ML20003E296
| ML20003E296 | |
| Person / Time | |
|---|---|
| Site: | 07001308 |
| Issue date: | 01/12/1981 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20003E291 | List: |
| References | |
| NUDOCS 8104020807 | |
| Download: ML20003E296 (23) | |
Text
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- 6 O
ATTACHMENT H PROPOSED TECHNICAL SPECIFICATIONS
1.0 INTRODUCTION
The technical specifications and other license conditions in this document govern the receipt, possession, storage, and transfer of irradiated fuel from light water reactors by Morris Ope-ation.
Operation of the Morris fuel storage facility cannot result in a sudden, large release of radioactive materials or high radiation levels 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 environmental effects.'
In summary, there are no credible events that could cause a release of radioactivity that would pose a danger to the public.*
1.1 DEFINITIONS The following definitions apply for the purpose of these technical specifications:
a.
Administrative Controls: Provisions relating to organiza. tion and management, procedures, record keeping, review and audit, and report-ing-necessary to conduct activities in a manner consistent with
. technical specifications and applicable government regulations.
b.
Design Features: Features of the facility associated with the basic design such as materials of construction, geometric arrange-ments, dimensions, etc., which, if altered or modified, could have a detrimental effect on safety.
c.
Functional and Ooerating Limits: Fuel handling and storage conditions required for normal operation of the facility.
d.
Fuel Bundle: The unit of nuclear fuel in the form that is charged or discharged from the core of a light water reactor (LWR).
- mally, 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 bundles),
s
- e. -Limiting Conditions: The lowest functional capabilities or per-formance levels of equipment and systems for safe operation of the facility.
- See Section 7.0 of-this doc: cent for weferences and notes 810.4020 M N
Surveillance Recuirements: Require-ents for :cnitoring, sampling, testing, calibrating, or inscecting fuel in storage, equipment anc systems to de-onstrate that functional capabilities or perfor ance levels are maintained as required for normal operation of the facility.
g.
Tonne.Te): One metric ton, equivalent to 1000 kg or 220A.6 lb.
fuel quantity is expressed in terms of the heavy metal content of the fuel measured in metric tons and written TeU: forrerly fiTU.
1.2 GENERAL LICENSE CONDITIONS 1.2.1 Authorized Place of Use The irradiated nuclear fuel, as described in Section 2.0, is to te received, possessed, and stored at the Morris Operation located in Grundy County, Illinois, near Morris, Illinois. This site is described in Chapters 1 and 3 of the
" Consolidated Safety Analysis Report for Morris Operation", NED0-21326C-series revisions.
- 1. 2. 2 Quality Assurance,
~
Activities at Morris Operation shall be conducted in accordance with require-ments of Appendix B,10CFR Part 50, as described in Spent Fuel Services Operation Quality Assurance Plan, NEDO 20776, as revised.
(See Acpendix S.S of the CSAR).
1.2.3 Fuel Transfer Canal Closure The upper end of the transfer canal (CSAR Figure 1-5) has been sealed by welding a stainless steel plate, 1/4 inch thick, to imbedded steel angles framing 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 movenent, and by disabling the arm hydraulic system. These conditions shall not be changW without prior approval by the Commission Staff.
- Hereafter refer ~ed :o ca Y.e C5.:R
H-3 1.2.4 Fluorine Facilit, The fluorine facility, part of the fuel reprocessing facilities, has been dis-mantled and shall remain inoperative.
1.2.5 Radiological Emergency Ple.n A radiological emergency plan for the facility shall be established and main-tained in accordance with 10CFR72.33(f).
1.2.6 Physical Security Plan A physical security plan for the facility, including a contingencv plan and a plan for qualification and training of security personnel, shall be established and maintained in accordance with 10CFR73.
2.0 FUNCTIONAL AND OPERATING LIMITS Functional and operating limits applicable to Morris Operation are founded on the basic assumptions of the safety analysis and the design of the facility.
2.1 AUTHORIZED MATERIALS 2.1.1 Specification a.
Light water reactor nuclear fuel to be received and stored at Morris Operation shall meet the following requirements:
(1) Fuel shall contain uranium as uranium dioxide (U0 )
2 enriche6 not greater than 5 percent U-235.
(2) Fuel shall be clad with stainless steel, zirconium or zir-conium alloys.
(.3) Maximum average exposure of reactor discharge batch (fuel) shall be 44,000 mwd /TeU.
(4) Fuel shall have cooled a minimum of one year after reactor shutdown and prior to receipt at Morris Operation.
(5) Rod lattice k limits without allowance for burnup shall not exceed:
o 1.37 for 15x15 PWR (<8.55 inches square) o 1.38 for 10x10 BWR (<5.65 inches square)
H-4 o 1.40 for 7x7 or 8x8 BWR o 1.41 for 14x14 PWR (<7.80 inches square) b.
Fuel parameters shall be within the ranges defined in Figures 2-1 and 2-2, or as otherwise specified in this specification.
(1) Morris Operation is authorized to store stainless steel clad Lacrosse 10x10 BWR fuel pellet diameter of 0.35 inch, a pitch of 0.565 and enriched to a maximum of 3.93% U-235.
c.
The combined quantity of unirradiated natural uranium and un-irradiated depleted uranium at the Morris Operation facility shall not exceed 42 Te.2 d.
Instrument, calibration, and laboratory sources may be possessed within the limiting amounts given in Table 2-1.
Tools and equipment incidental to the conduct of General Electric's nuclear and nuclear-related business which have become contaminated with radioactive materials may be possessed.
Items bearing smearable contamination shall be packaged for stor-age. The total contamination of all tools and equipment shall not exceed 10 Ci as determined by external exposure from the items as packaged for storage.
f.
Tools and equipment specifically related to the conduct of fuel storage operations, such as shipping cask internals, which have become contaminated with radioactive materials may be possessed.
Table 2-1 AUTHORIZED MATERIALS - INSTRUMENT, CALIBRATION, AND LABORATORY SOURCES Chemical and/or Physical Form Quantity Material Solution or Total aggregrate Radionuclides with atomic numbers ranging calibration disc of 5 curies from 1 to 83 Cobal t-60 Sealed source 10 curies Cesium-137 Sealed source 10 curies Thorium-230 Any 1 millicurie Neptunium Any 20 grams Plutonium Any 50 grams Uranium-235 Any 250 grams (in uranium of any enrichment)
Americium-241 Any 200 uCi Americium-241 Sealed source 40 curies Plutonium-Beryllium Sealed source 2 curies Uranium-natural Any 15 kilograms 2.1.2 Basis The design criteria and subsequent safety analyses of the Morris Operation assumed certain characteristics and limitations for the fuels that are to be,
received and stored.
Specification 2.1.la assures that these bases remain valid by defining the allowable fuel form, cladding, k and _ irradiation history.
j
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H-6 Specification 2.1.lb establishes fuel parameters referencing graphical and other criteria.
The fuel requirements establish criteria (including k ) for fuel to be stored to protect against an accidental criticality.
For the most reactive conditions credible, k for any array of stored fuel must be less than 0.95 eff at the 95% confidence level.
The design bases for criticality analyses were selected from detailed analytical studies which were based on the physical parameters of specific fuel designs (see Table A.10-1, CSAR Appendix A.10). The largest bundle cross-sectional areas and infinite bundle length were assumed in the calculations. These limits were based on unirradiated, clean fuel and include allowance for the poisoning effect of the stainless steel baskets.
Fuel centerline locations and other orientations were assumed to be those giving the maximum system reactivity.
Figures 2-1 and 2-2 provide k, as a function 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. Other fuel con-
~ figurations that have been analyzed 6nd reviewed separat'ly may be excepted from e
the limitation of Figures 2-1 and 2-2, as referenced in Specification 2.1.lb.
Specification 2.1.lc defines the allowable quantity of unirradiated natural' and depleted uranium to be received and stored.
Specification 2.2.ld authorizes possession of various isotopes to be used for instrument and calibration sources.
Specification-2.1.le provides for storage of tools and equipment incidental to the conduct of General Electric nuclear businesses while awaiting decontami-nation, reuse, or ultimate disposal. Activity will be back-calculated from measurements of the highest exposure rate at 3 ft from a package, assuming that the radiation originates from a uniform volumetric source having approximately the same dimensions as the package. Unless otherwise determined, gamma emissions of 1 MeV/ disintegration will be assumed.
H-7 Specification 2.1.lf provides for storage of tools and equipment specifically related to the conduct of General Electric fuel storage operations, such as cask internals and yokes, while awaiting decontamination, reuse, or ultimate disposal.
These tools and equipment may be contaminated with Co-60, Cs-137, or other isotopes as encountered in fuel handling and storage activities.
3 2.2 FUEL STORAGE-PROVISIONS 2.2.1 Specification Irradiated fuel bundles shall be stored in authorized fuel storage baskets, mounted in a support grid, under water in a fuel storage basin.
2.2.2 Basis The design criteria and subsequent safety analysis for Morris Operation assume irradiated fuel is stored under water in fuel storage baskets, mounted in a support grid in a fuel storage basin. Specification 2.2.1 assures that these assumptions remain valid. The fuel storage baskets and support grid are those described in CSAR Chapter 5, 3.0 LIMITING CONDITIONS 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.
3.1 LIMITING CONDITION - WATER SHIELD 3.1.1 Specification j
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 ft.
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,-see Section 5.2).
l I
L_ _
H-8 Tests have shown that the dose rate at the water surface does not increase above background a til the water thickness is decreased to about 7 ft.
A conservative water shield thickness of 9 ft (2.74m) has been chosen to provide an increased margin of safety.
3.2 LIMITING CONDITION - CRITICALITY 3.2.1 Specification A structure (unloading pit doorway guard: CSAR Figure 5-3)4 shall be used at the doorway between the unloading basin and Storage Basin No. I to prevent a basket from tipping in a manner such that its contents may be emptied into the unloading basin.
3.2.2 Basis The analysis of a. fuel basket drop accident (CSAR Chapter 8) indicates that a basket dropped or tipped over in Basin No. 1, near the doorway to the cask un-loading basin, could empty its contents into the unloading basin.
It is assumed that the fuel could conceivably fall into an uncontrolled, potentially critical configuration in the bottom of the unloading basin. The unloading pit doorway guard assures that a basket cannot empty its fuel into the unloading basin.
4.0 SURVEILLANCE REQUIREMENTS Requirements for surveillance 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 sunnarized in Tables 4-1 and 4-2, from details contained in Sections 4.1 through 4.6.
H-9 Table 4-1 SURVEILLANCE REQUIREMENTS
SUMMARY
a b
Section Quantity or item Pericd Value 4.1.1 Effluent air W
S: 4x10-8 Ci/ml
- 4. 2.' 1 Water-evaporation pond M
S: 10-5 uCi/mi 5x10-6 uCi/ml and sanitary lagoons 2:
4.3.1 Sealed sources, S, y, n, x SA a 8: 0.005 uCi Sealed sources - a Q
a: 0.005 uC1 4.4.1 Instruments (see Table 4-2) 2 4.5.1 Basin water coolers M
2200 dpm/100 cm smearable 4.6.1 Process steam bypass b
10-5 Ci/ml 4.7.1 Cask coolant identification b
Water 4.8.1 Cask coolant b
Not greater than 10CFR71.35(a)(4) 4.9.1 Basin water W
pH:4.5 to 9.0 NANO : <200 ppm 3
Cl~: <10 ppm 4.10.1 Basin water W
0.1 <Ci/ml aAnalyses frequency:
W:
Weekly Q: Quarterly NR: Not Required SA: Semiannual
-M:
Monthly A: Annual bSee text for. requirements
=_
H-10 l
Table 4-2
SUMMARY
REQUIREMENTS SYSTEM AND EQUIPMENT TEST CALIBRATION I
Operability a
System or Equipment Testa Calibrate Basin Leak Detection System W
M i
LAW Vault Leak Detection System Q
NR LAW Vault Intrusion System M
NR Cladding Vault Leak Detection System Q
NR Area Radiation Monitors Q
Q Criticality Monitors A
Q a0perability test / calibration frequency:
W:
Weekly Q: Quarterly NR: Not Required M:
Monthly A: Annual SA: Semiannual =
H-ll 4.1 EFFLUEllT AIR SAfiPLIflG 4.1.1 Specificatio.'
Effluent air shall te 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 maximum value shall be a weekly average of 4x10-8 uCi/ml.
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 off-site concentrations (calculated) within limits established by 10 CFR 20. The effluent air concentration limit established in Specification 4.1.1 assures that off-site concentration will be within 10 CFR 20 limits. The sampling and analysis program provides data for estimating the amounts of radio-active material released to the environment during routine or accident conditions.
4.2 EFFLUENT WATER SAftPLING 4.2.1 Specification Water in the sanitary holding basin and the evaporation pond shall be sampled at least once,each month and analyzed for gross alpha and beta radiation.
-5
' The maximum concentrations shall not exceed 10 uCi/ml beta and 5x10-6
~
uCi/ml.' alpha radiation.
If either pond is dry, no sampling of that pond is required.
4.2.2 Basis Periodic sampling and analysis of Morris Operation effluents is prudent, even
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though it is very unlikely that v ' adioactive material would be present in sewer effluent. The limits selected are for isotopes that are present at the Morris Operation.
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H-12 4.3 SEALED SOURCES 4.3.1 Specification Each licensed sealed source (not irradiated fuel) containing radioactive material in excess of 100 uCi of beta-gamma emitting material or 10 uCi of alpha-emitting material shall be tested for leakage at least once every 6 months, except that each source designed for the purpose of emii.cing alpha particles shall be tested at intervals not to exceed 3 months. The maximum level of removable (non-fixed) contamination shall be less than 0.005 uCi total for each source, using dry-wipe testing techniques.
4.3.2 Basis Surface contamination is measured to determine that a sealed source nas not de-veloped a leak. The limitations on removable contamination are based on 10 CFR 70.39(c) limits for plutonium, but other provisions of this reference are not applicable.
4.4 INSTRUMENTATION 4.4.1 5pecification Systems and equipment shall be tested for operability and calibrated at least once during the intervals specified in Table 4-2.
Calibration shall be performed in accordance with manufacturer's recommendations, and operational tests shall be performed to check alarm functions and demonstrate other operational features of the system or equipment.
4.4.2 Basis 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 correc-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 is checked and instruments receive periodic calibration.
H-13 l
l 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 groundwater leakage through the concrete structure of the vault, and initiates pump-out action to prevent LAW vault flooding. Since a spe-cific level is not involved, calibration is not required.
d.
Cladding Vault Leak Detection System: Operation of this system provides for detection of water between the vault liner and the concrete structure, with subsequent pump-out action. Since a specific level is not involved, calibration is not required.
e.
Area Radiation Monitors: The audible alarm system for these monitors is tested (operated), and the alarm set point calibrated periodically to provide assurance of reliable operation within equipment specifica-tions, to' alert personnel to radiation above preset le"?ls.
Crit cality Monitors: The audible alarm systems for these monitors, i
f.
which warn personnel of a criticality, are tested (operated) and the alarm set point calibrated periodically to provide assurance of reliable operation within equipment specifcations.
14.5 COOLERS i
425.1 Specification-sin water coolers that are_in service shall be inspected at least once each i
month:
a.
The equipment shall be visually inspected for signs of leakage with i
the fans off.
'b.
Random snear surveys for removable beta contamination. shall be no more than 2
2200 dpm/100 cm,
i
_m.
l H-14 4.5.2 Basis Leakage coulo occur in th coils or piping of the fin-fan coolers, releasing contaminated basin water to the environs. Rcutine visual and smear tests are made to detect leakage.
4.6 PROCESS STEAM BYPASS 4.6.1 Specification Whenever the process steam generator is bypassed and utility steam is sub-stituted for process steam to operate the low activity waste evaporator, con-densate from the process steam condensate system returning to the utility boiler shall be sampled at least once each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of such operation and analyzed for gross beta activity. The highest acceptable concentration so measured shall not exceed 10-5 uCi/ml.
4.6.2 Basis The sampling requirement helps assure that if radioactive material is released in the condensate, it would be discovered quickly.
6 j
4.7 CASK C00LANTS 4.7.1 Specification Water shall be the only liquid coolant permitted in all casks received by Morris i-Operation.
Chemical additives to prevent freezing of the water are prohibited.
Gas-cooled casks may be accan*ed providing that they can be flushed and other-wise handled as a cask u-
, water coolant.
4.7.2 Basis l
The Morris Operation is not normally equipped to accommodate liquid coolants other than water.
4.8 CASK COOLANT SAMPLING 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 failed, and action shall be taken in accordance with established procedures.
t
e i
H-15 4.8.2 Basis This specification provides for detection of off-standard conditions within a cask so that the need for special handling or other considerations can be evaluated.
4.9.
BASIN WATER CHEMICAL CHARACTERISTICS 4.9.1 Specification Basin water chemistry shall be maintained as follows:
Item Acceptable Analysis pH 4.5 to 9.0 NANO 3
<200 ppm Cl-
<10 ppm 4.9.2 Basis Basin water chemical characteristics are selected to maintain a benign environ-ment for stored fuel and equipment in the basin water.
4.10 BASIN WATER RADI0 ACTIVITY SAMFLING 4.10.1 Specification Additional basin water cleanup measures shall be initiated if the concentration of radioactive mate-ials in the water exceeds 0.02 uCi/ml beta.
Fuel receiving opera-tions shall be stopped if the concentration exceeds 0.1 uCi/ml beta. The USNRC shall be notified and immediate measures taken to reduce concentrations below 0.1 uCi/ml prior to continuation of fuel receiving operations.
4.10.2 Basis Periodic sampling of the basin water is required to assure that radioactivity levels remain as low as reasonably achievable. The values selected are consis-tent with current decontamination practices.
5.0 DESIGN FEATURES The design features in the following section are those incorporated in the Morris Operation facility for the safe handling and storage of irradiated fuel.
5.1 FUEL STORAGE BASIN The energy-absorbing pad on the cask set-off shelf shall not be altered without appropriate safety review and documentation.
J
H-16 5.1.1 Basis The cask drop accident was analyzed for the IF-300 cask with the energy-absorb-ing pad in place (CSAR Chapter 8).
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.
5.2.1 Fuel Storage Baskets 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 is related to nuclear safety.
c.
The heat transfer properties are considered in fuel cooling thermal analyses and are related to nuclear safety.
5.2.2 Basket Suoport Grids 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 design seismic and tornado conditions, and therefore related to nuclear safety.
5.2.3 Fuel Grapples 5.2.3.1 Basis Fuel grapples used with the fuel handling crane and those used with the basin crane are designed to preclude lifting a fuel bundle closer than 9 ft to the normal water level of the tasin.
~
5.2.4 Fuel Basket Grapple.;
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 level of the basin.
H-17 J
6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY The Manager - Morris Operation shall be responsible for overall facility opera-tion in accordance with these specifications and applicable government regula-tions, and shall delegate in writing the succession of this responsibility during his absence. Operations involving licensed materials shall be performed by, or under the supervision of individuals desigr.RA by the Manager - Morris Operation, or his delegate.
6.2 ORGANIZATION 6.2.1 The facility staff organization is shown in the CSAR, Figure 9-2.
6.2.2 Staff Qualifications Minimum qualifications for members of the facility staff shall be the following:
a.
Manager - Morris Operation o BS degree in engineering, or related physical science, or equivalent in nuclear industrial experience.
o Demonstrated competence in the technologies and control methods applicable to nuclear energy business activities, including radio-active materials handling and radiation and criticality safety considerations.
o Ten years of industrial experience with at least five years in nuclear facility management.
b.
Manager - Plant Operations o BS degree in engineering or equivalent in nuclear industrial experience.
o Demonstrated competence in the technologies and control methods i
applicable to nuclear energy business activities, including radio-active materials handling and radiation and criticality safety considerations.
~
o Eight yea-s of prior manufacturing or engineering experience, with at least five years in the nuclear industry.
c.
Manager - Plant Engineering and Maintenance o BS degree in engt 9ering, or equivalent technical experience.
o Thorough knowledge of radiation and criticality safety requirements l
and practice, including safety requirements specifically related to maintenance operations under radioactive contamination conditions.
t
H-18 o Five years of industrial experience, with at least three of these in the nuclear industry.
d.
Manager - Quality Assurance and Safeguards BS degree in Engineering, or equivalent technical experience o
Thorough knowledge of nuclear materials handling, safeguards, o
and quality assurance methods and procedures.
Five years of experience in manufacturing and quality assurance o
fields, with at least three years of these in the nuclear industry.
Senior Engineer - Licensing & Radiological Safety e.
BS degree in Engineering, or equivalent technical experience.
o Specialized knowledge of health physics, and thorough knowledge o
of radiation and criticality safety requirements for nuclear industr. ' operations, and knowledge of regulatory requirements and practice.
Eight years of Engineering experience with at least five years o
in the nuclear field.
6.3 PLANS AND PROCEDURES Plans and procedures shall be established and implemented to assure compliance with Technical Specifications and applicable governmental regulations.
6.3.1 Changes to Plans and Procedures All changes or revisions of established plans or p acedures required by this section shall be made in accordance with facility modification control practices as described in the CSAR, Sections 9.4.3 and 9.4.4.
6.3.2 Plans and Procedures - Minimum Requirement Plans and procedures required by this section 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 o'f employees and others while on site, safety training programs as appropriate, and other measures to main-tain exposures as low as reasonably achievable, b.
A radiological erergency plan that defines responsibilities and specifies actions, including channels of communication required to cope with credible emergencies on site (Radiological Emergency Plan for Morris Operation, NEDE-21894, as revised. This plan requires Commission approval.)
3
H-19 c.
Facility change or modification control procedures for facility structures, systems and components.
d.
Procedures for determining certain characteristics of fuel to be stored, and to verify that fuel meets storage criteria.
e.
Plans requiring analyses of cask drop accidents for types of casks not previously received or unloaded.
f.
Procedures for the conduct of routine fuel storage operations.
g.
A preventative maintenance system for structures, systems, and com-ponents important to site radiological and criticality safety.
h.
Arrangements for providing makeup water to the storage basins under normal and emergency conditions.
1.
Arrangements for an invironmental monitoring program to demonstrate compliance with techr.ical specifications for effluents.
J.
Plans and implementing procedures for the training and certifi-cation of emoloyees, as described in the CSAR, Section 9.3.
These plans require Comission app-oval.
k.
Plans for decomissioning the facility, including decontamination of structures and equipment and removal of radioactive wastes and contaminated materials. These plans require Commission approval, 1.
Physical security and contingency plans, and other plans and cro-cedures as may be required under 10CFR73 to provide for the physical security of stored fuel. These plans require Comission approval.
m.
Procedures sufficient to account for spent fuel in storage.
64 REVIEW AND AUDIT 6.4.1 Plant Safety Comittee Plans, procedures, and the operations carried out under established plans and procedures involving elements of radiological safety shall be reviewed and approved by a Plant Safety Comittee. This Comittee shall consist of the following members, as a minimum:
a.
Manager - Morris Operation b.
Manager - Plant Operations c.
Manager - Plant Engineering and Maintenance
- d. ' Manager - Quality Assurance and Safeguards e.
Supervisor - Plant Safety f.
Senior Engineer - Licensing and Radiological Safety
H-20 The Committee shall normally meet on a monthly basis, but at no less than 45-day intervals. The Manager - Morris Operation shall establish appropriate procedures and practices for the conduct of Committee responsibilities.
6.4.2 Audit of Operations Activities of Morris Operation shall be audited to ascertain the degree of compliance with specifications, standards and procedures. Audits shall be con-ducted by organizations and persons and at such times as may be designated by Manager - Spent Fuel Services Operation and General Manager - Nuclear Fuel and Services Division. Audits and audit response shall be performed in accordance with procedures established by General Electric.
6.5 ACTION REQUIRED FOR SPECIFICATION NON-COMPLIANCE 6.5.1 Functional and Operating Limits The following actions shall be taken if a functional or operating limit (Sections 2.1 and 2.2) is found to have been exceeded:
a.
The Plant Safety Committee shall be promptly notified of the non-compliance.
b.
When feasible, prompt action shall be taken to assure timely return of operations to specification compliance.
c.
Notification of NRC Inspection and Enforcement Regional Offices, Region III, shall_be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, advising them of
- events that resulted in a non-compliance condition.
d.
A review of the incident shall be made by the Plant Safety Comittee i
to establish the cause and to define means to prevent re-occurrence.
6.5.2 Limiting Conditions l
The following actions shall be taken if a limiting condition is found to have been exceeded:
a.
Prompt corrective action shall be taken to assure timely return l
of operations to specification compliance.
b.
The Plant Safety Committee shall be advised of the non-compliance within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l i
l
[
. = _.
H-21 c.
Notification of NRC Inspection and Enforcement Regional Office, Region III, shall be made at the time of the next inspection to advise them of events resulting in limiting conditions being exceeded.
d.
A review of a noncompliance situation shall be made by the Plant Safety Committee whenever a given limiting condition 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.
6.5.3 Surveillance Requirements The following actions shall be taken if surveillance requirements are not satisfied:
The Manager - Morris Operation, or his delegate, shall take such action a.
as may be required to assure future compliance with surveillance require-I ments, 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 at the time of the next inspection, advising them of events that resulted in a surveillance requirement being violated.
b 1
- _. -. _ -. _ _. _. - _--. -. - _ = _ _ - -
i H-22 i
6.5.4 Design Features Design features shall only be changed in accordance with Specification 6
6.3.1, and CSAR Sections 9.4.3 and 9.4.4.
Unauthorized modifications of spect-fied design features, or unauthorized introduction of unapproved tools, fixtures, or other equipment shall require action as spceified for limiting conditions in Specification 6.5.2.7 6.6 LOGS, RECORDS AND REPORTS 6.6.1 Logs and Records a.
A shift log shall be maintained to record non-routine and sig-nificant events that may occur during a shift.
- b. Minutes of the Plant Safety Committee shall be documented, includ-ing copies of reports required in Section 6.5.1, and other actions of the Comittee.
c.. Records of facility changes, and changes in procedures described in the CSAR shall be maintained throughout the lifetime of the facility.0
- d. Records of tests or experiments conducted under provisions of CSAR Section 9.4.4 shall be maintained throughout the lifetime of the facility, and shall include written safety evaluations that provide the bases for determining that the test or experiment did not in-volve unreviewed safety or environmental questions.
f
. e. Records of spent fuel receipt or transfer, and annual inventory shall be maintained in duplicate, with one set of records at a separate location, as specified in 10CFR72.51.
- f. Other records shall be established and maintained in accordance with 10CFR72.55.
(
. 6.6.2 Reports-
- a. An annual report shall be made to the Comission sumarizing changes, tests and experiments,. including safety evaluations, conducted with-(
out prior approval by the Comission.
- b. Accidental criticality shall be. reported imediately to the NRC regional office by telephone and telegram or teletype.
1
- c. Any loss of-special nuclear material shall be reported imediately to the NRC regional' office by telephone and telegram or teletype.
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- d. Material status reports shall be made to the Commission in accordance with instructions on Form NRC-742 as of March 31 and September 30 of each year, as specified in 10CFR72.53.
- e. Upon-receipt or transfer of spent fuel, a Nuclear Material Transfer Report (Form NRC-741) shall be made in accordance with 10CFR72.54.
- f. Other reports shall be made in accordance with 10CFR72.33 and 72.55.
6.6.3 CSAR Annual Review The CSAR shall be reviewed and updated annually in accordance with 10CFR72.50.
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7.0 REFERENCES
AND NOTES 1.
See analyses in Chapters 7 and 8, CSAR.
2.
This limitation does not include uranium in stored fuel, or uranium used 7
in construction of shipping casks such as the GE IF-300.
y, 3.
Natural UO, UO, UNH and UF used during MFRP testing may be stored 3
2 6
in process vessels in the canyon area, or in the site warehouse.
4.
The use of the unloading pit doorway guard is described in the CSAR, Chapters 1 and 5; see Section 5.4.3.3.
5.
Dry to the extent that water samples cannot be obtained in the usual manner.
6.
" Coolant" refers to the heat transfer medium used within the cask.
~7.
Authorized modifications and approved tools, fixtures, or other equip-ment are those processed under the provisions of CSAR Sections 9.4.3 and 9.4.4.
8.
Significant changes in structures, systems and procedures described in the CSAR, when of a permanent nature shall be recorded by incorporation in the CSAR.
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