ML20148P073
| ML20148P073 | |
| Person / Time | |
|---|---|
| Site: | 07001308, 07001309 |
| Issue date: | 10/15/1978 |
| From: | Dawson D GENERAL ELECTRIC CO. |
| To: | Cunningham R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20148P079 | List: |
| References | |
| NUDOCS 7811280107 | |
| Download: ML20148P073 (5) | |
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', i r.g 9 NUC R ENERGY GENERAL' ELECTRIC PROGRAMS DIVlSION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 DMD-247 Docket 70-1308d
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License SNM-1265 Docket 70-1309 License SNM-1281 October 15, 1978 Office of Nuclear Material Safety and Safeguards Attention:
Mr. R.E. Cunningham, Acting Director i
c' Division of Fuel Cycle & Material Safety s _ s}
U.S. Nuclear Regulatory Commission t -- }
Washington, D.C.
20555 H
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u, Gentlemen:
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SUBJECT:
REQUEST TO AMEND CONDITIONS OF LICENSE SNM-1265, MY#;)'
o AND CANCEL LICENSE SNM-1281 General Electric Company requests that the conditions of License Number SNM-1265 be amended as specified herein, and that concurrent with the requested amendments of License No. SNM-1265, License SNM-1281 be canceled.
The requested amendment constitutes administrative action, having no bearing on matters affecting the health and safety of the public or employees of Morris Operation.
A fee of $150 is therefore enclosed.
Physical security conditions of License SNM-1265 are discussed in a concurrent letter to Mr. R. Burnett, Division of Safeguards; a copy is attached for your information.
Please note that this licensing action constitutes a single action; consequently a single fee is involved.
The documentation submitted with this application includes:
a.
Request for amendment - this letter b.
Attachment A - comparison of current license conditions with contents of Chapter 10, NED0-21326.
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GENER AL @ ELECTRIC
. c.
Attachment B - proposed amended license d.
Draft - Revision A4, NED0-21326. After approval by USNRC, replacement pages will be submitted.
e.
General Electric Company check - $150 fee.
Chapter 10 of the Consolidated Safety Anatyaie Report for idorric Operation, NED0-21326 including A-series revisions through Revision A4, October 1978 (attached) incorporates the existing license conditions of License No.
SNM-1265 as well as the conditions of License No. SNM-1281.
General infor-mation such as " place of use" contained in the existing licenses is included in Chapter 10 by reference to other chapters of NED0-21326.
Revision A4 also deletes the term " proposed" from the title of Chapter 10, placing the Operation Specifications in use at Morris Operation upon approval of this request.
A comparison of existing license conditions and equivalent or corresponding Operation ~ Specifications in NE00-21326 is contained in Attachment A.
Attach-ment B contains proposed conditions of License No. SNM-1265, including a correction of address and a new Condition 6, which is based on the wording of Amendment 4 to License No. SNM-1265, dated April 1977. The proposed Revision A4, October 1978, to NED0-21326 is also attached.
General Electric Company requests this licensing action to reduce adminis-trative effort associated with two licenses and to obtain the advantages of the single set of Operation Specifications.
Two licenses require separate maintenance activities, including corrections for changes in organization titles and nomenclature and license renewal processes, as well as filing, correspondence, and other administrative re-quirements.
Since the two licenses in question can be easily combined, these redundant activities can be eliminated.
The removal of the detailed conditions from the license and the adoption of Operating Specifications in Chapter 10 of NED0-21326 removes the possibility of conflicts between these documents without in any way reducing the effective-ness of regulatory control, or licensee responsibility. Again, only one dccument (NED0-21326) will need to be revised to accommodate operational changes.
The single set of operation specifications, similar to the technical speci-fications of reactor facilities, will also simplify inspection and audit activities for USNRC I&E functions, state inspectors and internal General Electric auditors.
Please contact H.A. Rogers or C.C. Herrington of tSis office if there are questions regarding this matter.
gettfullysubmitted,
'E%EPAL ElfqTRIC COMPANY cc:
R. Burnett, Director A A Division of Safeguards
[ b' O fM,k12.
U.S. Nuclear Regulatory Comm.
Washington, D.C.
20555 D.M.
- Adon, nager Licensi
& ransportation 408*925-6 MC 861 DMD:HAR:bn
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ATTACHMENT A EQUIVALENT COVERAGE OF LICENSE CONDITIONS LICENSE N0. SNM-1265 & SNM-1281 License No. SNM-1265 Equivalent Coverage Condition (*)
NED0-21326 6.
Material descriptions Section 10.2.1.1 7.
Chemical and/or physical forms Section 10.2.1 8.
Amount possessed Section 10.2.1.1:
9.
Possess and store, only Section 10.1, Revision A4 attached l
- 10. Authorized' place of use Section 10.1.2, Revision A4 attached.
11.
Liquid coolant in received casks Section 10.4.7, Revision A4 attached 12.
Cask coolant sampling Section 10.4.8, Revision A4 attached 13.
Basin leak detection Section 10.4.4.1 Table 10-3 14.
Reactivity limits for stored fuel Section 10.2.1.1-15.
Restriction on fuel storage system Sections 10.5.2.1 & 10,5'.2.2 16.
Parameters for fuel to be stored Section 10.2.1.1. and Figures 10-1 and 10-2 17.
Basin radioactivity l'fil1ts & sampling Section 10.4.10.1, Revision A4 attached 18, Quality assurance provisions Section 10.1.3, Revision A4 also Appendix B.8
- 19. - Transfer canal closing Section 10.1.4, Revision A4 attached 20.
Release of hydrogen fluoride Section 10.1.4.2, Revision 1 A4 attached 21.
Document references New Condition 6, Attachcant B.
- 22.. Storage of slightly irradiatcd Requirement deleted.
enriched uranium fuel assemblies 23.
Basin water chemical composition Section 10.4.9,' Revision A4 attached 24.
Exemption from neutron monitoring Section 10.2.2.1 devices
(*) See Attachment B for new Conditions 1 through 5
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l License No. SNM-1265 Equivalent Coverage Condition (*)
NED0-21326 25.
Sand cilter monitoring Section 10.4.1.1 and Table 10-2 26.
Sampling when bypassing process Section 10.4.6.1 steam generator 27.
Personnel training and safety practice Sections 10.6.3; also Section 9.3, 9.4 License No. SNM-1281 Equivalent Coverage Conditions (**)
NED0-21326 6.
Material descriptions Table 10-1 and Section 10.2.1.1 7.
Chemical and/or Physical forms Table 10-1 and Section 10.2.1.1 8.
Amount possessed Table 10-1 and Section 10.2.1.1 9.
Authorized use Section 10.2.1.1 10.
Authorized place of use Section 10.1.2 11.
Reference to 10CFR20 Section 10.6.3 through Section 10.6.3.2 12.
Designated individual Section 10.6.1, Revision A4 attached 13.
Possession and use of licensed See Amended SNM-1265, New materials Condition 6, Attachment B 14.
Provisions for testing sealed sources Section 10.4.3.1
(**)
See Attachment B for new Conditions 1 through 5
ATTACHl!ENT B LICENSE' CONDITIONS - LICENSE N0. SNM-1265 Pursuant to the Atomic Energy Act of 1954, as amended, and Title 10, Code of Federal Regulations, Chapter 1, part 30, " Rules of General Applicability to Licensing of Byproduct Material, Part 40, " Licensing of Source Material,"
and Part 70, "Special Nuclear Material," and in reliance on statements and representations made by the licensee, a license is hereby issued authorizing the licensee to receive and possess byproduct, source, and special nuclear material designated below; and to transfer such materials to persons authorized to receive them in accordance with the regulations in said Parts.
This license shall be deemed to contain the conditions specified in the Atomic Energy Act of 1954, as amended, and is subject to all applicable rules, regulations, and orders of the Nuclear Regulatory Commission now or hereafter in effect and to the conditions specified below:
1.
Licensee:
General Electric Company 2.
Address:
Spent Fuel Services Operation Nuclear Energy Programs Division General Electric Company 175 Curtner Avenue San Jose, CA 95125 3.
License No:
SNM-1265 4.
Expiration Date:
August 31, 1979 5.
Docket No:
70-1308 6.
The licensee shall possess and store licensed material in accordance with statements and representations contained in the Consolidated Safety Analysis Report for Norris operation, NED0-21326, and subsequent revisions; Chapter 10, operation ~ Specifications shall contain specifi-cations and criteria for routine operation and maintenance of the facility, and these specifications and criteria shall not be changed without prior approval of the Nuclear Regulatory Commission.
7.
The licensee shall comply with requirements of the Physical Security Plan -
Norrio Operation, NEDS-14507, including subsequent revisions, and require-ments of 10CFR70.32(d) or (e).
A.
Special provisions for surveillance of that portion of the Administration Building which constitutes a part of the protected area physical barrier and the isolation zone thereof, are specified in NEDS-14507, Sections 2.1.2.1, 4.1.3.1 and 5.5 and constitute compliance with 10CFR73.50(b)(4).
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- NEDO-213R6 REVISION A4 -' OCTOBER 1978' Lf
. REVIEW DRAFT
~ CHAPTER 10.
GPERATION SPECIFICATIONS b
This draft contains additions to the' Operation Specification' originally submitted _as proposed specifications in June 1977.
4 Pages containing'new material are noted by the phrase " review
't draft"_in the upper right corner of the page. A complete Chapter is furnished to facilitate rev'iew.
t Upon approval of this revisien,-change pages will be issued for NEDO-21326.
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e NED0-21326-2A4 OCTOBER 1978 10.
OPERATION SPECIFICATIONS
10.1 INTRODUCTION
The specification in this chapter establish conditions governiag E
the receipt and. storage of irradiated fuel from light. water reactors.
Eby Morris Operation. These Operation Specificatioca define require-ments that protect the health and safety of the public and employees.
Cpcration of the Morris fuel storage facility cannot result in a sudden large release of radioactivity to the' environs, even under those credible meteorological and seismic conditions that have'been considered in the design basis of the facility. The consequences of accidents have been analyzed and found to have insignificant en-vironmental effects.1 In summary, there are no credible events that could cause a release of radioactivity that would pose a danger to E
the public.
t 10.1.1 Definitions The'fo11owing definitions apply for the purposes of these Operation Specifications:
Safety Limits - Those bounds which if exceeded may affect the a.
health and safety of the public and employees.
b.
Limiting Conditions - The appropriate functional capabilities or performance levels of equipment ana systems for normal operation ~
of the facility, Surveillance Requirements - Requirements for monitoring.. sampling, c.
testing, p3.11brating, and inspecting equipment and systems to demonstrate that functional'capabili. ties or performance levels are maintained as required for normal operation of the f acility.
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1 See an-)2ses in; Chapters 7 and 8, NED0-21326-2 D10 E en -.
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~ June 1977
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Design Features - Features of'the facility associated with the basic design such as materials of construction, geometric arrangements, dimensions, etc., which, if' altered or modified, could have a detri-
. mental effect on safety.
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Administrative Controls.- Provisions relating to organization and management, procedures, record keeping, review and audit, and report-ing necessary to conduct activities in.a manner consistent with operation specifications and applicable government regulations.
i f.
Fuel Bundle - The unit of nuclear fuel in the form that it'is charged or discharged from the. core of a light water reactor (LWR). Normally, it will consist of'a rectangular arrangement of fuel rods held together
-by end fittings, spacers, and tie rods.
The BWR fuel bundle does not include the fuel channel (which is reusable and not shipped with fuel t
bundles).
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Tonne (Te) - One metric ton, equivalent to 1000 kg :or 2204.6 lb.
Fuel quantity is expressed in terms of the uranium content of the fuel measured in metric tons and-written TeU, formerly MTU.
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i 10-2
REVIEW DRAFT i
NED0-21326-2A4
+
OCTOBER 1978 i
10.1.2 Authorized Place.of..Use The irradiated nuclear fuel, as described in Section 10.2, is to be possessed and stored at the Morris Operation located in Grundy N
County, Illinois near Morris, Illinois. This site is described in Chapters 1 and 3 of this document.
10.1.3 quality Assur_ang L Activities at Morris Operation shall be conducted in accordance with
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requirements of Appendix B, 10CFR Part 50, as described in Spent Fuel Services Operation Quality Assurance Plan, NEDO-20776, a's revised (see Appendix B.8).
10.1.4 General Considerations The general considerations of the following subsections are in effect at Morris Operations; change in these considerations shall require prior approval of the U.S. Nuclear Regulatory Commission:
10.1.4.1 Fuel Transfer Canal Closure The upper end of the transfer canal (Figure 1-5 and 1-27) has been sealed by welding a'stainices steel plate, 1/4 inch thick, to imbedded steel angles f raming the opening. There are no protrusions from the plate that could be used to facilitate removal. The fuel basket transfer arm has been rendered inoperative by welding a block in place to prevent arm mevement, and by disabling the arm hydraulic system.
10.1.4.2 Fluorine Facility The fluorine facility, part of the fuel reprocessing facilities, N
shall be inoperative (the majority of fluorine generation equip-ment has been sold and removed).
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10.2 SAFETY LIMITS Safety limits applicable to Morris Operation are found on the basic assumptions of the safety analysis.
If a safety ifmit is exceeded, plant procedures re-quire action to return operations to within Specification requirements.
10.2.1 Authorized Materials 10.2.1.1 Specification Light water reactor nuclear fuel to be received and stored at Morris a.
Operation shall meet the following requirements:
(1 ) Fuel shall contain uranium as uranium dioxide (UO )
2 (2) Fuel shc11 be clad with stainless steel, zirconium or zirconium alloys.
(3) Average exposure of reactor discharge batch (fuel) shall not exceed 44,000 mwd /TeU.
(4) Fuel shall have cooled a minimum of 90 days af ter reactor shut-down and prior to shipping.
(5) Rod lattice k, limits without allowance for burnup shall not exceed:
o 1.40 for 7x7 or 8x8 5'4R o
1.37 for 15x15 PWR (<8.55 inches square) o 1.41 for 14x14 PWR (<7.80 inches square) 6 (6) Fuel parameters shall be within the ranges defined in Figures 10-1 and 10-2.
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NEDO-21326-2A1 June 1977
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b.
The combined quantity of unirradiated natural uranium and unitradiated depleted uranium at the Morris Operation facility shall not exceed 50 Te.
c.
Instrument, calibration, and laboratory sources may be possessed within the limiting amounts given in Table 10-1, 10.2.1.2 Basis The design criteria and subsequent safety analyses of the Morris Operation assumed certain character 1stics and limitations for the fuels that are to be received and stored.
Specification 10.2.1.1.a assures that these bases remain valid by defining the allowable fuel form, cladding, k, and irradiation history. The fuel requirements establish criteria (including k,) for fuel to be stored to protect against an accidental criticality. For the most reactive credible conditions, k,ff for any array of stored fuel must be less than 0.95 at the 95% confidence level.
The design bases for criticality analyses were selected from detailed analyti-cal studies which were based on the physical parameters of specific fuel designs (see Table A10-1, Appendix A.10, NEDO-21326-2).
The largest bundle cross-sectional areas and infinite bundle length were assumed in the calcula-tions.
These limits were based on cold, clean fuel and include allowance for the poisoning ef fect of the stainless steel baskets. Fuel centerline locations and other orientations were assumed to be those giving the maximum system reactivity.
Figures 10-1 and 10-2 provide k, as a f anction of fuel enrichment and reactor type, as well as correction factors for principal variables affecting k,:
the pellet diameter, the water-to-fuel ratio, and the cladding material.
2This limitation does not include uranium in stored fuel, or uranium used in construction of shipping casks such as the GE IF-300.
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Table 10-1 Authorized Materials - Instrument, Calibration, and Laboratory Sources CHEMICAL AND/OR PHYSICAL FORM QUANTITY MATERIAL l
Radionuclides with Solution or Total aggregate atomic numbers ranging calibration disc of five curies from 1 to 83 Cobalt-60 Sealed source 10 curies Cesium-137 Scaled source 10 curies 1 millicurie Thorium-230 Any Neptunium Any 20 grams gj
&8 Plutonium Any 50 grams g g,
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Uranium-235 Any 250 grams 3g(
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enrichment)
Americium-241 Any 200 p Ci Americium-241 Scaled source 40 curies Plutonium-Beryllium Sealed source 2 curies Uranium-natural Any 15 kilograms i
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NEDO-21326-2A1 June 1977 Specification 10.2.1.1.b defines the allowable quantity of unirradiated natural i
and depleted 'itranium to be received and stored.
l Specification 10.2.1.1.c authorizes possession of various isotopes to be used for instrument and calibration sources.
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Juna 1977 10.2.2 Fuel Storage Provisions i
10.2.2.1 Specification Irradiated fuel bundles shall.be stored in authorized fuel storage baskets, mounted in a support grid, in a fuel storage basin.
Fuel storage locations of this specificction shall be exempt from requirements of Section 70.24 of 10CFR70 (see 10.2.2.2, below).
10.2.2.2 Basis The design criteria and subsequent safety analysis for Morris Operation assume irradiated fuel is stored in fuel storage baskets, mounted in a support grid in a fuel storage basin.
Specification 10.2.2.1 assures that these assumptions remain valid.
The last sentence of 10.2.2.1 exempts th M0'from the' requirement to have neutron detectors for criticality monitoring. T rength of neutron radiation from the fuel at the surface of the basin water is below detection levels, even in the unlikely event of a criticality.
3Natural UO, U0, UNH,, and UF6 useJ Juring MFRP testing may be stored in process 3
2 vessels in the canyon area, or in the site warehouse.
4 10-9
NEDO-21326-2A1 Juna 1977 10.3: LIMITING CONDITIONS I
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The limiting conditions described in this section apply to normal operation of '
the Morris Operation facility.
If a limiting condition is exceeded, plant pro-cedures require action to return operations to within specification requirements.
None of the limiting conditions are crucial to public health and safety, or the health and safety of site personnel.
10.3.1 Limiting Conditions - Water Shield 10.3.1.1-Specification The depth of water between the uppermost part of a fuel bundle and the surface of the basin water shall be a minimum of 9 f t.
10.3.1.2 Basis This specification establishes a minimum thickness of water shielding to limit radiation from the fuel stored in the basin area. This specification applies to all fuel in storage or being transferred from cask to. storage location (also, sea 10.5.2).
Tests have shown that the dose rate at the water surface does not increase above background until the water thickness is decreased to about 7 ft.
A conservative water shield thickness of.9 f t (2.74 meters) has been chosen to provide an increased margin of safety.
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NED0-21326-2A1 June 1977 10.3.2 Limiting Condition - Criticality i
10.3.2.1 Specification A atructure (unloading pit doorway guard; Figure 5-3, NED0-21326-1)' shall be used at the doorway between the unloading basin and Storage Basin No. 1 to prevent a basket f rom tipping in a manner such that its contents may be emptied into the unloading basin.
10.3.2.2 Basis The analysis of a fuel basket drop accident (Chapter 8, NEDO-21326-2) indicates that a basket dropped or tipped over in Basin No. 1. near the doorway to the cask unloading basin, could empty its contents into the unloading basin.
It is assumed that the fuel could conceivably fall into an uncontrolled and potentially critical configuration in the bottom of the unloading basin. The unloading pit doorway guard assures that a basket cannot empty its fuel into th'e unloading basin.
The use of the unloading pit doorway guard is described in NED0-21326-1, Chapters 1 and 5; See Section 5.3.4.5.
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,a NED0-21326-2A1 June 1977 10.4 SURVEILLANCE REQUIREMENTS
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Requirements for e rveillance of various radiation levels, water levels, and other physical quantities, as well as inspections and other periodic activities to provide assurance of specification compliance, are contained in this section.
These requirements are samarized in Tables 10-2 and 10-3, from details con-tained in Subsections 10.4.1 through 10.4.6.
10.4.1 Effluent Air Sampling 10.4.1.1 Specification Ef fluent air shall be continuously sampled for particulates at a location between the main stack and the sand filter.
Samples shall be analyzed weekly for gross beta (S) activity. The highest acceptable value shall be a weekly average
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of 4 x 10 pCi/ml.
10.4.1.2 Basis This specification requires sampling of ventilation air leaving the sand filter to provide assurance that effluent concentrations meet regulatory requirements, with resultant offsite concentrations (calculated) within limits established by 10CFR20. The effluent air concentration limit established in Specification offsite concentration will be within 1'0CFR20 limits. The 10.4.1.1 assures that sampling and analysis program provides data for estimating the amounts of radio-active material released to the environment during routine or accident conditions.
10-12 i
REVIEW DRAFT
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NEDO-21326-2A4 OCTOBER 1978 Table 10-2 SURVEILLANCE REQUIREMENTS
SUMMARY
Subsection Quantity or item Period Value
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10.4.1.1 Effluent air W
6: 4 x 10 pCi/ml
-5 10.4.2.1 Effluent air M
B:10 C1/ml a: 5 x 10-6 pCi/ml l(
10.4.2.3 Cask coolant 10.4.3.1
. Sealed sources SA 0.005 pCi 10.4.4.1 Instruments (see Table 10-3) 10.4.5.1 Basin water coolers W
-5 10 Ci/ml a or s 10.4.6.1 Process steam bypass 10.4.9.1 Basin wcter chemical W
pH 4.5 to 9.0 analysis NANO 3 <200 ppm Cl
<10 ppm N.
10.4.10.1 Basin water radioactivity W i
6 key:
- Analysis of samples to occur W: Weekley Q:
Quarterly NR:
Not Required M: Monthly A:
Annual SA:
Semiannual 1
- See text for requirement D10-13 u
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NEDO-21326-2A1 June 1977 Table 10-3
SUMMARY
REQUIREMENTS SYSTEM AND EQUIPMENT TEST AND CALIBRATION Operability Test Calibrate System or Equipment W
M Basin Leak Detection System LAW Vault Leak Detection System Q
NR LAW Vault Intrusion System M
NR Clad Vault Leak Detection System Q
NR Area Radiation Monitors Q
Q A
Q Criticality Monitors Key: Operability test / calibration to occur:
W: Weekly Q: Quarterly NR: Not Required M: Monthly A: Annual SA: Semiannual l
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NEDO-21326-2A 1 June 1977 10.4.2 Effluent Water I
10.4.2.1 Specification Water in the sanitary effluent holding basin and the evaporation pond shall be sampled at least once each month and analyzed for gross alpha and beta radia-tion. Maximum acceptable concentrations shall not e::ceed 10- uC1/ml beta and
-6 5 x 10 pCi/mi alpha radiation. If either pond is dry, no sampling of that pond is required.
10.4.2.2 Basis Periodic sampling and analysis of Morris Operation effluents is prudent, even though it is very unlikely that any radioactive material would be present in sewer effluent. The limits selected are for isotopes that are present at the Morris Operation.
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v NEDO-21326-2A 1 June 1977 10.4.3 Sealed Sources i
10.4.3.1 Specification 1
Each sealed source (not irradiated fuel) containing radioactive material in excess of 100 uCi of beta-gamma emitting material or 5 uCi of alpha-emitting material, shall be free of removable (non-fixed) contamination. The~ maximum acceptable level shall be 0.005 uCi (total, each source), with dry wipe testing to occur at least once every 6 months.
10,4.3.2 Basis Surface contamination is measured to determine that a sealed source has not developed a leak. The limitations on removable contamination are based on 10CFR70. 39(c) limits for plutonium.
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NEDO-21326-2A1 June 1977 10.4.4 Instrumentation 10.4.4.1 Specification Systems and equipment shall be tested for operability and calibrated at least once during the intervals specified in Table 10-3.
Calibration is performed in accordance with manufacturers' recommendations, and operational tests are performed'to check alarm functions and demonstrate other operational features of the system or equipment.
10.4.4.2. Basis t
Bases for these test and calibration requirements are as follows:
a.
Basin Leak Detection System - Operation of this system ensures that a leak in the basin liner will be promptly detected, so that corree,
tive action can be initiated.
Since the operation of the system is related to the level of water in the detection system, the level set
. point requires periodic calibration.
b.
LAW Vault Leak Detection System - Operation of this system ensures that a leak in the LAW vault inner container will be promptly detected.
Since a specific level is not involved, calibration is not required.
c.
LAW Vault Intrusion System - Operation of this system detects external,.
ground water leakage through the concrete structure of the vault, and initiates pumpout action to prevent 1AW vault flooding.
Since a specific level is not involved, calibration is not ' required.
d.
Clad Vault Leak Detection System - Operation of this system provides for detection of water between the vault liner and the concrete structure,' with subsequent pumpout action.
Since a specific level is not involved, calibration is not 'equired.
e.
Area Radiation Monitors - The audible alarm system for these monitors are tested (operated), and the alarm set point calibrated periodically 10-17
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-0 NEDO-21326-2A1 Juns 1977 to provide assurance of reliable operation within equipment specifica-(
tions, to alert personnel to radiation above preset levels.
f.
Criticality Monitors - The audible alarm system for these monitors are tested (operated), and the alarm set point calibrated periodically to provide assurance of reliable operation within equipment specifications, to warn personnel of criticality.
10-18 m
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NEDo-21326-2A1 June 1977 10.4.5 Basin Coolers 10.4.5.1 Specification Basin water crolers that are in service shall be inspected at least once each week, including:
S a.
The equipment shall be visually inspected for signs of leakage with the fans off.
b.
Random smear surveys for removable contamination.
10.4.5.2 Basis Leakage could occur in the coils or piping of the fin-fan coolers, releasing contaminated basin water to the environs.
Routine visual and smear tests are made to detect leakage.
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NEDO-21326-2A1 e
June 1977 10.4.6 Process Steam Bypass i
10.4.6.1 Specification Whenever the process steam generator is bypassed, and utility steam is sub-stituted for process steam, condensate from the process steam condensate system returning to the utility boiler shall be sampled at least once each 12-hr period and analyzed for gross beta activity. The highest acceptable concentration so measured shall not exceed 10-5 pCi/ml.
10.4.6.2 Basis The sampling requirement helps assure that if radioactive material is released in the condensate, it would be discovered quickly.
l l
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REVIEW DRAFT j
NED0-21326-2A4 OCTOBER 1978 10.4.7 Cask Liquid Coolants N
10.4.7.1
_Snecif ication Water shall be the only liquid coolant permitted in all casks received by Morris Operation. Chemical additives to prevent freezing of the water are prohibited. Air-cooled casks may be accepted providing that they can be flushed and otherwise handled as a cask using water coolant.
10.4.7.2
" anis The Morris Operation is not equipped to accommodate liquid coolants other than water.
Chemical anti-freeze additives cannot be accepted because of the effect of such additives on water chemistry and the basin filter system.
10.4.8 Cask Coolant Sampling-N 10.4.8.1 Specification The concentration of radioactive material in the cask coolant, as determined by analysis of the coolant or first cask flush of an air-cooled cask, shall be less than limits specified in 10CFR Part 71.35 (a)(4).
If these limits are exceeded, the fuel in the cask shall be assumed to have f ailed, and action shall be taken in accordance with established procedure.
l'.4.8.2 -Basis 0
Undetected failed fuel could cause complications in operation of the basin filter and cooling systems and, in the extreme, could N
result in radiation exposures greater than ALARA.
r 6 " Coolant" refers to the heat transfer medium used within the cask 3
l 7 Refer to Section 5.3.3.1 D10-20A
.._ _, ~.
REVIEW DRAFT NED0-21326-2A4 OCTOBER 1978 N
10.4.10 Basin Water Radioactivity Sampling 10.4.10.1 Specification Additional basin water cleanup measures shall be initiated if the concentration of radioactive materials in the water exceeds 0.02 pCi/ml.
Fuel receiving operations shall be stopped if the con-centration exceeds 0.1 pCi/ml. The USNRC shall be notified, and immediate measures taken to reduce concentrations below the 0.1 p Ci/ml.
10.4.10.2 Basis Periodic sampling of the basin water is required to assure that radioactivity levels remain as low as reasonably achievable. The N
values selected are consistent with current decontamination practices.
D10-20C
REVIEW DRAFT NED0-23126-2A4 OCTOBER 1978 10.4.9 Basin Water Chemical _ Characteristics 10.4.9.1 Specification Basin water ehemistry shall be maintained as follows:
Item Acceptable Analysis pH 4.5 to 9.0 NANO
< 200 ppm 3
Cl-
< 10 ppm 10.4.9.2 Easis Basin water chemical characteristics are selected to maintain a N
benign environment for stored fuel.
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NEDO-21326-2A1 Juna 1977 10.5 DESIGN FEATURES
(
The design features in the following subsection are those incorporated in the M0 facility for the safe handling and storage of irradiated fuel.
10.5.1 Fuel Storage Basin The energy-absorbing pad on the cask set off shelf shall not be altered with-out appropriate safety review and documentation.
10.5.1.1 Basis The cask drop accident was analyzed for the IF-300 cask with the energy-absorbing pad in place (Chapter 8, NEDO-21326).
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I 10-21 i
NEDO-21326-2A1 June 1977 10.5.2 Fuel Storage System The following pieces of equipment employ favorable geometry, specific materials, and methods of construction to assure nuclear criticality safety. Modifications to the design in dimensions, materials of construction, or construction methods shall not be made without appropriate safety review and documentation.
10.5.2.1 Fuel Storage Baskets 10.5.2.1.1 Basis a.
The neutron attenuation properties of stainless steel are considered in the nuclear safety analysis.
b.
The structural strength, as fabricated, is considered in seismic and tornado accident analyses and related to nuclear safety.
c.
The heat transfer properties are considered in fuel cooling thermal analyses and related to nuclear safety.
10.5.2.2 Basket Support Grids 10.5.2.2.1 Basis a.
The spacing of the grids determines the spacing of fuel that was used in the nuclear safety analysis, b.
The structural strength of the grids and grid-to-wall intertie are integral to the strength of the system during the seismic and tornado conditions, and therefore related to nuclear safety.
10.5.2.3 Fusi Grspples 10.5.2.3.1 Basis Fuel grapples used with the fuel handling crane and those used with the basin s
i 10-22
NEDO-21326-2A1 Juns 1977 i
crane are designed to preclude lifting a fuel but.dle closer than 9 ft to the I
normal water level of the basin.
10.5.2.4 Fuel Basket Grapples 10.5.2.4.1 Basis Basket grapples are designed for use with the basin crane, and are designed to preclude lifting a basket closer than 9 ft to the normal water'1evel of the basin.
i 10-23
e REVIEW DRAFT NED0-21326-2A4 OCTOBER 1978 10.6 ADMINISTRATIVE CONTROLS 10.6.1 Responsibility.
The Manager-Morris Operation shall be responsible for overall facility operation in accordance with these specifications and applicable government regulations, and shall delegate in writing the succession of this responsibility during his absence. Licensed material shall N
be used by, or under the supervision of individuals designated by the Manager - Morris Operation, or his delegate.
10.6.2 Organization 10.6.2.1 'The facility staff organization, as a minimum, shall be as shown in Figure 6-1, 10.6.2.2 Staff Qualifications Minimum qualifications for members of the facility staff shall be the following:
Manager - Morris Operation a.
BS degree in engineering, or related physical science; or o
equivalent in nuclear industrial experience, Demonstrated competence in the technologies and control o
methods applicable te nuclear energy business activities, including radioactive materials handling and radiation and criticality safety considerations.
Ten years of industrial experience with at least five years o
in nuclear facility management.
b.
Plant Operations and Services BS degree in engineering or equivalent in nuclear industrial o
experience.
D10-24
-d s
f NEDO-21326-2A1 June 1977-o-
- Demonstrated competence in the technologies' and control methods
' ('
applicable to nuclear energy business activities, including radioactive materials handling and radiation and criticality
(
saf ety considerations.
Eight years of prior manufacturing or engineering experience, o
. vith at least five. years in the nuclear industry.
i Manager - Plant Engineering and Safety c.
3 BS degree in engineering, or equivalent technical experience.
o o
Thorough knowledge of. radiation and criticality safety require-ments and practice, including safety requirements specifically i
- related to maintenance' operations under radioactive contaminations conditions.
e Five years of industrial experience, with at least three of-o these in the nuclear indastry.
- d.
Manager,-Quality Assurance and Safeguards o
BS degree in engineering, or equivalent technical experience.
s Thorough knowledge of nuclear materials handling, safeguards, and o
quality assurance methods and procedures.
Five' years of experience in manufacturing and quality assurance o
l fields, with at least three, years o~ these in the nuclear industry.
10.6.3 Plans and Procedures
-Plans and:procedur.es shall be established and. implemented to assure compliance with Operation Specifications and appliceLle governmenta1' regulations.
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e NEDO-21326-2A1 June 1977 10.6.3.1 Changes to Plans and Procedures All changes or revisions of. established plans or procedures required by Sub-section 10.6.3 shall be made in accordance with facility modification control practices as described in Subsection 9.4.3, NEDO-21326-2.
10.6.3.2 Plans and Procedurek - Minimum Requirement Plans. 3 procedures required by Subsection 10.6.3 shall include, but need not be limited to, the following:
A safety manual defining responsibilities and specifying actions to a.
protect the health and safety of employees and others, while on site.
b.
An emergency plan that defines responsibilities and specifies actions, including channela of communication required to cope with credible emergencies on site, and during transportation of irradiated nuclear fuel.
Facility change or modification control procedures for facility c.
structures, systems, and components.
d.
Procedures for determining certain characteristics of fuel to be stored, and to verify that fuel me;ts storage criteria.
Plans requiring analyces of cask drop accidents involving types of e.
casks not previously received or unloaded.
f.
Procedures for the conduct of routine futi storage operations.
A preventative maintenance system for structures, systems, and g.
components important to site radiological and criticality safety.
h.
Arrangements and procedures f or providing makeup water en the storage basins under normal and emergency conditions.
10-26
e NEDO-21326-2A1 June 1977 10.6.4 Review and Audit
(
10.6.4.1 Planc Safety Committee Review and audit of plans and procedures, and of operations carried out under established plans and procedures involving elements of radiological safety, shall be conducted by a Plant Safety Committee. This Committee shall consist of the following members, as a minimum:
a.
Manager - Morris Operation b.
Manager - Plant Operation and Services Manager - Plant Engineering and Safety c.
d.
Manager - Quality Assurance and Saf eguards Supervisor - Plant Safety e.
f.
Senior Engineer - Plant Safety and Licensing The Committee shall normally meet on a monthly basis, but at no less than 45-day intarvals. The Manager, Morris Operation, shall establish applicable procedures and practices for the conduct of Committee responsibilities.
10.6.4.2 Audit of Operations Activities of Morris Operation shall be audited to ascertain the degree of com-pliance with specifications, standards, and procedures.
Audits shall be conducted by organizations and persons and at such times as may be designated by General Manager, Fuel Recovery and Irradiated Products Department, and General Manager, Nuclear Energy Programs Division. Audits and audit response shall be performed in accordance with procedures est<blished by General Electric.
I 10-27 i
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- ' 4 NEDO-21326-2A1 June 1977~
10.6.5 Action Required For Specification Noncompliance 9 '.,'
g
'10.6.5.1
' Safety Limits
- The' following actions shall be taken if a safety limit (Subsection 10.2.1 and.10. 2. 2, HED0-21326-2) is found to have been exceeded:
a.'
Prumpt action shall be taken to assure timely return of operations to 1
specification compliance.
j l
i b.
The Plant Safety Committee shall be promptly notified of the non -
t c ompliance.
c.
Notification of NRC Inspection and Enforcement Regional Offices, Region.III, shall be made within 24 hr, advising them of events that resulted-in a safety limit being exceeded.
.3 d.
A review of the incident shall be made by the Plant Safety Committee to establish the cause, and'to define means to prevent reoccurrence.
10.6.5.2 Limiting Conditions i
The following actions shall be taken if a limiting condition is found to have a
been exceeded:
f a.
Prompt corrective action shall be taken to assure timely return of operations to specification compliance.
b.
The Plant Safety Committee shall be advised of the noncompliance l
within 24 hr.
Notification' of NRC Inspection and Enforcement Regional Office, i
c.
Region III, shall be msde-quarterly ~to advise them of events resulting j
in limiting conditions being exceeded, i
l N
'l:
10-28 h
I
NEDO-21326-2A1 1
June 1977 d.
A review of a noncompliance situation shall be made by the Plant Saf ety Committee whenever a given limiting conditon has been exceeded more than once in a period of 3 months, or more than twice in any 12-month period. In these situations, the Committee shall establish the cause and define means to eliminate or reduce the frequency of occurrence.
10.6.5.3 Surveillance Requirements The following actions shall be taken if surveillance requirements are not satisfied:
a.
The Manager, Morris Operation, or his delegate, shall take occh action as may be required to assure future compliance with surveillance requirements, and - if necessary - to assure return of operations to specification compliance in minimum time, b.
The Plant safety Committee shall be advised of any event, or sequence of events, involving surveillance requirements that involve systems directly related to radiological safety. The Committee shall inves-tigate such events, and recommend corrective action.
c.
Notification of NRC Inspection and Enforcement Regional Office, Region III, shall be made quarterly, advising them of events that resulted in a surveillance requirement being violated.
10.6.6.5.4 Design Features Design features shall only be changed in accordanc'e with Subsection 10.6.3.1, and Subsection 9.4. 3, NEDD-21326-2. Unauthorized modificat1ons of specified design features (per Section 10.5), or introduction of unapproved tools, fix-tures, or other equipment, shall require action as specified for limiting con-ditions in Specification 10.6.5.2.
'i-10-29
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=c NEDO-2131 Juna 1' 10.6.6 Logs, Records, and Reports I
10.6.6.1 Logs and Records A shift log shall be maintained to record nonroutine and significant a.
events that may occur during a shift.
b.
Logs, or other records shall be maintained to document essential site operations, such as sample logs, fuel storage locations, and SN>1 Accountability Records.
c.
Minutes of the Plant Safety Committee shall be recorded, including specification noncompliance reports.
d.
All logs or records required by applicable government regulations shall be maintained.
10.6.6.2 Reports Reports shall be prepared and submitted as required by applicable governmental regulations.
a.
Reports of noncomplianc, ith safety limits (Subsection 10.6.5.1) chall be written and submitted within 30 days of the event.
b.
Reports of noncompliance with limiting conditions shall be made quarterly (Subsection 10.6.5.2).
10-30
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