB (with the o
l exception of some areas over the floor drains in the decay and cut-up pool rooms), the machine pit in the FARB, cask car trestle shed, operations floor in the Reactor Dome, second floor of the Sodium Building and all outdoor areas are within the release criteria as defined in letter from James R. Miller (USNRC) to Dr. Roland A.
Finston (Stanford University) Docket No. 50-141, dated April 21, 1982
("... 5 Micro Rea per hour at one meter for reactor generated, gamma esitting isotopes").
o In forty years, the decay and cut-up pool rooms and Reactor Building basement area will be at the unrestricted release dose rate criteria.
{
o The bottom of the decay pgol will be at or near unrestricted release criteria limit in forty years.
o Dose rate reductions in the systems that could be monitored (outside of the primary biological shield wall) will be reduced by 93% to 995 i
in the liquid waste systems and about 89% in the sodium and gas decay systems.
4 CONCLUSION:
Based on this review of radiological conditions at Fermi 1, it is concluded that deferral of decommissioning for forty years will result in tangible benefits and cost savings without significant adverse impact upon the l
health and safety of this public or Detroit Edison employes, f
1.
Resultant potential exposure to the public from a release of the available radionuclide inventory is estimated to be much less than received from natural background radiation.
2.
Resultant exposure to employes from routine maintenance and housekeeping activities for the forty year period is estimated as less than 1.6 man-rem.
3 Resultant reduction in radionuclide inventory from radioactive decay during the forty year period is estimated at 80%, with exception of Na-22, which is estimated at 65%.
4.
Associated with the reduction of radionuclide inventory, commensurate reductions are expected fcr:
a.
radiation levels b.
radiation exposure (man-res) c.
radwaste volume d.
work scope e.
decommissioning costs.
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4 DETAILED RESPONSES 1
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QUESTION 1 - ALARA:
1 Discuss specific measures which will be utilized during the forty cdditional years to maintain radiation exposures and releases of radioactive materials to unrestricted areas as low as reasonably achievable (ALARA).
Discuss, also, the potential for a release of radioactivity during the safe st rage period, and estimate the potential exposure to the public through v;rious radiological pathways during this period.
l
RESPONSE
Specific measures for the maintenance and control of radiation exposures and releases of radioactive materials to unrestricted areas are contained in th2 manual " Decommissioned Enrico Fermi Unit-1 Reactor and Associated Building cnd Equipment-Administrative and Surveillance Procedures". The manual is j
rsviewed annually by the Enrico Fermi Unit-1 Site Review Committee. Any rsvisions to the manual are reported in the Annual Report to the NRC.
Every six months the Audit Subcommittee performs an inspection and review of records and evaluates compliance with connitaents for periods of curveillance.
4 The Plant Manager of Fermi 2 has been appointed by Corporate Management as i
Custodian for Feral 1 during its SAFSTOR status. Two Custodial Delegates are cppointed or reaffirmed at the annual Review Committee meeting to act on behalf of the Custodian.
The initial implementation of an ALARA program for the SAFSTOR period of j
Fsrai-1 was the removal and offsite disposition of the fuel elements, blanket l
naterials, sodium cold trap system and all equipment used in the cut-up and j
shipment. The reactor vessel was welded shut and all access points to the below operating floor primary biological shield and primary reactor containment were sealed by welding. Figure 1 provides a plan view of the Fermi 1 site.
The ALARA program is maintained by the following Administrative Procedures.
It should be noted that in some instances the following procedures exceed the requirements of the Fermi 1 Technical Specifications. These procedures may be revised in the future to reflect surveillance frequencies required by the Technical Specifications.
1.
All access points to the Controlled area are kept locked. The access
[
to the secondary reactor shell is also controlled with a padlock.
Only two keys are available for the locks. One key is kept in the Plant Managers safe and the other in the critical key cabinet in the Ferai-1 Control Room. Unescorted access to the controlled area aust be approved by the Custodian or his delegate.
i 2.
Both Thermoluminscent and Direct Reading Dosimeters (DRD) are supplied for all persons who have been granted permission for unescorted access. Visitors when properly escorted and approved are issued DRDs.
I 1-1
3 Periodic test and commitment compliance activities are initiated by surveillance work orders (PN-21's) which are issued by the Custodial Delegate.
4.
. The following weekly tests are performed:
a.
General walk-through and inspection of the controlled (or protected) area.
b.
Continuity test of the water intrusion alarm circuits.
(Detectors are in the containment building lower level sump, the waste water sump and the biological shield annulus around the containment building).
c.
Observation of the CO2 cover gas pressure over the essentially empty sodium storage tanks.
(1,344-55 gallon drums of primary sodium were shipped to Argonne National Laboratory-West during the last few months of 1984. Small remnants of sodium remain in the pipes, tanks, pumps and reactor.)
The cover gas (CO ) pressure in the reactor is also checked and d.
2 recorded in the containment building.
If radioactive liquid waste were to be discharged (a management e.
decision was made to not use this option), surface waters must be monitored (see semi-annual test).
5 The following monthly surveillances are performed:
a.
Check and record the volume of liquid in the liquid waste tanks.
(Intrusion of rain water into the FARB caused one tank to fill, this was corrected over three years ago by the placement of a concrete apron on the West Side of the FARB).
b.
Detailed inspection of protected area and report to the Custodial Delegate of potential problems.
c.
Check water levels in all active sumps.
6.
The following quarterly surveillances are performed:
a.
Radiation and Smear Survey of FARB rooms.
b.
Radiation and Smear Survey of Reactor Building.
7.
The following semi-annual surveillances are performed:
Twenty radiological environmental sample analyses of raw surface a.
water and sediment around the plant environs and raw city water are performed by an outside contractor.
b.
Physical tests (wet compress application) are performed of the water intrusion alarms at the detectors.
The Hi/Lo pressure alarms for the reactor cover gas are tested.
c.
1-2
8.
The reactor carbon dioxide cover gas pressure relief valve is tested annually.
t Our ALARA program is supported by application of the associated procedures cf Fermi-2 which apply to Fermi-1. With the exception of drains and the sump, cll activity in Fermi-1 is sealed from general access of personnel.
The only possible pathways of exposure to the public would be through the disruption of the radwaste system and loss of fluid to the lake or total loss of the sodium systems (reactor vessel, pumps & heat exchangers, piping, and' stcrage tanks) with exposure of the sodium to water. Assuming a total lost of inventory, the potential exposure to the public could be determined through use of the Fermi-2 offsite dose calculation methodology. The results are gNen below:
Two primary sources of potentially mobile radioactive material are left in pitce, at Fermi 1:
1.
The MK9 and MK15 liquid waste tanks, containig a total of 7550 lons (2.86 E7 al) of liquid with 6 mci of Co and 6 mci of gCs.
22 residues containing a total of 0 98 mci of Na and 4.84 aCi SodgCs.
2.
of Two worst case accident scenarios are postulated:
1.
The simultaneous rupture of both tanks, resulting in the release of the contents to Lake Erie.
2.
A fire or other catastrophic event, resulting in the atmospheric release of the residual sodium, including the entire radionuclide inventory, over the course of 1 day.
Th::se scenarios are reviewed, below:
I.
Liquid Release f
In this case, it is assumed that the contents of the tank, after dilution in Lake Erie, reach the Monroe water intake, approxiinately 2 i
miles south of Fermi 1, and that an individual drinks the diluted effluent as his sole source of drinking water for one day.
The concentration of each radionuclide in the tank
- 2.1 E-4 uCi/ml.
The established dilution factor to the Monroe water intake (Fermi 2 FSAR, Appendix 11A, page 9) is 77 Hence the concentration of each radionuclide at the water intake is 2.73 E-6 uCi/ml.
I
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1-3 1
From " Standard Man" data (Radiological Health Handbook, page 217), an adult man drinks 150 ml of tap water per day. Hence, if his sole source of tap water consists of diluted effluent, his total intake of each radionuclide would be 409 pCi.
Using the " Adult Ingestion Dose Factors", DF, of Table 1.2-2, of i
the Fermi 2, Offsite Dose Calculation Manual:
A. 60Co 0
DFi (ares /pC1)
Dose (ares) l 3
Liver 2.14 E-6 8.75 E-4 GI 4.02 E-5 1.64 E-2 Whole Body 4.72 E-6 1 93 E-3 B. 137Cs O
Dg (aren/pCi)
Dose (arem) g Bone 7 97 E-5 3 26 E-2 Liver 1.09 E-4 4.46 E-2 Kidney 3 70 E-5 1.51 E-2 Lung 1.23 E-5 5.03 E-3 GI 2.11 E-6 8.63 E-4 Whole Body 7.14 E-5 2 92 E-2 Hence, considering both radionuclides, the total whole body dose would be 0.031 arem, while the highest organ dose, to the liver, would be 0.045 area. This compares favorably with the dose of 0.24 mrem that this same individual would receive from natural background, assuming a natural background dose rate of 10 ures/hr.
II. Airborne Release I
In this case, the accident is assumed to release the entire inventory of radionuclides contained in the residual sodium, at an even rate, overthecourseof1 day (8.64E4sggonds). The resulting release gCs.es, Qi are 1.13 E-2 uCi/sec for Na, and 5.60 E-2 uCi/sec for Assuming the unlikely, but conservative, scenario that an individual spends the entire 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the site boundary in the sector with the most conservative dispersion factor, then the Fermi 2 Offsite Dose Calculation Manual (ODCM), Section 2.2.1.b, determines the l
organ dose Do, to that individual as:
Do = (x/q)(P o)(Qi) i where x/qigtheatmosphericdispersionfactor=4.186E-6 sec/m Qi is the isotopic release rate, as shown, above, in uCi/sec Po is the organ dose parameter for radionuclide, i i
= (K)(BR)(DF o) i l
1-4 l
where K = 1 E6 pCi/uCi BR is the adult breathing rate = 21.9 m3 ay
/d DF o is the adult inhalation dose factor in terms of i
area /pCi TheDFofactorisgenegilytakenfromtheODCM-Table i
2.2-4 for adults. For Na, however this factor is not available, so this information is obtained from ICRP Publication 30.
1 Substituting, Do = (91.7)(DF o)(Qi) i A. 22,
y Qi = 1.13 E-2 uCi/sec Do = 1.04 DF o i
Then:
9E.gan_
gio(aren/pC1)
Do(aren)
Gonads 6.7 E-6 7 0 E-6 Breast 5.9 E-6 6.1 E-6 Bone 1 3 E-5 1.4 E-5 Lungs 9 3 E-6 9.7 E-6 Thyroid 5.9 E-6 6.1 E-6 GI 7.0 E-6 7 3 E-6 Whole Body 8.1 E-6 8.4 E-6 B. 137Cs l
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Qi = 5.6 E-2 uCi/sec l
Do = 5.14 DF o i
Then:
Organ gi.o(arem/pC1)
Do(area)
Bone 6.0 E-5 3 1 E-4 Liver 7.8 E-5 4.0 E-4 Kidney 2.8 E-5 1.4 E-4 Lung 9.4 E-6 4.8 E-5 GI 1.1 E-6 5.7 E-6 Whole Body 5.4 E-5 2.8 E-4 This compares favorably with the dose of 0.24 mrem that the individual would receive from natural background during that same time period, assuming a natural background dose rate of 10 ures/hr.
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NOTE A: BUILDING HAE BEEN DISMANTLED ONLY SLAB REMAINS FIGURE 1 l
FACILITY PLAN l
QUESTION 2 - DOSE ASSESSMENT:
a) Provide a table showing Fermi Unit 1 personnel exposure experience for the years 1973 thru 1985 showing the man-rem exposure by facility structure (i.e., reactor building, etc.), regardless of how these exposures were obtained (e.g., decontamination, repair, etc.) or by whom (e.g., by plant personnel, plant maintenance personnel, contractor / vendor personnel, etc).
b) Provide a similar table of anticipated exposures for the years 1985 to 2025 from expected decontamination, decommissioning, additional maintenance and repair operations.
RESPONSE
Exposure experience for 1973 through 1975 involved all plant personnel (operations, maintenance and health physics) for removal of all high activity material and equipment and area decontamination afterward. Most of work resulting in personnel exposure (20.54 man-rem) was performed in the Fuel and Ripair Building (FARB) and the cold trap room in the Sodium Storage Building.
Monthly Man-rem is tabulated for this period in Table 2-1.
Total Man-Rem for tha period was 20.54.
Table 2-1 Man-Rem Month 1973 1974 1975 January 0.73 0.08 0.44 February 0.86 1.07 0.49 March 1.07 0.84 0.49 l
April 0.21 0.22 0.48 May 0.24 0.43 0 38 June 0.08 0.29 2.28 July 0.42 0 31 0.77 August 0.86 0.10 1 30 September 0.17 0.07 2 39 l
October 0.03 0 39 0.48 November 0.98 0.76 0.18 l
December 0.14 0.49 0.02 1
Total 5.79 5.05 9.70 l
During the period from 1976 through 1981, surveillance and routine maintenance activities were performed. Han-Rem per month as tabulated in Table 2-2 for th:se years.
2-1 l
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Table 2-2 Man-Rem Month 1976 1977 1978 1979 1980 1981 J nuary 0
0 N/A#
0.042 0
N/A Fcbruary 0.010 0.030 0
0 0
0 March 0
0 0
0 0
0 April 0.015 0
0.031 0
0 0
May 0
0 0
0.039 0
0.011 June 0
0 0
0 0
0 July 0
0 0
N/A 0
0 August 0
0.011 0.012 0
0 0
September 0
0 0
0 0
0 October 0
0 0
0 N/A N/A November 0
0 0
0 0
N/A December 0
0 N/A 0
0 0
Total 0.025 0.041 0.043 0.081 0
0.011 0 N/A = not available Preliminary evaluations and planning were made in 1982 for the disposition of th3 primary sodium stored in Fermi 1.
Drumming operations were performed in 1983 and the sodium was shipped to Idaho Falls in 1984.
A direct reading dosimeter calibration facility was established in the old machine shop located in the FARB. This area was selected because of the low background radiation and the access control maintained over the area. Most of the exposure for 1985 w.s the result of activity in this facility doing work for Fermi 2.
The calibrations are performed at night to keep the dose received during this activity ALARA. Man-Rem for these years is given in Table 2-3 Table 2-3 Man-Rem Month 1982 1983 1984 1985 January N/A 0.011 0
0.010 February 0.012 0
0 0.020 March 0.044 0.071 0
0 April 0
0 0
0.010 May 0.010 0.030 0
0.010 June 0
0.024 0
0 July 0.016 0
0 0
August 0.045 0.136 0
0 September 0
0.013 0
0 October 0
0 0
0 November 0.015 0
0.020 0.040 December 0.024 0
0 0.020 Total 0.166 0.285 0.020 0.110 2-2
~.
Exposures for the drumming operations were experienced in the sodium building which involved direct handling of the hot drums. All other drum handling was d:ne with forklifts and the drums on pallets. Total Man-Rea for the 1973 to 1985 per.'.od was 21 322.
Maintenance, repair and surveillance operations over the next 40 years will tv: rage about 0.035 Man-Ren per year. Fermi-2 dosimetry calibration activities in the FARB might result in personnel exposures as high as 0.20 Man-Rem per y:er.
Man-Rea resulting from decontamination and decommissioning activities will d pend on future factors not presently available; i.e., avails bility of low lsvel burial sites, adoption of de minimus levels, development of new dismantlement and decontamination techniques and acceptance in burial sites of material that has been contaminated with sodium. All of these factors if rcalized along with normal decay will provide a lower Man-Rem exposure 40 years from now as opposed to an immediate dismantlement.
2-3
QUESTION 3 - PROCEDURES:
Describe the methods and procedures for personnel monitoring (external and internal), including methods of recording, reporting and analyzing results.
Describe the program for internal radiation exposure assessment (whole-body counting and bioassay), including the basis for selecting personnel who will be in the program, the frequency of their whole-body count and bioassay, and any nonroutine bioassay that will be performed.
Describe the methods and procedures for evaluating and controlling potential airborne radioactive concentrations. Discuss any requirements for special air sampling and the issuance, selection, use, and measurement of respiratory protective devices, including training programs and respiratory protective equipment including program.
Methods of handling and storage of sealed and unsealed byproduct, source and special nuclear material should be described.
RESPONSE
Radiological protection is provided to personnel entering and/or working in the Fermi 1 Radiation Controlled Area through the implementation of Fermi 2 protective documents. The following list identifies those documents which
. are typically used to meet the requirements of the Fermi 1 Technical Specifications. The identification of these documents does not constitute a cosaitment to implement these documents for the life of the Fermi 1 license, but solely to illustrate how Detroit Edison implements the requirements of the license. These documents include Nuclear Operations Program Descriptions, Nuclear Operations Directives, Enrico Fermi 2 Plant Orders and Procedures.
In response to the item concerning personnel selection in the NRC question, personnel utilized in the conduct of Fermi 1 activities will have been trained in accordance with the applicable documents below, prior to participating in the subject Fermi 1 activity.
Those documents from Fermi 2 that effect operations at Fermi 1 and selected summaries related to specific questions are listed below:
1.
Nuclear Operations Interfacing Procedure 11.000.129, " Reporting Requirements - Personnel Training History System".
The online portion of the personnel training history system is used for immediate updating or displaying of training information and to store, update and generate requested reports from batch processing during off hours. The system is available to all personnel required to review qualifications of a person desiring to enter an area or to work on a radiation work permit.
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2.
Fermi 2 Precedure 12.000.13 " Radiation Work Permit".
The radiation work permit is an administrative method of controlling personnel access to areas for the purpose of minimizing internal and external radiological hazards, maintaining the total dose equivalent i
as low as reasonably achieveable (ALARA) and working with maximum radiological safety. Two types of radiation work pernits (RWP) are used:
- a. A general RWP for performance of routine duties or to cover narrow, well defined activities of a repetitive nature, and are valid for an extended period of time, not to exceed one year.
- b. The specific radiation work permit is issued for performance of a specified job in a specified location, and is valid only during the time it takes to complete this specified job.
3 Fermi 2 Procedure 12.000.62, " Radiologically Controlled Area Rules of Practice".
This procedure describes rules to be followed by all persons while in the radiologically controlled area of Fermi 2.
Although another procedure covers access to Fermi 1, the rules imposed in this procedure do apply to Fermi 1.
4.
Fermi 2 Procedure 12.000.108, "Offsite Dose Calculation Manual and Process Control Program Control".
The offsite dose calculation manual will probably not be required for use at Fermi 1 until it is decided that dismantlement is required.
5 Fermi 2 Procedure 61.000.07, " Bioassay Program".
Wholebody counts are required for all personnel who will be working in areas where contamination and/or airborne controls are exercised. The frequency is as follows:
a)
Prior to being granted access into the radiologically controlled area i
b)
Annually c)
Upon termination of employment or end of assignment d)
At the discretion of Health Physics supervision The initial non-routine bioassay shall be performed in accordance with the following:
4 a)
Exposure to airborne concentrations in excess of 10 MPC-hrs in any seven day period.
b)
The nasal smears indicate 1000/DPM or greater, c)
Whenever a significant internal uptake of radiological material is suspected, d)
At the discretion of Health Physics supervision (e.g.,
during the performance of extensive maintenance activities).
3-2 I
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All records generated by non-routine bioassay shall be reviewed and signed by Health Physics supervision and maintained as records in accordance with Regulatory Guide 8.26.
6.
Fermi 2 Procedure 63 000.21, " Contamination Survey Techniques, Personnel" Personnel contamination surveys are performed by Health Phyics to determine the extent of external personnel contamination, the possibility of internal contamination, the effectiveness of protective devices and the necessity for decontamination. The surveys include general body survey, nasal surveys, and eye and ear surveys.
7 Fermi 2 Procedure 61.000.50, " Health Physics Records".
The General Supervisor-Health Physics shall be responsible for the proper maintenance of Health Physics records.
In accordance with Title 10 CFR 20, the following records are considered to be permanent records:
a)
Personnel radiation exposure records and reports, and Health Physics respiratory protection records, b)
Instrument calibration and maintenance records.
c)
Radioactive source records.
d)
Radiological survey records.
e)
Radiation work permit records.
f)
ALARA review records, i
In addition to the records specified above, the General l
Supervisor-Health Physics may designate, as required, any record generated by Health Physics as a permanent record.
8.
Fermi 2 Procedure 61.000.51, " Personnel Radiation Exposure Records and Reports".
This procedure provides for compliance with 10 CFR 19, and 10 CFR 20 Section 102 and Section 401.
9 Fermi 2 Procedure 62.000.41, " Processing TLDs on the Panasonic Manual TLD Reader".
This procedure describes the method by which radiation doses are determined using the Panasonic UD-800 Series Thermoluminescent Dosimeters (TLDs) and the Panasonic UD-702 E Manual TLD Reader. The procedure provides operative instructions and equations needed for converting the raw TLD readings into a format suitable for input into various algorithms which calculate dose (mrem).
3-3
- 10. Fermi 2 Procedure 62.000.43, " Bioassay Sample Collection and Processing".
This procedure provides instructions for collecting and processing indirect bioassay samples such as urine or fecal samples. There are four situations in which bioassay samples may be collected.
a)
A random sampling program in which 10% of the individuals from the respiratory issue log are selected for the desired sampling period (Normally monthly),
b)
Positive body count or confirmed uptake.
c)
Suspected uptake of activity not detectable by direct analysis.
d)
Unavailability of any wholebody counters.
- 11. Fermi 2 Procedure 62.000.46, " Direct Reading Dosimeter - TLD Comparison".
The method used to compare direct reading dosimeter results with Thermoluminescent Dosimeter results are described in this procedure.
Comparisons are to be performed monthly.
Investigations will be conducted for individual cases of unacceptably large discrepancies between TLD and DRD results.
- 12. Enrico Fermi 2 Plant Order EFP-1037, " Controlling Access to EF1".
Specific requirements for escorted or unescorted access to Fermi 1 is covered in this procedure. Written authorization is required for all access.
13 Enrico Fermi 2 Plant Order EFR-6003, " Training of RadChen Contract Personnel".
The General Supervisor will require contract personnel to complete general training courses such as EF-2 Orientation, Radiation Worker Training, and any other course as required by Detroit Edison. The maintenance of all contract personnel documents is the responsibility of the General Supervisors.
In this manner, the contract personnel are advised of their responsibilities and other responsibilites of Detroit Edison to their personal monitoring protection.
- 14. Fermi 2 Procedure 61.000.05, " Radiation, Contamination and Airborne Guides, Limits and Controls".
This procedure defines the radiation, contamination and airborne radioactivity limits and guides under which radiological work activities at Fermi 2 are to be conducted. At the present time these guidelines have not been applicable to Fermi 1 because the contamination and airborne activity and radiation are so much lower than the guidelines. The only airborne contamination found at Fermi 1 that is measurable has been radon.
3-4
.~
15 Fermi 2 Procedure 61.000 36, " Respiratory Protection Training".
Prior to any perscnnel using respiratory equipment, potential users must receive respiratory protection training. The prerequisites for this training are:
a)
Trainees shall have successfully completed (and have documentation of same) a respirator medical examination conducted by the Detroit Edison Medical Department or their 3
stated designee.
b)
Trainees and users shall be clean shaven in the seal area of the respirator face piece.
c)
Trainees shall not wear contact lenses while using the respirator.
Each potential respirator wearer shall be retrained at least annually.
- 16. Fermi 2 Procedure 63 000 32, " Air Sample Collection, Anaylsis".
The objective of this procedure is to establish the proper method for i
obtaining, analyzing and logging of air samples, and provide a method I
for tracking plant personnel's exposure to airborne radionuclides.
Individual exposure is normally restricted to 25% of allowed limits.
I 17 Fermi 2 Procedure 65.000.21, " Qualitative Respirator Fit Test",
This procedure provides a methodology for performing certain qualitative fit testing (QLFT) on individuals wearing tight-fitting full face piece respirators for the purpose of investigating and assuring an effective face to face piece seal with the respirator.
For this procedure, test booths are used.
- 18. Fermi 2 Procedure 65.000.22, "Use of Respiratory Protection Equipment".
I In order to optimize the effectiveness of respiratory protection equipment, it is se,essary for the wearer to properly use (user inspection, donning, wearing, removal, and disposition) the device prescribed for them. This procedure provides a step by step method i
for the various types of equipment.
- 19. Fermi 2 Procedure 65.000.20, " Selection and Prescription of Respiratory Protection Equipment".
This procedure provides guidelines for the selection of the various types of respiratory equipment available.
It also gives the range of protection factors that can be accomplished with the various equipment.
It does not provide for use of the respirator in an atmosphere that is immediately dangerous to life.
3-5 l
- 20. Fermi 2 Procedure 93 000.01, " Operation of a Respiratory Protection Equipment Facility".
The respiratory protection equipment facility, located in the Radwaste Building of Fermi 2, is utilized to clean, decontaminate, sanitize, inspect, maintain, repair, test, store and issue respiratory protection equipment.
- 21. Fermi 2 Procedure 93 000.02, " Issuance, Accountability, and Collection of Respiratory Protection Equipment".
The following requirements must be met for issuance of respiratory protection equipment:
a)
Personnel issuing respiratory protection equipment must be qualified to do so.
b)
Respiratory protection equipment must be prescribed by a radiation work permit.
c)
A positive means of identifying an individuals qualifications to wear respiratory protection equipment shall be accomplished.
d)
Respirator wearers are required to be clean shaven in the seal area of the respirator.
e)
Equipment must have a completed respirator protection equipment inspection certification, f)
Personnel performing any part of this procedure must have read and understood the applicable radiation work permit.
- 22. Fermi 2 Procedure 93 000.03, " Inspection of Respiratory Protection Equipment".
All respirators must be inspected routinely before each use. The devices in storage are considered ready for normal issue or emergency use and must be reinspected at least monthly (not to exceed 30 days).
23 Fermi 2 Procedure 93 000.04, " Maintenance and Repair of Respiratory Protection Equipment".
This procedure provides guidelines for the maintenance and repair of respiratory protection equipment and is intended to categorize components of the respirator for the purpose of guidance regarding maintenance and repair.
- 24. Enrico Fermi 2 Plant Order EFP-1052, " Facial Hair Policy for Respiratory Protection Users".
Employees reporting to work who are respirator certified will report to work clean shaven. The employe's supervisor will assure that the employes facial hair will not prevent the proper fit of a respirator.
3-6 l
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- 25. Fermi 2 Procedure 63 000 50, " Source Leak Testing".
Each sealed source containing licensed material other than H-3 with a half life of greater than 30 days and in any form other than gas and containing radioactive material either in excess of 100 micro curies of beta and/or gamma emmiting material or five micro curies of alpha enmiting material shall be free of 0.005 micro curies of removable contamination.
l Sealed sources found to be leaking in excess of the above limits shall be withdrawn from use immediately and will be either decontaminated and repaired, or disposed of in accordance with NRC regulations.
- 26. Fermi 2 Procedure 66.000 31, " Removal of Material from the Radiologically Controlled Area",
j Before vehicles or materials (tools, equipment, etc.) are removed from the radiation controlled area, they must be surveyed by Health Physics personnel for contamination. Any materials / vehicles found to be contaminated above the administrative limits vill be either decontaminated or controlled in accordance with the special directions of Health Physics.
27 Fermi 2 Procedure 67.000.21, " Accountability of Radioactive Sources".
Licensed radioactive material in the form of solid, liquid or gas must be accounted for from the time a purchased source is received or a source is fabricated on-site, until the source is disposed or completely used. Accountability occurs in the form of a documented inventory performed semi-annually in May and November with interim location changes tracked to update files.
All source and special nuclear material have been removed from Fermi-1. The only remaining activity is in the byproduct and activation material found in the pipes and tanks in Fermi 1.
3-7
QUESTION 4:
Attachments 1 and 2 of your application are difficult to review and to cross-reference because of the lack of page numbers and proper formatting with sub-section numbers. A revised application with modified attachments should be submitted to correct these deficiencies.
RESPONSE
Detroit Edison has reviewed and revised, as appropriate, those documents prsviously transmitted to the NRC as Attachments 1 and 2 to Edison letter NE-85-0714, dated May 17, 1985 The revised documents are provided as.
It should be noted that the text of Enclosure 2 has been modified only to airplify its review, and has not been revised to reflect updated dose rate and contamination data. Current dose rate and contamination data is provided in response to questions 1, 2, 5 and 6.
l l
l 4-1 l
l
. QUESTION 5 - INVENTO M of RADIONUCLIDES:
Your submittal indicates that the only radionulclide now present at Fermi 1 in cobalt-60 which has a 5.2 year half-life. Your submittal states that the cobalt-58, iron-59 and chromium-51 present earlier are now essentially
- dscayed, i
Wa agree that cobalt-58, iron-59 and chromium-51 are absent because of their I
short half-lives (71 days or less). We have found, however, the presence of other. radionuclides in most reactor decommissioning situations. Europium-152 is generally present in concrete shielding structures around reactor vscsels. Since europium-152 has a 13 4 year half-life, it might still be present at Fermi 1.
Cesium-137, with a 30 year half-life, is often found as a contaminant on surfaces and in reactor systems. Also, nickel-63, with a 100 year half-life would likely be present in the Fermi 1 reactor vessel and i
vsssel components.
a) We, therefore, request that you provide a more complete inventory of the residual radionuclides which have half-lives equal to or greater than that of cobalt-60. Include inventory estimates of cobalt-60, europium-152, cesium-137, nickel-63, niobium-94 and nickel-59, in particular, and discuss the potential health and safety impact of these l
radionuclides at Fermi 1 on the removal of residual radioactivity at the end of the additional 40-year storage period.
b) Provide graphs of direct radiation exposure vs. time (zero to 40 years) as it may impact on workers involved in dismantling.
c)
Identify the benefits that result from the proposed additional 40-year delay in removal of residual radioactivity with respect to:
the methods that would be used in removing radioactivity; reduction in the total man-ren exposure to workers; and reduction in the volume of radioactive waste produced during decommissioning.
RESPONSE
Activation analyses were performed for the reactor vessel, the primary shield tank, and the concrete. Even with conservative assumptions for neutron flux, tha total amount of activity for several radionuclides was small.
Tha following is a discussion of the activation analyses performed for Fermi 1 structures. The calculational methodology, assumptions and results are presented. Three major components were analyzed for radionuclide inventories, these were the reactor vessel, the primary shield tank, and the biological shield (concrete). Reactor vessel internals, sodium residuals and liquid waste samples are also discussed below.
f 5-1
4 REACTOR VESSEL:
Th3 reactor vessel which is primarily 304 stainless steel was analyzed for th3 following radionuclides:
Nb-94 Co-60 Ni-59 Ni-63 C-14 Fe-55 Th3 reactor was assumed to be at full power for 30 days. A decay time of 164 Eonths was used. (It has been over 13 years since Fermi 1 operation.) The rtctogvesselvolumeinwhichactivationcouldoccurwasestimatedat131x 10 cm A constant flux was assumed for the 30 day period which was 3 x b
10 n/ca /sec. (See Table 5-1 and Figure 5-1.)
R:sults from the calculations are as follows:
July 1, 1986 TOTAL CONCENTRATION TOTAL ACTIVITY TOTAL ACTIVITY AFTER AFTER 40 YEARS AFTER 100 YEARS ISOTOPE ACTIVITY (C1) 40 YEARS (Ci)
(C1/cm3)
(Ci)
Nb-94 1.87E-06 1.87E-06 1.43E-12 1.87E-06 Co-60 1.29E-00 6.73E-03 5.14E-09 2.53E-06 Ni-59 2.12E-03 2.12E-03 1.62E-09 2.12E-03 Ni-63 2.50E-01 1.87E-01 1.43E-07 1.21E-01 C-14 9.50E-10 9.50E-10 7.25E-16 9.50E-10 Fe-55 5.09E-01 1.77E-05 1 35E-11 3 61E-12 From the above table Ni-63 (a beta emitter) has the greatest total activity of 1.87E-01Cibasedontheconsegvativeassumptionsusedinthesecalculations.
in the reactor vessel after 40 years.
This represents 1.43E-07 Ci/cm PRIMARY SHIELD TANK:
The primary shield tank which is ASTM-A-285 Grade C carbon steel was analyzed for the following isotopes:
l C-14 Fe-55 All assumption were the same except for volume and flux estimates. Theprimagy cujeldtankvolumeinwhichactivationcouldoccurwasestimatedat2.16x10 sh
.g Acogstantfluxwasassumedforthe30dayoperationperiod,whichwas2 x 10 n/cm /sec.
5-2
R; cults from the calculation are as follows:
July 1, 1986 TOTAL CONCENTRATION TOTAL ACTIVITY TOTAL ACTIVITY AFTER AFTER 40 YEARS AFTER 100 YEARS ISOTOPE ACTIVITY (C1) 40 YEARS (Ci)
(Ci/cm3)
(C1)
C-14 3 92E-10 3 92E-10 1.81E-16 3 92E-10 Fs-55 8.13E-02 2.82E-06 1.31E-12 5.77E-13 BIOLOGICAL SHIELD:
ThD biological shield which is concrete was analyzed for the following isotope:
Eu-152 Since the amount of Eu-151 in concrete was not known, the quagtity was 0
conservatively assumed to be.01%.
A volume of 1.16 x 10 cm3 of concrete5 wps,gsed. A constant flux was assumed for the 30 day period, which was 10
/sec.
Results for the calculations are as follows:
July 1, 1986 TOTAL CONCENTRATION TOTAL ACTIVITY TOTAL ACTIVITY AFTER AFTER 40 YEARS AFTER 100 YEARS ISOTOPE ACTIVITY (C1) 40 YEARS (C1)
(Ci/cm3)
(Ci)
Eu-152 2 75E-03 3.44E-04 2.97E-12 1.52E-05 REACTOR INTERNALS The data for the reactor internals, with the exception of niobium-94 and nickel-59, were extracted from NP-20047, " Retirement of the Enrico Fermi Atomic Power Plant." The values were obtained using the total flux (both thermal and epithermal) in the region of interest and documented for July 1,1986.
July 1, 1986 TOTAL TOTAL ACTIVITY TOTAL ACTIVITY AFTER AFTER 100 YEARS ISOTOPE ACTIVITY (C1) 40 YEARS (Ci)
(Ci)
Nb-94 4.06E-02 N/C#
N/C Co-60 274 1.41 5.2E-04 Ni-59 3.92 N/C N/C Ni-63 87 64 41 Fe-55 110 2.5E-03 2.92E-10
- N/C = Essentially no change 5-3
SODIUM RESIDUALS Th;re are approximately 550 gallons of residual sodium within the Reactor VI:sel, Primary Sodium Piping, Primary Sodium Heat Exchangers, Sodium pumps and th) three Sodium Storage Tanks. During the 1983 sodium drumming operations sodium samples were taken and analyzed for isotopic concentrations and totivities.
These results were used to perform our total activity calculations for the remaining residual sodium.
July 1, 1986 TOTAL ACTIVITY TOTAL ACTIVITY TOTAL ACTIVITY AFTER 40 YEARS AFTER 100 YEARS ISOTOPE (MICR0 CURIES)
(MICR0 CURIES)
(MICR0 CURIES)
Na-22 9.8E+02 2.5E-02 3 2E-09 Cs-137 4.8aE+03 2.04E+03 5.05E+02 LIQUID WASTE Ths estimated activity calculations in the waste water tanks were based on the activities found in the floor sump, the dose rate on the tanks and sample results from the tanks last discharge.
July 1, 1986 TOTAL ACTIVITY TOTAL ACTIVITY TOTAL ACTIVITY AFTER 40 YEARS AFTER 100 YEARS ISOTOPE CURIES CURIES CURIES Co-60 6E-3 3.08E-05 1.1E-08 Cs-137 6E-3 2 38E-03 5.95E-04 ENRICO FERMI UNIT 1 TOTAL NUCLIDE INVENTORY (CURIES)
Total Total Total Activity Activity Activity Nuclide 1986 2026 2086 Nb-94 4.06E-02 NC#
NC Co-60 2.75E+02 1.42E+00 5.21E-04 Ni-59 3 92E+00 NC NC l
Ni-63 8.72E+01 6.42E+01 4.12E+01 C-14 1 34E-09 NC NC Fe-55 1.11E+02 2.50E-03 2.92E-10 Na-22 9.80E-04 2.50E-08 3 20E-15 Cs-137 1.08E-02 4.52E-03 1.10E-03 Eu-152 2.75E-03 3 44E-04 1.52E-05 Total 4.77E+02 6.95E+01 4.51E+01
- NC = Essentially no change.
5-4
Observations Only Cobalt-60, Cesium-137 and Sodium-22 have been identified by gamma tetivation analyses of materials and equipment outside of the secondary shield wall (primary biological shield). All access to the Reactor Machinery dome, primary shield tank and concrete secondary shield wall were welded shut after the fuel, blanket material and sodium were removed and the piping cut and caps welded to the ends. We were unable to verify the calculations without violation of the sealed containments which would not be consistent with our ALARA cormitments.
The radiation exposure for removal of the fuel, blanket material, cold trap t
equipment, equipment used for cutup and shipment, and liquid waste generated was less 21 man-rem. The response to Question 6 demonstrates an eighty-nine to ninety-eight percent reduction in direct radiation exposure from the support piping and equipment.
The reduction in activity of the total isotopic inventory is provided in Figure 5-2 for the next 100 years. The exposure reduction will be proportional to this slope with an added reduction caused by the reduction in the gamma ensrgy profile.
(
An extension of forty years will result in the following benefits:
1.
About an 85% reduction in activity.
l 2.
From 89 to 98% reduction in exposure.
3 About 90% reduction in liquid waste activity.
4.
Development of sodium neutralization techniques from experiments proposed at Idaho Falls.
5-5 1
DATA TABLE 5-1
% IN REACTOR
% IN PRIMARY
% IN DENSI Y ISOTOPE VESSEL SHIELD TANK CONCRETE (gm/cm )
]
8.4 Nb-93
.01 8.9 Co-59 0.20 8.9 Ni-58 9.25 8.9 Ni-62 9.25 2.25 C-13
.08 0 30 7.86 Fe-14 68.40 99 32
.01 2.40 Eu-151 HALF LIFE OF CROSS ACTIVATION ACTIVATION ISOTOPIC SECTION ISOTOPE PRODUCT PRODUCT (sec)
ABUNDANCE (%)
(barns) 11 Nb-93 Nb-94 6.4 x 10 100.0 1.15 8
Co-59 Co-60 1.66 x 10 100.0 37 2 12 Ni-58 Ni-59 2.37 x 10 67.77 4.6 9
Ni-62 Ni-63 3 03 x 10 3 66 14.2 11 C-13 C-14 1.81 x 10 1.11 9E-04 I
Fs-54 Fe-55 8.51X1Q 5.82 2.25 Eu-151 Eu-152 4.2 X 10 47.82 9200 3
2 COMPONENT VOLUME (cm )~
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ACTIVATION TIME (days) l l
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QUESTION 6:
ENCLOSURE NO. 2 (Letters from the NRC to Stanford University) states our present criteria for release of a facility to unrestricted access.
a)
Provide an evaluation of the predicted radiation levels after the additional 40 years in buildings, rooms and structures, relative to the Enclosure No. 2 criteria.
b)
Estimate the additional interval of time that would be required for all buildings, rooms and structures to meet the release criteria presented in Enclosure No. 2 without removal of equipment or structures.
RESPONSE
Detroit Edison wishes to clarify that the subject license amendment request does not include a request for unrestricted access to Fermi 1.
However, the following discussion indicates which Fermi I areas meet the criteria dalineated in the Stanford letter for unrestricted access.
RADIOLOGICAL SURVEYS Several radiological surveys were performed to assess exposure rates and
~
contamination levels within the Fermi 1 Radiological Controlled Area. Areas curveyed included:
1)
Fuel and Repair Building 2)
Reactor Building 3)
Contamination Cask Car Trestle Shed 4)
Ventilation Building 5)
Outside Areas within the RCA l
6)
Primary & Secondary Sodium & Steam Tunnels 7)
Sodium Service Building FUEL AND REPAIR BUILDING Elevation 590' General area exposure rates for 590' elevation of the fuel and repair building were 6-10 uR/hr with the exception of the decay pool and cut-up pool. General area exposure rates around the decay pool and the cut-up pool wtre 10-20uR/hr. Contact readings on floor drains near the decay pool and cut-up pool ranged from 15-250 uR/hr. Smears were taken throughout the area.
Begaandgammacontaminationareasweresignificantlylessthan1000dpm/100 cs. Approximately 10% of the smears were counted for alpha, all levels w:
less than 20 dpm/100 cD 6-1
- - - -,. ~ - _ _ _ _ _ _ _, _ _. _ -
Eltvation 611'6" General area exposure rates for the mezzanine of the fuel and repair building were 5-8 uR/hr.
Contactreadingsontheboroscopeswere12gR/hr. Beta and gammacontaminationlevelswerealllessthan100gdpa/100cm. Alpha contamination levels were less than 20 dpm/100 cm.
Decay Pool General area exposure rates inside the decay pool were 10 - 80 uR/hr and 0.1
-4mR/hrinthedecaypooltunngl. Beta and gamma contamination levels rangedfrom1500-3g00dpa/100cm. Alpha contamination levels were less than 20 dps/100 cm Cut-up Pool General area exposure rates inside the cut-up pool were 80-200 uR/hr and 100 uR/hrinthecut-uppoolgunnel. Beta and gamma contamination levels ranged from1500-g000dps/100cm. Alpha contamination levels were less than 20 dpa/100 cm R: pair Pit Currently, the repair pit at Fermi 1 is being used to store the turbine bypass line which was removed from Fermi 2.
General area exposure rates were 6-30uR/hr. Betaandgammacontaminationlevelswerelessthgn1000dps/100 en. Alpha contamination levels were less than 20 dpm/100 cm W:.ste Tanks General area exposure rates for the waste tanks in the fuel and repair building (elevation 559') were 0.5 - 2 mR/hr. The liquid waste dump tank (MK-15) had the highest contact exposure rate of 80 mR/hr. This tank is currently full. Waste water is now pumped to waste tank MK-9 All other tcnks had contact readings which ranged from 5-10 mR/hr with the exception of thesupplytankwhichhadacontactreadingof50mg/hr. Beta and gamma contaminationlevelswerelgssthan1000dpm/100cm. Alpha contamination was less than 20 dpm/100 cm Pump Room General area exposure rates for the pump room (elevation 576'6") were 10-100 uR/hr. Contact exposure rates for piping, pumps, gas dryer and gas receiver rangedfrom50to60guR/hr. Beta and gamma contamination levels were less than1000gpm/100cm. Alpha contamination levels were less than 20 dpm/100 cm 6-2
Sumpa General crea cupo:ura rat:0 for thn sump in th3 fusi and r3 pair building w:ro 0.1 - 2 tR/hr. Contact r:adings on a recoved pump wira 2 15 IR/hr. Beta and "g
^ Ph*
gammacontaminationlevelswerelessthan1000dp5'.'Waterandsludge contamination levels were less than 20 dpa/100 cm samples from the sumps were counted on an intrinsic germanian detector.
Co-60 and Cs-137 were the only isotopes identified. Co-60 levels were t =sured at 2.75E-04 uCi per 500 ml of sample and Cs-137 levels were measured at 2.65E-04 uCi per 500 ml of sample.
REACTOR BUILDING Elevation 557' General area exposure rates for the 557' elevation of the reactor building w:re 0.1 - 0.3 mR/hr. Contact exposure rates ranged from 0.1 - 4 mR/hr on thasodiumheatexchangers,angpiping.Betaandgammacontaminationlevels wirelessthag1000dps/100cm. Alpha contamination levels were less than 20 dpm/100 cm l
Elsvation 590' General area exposure rates for the 590' elevation of the reactor building t
were 2-10 uR/hr. A posted contaminated area on the floor had a contact exposure rate of 20 uR/hr. A " hot spot" was measured on the north side of primary contaminnent 50 uR/hr. Beta and gamma contamination levels were less 2
than1000gpa/100cm. Alpha contamination levels were less than 20 dpm/100 cm CONTAINMENT CASK CAR TRESTLE SHED General area exposure rates for the containment cask car trestle shed were 6-j0uR/hr. Beta and gamma contamination levels were less th 1000 dpm/100 cs. Alpha contamination levels were less than 20 dpa/100 cm VENTILATION BUILDING General area exposure rates for the ventilation building were 8-10 R/hr.
2 Betaandgammacontaminationlevelswerelessthag1000dpm/100cm. Alpha l
contamination levels were less than 20 dpm/100 cm OUTSIDE AREAS WITHIN THE RCA I
General area exposure rates for outside areas within the RCA were 5-10 uRfhr. Beta and gamma contamination levels were less than 1000 dpm/100 ca. No smears were counted for alpha contamination for the outside areas.
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PRIMARY AND SECONDARY SODIUM AND STEAM TUNNELS Primary Sodium Tunnel General area exposure rates for the primary sodium tunnel were 10 - 90 uRfhr. Betaandgammacontaminationlevelswerelessthan10pdpa/100 ca. Alpha contamination levels were less than 20 dpa/100 cm.
Secondary Sodium Tunnel - West General area exposure rates for the secondary sodium tunnel were 8 10 uR/hr.
g Betaandgammacontaminationlevelswerelessthag1000dpa/100cm. Alpha contamination levels were less than 20 dpa/100 cm Secondary Sodium Tunnels - East General area exposure rates for the Secondary Sodium Tunnel-East were 8-10 uRfhr. Betaandgammacontaminationlevelswerelessthan10g0dpa/100 c3. Alpha contamination levels were less than 20 dps/100 cm.
Gas Tunnel General area exposure rates for the gas tunnel were 10 - 50 uR/hr. A contact cxposure rate taken on a capped waste gas line was approximately 350uR/hr.
Betaandgannacontaminationlevelswerelessthag1000dpa/100cm. Alpha contamination levels were less than 20 dps/100 cm SODIUM SERVICE BUILDING Cold Trap Cell General area exposure rates for the cold trap cell were 20 - 50 uR/hr.
Contact readings on a drum, a hoteell, and washers were 0 1 mR/hr. Beta and 2
nana contamination levels were less than 1000 dpa/100 cm. No smears were caunted for alpha contamination in this area.
Sodium Storage Roon General area exposure rates for the sodium storage room were 0.2 - 0.8 tR/hr. Contact readings for tanks Nos. 1, 2 and 3 were 2 sR/hr. Beta and 2
ganna contamination levels were less than 1000 dps/100 cm. No smears were counted for alpha contamination in this area.
Covered Storage Area General area exposure rates for the covered storage area were 8-10 uR/hr.
Beta and ganna contamination levels were less than 1000 dpa/100 cm. No s ears were counted for alpha contamination in this area.
6-4
Empty Drum Storage Mezzanine The general area exposure rate for the empty drum storage mezzanine was 8 uRfhr. Beta and gamma contamination levels were less than 1000 dpm/100 cs. No smears were counted for alpha contamination in this area.
Sodium Transfer Room General' area exposure rates for the Sodium Transfer Room were 8-12 uR/hr.
Contact exposure rates for piping and a vent box ranged from 100 - 150 uRfhr. Beta and gamma contamination levels were less than 1000 dpa/100 c;.
Sodium Service Building-616' Elevation Th3 general area exposure rates for the sodium service building 616' elevation was 6 10 uR/hr. Beta and gamma contamination levels were less than g
10g0dpm/100cm. Alpha contamination levels were less than 20 dpm/100 c3 Wmte Gas Building General area exposure rates for the waste gas building (hold-up tank room) w:re 20-50 uR/hr. Contact exposure rates on the tanks and a vapor trap rngedfrom50-300uR/hr. Beta and gamma contamination levels were less than 10 Alpha contamination levels were less than 20 dpm/100 c;g0dpm/100cm.
SUMMARY
Mo significant removable contamination was detected in any of the above surveyed areas at Fermi 1.
The only detectable removable contamination was found in a few spots f the decay and cut up pools which ranged from 2
l 1500-4000 dpm/100 cm. Aglotherbetaandgammacontaminationlevelswere Ices than 1000 dpm/100 cm, which was the minimum detectable level for the purpose of this survey. Alpha contamination levels were all less than 20 2
dpm/100 cm which represents the alpha contamination minimum detectable level.
Presently the first floor and mezzanine of the FARB (with the exception of sole areas over the floor drains in the decay and cut-up pool rooms), the tachine pit of the FARB, cask car trestle shed, operations floor in the R=ctor Dome, second floor of the Sodium Building and all outdoor areas are within the release criteria as defined in letter from James R. Miller (USNRC) to Dr. Roland A. Finston (Stanford University) Docket No. 50-141, dated April 21, 1982 ("... 5 Micro Rem per hour at one meter for reactor generated, gamma caltting isotopes").
6-5 l
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Isotopic Analysis Isotopic analyses were performed at selected locations with enough activity to provide significant data. The inplace analyses were performed with a Quantum Technology transportable gamma spectroscopy system. Selected locations were those areas in which the general radiation level was higher than normal background.
Only Cesium-137, cobalt-60 and Sodium-22 were found. The reasured totivities, present dose rate at the point of measurement, dose rate in forty y; rs, forty year dose rate corrected for gamma energy profile change, and percent reduction over the forty year span are shown on Table 6-1.
Background Measurements Background measurements were made near our Southwest boundary (about one mile from Fermi 1). Values ranged from 10 to 11 uR/hr. Measurements were also obtained from the grouria and interior of a home eighteen miles north of Fcrai 1.
Values in the yard were 9 uR/hr overall and ranged from 8 to 12 uR/hr within the house.
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I:ble 6-1
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Isotopic Activity and Dose Rate Projections #
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CORRECTED PERCENT AREA Co-60 Cs-137 40 YR 40 YR/ CURRENT DS. RATE SURVEY LOCATION DS. RATE ACT.
ACT.
DS. RATE DS. RATE REDUCTION Decay Pool 8.5E+01 8.49E-01 9.86E-02 3 92E+00 1.28E+00 98.50%
S. Drain Decay Pool 8.0E+01 1.12E+00 2.42E-02 1.08E+00 5.76E-01 99.28%
SW Pipe Cutup Pool 1.8E+02 1.89E+00 1.66E-01 6.66E+00 2 31E+00 98.72%
S. Drain Fu21 Pool 3 0E+01 4.85E-01 4.22E-02 1.10E+00 3 83E-01 98.72%
Ex. Fan Pump Room 6.0E+02 2.90E+00 1.58E-01 1.53E+01 6.05E+00 98.995 MK 15 1.5E+03 3.87E+00 1.02E+00 1.31E+02 3 74E+01 97.51%
T nk Room RX Bldg 5.0E+02 7.16E+03 1.14E+00 2.63E+00 2.61E+00 99.48%
Basement Decay Pool 1.0E+02 4.99E-01 8.12E-01 2.49E+01 6 38E+00 93.62%
Tunnel AVERAGE PERCENT DOSE RATE REDUCTION (Co/Cs ACTIVITY) 98.10%
CORRECTED PERCENT SURVEY AREA Na-22 Cs-137 40 YR 40 YR/ CURRENT DS. RATE LOCATION DS. RATE ACT.
ACT.
DS. RATE DS. RATE REDUCTION NA Storage 2.00E+03 3 77E-02 7.54E+00 7.94E+02 2.18E+02 89.08%
Tank Gas Decay 1.2E+02 7.23E-03 2.13E+00 4.77E+01 1 31E+01 89.06%
AVERAGE PERCENT DOSE RATE REDUCTION (Na/Cs ACTIVITY) 89.07%
- All dose rates are in microrem/ hour.
6-7
Enclosura 2 to VP-86-0092 l
i RESPONSE TO QUESTION 4: MODIFIED ATTACHMENTS 1 AND 2 l
l FROM DETROIT EDISON LETTER NE-85-0714 i
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ATTACHMENT 1 - FERMI 1 SAFETY EVALUATION #
CRevised as noted in Enclosure 1, Question 4
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SAFETY EVALUATION I.
INTRODUCTION 4
Retirement of the Fermi 1 plant was accomplished through provisions of 10CFR50.59 and through a series of Technical Specification changes. The original decommissioning plan submitted through ERDA document NP-20047, consisted of the following major elements:
1.
Shipping all fuel and blanket elements offsite.
2.
Shipping all bulk radioactive and non-radioactive sodium of fsite and passivating the residual sodium.
3.
Disposing of all other contaminated or irradiated materials by i
shipping offsite or placing in restricted areas whose access is limited only to authorized personnel.
4.
Securing some of the Reactor Building electrical, instrumentation, piping, ventilation, personnel and equipment penetrations.
5.
Sealing the primary system, comprised of the reactor vessel, primary sodium piping, primary shield tank, machinery dome, primary sodium service, and secondary sodium system out to welded pipe caps, and passivating the residual sodium therein.
6.
Revising the site boundary to a very limited area to include only those areas containing radioactivity.
7.
Implementing a post-retirement surveillance plan.
All elements of the plan have been implemented with the shipment of primary l
sodium to the Argonne National Laboratory-West between October 29 and i
November 12, 1984.
The facility is currently in a safe storage condition. Detroit Edison is proposing to continue maintaining the facility in its present condition until the year 2025 at which time all residual radioactivity is currently intended to be removed and the license terminated.
II.
SAFETY EVALUATION The subject license amendment request has been reviewed against the criteria of 10 CFR 50.92, and was determined not to represent a significant hazards consideration as delineated below, a.
The proposed amendment does not involve a significant increase in the probability of consequences of an accident previously evaluated.
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The radiological status of the facility is progressively becoming more benign due to the decay of residual radioactivity. In June 19,J, the main source of radioactivity in the facility was estimated to be 1550 Curies, of which slightly more than 1500 Curies was due to Cobalt 60.
The Cobalt 60 presently is calculated to have decayed to approximately 320 Curies, and by the year 2025, will have decayed to approximately 1.6 Curies.
Other contributors originally present such as Co-58, Fe-59 and Cr-51 are now essentially absent. The safe storage condition of the facility has been and will continue to be maintained on the basis of the Technical Specifications. This will assure that the barriers to residual radioactivity will retain their physical integrity until such radioactivity is removed.
Therefore, it can be seen that the probability of an accident is not increased by this license amendment request, and that theconsequences of an accident will decrease as the levels of radioactivity decrease.
b.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The amendment does not involve the reconfiguration of any of the stored materials and, therein, will not introduce the potential for any new accident scenarios.
c.
The proposed amendment does not involve a significant reduction in a margin of safety.
The requested amendment requests that Detroit Edison be allowed to continue to maintain the Fermi 1 facility in the same SAFSTOR condition that it has used for the past several years. This will not degrade the current margin of safety since the facility's status will be continuously monitored by periodic walk-throughs and surveys.
III. CONCLUSIONS Based on the above, the maintenance of the Fermi 1 facility in a safe storage condition until the year 2025 presents no increased risk to the health and safety of the employees or the public, and is within the scope i
of current regulation.
In addition, the proposed amendment creates no irreversible conditions.
Options for final decommissioning of the facility remain open and flexible as the residual activity decreases.
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ATTACHMENT 2 - FERMI 1 STATUS DESCRIPTION
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- Revised as noted in Enclosure 1, Question 4
TABLE OF CONTENTS Page 1.0 ' INTRODUCTION 1
2.0 FACILITY DESCRIPTION 1
3.0 ADMINISTRATION 6
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4.0 ACCESS CONTROL 7.
5.0 MONITORING AND ALARMS 7
6.0 SURVEYS, INSPECTIONS AND TESTING 8
7.0 PROCEDURES 9
Table 1 Environmental Survey Regimes Figure 1 Facility Plan i
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1.0 INTRODUCTION
The retirement plan for the Enrico Fermi 1 Power Plant was previously presented to the Atomic Energy -Commission via a September 24, 1973 letter from the Power Reactor Development Company. The following discussion is an update of that information which reflects the current status of Fermi 1.
As reflected in the September 24, 1973 submittal, the core fuel sections of T all 214 Core A fuel subassemblies composed of 25.6 w/o enriched uranium molybdenum alloy were shipped to the Savannah River Plant facility for reprocessing.
The complete inventory of approximately 70,000 gallons of primary sodium was atored frozen in 1344, 55 gallon drums stored in the reactor containment dome and, just prior to shipment, in the cask car trestleway.
This sodium was shipped to Argonne National Laboratory-West between October 29 and November 12, 1984. Though virtually all sodium has been removed from the systems within Fermi 1, a residual heel of sodium [approximately 450 gallons (estimated)] is retained in vessels and piping formerly used for sodium ~.
The main source of activity in the facility, concentrated in the lower elevation of the reactor building, was estimated in June 1973 to be 1550 curies, of which slightly more than 1500 curies was due to Cobalt 60 present in the reactor support plates, holddown mechanism, and -shield bars.
At the present time, the Cobalt 60 is calculated to have decayed to approximately 320 curies, and in 2025, the time that " safe store" condition is intended to be terminated, it will have decayed to approximately 1.6 curies. Other contributors present in 1973 (Co-58, Fe-59, and Cr-51) are now essentially absent. Typical radiation levels in the high radioactivity regions within the Protected Area presently average between 1.0 and 15 mrem /hr. Areas outside the protected area are less than 5 microrem/hr above natural background.
The safe storage condition of the facility is maintained on the basis of the Technical Specifications and the Administrative and Surveillance Procedures which prescribe the administration, access control, monitoring, periodic environmental and radiological surveys, long term maintenance provisions, and record keeping requiremente.
2.0 FACILITY DESCRIPTION The Fermi 1 facility is located within the owner controlled area and outside the protected area of the nearby Fermi 2 Unit. Figure 1 presents the current facility plan. As shown in the figure, the following buildings are identified:
2.1 Reactor Building The Reactor (or containment) Building contains, below floor, the empty radioactive reactor vessel (which itself is contained in the Primary Shield Tank), heat exchangers, primary sodium pumps,
1
and the primary sodium overflow tank (which has been passified and is open to the atmosphere). A moisture detector is located in the area sump that alarms in the manned control station.
Operators enter this area twice a year to check the moisture,
detector operation in accordance with the Technical Specifica-tions. No moisture has been detected in this area since the retirement of the facility.
Above. floor level, the Reactor Building contains the machinery dome, containment crane, and other machinery. The primary system has been capped with carbon dioxide at approximately two inches water pressure. Though virtually all sodium has been removed from the system an estimated 450 gallons of residual heel of sodium remains in the vessels or pipes formerly used for sodium.
This residual sodium contains an estimated 1.7 mci of Cs-137 and 0.34 mci of Sr-90.
l The Reactor Building outer air lock door is kept locked, except when occupied, to provide additional security. The interlock on i
the emergency exit has been removed and the outer door can be operated from the inside in case of an emergency.
2.2 Fuel and Repair Building (FARB) l This building houses the fuel storage pool and fuel cut up pool l
which were drained out, cleaned and painted with strippable paint.
In addition, above floor, there is the sealed containment steam cleaning chamber and below floor, the sealed transfer tank room, mechanical equipment room, liquid radioactive waste tank rooms, and the " hot" drains sump. The transfer tank room houses a i
drained and sealed transfer tank containing a heel of passivated i
residual sodium. The radioactive liquid waste tanks were originally drained. The " hot" sump has been left active. A moisture detector in the sump alarms at the manual control station when water gets to the applicable level.
The sump pumps preaently pump into the liquid waste tanks: MK 7, 8, 9, or 15.
In the last 10 years, during which Fermi 1 has been in a safe storage condition, approximately 15,000 gallons of water have been discharged to the tanks. The quantity of liquid in the tanks is monitored from level indicators located in the upper floor of the Fuel and Repair Building. The total capacity of the tanks is 37,500 gallons. Much of this water came from water I
in-leakage from pipe penetrations to the Health Physics Building.
The situation causing the in-leakage has been corrected, t
Radiation indications within the general area of the ground level floor of the Fuel and Repair Building are below detectable levels.
Slight contamination has been found in the fuel and cut-up pools and the " hot" sump.
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As of May 1985, the activity at the bottom of the sump pump was 2
100,000 dpm/100cm and at the shaft of the pump was 20,000 Z
dpm/100cm. One corner of the bottom of the sump measured 15 mrem /hr.
2.3 Sodium Storage Building The primary Sodium Storage Building contains three 15,000 gallon tanks. The sodium in these tanks was removed and placed in 55 gallon drums. These drums were shipped to Argonne National Labora tory-Wes t in Idaho. The three primary sodium storage tanks have been passivated with a carbon dioxide cover gas which is maintained on the tanks.
Entrance to the sodium tank room is through a steel door in the north wall of the building which has been locked.
The Sodium Storage Tank room was surveyed for radiation levels on April 15, 1985. The radiation levels between tanks range between 0.5-1.0 mrem /hr.
2.4 Waste Gas Building This building contains two deactivated waste gas tanks and is nonradioactive.
2.5 Waste cas Tunnel The tunnel runs east to west from the Reactor Building along-side the south wall of the Waste Gas Building. The tunnel steps down as it approaches the Waste Gas Building and becomes too congested and small for access along its whole length. Access is via a steel cover southeast of the Waste Gas Building.
2.6 Inert Gas Building The south end of this building contains a 500 cubic foot vacuum tank, a 200 cubic foot vapor trap, and a 500 cubic foot hold-up tank. The north room contains th ree inert gas compressors. Both rooms are nonradioactive.
2.7 NaK Room The room houses the NaK equipment (used in conjunction with the sodium cold trap) which has been disconnected. The sodium drumming facility is also located in this room. Radiation surveys indicate only trace amounts of radioactive material present in equipment and structures in this room.
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2.8 cold Trap Room i
l This room contains sodium piping needed formerly for sodium l
drumming, and the piping and valves to and from the primary sodium tunnel. _ Some of the latter has been cut and capped.
Others have had the valves closed and the handwheels disconnected.
Some miscellaneous radioactive materials are being stored in the room.
Not shown in Figure 1 is a second story above the Cold Trap Room, NaK Room, and part of the Inert Gas Building. This second story 2
contains the handwheels to the sodium valves at the north end of the room (some disconnected) and the electrical feeds for sodium heating.
4 2.9 Vent Building This building has been emptied of all equipment and the fence has been modified to be continuous past the east doors of the building.
2.10 Primary Sodium Tunnel This tunnel is steel lined and runs from the northwest corner of the Reactor Building to the Cold Trap Room. The piping in this tunnel has been drained and capped at the Reactor Building, and either capped or isolated with closed disconnect valves in the Cold Trap Room. Access to this tunnel is via a manhole near the transfer corridor.
2.11 Fission Product Detector Building This is a small building, partly below ground level, to the east i
of the Reactor Building.
It contains some slightly radioactive Ii P P ng.
2.12 Sodium Piping Galleries i
There are two below ground piping galleries that were used to house secondary sodium piping.
The west gallery consists of two chambers (i.e., north and south chambers) which hold the secondary sodium lines that supplied the No. 3 steam generator. The sodium lines have been capped where they exit and enter the Reactor Building.
1.
l The ground level entry to the north chamber was sealed of f as part of the decommissioning program. Entry is now made via a short 30" diameter tunnel which runs between the biological shield wall space and the north chamber, i
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Access to the south chamber is via a steel door just above ground level at the north wall of the Steam Generator building.
The east gallery consists of three separate chambers and contains the secondary sodium pipe lines that supplied the Nos. 1 and 2 steam generators. These lines have been capped as in the west gallery.
Access to the three chambers is by means of steel doors just above ground level outside the southeast quadrant of the Reactor Building.
2.13 Biological Shield Wall Area This is approximately a three foot wide annulus that surrounds the building below floor level to about three feet below the concrete pedestal on which the steel Reactor Building stands.
Various service piping systems are located in the annulus.
It is entered via a bolted-in place cover at the west azimuth outside of the containment shell. An access ladder has been left in place.
In addition, at about 30' north of the entrance, at just below floor level, there is an access port to the northwest secondary sodium pipe gallery chamber.
The annulus has four floor drains that drain into a collection tank and sump pump system located in the basement of the Steam Generator Building.
In additioa, there is a moisture detector that alarms at the manned control station. The water drained into the tank is nonradioactive.
2.14 Fuel Transfer Corridor This is a covered way covering the tracks of the former fuel 7
handling cask car. The end adjacent to the Reactor Building was modified af ter 1966 to provide a covered access to the Reactor Building air lock. The area is free of contamination.
2.15 Health Physics Building This building has been removed. Only the foundation slab remains.
There are potentially radioactive drains and lines to the hot sump in the FARB which have been permanently plugged and marked.
Due to some problems with rain water in-leakage via these drains and where they penetrate the FARB, the entire area between the l
foundation sinb and the FARB has been covered by a concrete slab.
Since this has been done, there has been no consequential in-leakage. Radiation levels above the slab are in the same range as those of natural concrete.
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2.16 Primary System Cover Gas The primary system is defined as the reactor vessel and all connecting volumes. These include the three primary loops, the machinery dome, and the primary sodium service, and secondary sodium systems out to the welded fittings. The primary system is connected to reserve and backup supplies of carbon dioxide to passivate the residual sodium and is kept at approxirately two inches water pressure with a relief valve set for approximately 5 psig. The relief valve is checked annually for proper operation.
The cover gas is instrumented to alarm in the manned control station at low (1/2-in water gauge) and high (2 psig increasing pressure) pressure. Primary cover gas alarm tests are performed every six months.
2.17 Liquid Waste Discharge System All potentially contaminated drains and sumps collect in the hot sump L. Une Fuel and Repair Building. The liquid waste collected
. in the hot sump is discharged to the liquid waste tanks (MK 7, 8, 9, 15). Liquid quantities in these tanks are monitored and recorded.
If there is a need to discharge the liquid from the tanks, the water will be processed via a portable liquid radwaste processing system, such as is typically used at many power reactor facilities, until it is acceptable for discharge in accordance with the Technical Specifications.
I Liquid waste collected in non-contaminated sump systems are not part of the Liquid Waste Discharge System. Non-contaminated sumps collect underground water or rain water intruding in non-contaminated areas. These sumps discharge to the plant drain system which drains to Lake Erie.
2.18 Other The remaining Fermi 1 buildings (e.g., the steam generator, control, and of fice buildings) were not exposed to the radioactivity resulting from the operation of Fermi 1.
Due to this, they have been used for various activities affiliated with the operation of the oil peaker unit (now decommissioned) and Fermi 2.
3.0 ADMINISTRATION The Detroit Edison Company has the responsibility for maintaining a continuing administrative and surveillance program in compliance with current Nuclear Regulatory Commission requirements to ensure that the health and safety of the public and employees are not threatened or injured.
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l The Vice President, Nuclear Operations, who reports to the Group Vice President has overall responsibility for the Enrico Fermi Unit I reactor facility. Responsibility for the decommissioned Enrico Fermi Unit I reactor ~ facility is delegated through the line organization of the Vice President, Nuclear Operations, to a qualified Custodian selected from the staff of the adjacent Fermi 2 Atomic Power Plant. The Custodian is assisted in his duties by Custodial Delegates and Custodial Agents.
In addition, the facility administrative and surveillance program is audited by means of a Review committee which will also review and approve all matters of safety associated with any maintenance activities in the facility.
Written procedures delineate the qualification, selection and responsibilities of the Custodian, Custodial Delegates and Agents, and members of the Review Committee.
4.0 ACCESS CONTROL The area encompassed by physicci barriers and to which access is controlled is the Protected Area. The Protected Area, is enclosed by either a chain link fence or building walls which provide equivalent degree of resistance to penetration. The fence is topped by three or more strands of barbed wire or brackets angled outward with an overall height of no less than seven feet. Normal entry to the Protected Area is through a normally locked gate in the fence adjacent to the Sodium Building. Other doors in walls which act a part of the Protected Area boundary are locked or permanently sealed.
Access to the Protected Area is controlled, limited and recorded. The access key is in the manned control station. A second key is held in safe keeping by the Custodian for use only in extenuating circumstances.
Written procedures delineate the requirements associated with entry into the Protected Area and specific areas withir. the Protected Area to prevent unauthorized entries and to protect the safety and health of authorized personnel.
5.0 MONITORING AND ALARMS Monitoring detectors for water intrusion are located in three areas:
(1) the Fuel and Repair Building basement hot sump, (2) the lower reactor building overflow tank pit, and (3) the Biological Shield Wall Area.
Accumulation of water in these areas activates an alarm in the manned control station.
7
The primary system cover gas pressure is also monitored with high and low ala rms. The monitors and the alarm circuitry are periodically checked and calibrated in accordance with the Technical Specifications and written procedures.
6.0 SURVEYS, INSPECTIONS, AND TESTING Two types of surveys are identified, environmental and radiological, in addition to periodic facility inspections and instrumentation testing.
For the environmental surveys, a number of stations have been established where it is estimated that maximum concentrations of radioactive material discharged from the facility may occur. Two dif ferent regimes of sampling and analysis are utilized. A summary of these regimes is given in Table 1.
1.
Regime I is followed if activity is released.
2.
Regime II is followed if no activity has been released during the previous 90 days.
Periodic radiation surveys are performed to check for the presence of gamma radiation and transferable contamination at the frequency specified in the Technical Specifications. Gamma radiation measurements using portable survey instruments and contamination checks using smears are made of the following areas:
Reactor Building - Operating floor, doors and seals around machinery dome, breather pipe, sump pump serving Reactor Building annulus.
Fuel and Repair Building - Pool area, operating floor access points to contamination areas, Steam Cleaning Room access plug.
Environmental and radiological surveys are performed by or under supervision of qualified personnel having parallel duties and responsibilities at Fermi 2.
A monthly visual inspection of the Protected Area is performed by Custodial Agents. The inspection consists of a visual inspection of the fence, gates and all exterior doors at the Protected Area, a check and recording of the level of liquid in the liquid water tanks, a check on the condition of the strippable paint on the decay and cut-up pools in the Fuel and Repair Building, and verifications of the operation of the sump pumps which serve the Protected Area. All abnormal conditions observed are reported to the Fermi 2 Nuclear Shift Supervisor so that corrective measures may be taken.
The Custodian is also notified of all abnormal conditions immediately, and of the corrective action taken.
8
Testing and calibration of the water intrusion monitors, and testing the primary cover gas pressure alarms, is performed every six months. Testing of the carbon dioxide pressure relief valve is performed annually. All testing is performed in accordance with written and approved procedures.
7.0 PROC EDURES Procedures ensure that the requirements of the Technical Specifications are carried out in a proper and timely manner. They also serve as training and reference units for future Custodians, Custodial Delegates and Custodial Agents. Administrative procedures include Custodial qualifications, responsibilities and authority, Procedure Manual control, Custodial Delegate and Custodial Agent selection and function, reporting procedures, Review Committee functions and financial accounting procedures. In addition, there are appropriate procedures for details of inspections, surveillances and operation.
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TABLE 1 ENVIRONMENTAL SURVEY REGIMES Number of Stations Regime Sample Media Indicator Background I
II Water South Lagoon 1
0 G26b G26b River Water 1
1 Gib G26b Lake Water 1
0 Gib G26b Raw City Water
- 0 3
G4b G26b Sediment South Lagoon Sediment 1
0 G26g G26g River Sediment 1
1 G26g G26g Symbols:
G - Grab Sample Frequency of Sampling:
1 - one week interval 4 - four week interval 26 - twenty-six week interval Type of Analysis:
b - beta g - gamma Example : Gib - Sample is collected at one week intervals and analyzed for beta radioactivity
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LEGEND: FENCE ///////////
NOTE A: BUILDING HAS BEEN DISMANTLED ONLY SLAB REMAINS
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FIGURE 1 FACILITY PLAN 1
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