ML20148P094

Review Draft of Chapter 10 to Oper Specs Containing Additions to Oper Specs Originally Submitted as Proposed Specs in June 1977

ML20148P094
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Site:	07001308, 07001309
Issue date:	10/15/1978
From:	GENERAL ELECTRIC CO.
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References
NEDO-21326, NUDOCS 7811280112
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NEDO-213R6 REVISION A4 - OCTOBER 1978 REVIEW DRAFT CHAPTER 10.

OPERATION SPECIFICATIONS This draf t contains additions to the Operation Specification originally submitted as proposed specifications in June 1977.

Pages containing new material are noted by the phrase " review draft" in the upper right corner of the page. A complete Chapter I

is furnished to facilitate review.

Upon approval of this revision, change pages will be. issued for NEDo-21326.

1 11C15 7,qj1280\\l,,y q

L REVIEW DRAFT s-NEDO-21326-2A4 OCTOBER-1978 10.

OPERATION SPECIFICATIONS

10.1 INTRODUCTION

The specification in this chapter establish conditions governing E

the receipt and storage of. irradiated fuel from light water reactors by Morris Operation. These Operation Specifications define require-ments that protect the health and safety of the public and employees.

1 Operation of the Morris fuel storage facility cannot retult in a sudden, large release of rcdioactivity to the environs, even under those credible meteorological and seismic conditions that have been considered in the design basis of the f acility.

The consequences of accidents have been analyzed and found to have insignificant en-vironmental effects.1 In summary, there are no credible events that' E

could cause a release of radioactivity that would pose a danger. to the public.

10.1.1 Definitions The following definitions apply for the purposes of these Operation Specifications:

Saf ety Limits - Those bounds which if exceeded may aff ect the a.

health and saf ety of the public and employees.

b.

Limiting Conditions - The appropriate functional capabilities or performance levels of equipment and systems for normal operation of the facility.

Surveillance Requirements - Requirements for monitoring, sampling, c.

testing, calibrating, and inspecting equipment and systems to demonstrate that functional capabilities or performance levels are maintained as required for normal operation of the f acility.

See analyses in. Chapters 7'and 8 NEDO-21326-2 D10-1

NEDO-21326-2A1 -

Juns 1977 d.

-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.

Administrative Controls - Provisions relating to organization and e.

management, procedures, record keeping, review and audit, and report-j l

ing necessary to conduct' activities in a manner consistent with operation specifications and applicable government regulations.

1 f.

Fuel Bundle - The unit of nuclear fuel in the form that it is charged

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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 bundles).

g.

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.

l l

l 10-2 t--

REUIES DRAFT f

NEDO-21326-2A4 OCTOBER 1978 f

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 Assurancq__

Activities at Morris Operation shall be conducted in accordance with -

requirements.of Appendix B,10CFR Part 50, as described in Spent Fuel Services Operation Quality Assurance Plan, NEDO-20776, as revistd (see Appendix B.8).

10.1.4 Ger. ral Considf.rarions The general considerations of the following subsections are in ef fect 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 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 f acilitate 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 f acility, part of the fuel reprocessing f acilities, N

i shall be inoperative (the majority of fluorine generation equip-ment has been sold and removed).

I l

D10-1A

a NEDO-21326-2A1 June 1977

(

10.2 SAFETY LIMITS l

Safety limits applicable to Morris Operation are found 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.

l l

I 10.2.1 Authorized Materials 10.2.1.1 Specification Light water reactor nuclear fuel to be re eived and stored at Morris a.

Operation shall meet the following requirements:

(1) Fuel shall contain uranium as uranium dioxide (UO )

y (2) Fuel shall be clad with stainless steel, zirconium or zirconium alloys.

(3) Average exposure of reacter discharge batch (fuel) shall not exceed 44,000 mwd /TeU.

(4) Fuel shall have cooled a minimum of 90 days after reactor shut-down and prior to shipping.

j (5)

Rod lattice k, limits without allowance for burnup shall not exceed:

o 1.40 for 7x7 or 8x8 BWR o

1.37 for 15x15 PWR (<8.55 inches square) o 1.41 for 14x14 PWR (<7.80 inches square)

(6) Fuel parameters shall be within the ranges defined in Figures 10-1 and 10-2.

10-3

4 NEDO-21326-2A1 Juns-1977 e

b.

The combined quantity of unirradiated natural uranium and unirradiated depleted uranium at the Morris Operation facility shall not exceed 50 Te.2 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 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. 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,gf for any array of stored fuel must be less than 0.95 at the'95 confidence level.

A The design bases for criticality analyses were selected from detailed analyti-cal studies which were based on the physica1' 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 effect 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 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.

l J

This limitation does not. include uranium in stored fuel, or uraniu= used in construction of shipping casks such as the GE IF-300.

10-4 a,__

.1 i

Table 10-1 Authorized Materials

. Instrument, Calibration, and Laboratory Sources CllEMICAL AND/OR l'lIYSICAL FORM QtlANTITY MATERIAL 4

i 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 g

"8 i

50 grams

!4 l'l u t on i um Any 2

i

_. w liranium-235 Any 250 grams ym 9e_

4 (in uranium of any h

enrichment)

Americium-241 Any.

200 p Ci Americium-241 Scaled source 40 curies 4

i P1utonium-lieryl1tum Scaled source 2 curles-ifrantum-natural Any 15 kilograms a

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NEDO-21326-2A1' Juna 1977 Specification 10.2.1.1.b defines the allowable quant 1ty of unirradiated natural and depleted uranium to be received and stored.

I Specification 10.~2.1.1.c authorizes possession of various isotopes to be used 3

for instrument and calibration sources.

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10-8

MED0-21326-2A1 Juno 1977 10.2.2 Fuel Storage Provisions (E

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 specification 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.

i 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 1

the unlikely event of a criticality.

2 1

'i 1

i 3Natural UO, UO, UNH, and U76 used during MTRP testing may be stored in process j

3 2

vessels in the canyon area, or in the site warehouse.

)

1 10-9

NEDO-21326-2A1 Juno 1977 10.3 LIMITING CONDr: IONS i

i 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, er 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, 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 f t.

A conservative water shield thickness of.9 f t (2.74 meters) has been chosen to provide an increased margin of safety.

4.

10-10

NEDO-21326-2A1 Juna 1977 10.3.2 Limiting Condition - Criticality 1.

10.3.2.1 Specification A structure (unloading pie doorway guard; Figure 5-3, NEDo-21326-1) shall be used at the doorway between the unloading basin and Storage Basin No. 1 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 The analysis of a fuel basket drop accident (Chapter 8, NED0-21326-2) indicates that a basket dropped or tipped over in Easin 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 en uncontrolled and potentially critical configuration in the bottom of the unloading basin.

The unloading pit doorway guard assures char 3 basket cannot empty its fuel into th'e unloading basin.

k r

4The use of the unloading pit doorway guard is described in NIDO-21326-1, Chapters 1 and 5; see Section 5.3.4.5.

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10.4 SURVEILLANCE RECUIREMENTS

(

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 swanarized 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 Effluent 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 pC1/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 10.4.1.1 assures that offsite concentration will be within l'0CFR20 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.

10-12

REVIEW DRAFT NEDO-21326-2A4 OCTOBER 1978 Table 10-2 SURVEILLANCE. REQUIREMEhTS

SUMMARY

Subsection Quantity or Item Period Value 10.4.1.1 Effluent' air.

W S: 4 x 10-8. Ci/ml

-5 10.4.2.1' Effluent air M

8:10 C1/mi-a: 5 x.10-6 pC1/mi li 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 C1/ml a or 8 10.4.6.1 Process steam bypass 10.4.9.1 Basin water 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 key:

• Analysis of samples to occur W:

Weekley Q: Quarterly-NR:

Not Required M:

Monthly A: Annual SA:

Semiannual

• • See text for requirement D10-13

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NEDO-21326-2A1

'Juns 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 M

NR LAW Vault Intrusion System Clad Vault Leak Detection System Q

NR Area Radiation Monitors Q

Q A

Q Criticality Monitors i

Key: Operability test / calibration to occur:

W: Weekly Q: Quarterly NR: Not Required M: Monthly A: Annual SA: Semiannual 5

1 P

b 10-14

NEDO-21326-2A 1 Juna 1977 10.4.2 Effluent Water i

10.4.2.1 Specification Nater in the sanitary effluent holding basin and the evaporation pond shall be 1

sampled at least once each month and analyzed for gross alpha and beta radia-

'taximum acceptable concencrations shall not e::ceed 10-5 tion.

e Ci/ml beta and

-6 5 x 10 C1/mi alpha radiation.

If either pond is dry,5 no sampling of tha.

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.

i SDry to the extent that water samples cannot be obtained in the usual manner.

10-15 i

)

NEDO-21326-2A1 a

Juna 1977 10.4.3 Sealed Sources 10.4.3.1 Specification Each sealed source (not irradiated fuel) containing radioactive ms.terial 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 sha2 a be 0.005 uC1 (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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10-16

NEDO-21326-2A1 Juna 1977 10.4.4 Instrumentation r

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 Bases for these test and calibration requirements are as follows:

a.

Basin Leak Detection Systes - Operation of this system ensures that a leak in the basin liner vill be promptly detected, so that correc,

t1ve 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 LAW vault flooding.

Since a specific level not involved, calibration is not required.

d.

Clad Vault leak Detection Systen - Operation of this system provides for detection of water between the vault liner and the concrete 4

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 10-17

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Criticality Monitors - The audible' alarm system for these monitors are l

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tested - (' operated), and the alarm set point calibrated periodically to provide assurance of reliable operation 'within equipment' specifications, to warn personnel-of. criticality.

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Juno 1977

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- 10.4.5 Basin Coolers' 10.4.5.1 specification Basin water coolers that are in service shall be inspected at least once each

. veek, including:

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a.

The equipment shall be visually inspected for signs of leakage with

.the fans off.

b.

Random smear surveys for removable contamination.

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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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- June 1977 1

10.4.6 Process Stes: Bypass

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10.4.6.1 Specificatioa Whenever the process stea= generator is bypassed, and utility stea= is.sub-stituted for process stea=, condensate from the process stea= condensate syste=

returning to the utility boiler.shall be sa= pled at least once each 12-hr period and analyzed for gross beta activity. The highest acceptable concentration so measured shall not~ exceed 10-5 uC1/ml.

10.4.6.2 Basis l

The sampling requirement helps assure that if radioactive nacerial is. released l

in the condensate, it would be discovered quickly.

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REVIEW DRAFT r

NEDO-21326-2A4 OCTOBER 1978 1

6

.10.4.7 Cask Liquid coolants N

10.4.7.1 Snacification 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 9 scia 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 Samolinz_

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 failed, and action shall be taken in accordance with established procedure.

10.4.8.2 Basis Undetected f ailed 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.

6 " Coolant" ref ers to the heat transf er medium used within the cask Refer to,Section 5.3.3.1 D10-20A

REVIEW DRAFT NEDO-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 pC1/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.

1 h

D10-20C

REVIEW DEAFT NEDO-23126-2A4 OCTOBER 1978 10.4.9 Basin Water Chemical Characteristics 10.4.9.1 Specification Basin water chemistry shall be maintained as follows:

Item Acceptable Analysis pH 4.5 to 9.0 NANO

< 200 ppm 3

10 ppm CI-10.4.9.2 Basis Basin water chemical characteristics are selected to maintain a 5

benign environment for stored fuel.

D10-20B

NEDo-21326-2A1 Juna 1977 10.5 ' DESIGN FEATURES I

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 analy:ed for the IF-300 cask with the. energy-e absorbing pad in place (Chapter 8, NEDO-21326).

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NEDO-21326-2A1 Juns 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.

The heat transfer properties are considered in fuel cooling thermal c.

analyses and related to nuclear safety.

10.5.2.2 Basket Support Grids 10.5.2.2.1 Basis The spacing of the grids determines the spacing of fuel that was a.

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 Fuel Grapples 10.5.2.3.1 Basis Fuel grapples used with the fuel handling crane ord those used with -he basin 10-22 i

IEDo-21326-2A1 -

a.,-

June 1977 crane are designed to preclude lifting'a fuel bundle closer.than 9 ft to the t

normal water levet 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.

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10-23

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REVIEW DRAFT NEDO-21326-2A4 OCTOBER 1978 10.6 ADMINISTRATIVE CONTROLS 10.6.1 Responsibility 1

The Manager-Morris Operation shall be responsible for overall f acility operation in accordance with these specifications and applicable government regulations, and shall delegate in writing the succession of this respotisibility during his absence.

Licensed material shall 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 f acility staff shall be the f ollowing:

a.

Manager - Morris Operation 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 to 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 f acility management, b.

Plant Operations and Services BS degree in engineering or equivalent in nuclear industrial o

experience.

D10-24

.NEDO-21326-2A1

~'

Juns 1977 o

Demonstrated competence in the technologies and control methods.

. applicable to nuclear energy business activities, including radioactive materials handling and radiation and criticality safety considerations.

Eight years of prior' manufacturing or engineering experience, o

with at least five years in the nuclear industry.

c.

Manager - Plant Engineering and Safety 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.

i

.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, and 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 P1ars and procedures shall be established and implemented to assure compliance v1th Operation Specifications and applicable governmental regulatiens.

10-25

REDO-21326-2A1 a

Juno 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-2132'-2.

10.6.3.2 Plans and Procedures - Minimum Require:er.t Plans and procedures required t y Subsection 'O.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 channels 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 componeuts.

d.

Procedures for determining certain characteristics of fuel to be stored, and to verify that fuel meets storage criteria.

Plans requiring 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.

i A preventative maintenance system for structures, systems, and g.

components important to site radiological and criticality safety.

h.

Arrangements and procedures for providing makeup water to the storage basins under normal and emergency conditions.

10-26

NEDO-21326-2A1

. Juns 1977-10.6.4 Review and Audit 10.6.4.1 Plant' Safety Committee Review and audit of plans and procedures, and of operations carried lout 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:

Manager - Morris Operation a.

b.

Manager - Plant Operation and Services Manager - Plant'Ingineering And Safety c.

d.

Manager - Quality Assurance and Safeguards Supervisor - Plant Safety e.

f.

Senior Engineer - Plant Safety and Licensing I

The Committee shall normally meet on a monthly basis, but at no less than 45-day intervals. The Manager, Morris ~0peration, 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-j 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 Genersi Manager, Nuclear Energy Programs Division. Audits and audit response shall be performed in accordance with procedures established by General Electric.-

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NEDO-21326-2A1 Juns 1977.

10.6.5 Action Required For Specification Noncompliance i

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-21326-2) is found to have been exceeded:

Prompt action shall be taken to assure timely return of operations to a.

specification compliance.

b.

The Plant Safety Committee shall be promptly notified of the non-compliance.

Notification of NRC Inspection and Enforcement Regional Offices, c.

Region III, shall 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 shall 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 KRC Inspection and Enforcement Regional Office, Region III, shall be made quarterly to advise them of events resulting in limiting conditions being exceeded.

10-26

'NEDO-21326-2A1 Juns 1977 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 j

requirements, and - if necessary - to assure return of operations to specification compliance in minimum time.

1 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 f eatures shall only be changed in accordance with Subsection 10.6. 3.1, and Subsection 9.4. 3,. NEDO-21326-2. Unauth'orized modifications of specified design features (per Section 10,5), or introduction of unapproved tools, fix-tures, or other equipment, shall require action as specified for ibniting con-dicions in Specification 10.6.5.2.

10-29

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NEDo-21326-2A1 Juno 1977 10.6.6 Logs, Records, and Reports

.l.

~10.6.6.1 Logs and Recteds A shif t 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 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 maintained.

10.6.6.2 Reports Reports shall be prepared and submitted as required by applicable governmental regulations.

Reports of noncompliance with safety limits (Subsection 10.6.5.1) a.

shall 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).

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