ML19274E379
| ML19274E379 | |
| Person / Time | |
|---|---|
| Site: | 07001308 |
| Issue date: | 02/23/1979 |
| From: | Dawson D GENERAL ELECTRIC CO. |
| To: | Cunningham R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 7903230234 | |
| Download: ML19274E379 (45) | |
Text
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cE m M y/ ELECTBiC
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- ., E; PROGRAMS DIVISION GENERAL ELECTRIC COMPANY,17s CURTNER AVE., SAN JOSE,gA5
- fiORNIA 9s12s a
SPENT FUEL SERVICES OPEP.ATIO:!
7 N "-
MMFEUC DOCUMENT KCQq V1 ~
DMD-299 L-
' "q W
Docket No. 70-1308 License SNT1265 ~
February 23, 1979 Office of Nuclear Material Safety and Safeguards Attn:
Richard E. Cunningham, Director Division of Fuel Cycle & Material Safety U.S. Nuclear Regulatory Commission Washington, D.C.
20555
SUBJECT:
CHAPTER 10, OPERATING SPECIFICATIONS; NED0-21326, CONSOLIDATED SAFETZ ANALYSIS PZPORT - MOPRIS OPMATION, REVISION A4, JANUARY 1979
Reference:
Ltr, D.'
M.
Dawson (GE) to R.E. Cunningham (NRC) dtd October 15, 1978; Request to Amend Conditions of License SNM-1265 and Cancel License SNM-1231 Gentlemen:
General Electric Company previously requested that the conditions of License SNM-1265 be amended, and that License Sf!Il-1231 be cancelled (reference above). A draft of Chapter 10, designated to become Revision A4, NED0-21326, was included in the referenced submittal for review by your staff.
During review, your Mr. F. Empson and our Mr. H. Rogers have dis-cussed various aspects of the draft specifications in Chapter 10 by telephone (such as on December 20,1978), finding no unresolved difference of opinion.
Consequently, we have prepared and submit herewith Revision A4 of NED0-21326.
Attachment A to this latter centains a comparison of the draft and final versions of this re-vision, with notations regarding differences.
Please contact H. A. Rogers or C.C. Herrington of this office if there are questions regarding this matter.
Respectfully submitted, qq5g N
- LELEE]ICCOMPANY GEt 5
{khfe h dlj *^ \\A Ipc M di' f
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v Y a W36 i
- 4. Dawson, Manager Licensing i Transportation.
408*925-6330 MC 861 DMD:HAR:bn Attachment M3MoGt C
1
ATTACHMENT A NED0-21326, CHAPTER 10 DRAFT 's FINAL COMPARISON v
Sections of the final version of Chapter 10 bnclosed) that differ in technical require-ments from those of the draft are discussed.
Editorial changes are not mentioned unless a change in technical requirement is involved.
Section Comment 10.1.4.2 The fluorine facility, in effect, no longer exists; language changed to reflect actual status 10.2.1.1,a.,(5)
Added data for Lacrosse 10x10 fuel
- b. (1) Same as above e.
Provided for storage of tools and equipment incidental to GE's business that are contaminated f.
Provided for possession of tools and equipment incidental to fuel storage (yokes, grapples, etc.)
10.2.1.2 Changes in bases for above 10.2.2.1 Added reference to neutron detectors 10.2.2.2 Added reference to Chapter 5 description of fuel storage system Table 10-2 Values have been augmented (e.g. sealed sources), and errors corrected (e.g.10.4.2.1 reference to " air" instead of water).
10.4.3.1 Language changed to be more directly related to existing license requirements, and to clarify reference to 10CFR70.39(c).
10.4.5.1 Inspection period changed to monthly to reflect practice, and contamination limit established.
10.4.8.2 Restated basis 10.6.2.2 Changed to reflect position title changes 10.6.3.2a.
Added reference to safety training programs b.
Changed to reflect new Radiological Emergenc; Plan, NEDE-2114 10.6.4.1 Changed to reflect position title changes
Attachment A '
I Section Comment
)
10.6.4.2 Changed to reflect practice 10.6.5.2c.
Changed to reflect current practice (i.e. notification of NRC inspectors during inspection).
10.6.5.3c.
Same as above 10.6.6.2 Changed to reflect current practice; see 10.6.5.2 and 10.6.5.3.
h T
NEDo-21326 January 1979 Consolidated Safety Analysis Report - Morris Operation NEDO-21326, Volumes 1 and 2 REVISION INDEX for Revision A4, January 1979 Incorporation of Operation Specifications, Chapter 10.
Pages *o be Removed New Pages to be Inserted Chapter Page Number Date Chapter Page Number Date Table of Table of Contents v
8/77 Contents v
1/79 List of List of Illustrations viii 8/77 Illustrations viii 1/79 List of Tables x
1/79 Volume 2 Volume 2 10.
Entire Chapter 6/77 10.
Entire Chapter 1/79 1
1 of 1
NEDO-21326-A4 January 1979 TABLE OF CONTENTS (Continued)
Page 3.6 Low Activity Waste Vault Drainage 8-15 8-15 8.7 Fuel Drop Accidents 8.8 Tornado-Oenerated Missile Accident 8-26 8-31 8.9 Cooling System Leak 8.10 Criticality Accident 8-33 9
CONDUCT OF OPERATIONS 9-1 7-1 91 Introduction 9-)
9.2 Corporate Organization 9-11 93 Training Programs 0-11 9.4 Normal Operations 4-18 9.5 Bmergency Plan 7-)?
9.6 Decocmissioning 10-1
- 10. OPERATION SPECIFICATIONS 13 1 10.1 Introduction 10 'i 10.2 Safety Limits
!0-11 10 3 Limiting conditions 10-12 10.4 Surveillance Requirements 10-?6 10.5 Design Features
'0-?"
10.6 Administrative Controls 11-1
- 11. QU ALITY ASSURANCE 11-1 11.1 IntroJuction and Summary APPENDICES A-1 A.
See Index following divider B-1 B.
See Index following divider v
NEDO-21326-A4 January 1979 LIS* OF ILLUSTRATIONS (Continued)
Figure Title Page 5-6 Typical Grid Assembly 9-24 5-7 Bottom View - PWR Module Showing Lattices and 7.lnkage 5-25 5-8 Details of Lock Mechanism 5-27 5-9 Excavation at Morris Operatien 5-30 5-10 Foundation Construction 5-31 5-11 Stainless Steel Basin Linara S-34 5-12 Water Intrusion c 49 5-13 LAW Vault Under Construction 5-50 5-14 Arrangement - LAW and Cladding Vault Equipment Pits 5-51a 7-1 History of the M0 P2e L Storage and the Basin Water Activity 7 -7 7-1A Radiation ProtectLon Monitor Locations 7-22a 7-2 Monitor and Sampling Locations (Within 2 Miles of Stte) 7-31 7-3 Environs Monitoring Locations (Within 5 Miles of Site) 7-32 74 Davironmental Monitoring Stations (Within 5 to 15 Miles of Site) 7-33 8-1
'ent Diagram for Postulated Accidents B-3 8-2 Kr-85 Activity as Function of Cooling Time for Different R2el Exposures (Total Inveritory in Fuel Rod) 3 1~
8-3 Iodine, Krypton and Xenon Decay S-18 S4 PWR Fuel Bundle Array at 2-Inch Separation 3-36 8-5 Close-Packed Array of Four BWR Bundles 8-39 9-1 Corporate Organization Chart - General Electric Company 9-2 3-2 Naclear Energy Programs Division Organization 1'
9-3 FR&IPD and FR0 Organization Chart 94 94 Morris Operations Organization Outline 98 9-5 Facility Modification Control Concepts 9-15 9-6 Emergency Organizations 9-23 10-1 Rod Lattice k. - FWR Fuel 10-7 10-2 Rod Lattice '<, - SWR Fuel 11-3 viii
NEDO-21326-A4 January 197o LIST OF TABLES (Continued)
Table Title Page 10-1 Authorized Materials - Instrument, Calibration, and Laboratory Sources 10-6 10-2 Surveillance Requi.mments Summary 10-14 10-3 Sumary Requirements System and Equipment Test and Calibration 10-15 X
N EDO-21326 January 1979 CONSOLIDATED SAFETY ANALYSIS REPORT NEDO-21326 Vnd 1 & Vol 2 REVISION
SUMMARY
Re vmon &
Amendment Date Summary NE DO-21326-1,77 Onqinai ssue - for USNRC Review N E DO-21326-2 1/77 Orgnal issue - for USNRC Revie<<
NEDO-21326-1 A 4/77 incorporate orrections N E DO-21326-2A 4e77 incorporate corrections rcposed Operation Spenu ations NE DO-21326-2A t 4,77 Chapter 10 a NE GO-21326-1 A2 8/77 incorporate torrections and new geological and hydrologicalinformation N DO 21J26-2A2 8/77 incorporate entrections and new Aopendices B 9 througn B '4 NF DG 213261 A3 2/78 incorporate changes and corrections NF Y 9
- 126-2 A3 2/78 incorporate enanges and corrections and new Appench sections B 15 and 8.16; responses to U5NRC questions (O t t NEDO-21326-2A4 1/79 incorporation of Operation Specifications Chapter 10 Revision Coding Key: New or cn6nged irdC-mation 's ndicated by vertical bars in the
,it-nano margin oppos;te the new or cnanged information N ~ indicates new information r indicates editonal changes or corrections Oct. men' Nomoer Key NEDO 21 326 1 A4 Prefix Senal
<Vo R
Amendment
NEDO-21326-2A4 January 1979 10.
OPERATION SPECIFICATIONS
10.1 INTRODUCTION
The specifications in this chapter establish conditions governing the receipt, E
possession, storage, and transfer of irradiated fuel from light water reactors by Harris Operation. These Operation Specifications define requirements that pro-tect the health and safety of the public and employees. Operation Specifications f
I may not be changed without prior approval or the U.S. Nuclear Regulatorv Commission (USNRC).
l Operation of the Morris fuel storage facility cannot result in a sudden, large release of radioactivity to the environs, even under those credible meteoro-logical and seismic conditions that have been considered in the design basis of the facility. The consequences of accidents have been analyzed and found i
to have insignificant environmental effects.1 In summary, there are no
'I credible events that could cause a release of radioactivity that would pose a danger to the public.
I 10.1.1 Definitions The following definitions apply for the purposes of these Operation Specifications:
Administrative Controls - Provisions relating to organiza son and a
management, procedures, record keeping, review and audit. and reporting necessary to conduct activities in a manner consistent wito operation specifications and applicable government regul tions.
b.
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.i det ri-mental effect on safety.
See analyses in Chapters 7 and 8.
10-1
NEDO-21326-2A4 January 1979 c.
Fuel Bundle - The unit of nuclear fuel in the form thrt 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 reusabic and not shipped with fuel bundles).
d.
Limiting Conditions - The appropriate functional capabilities or performance levels of equipment and systems for normal operation of the facility, Safety Limits - Those bounds which if exceeded may affect the health e.
and safety of the public or employees, f.
Surveillance Requirements - Regtnirements for monitoring, sampling, testing, calibrating, and inspecting equipment and systems to demon-strate that functional capabilities or performance levels are main-tained as required for normal operation of the facility.
g.
Tonne (Te) - One metric ton, equivalent to 1000 kg or 2204.6 lb.
Fuci quantity is expressed in terms of the uranium content of the fuel ueasured in metric tons and written TeU, formerly MTU.
10-2
NEDO-21326-2A4 January 1979 10.1.2 Authorized Place of Use The irradisted nuclear fuel, as described in Section 10.2, is to be received, possessed, and stored at the Morris Operation located in Grundy County, Illinois N
f near Morris, Illinois. This site is described in Chapter 1 and 3 of this document.
10.1.3 g_uality Assurance Activities at Morris Operation shall be conducted in accordance with require-7 Sarvve ments of Appendix 3, 10CFR Part 50, as described in Spent Fae Crerat.cn Ghality Assurance Plan, NEDO-20776, as revised (see Appendix B.8).
I 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) has been sealed by weld-ing 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 movement, and by disabling the arm hydraulic system.
10.1.4.2 Fluorine Facility The fluorine facility, part of the fuel reprocessing f acilit ies, has been dismantled and is inoperative. The majority of fluorine generation equip-N ment has been sold and removed.
10-3
NEDO-21326-2A4 January 1979 10.2 SAFETY LIMITS Safety limits applicable to Morris Operation are founded on the basic assumptions of the safety analysis.
If a safety limit 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:
(N (1) Fuel shall contain uranium as uranium dioxide (UO )
(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 after reactor shut-down ar.d prior to shipping.
(5) Rod lattice k. limits without allowance for burnup shall not exceed:
1.37 for 15x15 PWR (<8.55 inches square) o 1.38 for 10x10 BWR (<3.65 inches square) l N
1.40 for 7x7 or 8x8 BWR l
1.41 for 14x14 PWR (<7.80 inches square) i o
Fuel parameters shall be within the ranges defined in Figure 10-1 b.
and 10-2, or as otherwise specified in this section:
(1) Morris Operation is authorized to stor.e stainless steel clad La Crosse 10x10 BWR fuel with pellet diameter of 0.35 inch, a pitch of 0.565 and enriched to a maximum of 3.93 percent U-235.
f 10-4
NEDO-21326-2A4 January 1979 The combined quantity of unirradiated natural uranium and unirradiated c.
depleted uranium at the Morris Operation facility shall not exceed lE 42 Te.
d.
Instrument calibration and laboratory sources may be possessed within the limiting amounts given in Table 10-1.
e.
Tools and equipment incidental to the conduct of General Electric's nuclear and nuclear-related business which have become contaminated with low specific activity radioactive materials may be possesned.
Items bearing smearable contamination shall be packaged for storage.
The overall contamination 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.
10.2.1.2 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 10.2.1.1.a. assures that these bases remain valid by defining the allowable fuel form, cladding, k,and irradiation history.
Specification 10.2.1.1.b. 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 credible e.onditions, k for any array of stored fuel must be less g
than 0.95 at the 95% confidence level.
The design bases for criticality analyses were selected from detailed analytical studies which were based on t physical parameters of specific fuel designs (see Table A10-1, Appendix A.10).
The largest bundle cross-sectional areas and infinite bundle length were assumed in the calculations. These 1imits E
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 cl._ 2aximum system reactivity.
This limitation does not include uranium in stored fuel, or uranium used in construction of shipping casks such as the GE IF-300.
10-5
Tat ic 10-1 Authorized Materials - Instrument, Calibration, and Laboratory Sources
_CilF?t1 CAL AND/OR MATFRIAL PilYSICAL FORM QUANTITY 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 Sealed source 10 curies Thorium-230 Any 1 millicurie Neptunium Any 20 grams E' v.
gj E% p?
Plutonium Any 50 grams Nu Uranium-235 Any 250 grams
[f (in uranium of any M y!
enrichment)
Americium-241 Any 200 p Ci Americium-241 Scaled source 40 curies Plutonium-Bervilium Scaled source 2 curies Uranium-natural Any 15 kilograms
NEDO-21326-2A4 January 1979 Figure 10-1 and 10-2 provide k, as a function of fuel enrichment er.d reactor type, as well as correction factors for principal variables affec:ing k,:
the pellet diameter, the water-to-fuel ratio, and the cladding materio?. Other fuel configurations that have been analyzed and reviewed separately may be excepted from the limitation of Figures 10-1 and 10-2, as referenced in Specification 10.2.1.1.b.
Specification 10.2.1.1.c. defines the allowable quantity of unirradiated natural and depleted uranium to be received and stored.
Specification 10.2.1.1.d. authorizes possession of.arious isotopes to be used for instrument and calibration sources.
Specification 10.2.1.1.e. provides for storage of tools and equipment inci-dental to the conduct of General Electric nuclear businesses while awaiting decontamination, re-use, or ultimate disposal. Activity will be back-calculated from measurements of the highest exposure rate at 3 ft, from the package, assuming that the radiation originates from a uniform volumetric source having approximately the same dimensions as the package. Unless otherwise determined, N
gamma emissions of 1 MEV/ disintegration will be assumed.
l Specification 10.2.1.1.f. 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, re-use, or ultimate disposal.
These tools and equipment may be contamination with Co-60, Cs-137, or other isotopes as encountered in fuel handling and storage activities.
10-7
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NEDO-21326-2A4 January 1979 10.2.2 Fuel Storage Provisions 10.2.2.1 Specification Irradiated fuel bundles shall be stored in authorized fuel storage baskets, mounted in a support grid, a a fuel storage basin.
Fuel storage locations of this specification shall be exampf. from requirements of Section 70.24al of 10CFR70, regarding neutron detec ars.
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 fuel storage baskets and support grid are those described in Chapter 5, beginning with Section 5.3.5.
I The last sentence of 10.2.2.1 exempts the Morris Operation from the require-ment to have neutron detectors for criticality monitoring. The strength of neutron radiation from the fuel a the surface of the basin water is below detection levels, even in the ur'ikely event of a criticality.
3Natural UO, UO, UNH, and UF6 used during MFRP testing may be stored in process 3
2 vessels in the canyon area, or in the site warehouse.
10-10
NEDO-21326-2A4 January 1979 10.3 LIMITING CONDITIONS The limiting conditions described in this section apply to normal operation of the Morris Operation f acility.
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 - tJater 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 ft.
10.3.1.2 Basis This specification establishes a minimum thickness of watar shielding to limit radiation f rom 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 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.
10-11
NEDO-21326-2A4 January 1979 10.3.2 Limiting Condition - Criticality 10.3.2.1 Specification lE A structure (unloading pit doorway guard; Figure 5-3, NEDO-21326) 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.
10.3.2.2 Basis ll The analysis of a fuel basket drop accident (Chapter 8, NEDO-21326) 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 the unloading basin.
The use of the unloading pit doorway guard is described in Chapters 1 lE 0
and 5; See Section 5.3.4.5.
10-12
NEDO-21326-2A4 January 1979 10.4 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 summarized 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
-8 of 4 x 10 Ci/nl.
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 10.4.1.1 assures that offsite concentration will be within 10CFR20 limits. The sampling and analysis program provides data for estimating the amounts of radio-active material released to the environ nt during routine or accident conditions.
10-13
NED0-21326-2A4 January 1979 Table 10-2 SURVEILLANCE REQUIREMENTS
SUMMARY
Subsection Ouantity or Item Period
- Value**
10.4.1.1 Effluent air W
8: 4 x 10' uCi/ml
-5 10.'4.2.1 Water - Evaporation pond and M
8:
10 uCi/ml lE
-6 sanitary lagoons a: 5 x 10 uCi/ml 10.4.3.1 Sealed sources SA 8 or y: 0.005 uCi lN a: 0.005 uCi 10.4.4.1 Instruments (see Table 10-3) 10.4.5.1 Basin water coolers W
2200 dpm/100cm smearable
-5 10 uCi/ml a or 8 10.4.6.1 Process steam bypass 10.4.7.1 Cask coolant identification 10.4.8.1 Cask coolant radioactivity Not greater than 10 CFR 71.35 (a)(4) 10.4.9.1 Basin water chemical W
pH 4.5 to 9.0 analysis NANO 3 < 200 ppm Cl < 10 ppm 10.4.10.1 Basin water W
0.1 uCi/ml maximum radioactivity key:
- Anaysis frequency lE W: Weekly Q: Quarterly NR: Not Required M: Monthly A: Annual SA: Semiannual
- See text for requirements 10-14
NEDO-21326-2A4 January 1979 Table 10'3
SUMMARY
REQUIREMENTS SYSTEM AND EQUIPMENT TEST AND CALIBRATION Operability System or Equipment Test Calibrate Basin Leak Detection System W
M 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 Key: Operability test / calibration frequency lE W: Weekly Q: Quarterly NR: Not Required M: Monthly A: Annual SA: Semiannual 10-15
NEDO-21326-2A4 January 1979 10.4.2 Effluent Water 10.4.2.1 Specification Water in the sanitary holding basin and the eva oration pond shall be sampled at least once each month and analyzed for gross alpha and beta radiation.
Maximum acceptable concentrations shall not exceed 10' uCi/ml beta and 5 x
-6 10 uCi/ml alpha r..lation.
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 fa sewer effluent. The limits selected are for isotopes that are present at the Morris Operation.
/
s' Dry to the extent that water samples cannot be obtained in the usual manner.
10-16
NED0-21326-2A4 January 1979 10.4.3 Sealed Sources 10.4.3.1 Specification Each licensed sealed source (not irradiated fuel) containing radioactive material in excess of 100 uCi of beta-gamma emitting material or 10 pCi of alpha-emitting material shall be tested for leakage at least once every 6 months, except that each source designed for the purpose of emitting alpha particles shall be tested at intervals not to exceed 3 months. The maximum acceptable level of removable (non-fixed) contamination shall be less than 0.005 uCi total for each source, using dry-wipe testing techniques.
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 10 CFR 70.39 (c) limits for plutonium, but other provisions of this reference are not applicable.
10-17
NFD0-21326-2A4 January 1979 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' recommendatious, and operational tests are performed to check alarm functions and demonstrate other operational features of the system or equipment.
10.4.4.2 Basis Bases for these test and calibration requirements are as follows:
Basin Leak Detection System - Operation of this system ensures that a.
a leak in the basin liner vill be promptly detected, so that correc-tive action can be initiated.
Since the operation of the system is related to the icvel of water in the detection system, the level set point is checked and instruments receive 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 LAW
. alt 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 required.
e.
Area Radiation Monitors - The audible alarm system for these monitors are tested (operated), and the alarm set point calibrated periodically 10-18
NEDO-21326-2A4 January 1979 to provide assurance of reliable operation within equipment specifica-tions, to alert personnel to radiation above preset levels, f.
Criticality Monitors - The audible ala oystems for these monitors, which warn personnel of a criticality, are tested (operated) and the alarm set point calibrated periodically to provide assurance of reli-able operation within equipment specifications.
10-19
NEDO-21326-2A4 January 1979 10.4.5 Basin Coolers 10.4.5.1 Specification Basin water coolers that are in service shall be inspected at least once each month:
a.
The equipment shall be visually inspected for signs of leakage with the fans off.
b.
Random smear surveys for removable contamination shall be 2
no more than 2200 dpm/100 cm,
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. Routinc visual and emear tests are made to detect leakage.
10-20
NED0-21326-2A4 January 1979 10.4.6 Process Steam Bypass 10.4.6.1 Specification Whenever the process steam generator is bypassed and utility steam is sub-stituted for process steam to operate the '7w activity waste evaporator con-densate from the process steam condensate sy: tem returning to the utility boiler shall be sampled at least once each 8 hr of such operation and analyzed for gross beta activity. The highest acceptable concentration so measured shall not exceed 10-5 uCi/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.
10-21
NED0-21326-2A4 January 1979 6
10.4.7 Cask Liquid coolants 10.4.7.1 Specification N
Water shall be the only liquid coolant permitted in all casks received by Morris Operation. Chemical additives to prevent freezing of the water f
are prohibited. Air-cocied casks may be accepted providing that they can be flushed and otherwise handled as a cask using water coolant.
[
l 10.4.7.2 Basis The Morris Operation is not normally equipped to accommodate liquid coolants g
i other than water, t
6 " Coolant" refers to the heat transfer medium used within the cask.
10-22
NEDO-21326-2A4 December 1978 10.4.8 Cask Coolant Sampling l
10.4.8.1 Specification 3,
l 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 chan 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 procedure.
10.4.8.2 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.
!N l
10-23
NED0-23126-2A4 January 1979 10.4.9 Basin L'ater chemical characteristics S
10.4.9.1 Specification l
i Baain water chemistry shall be mair.tained as follows:
Item Acceptable Analysis pH 4.5 to 9.0 i
NANO
< 200 ppm l
Cl-3 10 ppm 10.4.9.2 Basis Basin water chemical characteristics are selected to maintain a benign environment for stored fuel and equipment in the basin water.
IN I
10-24
NEDO-21326-2A4 January 1979 I
10.4.10 Basin Water Radioactivity Sampling N
10.4.10.1 Specification Additional basin water cleanup measures shall be initiated if the con-centration of radioactive materials in the water exceeds 0.02 oCi/mi.
Fut i receiving operations shall be stopped if the concentrat ion exceeds 0.1 ut.1/ml.
The USNRC shall be notified, and immediate measures taken te reduce con-centrations below the 0.1 pCi/ml prior to continuation of fuel receiving operations.
10.4.10.2 Basis Periodic sampling of the basin water is required to assure that radio-activity levels remain as low as reasonably achievable. The values selected 3
are consistent with current decontamination practices.
10-25
NEDo-21326-2A4 January 1979 10.5 DESIGN FEATURES The design f eatures in the following subsection are those incorporated in the lE Morris Operation facility for the sate handling and storage of irradiated fuel.
l 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).
10-26
NED0-21326-2A4 January 1979 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 spr.cing of the grids determines the spacing of fuel that was used in the nuclear safety analysis.
b.
The steuctural 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.
10.5.2.3 Fuel Grapples 10.5.2.3.1 Basis Fuel grapples used with the fuel handling crane and those used with the basin 10-27
NEDO-21326-2A4 January 1979 crane are designed to preclude lifting a fuel bundle closer than 9 ft to the 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 t.han 9 ft to the normal water level of the basin.
10-23
NEDO-21326-2A4 January 1979 10.6 ADMINISTRATIVE CONTROLS 10.6.1 Responsibility The Ma: qer - Morris Operation shall be responsible for overall f acility operation in accordance with these specifications aad applicable govern-ment regulations, and shall delegate in writing the succession of this responsibility during his absence. Licensed material shall be used by, or under the supervision of, individuals designated by the Manager - Morris Operation, or his delegate.
N 10.6.2 Organization 10.6.2.1 The facility staff organization is shown in Figure 9-4.
10.6.2.2 Staff Qualifications Min' mum qualifications for members of the facility staf f shall be the following:
a.
Manager - Morris Operation o
BS degree in engineering, or related physical science; or equivaleat in nuclest industrial experience.
o
, Demonstrated competence in the technologies and control methods applicable co nuclear energy business activities, including radioactive materials handling and radiation and criticality safety considerations.
o Ten years of industrial experience with at least five vears in nuclear facility management.
I b.
Manager - Plant Operations F
o BS degree in engineering or equivalent in nuclear industrial experience.
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NEDO-21326-2A4 January 1979 Demonstrated competence in the cechnologies and control methods o
applicable to nuclear energy business activities, including radioactive materials handling and radiatic and criticality safety considerations.
Eight years of prior manufacturing or engineering experience, o
with at least five years in the nuclear industry, Manager - Plant Engineering ar,ct Maintenance E
c.
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 related to maintenance operations under radioactive contamination conditions, Five years of industrial experience, with at least three of o
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, cnd o
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.
10.6.3 Plans and Procedures Plans and procedures shall be established and implemented to assure ccmpliance with Operation Specifications and applicable governmental regulations.
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NEDO-21326-2A4 January 1979 10.6.3.1 Changes to Plans and Procedures All changes or revisions of established plans or procedures required by Subsection 10.6.3 shall be made in accordance with facility modification control practices as described in Subsection 9.4.3.
lE 10.6.3.2 Plans and Procedures - Minimum Requirement Plans and 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, and safety training programs as appropriate.
lN b.
A radiological emergency plan that defines responsibilities and specifies actions, including channels of communication required to cope with credible emergencies on site (Radiological Dnergency PL2n N
for Morris Operation, NEDE-21894, as revised).
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 meets storage criteria.
Plans requirir; analyses of cask drop accidents involving types of e.
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 components important to site radiological and criticality safety.
h.
Arrangements for providing makeup water to the storage basins under normal and emergency conditions.
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NEDU-21326-2A4 January 1979 10.6.4 Review and Audit 10.6.4.1 Plant 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 E
b.
Manager - Plant Operations c.
Manager - Plant Engineering and Maintenance d.
Manager - Quality Assurance and Safeguards e.
Supervisor - Plant Safety E
f.
Senior Engineer - Licensing and Radiological Safety The Committee shall normally meet on a monthly basis, but at no less than 45-day intervals. The Manager - Morris Operation, shall establish applicable procedures and practices for the conduct of Committee responsibilities.
4.
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 Manager - Spent Fuel Services Operation, and General Manager -
E Nuclear Energy Programs Division. Audits and audit response shall be performed in accordanca with procedures established by General Electric.
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NEDO-21326-2A4 January 1979 10.6.5 Action Required For Specification Nencompliance 10.6.5.1 Safety Limits The following actions shall be taken if a safety limit (Subsection 10.2.1 and 10.2. 2, NEDO-21326) is found to have been exceeded:
E a.
Prompt a2 tion shall be taken to assure timely return of operations to specification compliance, b.
The Plant Safety Committee shall be promptly notified of the non-compliance, c.
Notification of NRC Inspection and Enforcement Regional Offices, Region III, st.411 be made within 24 hr, advising them of events that resulted in a safety limit being exceeded.
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 The following actions ohall be taken if a limiting condition is found to have been exceeded:
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 within 24 hr.
c.
Notification of NRC Inspection and Enforcement Regional Office, Region III, shall be made at the time of inspection to advise them of events resulting in limiting conditions being exceeded.
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NEDO-21326-2A4 January 1979 d.
A review of a noncompliance situation shall be made by the Plant Safety 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 such 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 at the time of inspection, advising them of events that resulted in a surveillance requirement being violated.
10.6.5.4 Design Features Design features shall only be changed in accordance with Subsection 10.6. 3.1, and Subsection 9.4. 3, NEDO-21326. Unauthorized modifications of spec it ied lE 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.
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.a y,
NEDO-21326-2A4 January 1979 10.6.6 Loes, Records, and Reports 10.6.6.1 Logs and Records a.
A shift log shall be maintained to record nonroutine and significant 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 SNM 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 mair ined.
10.6.6.2 Reports Reports shall be prepared and submitted as required by applicable governmental regulations.
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