ML20066A378

Proposed Decommissioning Tech Specs Re Reactor Bldg Integrity,Reactor Bldg Ventilation Exhaust Sys,Radiation Monitoring Instrumentation & Pcrv Shielding Water Tritium Concentration

ML20066A378
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/21/1990
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20066A376	List: ML20066A374 ML20066A378
References
P-90367, NUDOCS 9101030222
Download: ML20066A378 (59)

Text

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DECOMMISSIONING TECHNICAL SPECIFICATIONS 1

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FORT ST VRAIN Unit No. 1 Docket No. 50-267 14 Appendix A

.i 4-to Facility License No. DPR-34 s

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DECOMMIS$10NING TECHNICAL SPECIFICATIONS TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 0:TINITIONS 3.0 GENERAL REQUIREMENTS 3.1 Reactor Building Confinement Integrity 3.2 Reactor Building Ventilation Exhaust System 3,3 Radiation Monitoring Instrumentation 3.4 PCRV Shielding Water Tritium Concentration 4.0 DESIGN FEATURES 5.0: ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Decommissioning Safety Review Committee 5.4 Procedures and Programs 5.5 Reporting Requirements 5.6 Record Retention 5.7 Radiation Protection Program 5.8-High Radiation Area 5.9 Process Control Program 5.10 Offsite Dose Calculation and Radiological Environmental Monitoring Program Manuals 5.11 Natural Gas Restriction L

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Fort St. Vrain 075 Page 1-1

1.0 INTRODUCTION

These Decommissioning Technical Specifications are applicable during the decommissioning of the Fort St. Vrain (FSV) reactor.

Decommissioning is considered to begin af ter all of the nuclear fuel has been removed from the FSV Reactor Building and after the NRC has approved the Decommissioning Plan.

The Fort St. Vrain Nuclear Generating Station originally operated as a High Temperature Gas-Cooled Reactor, which supplied steam to a turbine generator.

The facility may be converted to utilize a gas-fired boiler, Although some of the balance of plant systems will be retained for use after the conversion, many plant systems have been taken out of service and are not described in these Decommissioning Technical Specifications.

Activities that will be undertaken in accordance with these Decommissioning Technical Specifications include the dismantlement and decommissioning (DECON) of the radiologically activated and contaminated portions of the facility to release all site areas for unrestricted use.

There are two categories of FSV Technical Specifications:

" Decommissioning Technical Specifications (DTS)" include Amendment _ and all subsequent amendments.

" Operating Technical Specifications" refers to the historical Technical Specifications included in all previous amendments.

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Fort St. Vrain DTS Page 2-1 2.0 DEFINITIONS

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The defined terms in this section appear in capitalized type and are applicable throughout these Technical Specifications.

2.1 ACTIONS ACTIONS shall be that part of a specification which prescribes Required Actions under designated conditions, which shall be completed within specified Completion Times.

2.2 ACTIVATED GRAPHITE BLOCKS ACTIVATED LRApHITE BLOCKS shall include the reflector blocks and spacer blocks.

Other graphite items, such as defueling eier.ents, core support blocks, and core support posts, are not considered ACTIVATED GRAPHITE BLOCKS.

2.3 BASES The BASES shall summarize the reasons for the Limiting Conditions, Applicabilities, ACTIONS, and Surveillance Requirements.

In accordance with 10 CFR 50.36, the BASES are not considered part of the Decommissioning Technical Specifications.

2.4 CHANNEL CALIBRATION A

CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and with the required accuracy to known values of input.

The CHANNEL CALIBRATION shall encompass the entire channel, considering system design, including the sensors and alarm, interlock and/or trip functions, and may be performed by any series of sequer tial, overlapping, or total channel steps such that the entire channel is calibrated.

2.5 CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during its operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

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L' Fort St. Vroin OTS Page 2-2 DEFINITIONS (Continued) 2.6 CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable, considering system design, to verify OPERABILITY including alarm, interlock, and/or trip functions.

2.7 EXCLUSION AREA BOUNDARY The EXCLUSION AREA BOUNDARY shall enclose a rectangular area, 100 meters from the Reactor Building walls, as shown on Figure 4-1.

2.8 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with decommissioning the plant.

Individuals who are occupationally associated with the conversion of the plant, and persons who enter the site to service equipment or make deliveries, are included in this category. MEMBER (S) 0F THE PUBLIC also includes persons who use portions of the site for l

recreational, occupational, or other purposes not L

associated with the plant.

2.9 0FFSITE DOSE. CALCULATION MANUAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring l

Program. The ODCM shall also contain (1) tne Raoicactive Effluent Controls and Radiological Environmental Monitoring-Programs required by Specification 5.4.4 and (2) descriptions o f, the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent-Release Reports required by Specifications 5.5.1 and 5.5.2.

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DEFINITIONS (Continueo). _,,

<:.10 OPERABLE _- OPERABILITY i

A component or sy s t en. Snali ce OPERABLE or have OPERABILITY when it is capable of performing its intended 4

safety function within the required range.

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or system shall be considered OPERABLE when:

(1) it satisfies the Limiting Conditions defined in these Decommissioning Technical Specifications, and (2) it has been satisfactorily tested periodically in accordance I

with the Surveillance Requirements defined in these Decommissioning Technical Specifications.

I 2.11 PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas,

sampling, analyses,

tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as

• 1 assure compliance with 10 CFR Parts 20, 61, and 71, 49 CFR 100, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

2.12 RADIATION SAFETY RADIATION SAFETY shall refer to activities involving the final release of previously contaminated or activated site structures, systems, components, or materit... for unrestricted use; to activities that could rssult in exposures to project personnel or the public in excess of 10 CFR 20 limits; and to activities involving packaging and transportation of radioactive material.

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2.13 UNRESTRICTED AREA i

An UNRESTRICTED AREA shall be any area inside or outside the EXCLUSION AREA BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

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Fort St. Vrain DTS page 3.0-1 3.0 GENERAL REQUIREMENTS.

m.. ___m 3.0.1 Compliance with the Limiting Conditions (LC) contained in Section 3 of the Specifications is required during Fort St.

Vrain Oecommissioning; except that upon discovery of a failure to meet the LC, the associated Required Actions shall be met within the specified Completion Times.

3.0.2 Noncompliance with a specification shall exist when the requirements of the LC and associated Required Actions are not met within the specified Completion Times.

If the LC is restored prior to expiration of the specified Completion Time, the Required Actions need not be completed.

3.0.3 Surveillance Requirements shall be met as specified in the Applicability for '~

• Y 1 LCs unless otherwise stated in an indi"'

illance Requirement.

Failure to meet a o

a.rement, except as provided in 3.0.4, shall constitute failure to meet the LC.

Surveillance Requirements do not have to be performed on inoperable equipment.

3.0.4 Each Surveillance Requirement, any Required Actions whi :.h require the performance of a Surveillance Requirement, and any Required Action with a Completion Time requiring the periodic performance of an action on a "once per..." interval, shall be performed within the specified Frequency with a maximum allowable extension not to exceed 25% of the time interval.

3.0.5 For a surveillance Requirement not performed within

. CTIONS are A

the Frequency defined by 3.0.4, the applicable at the time it is identified that the Surveillance has not been performed.

The Required Actions may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the Surveillance when the Completion Time of the Required Action is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When a Surveillance is performed within t,he 24-hour allowance and the Surveillance Requirements are not

met, the Completion Times of the ACTIONS are applicable at that time.

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i FoM St. Vrain OTS i

Page 3.0 2 3.0 BASES _

m 3.0.1 and 3.0.2 3.0.1 and 3.0.2 establish the general requirements i

applicable to LCs.

These requirements are based on the requirements consistent with operating plants' j

Limiting Conditions for Operation per the Code of Federal Regulations, 10 CFR 50.36 (c)(2). 3.0.1 establishes the Applicability statement within individual specifications as the requirement for when conformance to the LC is required for safe decommissioning of the unit. The Required Actions establish those remedial measures that must be taken within specified Completion Times when requirements of a LC are not met, l

3.0.2 establishes that noncompliance with a specification exists when the requirements of the i

LC are not met and the associated Required Actions have not been met within the specified Completion Times.

The purpose of this general requirement is to clarify that:

(1) completion of the Required Actions within the specified Completion Times constitutes compliance with a speci f i ca tion, and (2) completion of the remedial measures of the Required Actions is not required when compliance with an LC is restored within the Completion Time specified -in the associated

ACTIONS, unless otherwise specified.

3.0.3 - 3.0.5 3.0.3, 3.0.4, and 3.0.5 establish the general requirements applicable to Surveillance Requirements.

-These requirements are based on the requirements consistent with operati ng plants'

- Surveillance Requirements stated in the Code of Federal Regulations,10 CFR 50.36 (c)(3).

3.0.3 establishes the requirements that Surveillance Requirements must be met during the

- conditions specified in the Applicability for which

- the requirements of the LC' apply unless otherwi se stated in an individual Surveillance Requirement.

The purpose of this general requirement is to ensure that Surve111ances are performed to verify the status of systems and components and that parameters are-within specified limits.

Surveillance Requirements do not have to be performed when outside of the Applicability of the LC unless otherwise specified, 1

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Fort St. Vraia DTS Page 3.0-3 3.0 BASES (Continued)_

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3.0.4 establishes the conditions under which the specified Frequency for Surveillance Requirements, Required Actions which require the performance of a specific Surveillance Requirement, and an) Required Action with a Completion Time requiring the periodic performance of an action vn a "once per interval may be extended.

3.0.4 permits an extension of the Frequency to facilitate Surveillance scheduling and consideration of decommissioning conditions that may not be suitable for conducting the Surveillance; e.g., maintenance activities.

The limit of 3.0.4 is based on engineering judgement and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is suf ficient to ensure that the reliability ensured throughout Surveillance activities is not significantly degraded beyond that obtained from the specified Surveillance Frequency.

3.0.5 establishes that the failure to perform a Surveillance within the allowed Surveillance Frequency, defined by the provisions of 3.0.4, is a condition that constitutes a failure to meet the OPERABILITY requirements for an LC.

Under the provisions of this general requirement, systems and components are assumed to be OPERABLE when tb9 associated Surveillance Requirements have not been met.-

However, nothing in this provision is to be construed as implying that systems or components are OPERABLE when they are found or known to be inoperable although still meeting the Surveillance Requi rement frequency.

This general requirement also clarifies that the ACTIONS are applicable when Surveillances have not been completed within the allowed Surveillance Frequency and that the Completion Times of the Required Actions apply from the point in time it is identified that a

Surveillance has not been performed and not at the time that the allowed Surveillance Frequency was exceeded.

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DTS Page 3.0-4 4

3.0 BASES (Continued)

If the Completion Times of the ACTIONS are less tnan 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a 24-hour allowance is provided to permit a

delay in implementing the Required Actions.

This provides adequate time to complete Surveillance Requirements that have not been performed.

If a Surveillance is not completed within the 24-hour allowance, the Completion Times of the ACTIONS are applicable at that time.

For the purpose of making the transition from the operating Technical Specifications to the Decommissioning Technical Specifications, surveillances performed under ths operating Technical Specifications may be utilized to satisfy the applicable surveillance requirements of the Decommissioning Technical Specifications, i

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Fort St. Vrain DTS Page 3.1-1 3.1~

REACTOR BUILDING CONFINEMENT INTEGRITY.

LC 3.1 Reactor Building confinement integrity snoil be maintained with:

a.

The kvactor Building overpressure protection system louvers closed *, and b.

Either:

1.

The outer truck bay closures closed, or 2.

The inner truck bay closures closed.

APPLICABILITY:

Whe.1ever ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV shielding water and remain inside the Reactor Building

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The Reactor Building overpressure protection system louvers may be open provided there are no activities in progress involving the physical handling of any ACTIVATED GRAPHITE BLOCKS.

Fort $t. Vrain DTS Page 3.1-2 SURVEILLANCE REQUIREMENT $_._.=.mm m.,.= m = m m m_. _

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BACKGROUND The integrity of the Reactor Building, in conjunction with operation of the ventilation exhaust system, limits the off-site doses u w normal and abnormal conditions during decommissioning activities.

In the unlikely event of a major release of activity from the Prestressed Concrete Reactor Vessel (PCRV) dismantlement (i.e.,

Heavy Load Drop Accident), the combination of the Reactor Building integrity and ventilation exhaust system would act to keep off-site doses well below 10 CFR 100 guidelines and within a small fraction of EPA guidelines (Reference 2).

The integrity of the Reactor Building confinement is normally maintained with the exterior closures and the overpressure protection system louvers closed.

The truck bay includes two redundant sets of closures.

The outer closures have historically included a truck door and the personnel access door in the truck door.

The inner closures have historically included the truck bay floor hatch, the truck bay overhead sliding hatch, and the internal personnel door.

During decommissioning, there will continue to be two redundant closures which may include the addition of new outer truck doors, external to the original truck doors, in an airlock-type configuration.

The Reactor Building shall be maintained subatmospheric at all times including normal access (see LC 3.2).

Subatmospheric conditions can be mair.tained with several louver banks open.

The overpressure protection system louvers may be opened on a controlled basis for various reasons (e.g.,

to provide extra ventilation cooling during hot weather).

The inner closures of the truck bay are closed to ensure integrity of the Reactor Building confinement prior to the opening of the outer truck doors to the truck bay.

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Fort $t. Vrain DTS Page 3.1-4 i

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3.1 BASES (Continued)

Reactor building confinement integrity is taken i

credit for in the Heavy Load Drop and the loss of AC Power accident analyses, as described in Se. tion 3.4 of the Decommissioning Plan. (Reference 1)

LC The

.ablishes the minimum conditions required to ense a that Reactor Butiding confinement integrity is maintained during applicable ricident scenarios (i.e., Heavy Load Drop and/or loss of AC Power).

The LC requirements are consistent with the accident analysis assumptions, and the criteria used during plant operation.

It should be noted that the Reactor Building overpressure protection system louvers may be open provided there are no activities in progress involving the physical handling of any ACTIVATED GRAPHITE BLOCKS.

For example,.the louvers may be open while ACTIVATED GRAPHITE BLOCKS are being dried or are in temporary storage within the Reactor Building, as long as they are not being moved,

cut, or otherwise physically handled.

APPLICABILITY The Reactor Building confinement integrity I

applicability is based on complying with the off-site dose requirements established in the 10 CFR 100 guidelines and the EPA Protective Action Guidelines in the event of a Heavy Load Drop accident and/or Loss of AC Power.

However, the Reactor Building overpressure protection system louvers may be open provided there are no activities in progress involving the physical l

handling of any ACTIVATED GRAPHITE BLOCKS.

Consistent with the Accident Analyses, ACTIVATED GRAPHITE BLOCKS include reflector blocks and spacer blocks, as defined in Specification 2.2.

The activation level of other graphite materials is l

significantly less than the reflector blocks and

. spacer blocks.

In the event of a load drop accident involving other graphite materials, the resultant doses are low enough that confinement integriij or ventilation are not required.

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DTS Page 3.1-S 2.1 BASES (Continued) _. m _,

ACTIONS M

When Reactor Building confinement integrity is

breached, suspend activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> completion time to suspend physical hanaiing vi the ACTIVATED GRAPHITE BLOCKS is reasonable as long as the Reactot-Building remains subatmospheric (per LC 3.2).

SURVEILLANCE REQUIREMENTS SR 3.1.1 The Reactor Building overpressure protection system J

-louvers are verified in their closed position datiy during activities when they are required to be closed, that is, during, although not necessarily contemporaneously

with, physical handling of ACTIVATED GRAPHITE BLOCKS.

l SR 3.1.2 Prior to opening the outer truck bay closures, the j

inner truck bay closures are verified closed.

While the outer truck-bay closures are open, locks or signs are rosted on the inner truck bay closures to prevent them from being opened. This ensures Reactor Building confinement integrity.

REFERENCES 1.

FSV Decommissioning PI'an l

2.

Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, EPA-520/1-75-001-A, January

1990, U.S.

Environmental Protection Agency I

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Fort St. Vtain DTS Page 3.2-1 3.2 REACTOR BUILDING VENTILATION EXHAUST SYSTEM._ _.

.m LC 3.2 The Reactor Building ventilation exhaust system shall be OPERABLE with:

a.

Reacta Building internal pressure subatmospheric, and b.

At least one of the three ventilation exhaust trains OPERABLE, with each train consisting of one exhaust fan (C-7301, C-7302, or C-73025) and the HEPA filter section of the associated filter assembly (F-7301, F-7302, or F-73025).

APPLICABILITY:

Whenever ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV shielding water-and remain inside the Reactor Building ACTIONS CONDITION l

REQUIRED ACTION COMPLETION TIME l

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Fort St. Vrain DTS Page 3.2-2

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DTS Page 3.2-3 3.2 BASES BACKGROUND The Reactor Building ventilation exhaust filter system is designed to filter the Reactor Building atmosphere prior to release to the vent stack during both normal and most accident concii t t ons during decommissioning, j

The system consists of three trains, one of which 1s normally in continuous operatim.

The design flow rate for each train is 19,000 cfm. Al wwing 10% for degradation, the minimum flow rate is 17,100 cfm.

One train is sufficient to maintain the Reactor Building subatmospheric and thereby minimize unfiltered fission product release from the b':lding. With only one exhaust fan operating, the ventilation system controls will throttle fresh air supply to the air handler in order to reduce the pressure.

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The Reactor Building is maintained in a

subatmospheric condition to ensure that all air ieakage will be inward and to minimize unfiltered fission product release from the building.

The ventilation system was designed to maintain a subatmospheric condition approximately 1/4 inch water gauge negative.

In actual practice, the Reactor Building pressure is normally 0.15 to 0.20 inches water gauge negative, depending on building activities and ventilation system configuration.

There is an alarm at approximately 0.08 inches water gauge negative, and the outside air supply will fully close if the building pressure increases t

to atmospheric.

The Reactor Building ventilation exhaust system is taken credit for in the Heavy Load Drop accident

analysis, as described in Section 3.4 of the Decommissioning Plan (Reference 1).

LCs The LC establishes the minimum conditions required to ensure the Reactor Building ventilation exhaust system is maintained while the potential exists for a drop of an ACTIVATED GRAPHITE BLOCK.

One train is sufficient to maintain the Reactor Building subatmospheric and thereby minimize unfiltered fission product release from the building.

l HEPA filters provide the required particulate l

filtration.

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Fort St. Vrain DTS Page 3.2-4 t

3.2 BASES ( N tinued)

APPLICABILITY The Reactor Building ventilation exhaust system will remain OPERABLE, providing filtration of effluents to the environment, while the potential exists for dropping an ACTIVATED GRAP41TE BLOCK.

ACTIONS M

i When the Reactor Building pressure is atmospheric or greater, suspend activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building.

The one hour completion time to restore the Reactor Building to subatmospheric conditions minimizes the time exposure of the Reactor Building to atmospheric or greater conditions and is a conservative time frame.

The suspension of physical handling activities is acceptabl.e because all analyzed accidents assume something active is happening no passive postulated accidents will result in radiological conditions where the need for ventilation and confinement exists.

B_,d The ability of the Reactor Building ventilation exhaust system to perform its filtering function during a Heavy Load (ACTIVATED GRAPHITE BLOCK) Drop is dependent on at least one exhaust train being OPERABLE.

With all the exhaust trains inoperable, restore at least one ventilation exhaust train to OPERABLE status.

A ventilation train may be operating but not OPERABLE, e.g.,

in the event a required Surveillance is not completed on time.

In this case, a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> completion time is reasonable since the Reactor Building vill still be maintained at subatmospheric conditions.

.C_d -

When Required Action B.1 cannot be completed within the. required. Completion Time, all activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building are suspended.

Twelve hours is reasonable to suspend handling activities.

The suspension of physical handling activities is accept'acie because all analyzed accidents assume something active is happening - no passive postulated accidents will result in radiological conditions where the need for ventilation and confinement exists.

I' Fort St. Vrain l

DTS Page 3.2-5 3.2 BASES (Continued).

SURVEILLANCE RE0VIREMENTS SR 3.2.1 Verification that Reactor Building pressure is subatmospheric ensures that the confinement integrity is i ntact.

The daily surveillance frequency is the same as the operating technical specification requirements.

SR 3.2.2 A pressure drop across the HEPA filter of less than 6 inches of water gauge at 90% of the filter design flow rate will indicate that the filters are not clogged by excessive amounts of foreign matter.

SR 3.2.3 Bypass leakage and penetration for High Efficiency Particulate Air (HEPA) filters are determined by dioctyi phthalate (00P) testing.

The filter penetration and bypass acceptance limits in the surveillances are applicable based on a HEPA filter efficiency of 95%.

The surveillance frequencies specified establish system performance capabilities.

Verification of the HEPA filter functions ensures system performance capabilities. The surveillance frequency is the same as the operating technical specifications.

REFERENCES 1.

FSV Decommissioning Plan 2.

Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, EPA-520/1-75-001-A, January

1990, U.S.

Environmental-Protection Agency =

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Fort St. Vraia DTS 3

Page 3.3-1 1

3,3 RADIATION MONITORING INSTRUMENTATION

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LC 3.3 The area radiation monitoring 'nstrumentatien :h;.nneli

,~.n in Table 3,3-1 shall be OPERABLE with their alarm setroints within the limits specified for the activities in progress,

~

depending on whether Radiation Work Permit (RWP) controis are in effect.

APPLICABILITY:

At all times, until all significantly contaminated or activated items tnat covid exceed alarm setpoints have been removed from the Reactor Building.

ACTIONS l

l 1

l l

CONDITION I

REQUIRED ACTION 1

COMPLET!0N TIME I

L l

l I

l l

I l

l-lA, One or more IA.1 Adjust alarm l

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> j

l radiation monitor l setpoint within l

1 channel alarm 1

11mit l

l setpoint exceeds l l

l value in Table l

OR l

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[A 2 Declare the i

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channel inoperable l l.

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lB. One or more

}B.1 Place a portable l

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> radiation monitor l monitor (with-l l

channels i

alarm) in the area l l

inoperable l

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l SVRVEILLANCE REQUIREMENTS-SR 3.3.1 Perform the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST,-and CHANNEL CALIBRATION surveillances as shown in Table 3,3-2.

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Fort St. Vrain OTS Page 3.3-2

)

TABLE 3.3-1 i

RADIATIO;. MONITORING INSTRUMENTATION j

INSTRUMENT ALARM SETPOINTS DURING ACTIVITIES OURING ACTIVITIES NOT CONTROLLED CONTROLLED j

BY AN RWP BY AN RWP l

j.

a.

Refueling Floor

< 15 mR/hr

< 100 mR/hr*

l b.

Truck Bay

< 15 mR/hr

< 100 mR/hr*

l 1

Monitors may be reset to alarm at a radiation level within a factor of 2 of the expected radiation level.

TABLE 3.3-2 SURVEILLANCE REQUIREMENTS e

CHANNEL CHANNEL FUNCTIONAL CHANNEL INSTRUMENT CHECK TEST CAllBRATION a,

Refueling Daily Monthly 18 months b.

Truck Bay Daily Monthly 18 months l'

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BACKGROUND The radiation monitoring instrumentation required by this specification at all times during decommissioning activities, until all significantly contaminated or activated items that could exceed alarm setpoints have been removed from the Reactor Building, includes two area radiation monitors, one on the refueling floor and one in the truck bay of the Reactor Building.

These monitors serve as accident monitors to detect unplanned radiation levels in the Reactor Building, that should be investigated and appropriately resolved.

Decommiss'ining of Fort St.

Vrain involves the removal of activated and contaminated material which inherently will result in increased radiation levels in the Reactor Building.

These increased radiation levels will normally be anticipated and planned for, with monitoring provided as required.

Individual work activities will be performed under Radiation Work Permits (RWPs), which will include monitoring provisions.

Also, gaseous effluent releases will ba monitored by the ventilation exhaust stack monitors, controlled by the Offsite Dose Calculation Manual (00CM) program.

Liquid releases will also be monitored and controlled, in accordance with the 00CM program.

The monitors required by this specification are not relied upon in any accident analysis, but they are provided to detect abnormal conditions that could indicate unplanned or accidental radiation levels.

LC The LC establishes the minimum conditions required to ensure the radiation levels are measured in the area served by the individual channels and that an alarm is initiated when the radiation level setpoint is exceeded.

Different alarm setpoints are allowable for the radiation monitors, depending on the activities in progress.

While Radiation Work Permit (RWP) controls are in effect, a 100 mR/hr setpoint will detect unplanned radiation levels.

This alarm setpoint may be raised during activities that are expected to exceed this :etpoint, but no greater than a factor of 2 of the expected radiation level.

At all other times, an alarm setpoint of 15 mR/hr is specified. These alarm setpoints will avoid nuisance alarms while still providing for detection of unplanned radiation levels.

l Fort St. Vrain DTS Page 3.3-4 3.3 BASES (Continued) _

APPLICABILITY This LC is applicable at all +4mes.

ACTIONS A.1 or A.2 When one or more radiation monitor channel alarm / trip setpoint exceeds the values i n Table 3.3-1 either adjust the alarm / trip setpoint within its limits or declare the channel inoperable.

The Required Action and Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is consistent and comparable with Standard Technical Specifications.

B.f.1 When one or more radiation monitor channels is

- inoperable, place a portable monitor with an alarm in the area.

The OPERABILITY of the radiation monitoring channels ensures that the radiation levels are measured in the areas served by the individual channels and an alarm is initiated when the radiation level trip setpoint is exceeded. A6 hour Completion Time is reasonable to complete the Required Action.

SURVEILLANCE REQUIREMENTS SR 3.3.1 The surveillance requirements frecuencies specified for CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION conform to industry practice and the surveillance frequencies given in Standard Technical Specifications and are adequate to ensure the proper operation of these detectors.

REFERENCES 1.

FSV Decommissioning Plan 2.

Offsite Dose Calculation Manual Program l

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Fort St. Vrain DTS Page 3.4-1 3.4 PCRV SHIELDING WATER TRITIUM CONCENTRATION LC 3.4 Tritium concentration in PCRV shielding water shall not exceed 62.4 pCi/cc.

APPLICABILITY:

Whenever there is shielding water within the PCRV, until all ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV.

ACTIONS I

I l

l l

CONDITION 1

REQUIRED ACTION l

COMPLETION TtME l

l l

l 1

l l

I l

lA.

PCRV shielding lA.1 Reduce tritium l

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l

]

water tritium l

" centration to i

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concentration l

.4 pCiicc l

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> 62.4 pCi/cc l

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OR l

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lA.2 Perform engineeringl i

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evaluation to 1

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1 tritium content l

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l s 1 E+5 Ci 1

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1 l

l lB.

Required Actions 1B.1 Prepare and submit l The next l

l not met within l

to the NRC a 1

30 days l

l required l

Special Report l

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Completion Time I

describing the l

l l

l safety concerns i

l l

l and the plans for l l

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restoring tritium l l

l l

concentration to l

l l

l within the limit l

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Fort St. Vrain DTS Page 3.4-2 SURVEILLANCE REQUIREMENTS I

SURVEILLANCE l

FREQUENCY l

l I

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I I

l SR 3.4.1 Verify PCRV shteiding water l

Daily, during l

l tritium concentration within i

initial filling i

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of the PCRV with l

l l

shielding water 1

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l SR 3.4.2 Verify PCRV shielding water l Weekly, after the l

l tritium concentration within l

PCRV has been l

l limits l

initially filled l

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Fort St. Vrain DTS Page 3.4-3 3.4 BASES-BAC& GROUND During Decommissioning of Fort St.

Vrain, the Prestressed Concrete Reactor Vessel (PCRV) cavity will be flooded with water to facilitate the removal of the reactor core components.

PCRV dismantlement activities will begin only after all spent fuel has been removed from the reactor building.

The water will be circulated, and purified by the PCRV water circulation system to gradually decrease the radioactivity, except tritium, in the water.

Thus, the flooding of the PCRV will provide shielding for the workers associated with PCRV dismantlement activities, There are a number of systems associated with the flooding of the PCRV to control radioactive material.

Their functions include filtration of the PCRV water inventory, partial demineralization for controlling dissolved solids, and " Feed and Bleed" for adding clean sakeup water and for removing contaminated (primarily tritium) water.

The initial fluctuating increase in the tritium concentration during the flooding of the PCRV will be controlled by the " Feed and Bleed" dilution process.

In accordance with the 00CM, released tritiated water will.normally be treated as normal liquid

.radwaste, diluted and released at a controlled rate.

A maximum PCRV shielding water tritium concentration is assumed in the Los; of PCRV Shielding Water accident analysis. as described in Section 3.4 of th_e Decommissioning Plan (Reference 1).

For this_ analysis, it is conservatively assumed that the theoretical maximum amount of tritium is transferred to the PCRV shielding water from the graphite -blocks, which is approximately 1 E+5 Curies.

The tritium concentration in the spilled water is calculated to be 62.4 pCi/cc.

Fort St. Vrain DTS Page_3.4-4 3.4 BASES (Continued)

LC The LC establishes the maximum concentration tolerable in the PCRV shielding water to ensure adequate protection to the MEMBERS OF THE PUBLIC.

The LC requirements are consistent with the accident analysis assumptions.

It should be noted that the accident analysis assumed 1 E+5 Curtes released.

The resulting tritium concentration of 62.4 pCi/cc was chosen as the LC requirement because it is easier to determine a tritium concentration for surveillance monitoring purposes.

APPLICABILITY This LC.is applicable whenever there is shielding water within the PCRV, until all ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV, af ter which there is no credible source of additional tritium.

ACTIONS A.1 or A.2 When the PCRV shielding water tritium concei,tration is greater than 62.4 pC1/cc it is prudent to either reduce the concentration to less than or equal to 62.4 pCi/cc or perform an engineering evaluation to veri fy that the total tritium content is less than or equal to 1 E+5 Curies. A completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable amount of time to change the concentration of large water volumes and to perform associated analyses.

B,d When a Required Action cannot be completed within the required Completion Time, a Special Report must be prepared and submitted to the NRC describing the safety concerns and the plans for restoring tritium concentration to within its safety analysis limit.

The preparation and submittal of a Special Report is an acceptable action because the 1 E+5 Curie analysis value results in doses far below the i

limits allowed by Reference 2.

The Special Report will te prepared as described in Specification 5.5.4.

Fort St. Vrain DTS Page 3.4-5 3.4 BASES (Continued)

SURVEILLANCE REQUIREMENTS SR 3.4.1 and 3.4.2 Verification of PCRV shielding water tritium concentration limits ensures adequate protection to the MEMBERS OF THE PUBLIC.

The daily surveillance frequency during the filling of the PCRV with the shielding water will detect any fluctuations in the tritium concentration during non-steady state conditions.

Once the PCRV has been filled with the shielding water, the 7 day surveillance frequency will ensure that a fluctuation in the tritium concentration during subsequent material handling activities will be detected.

REFERENCES 1.

FSV' Decommissioning Plan 2.

Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, EPA-520/1-75-001-A, January

1990, U.S.

Environmental Protection Agency

Fort St. Vrain DTS Page 4.0-1 4.0 DESIGN FEATURES 4.1 Site The Fort St. Vrain Nuclear Generating Station is located approximately 35 miles north of Denver and 3.5 miles northwest of the town of Platteville, in Weld County,

. Colorado.

The site consists of 2798 acres.

The EXCLUSION AREA BOUNDARY encloses a rectangular area, 100 meters from the FSV Reactor Building walls, as shown on Figure 4-1.

Points where radioactive gaseous and liquid ef fluents are released are shown on Figure 4-1.

4.2 Reactor Building The Reactor Building houses the prestressed concrete reactor vessel (PCRV), fuel handling area, fuel storage

wells, fuel shipment preparation facilities, decontamination and radioactive liquid and gas waste processing equipment, and most reactor plant process and service systems.

Decommissioning will not involve any major modifications to the Reactor Building structural steel without verification of the seismic' qualification, as described in Section 2.2.1 of the Decommissioning Plan.

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Fort St. Vrain DTS Page 5.0-1 5.0 ADMINISTRATIVE CONTROLS i

5.1 Responsibility The Program Manager for Decommissioning shall have overall onsite responsibility for all Fort St.

Vrain decommissioning activities, for both PSC and contractor personnel, The Program Manager shall delegate in writing the succession to this responsibility during absences.

The Vice President, Nuclear Operations shall have overtil executive responsibility for all Fort St. Vrain decommissioning activities.

5.2 Organization The decommissioning organization, functional requirements, and qualification requirements for key decommissioning personnel, for both PSC and contractor groups, shall be documented in the FSV Decommissioning Plan.

The organization responsible for quality assurance shall report to the Vice President, Nuclear Operations, on quality assurance matters, to ensure independence.

. An individual qualified in radiation protection procedures shall be present at the facility at all times during physical decommissioning activities.

5.3 Decommissioning Safety Review Committee (OSRC) 5.3.2 The DSRC shall be comprised of the following.

Program Manager for Decommissioning (Cheirman)

Facility Support Manager (Radiation Protection Manager)

Decommissioning Engineerir.g Manager Operations Manager Project Assurance Manager Westinghouse Project Director Consultants may be appointed as members, in writing, by the OSRC Chairman An alternate Chairman and alternate members, if required, shall be appointed in writing by the OSRC Chairman.

rort St. Vrain DTS Page 5.0-2 ADMINISTRATIVE CONTROLS (Continued) 5.3.3 The DSRC shall meet at least once per calendar quarter, or more frequently as convened by the DSRC Chairman or the Vice President, Nuclear Operations.

5.3.4 A quorum of the 05RC shall consist of the Chairman or alternate Chairman, and three members including alternates.

No more than two alternate members shall participate as voting members in DSRC activities at any one time.

5.3.5 The DSRC shall be responsible fer review of:

a, Administrative procedures,

plans, manuals, and programs required by Specifications 5.4.1 through 5.4.4, 5.7, and permanent changes thereto, that affect RADIATION SAFETY.

b.

Proposed tests and experiments that affect RADIATION SAFETY.

c.

The-following items, that have been evaluated to involve an unreviewed safety question as defined in 10 CFR 50.59:

1) Administrative procedures, plans, manuals, and programs required by Specifications 5.4.1 through 5.4.4, 5. 7., and permanent changes L

thereto, l

2)~ Proposed changes or modifications to plant systems or equipment, and

3) Proposed tests and experiments, d.

Proposed changes to the Decommissioning Technical Specifications or Facility License.

e.

Investigations of violations of Decommissioning j

Technical -Specifications, and of regulations or L

license requirements.

L f.

Reportable events as defined by 10 CFR 50.73, l

l g.

Unplanned release of radioactive material to the environs.

i 5.3.6 The DSRC shall':

a.

Advise the Program Manager for Decommissioning on matters that affect RADIATION SAFETY.

l

Fort St. Vrain DTS Page 5.0-3 ADMINISTRATIVE CONTROLS (Continued) b.

Recommend to the Program Manager for Decommissioning in fwriting, approval or disapproval of items considered under Specifications 5.3.5.a through 5.3.5 d above.

c.

Render determinations in writing with regard to whether or not each item considered under Specification 5.3.5.c constitutes an unreviewed safety question.

d.

Recommend to the Progran Manager for Decommissioning other areas of facility activities where additional aversight is prudent and/or where independent auditing is needed.

5.3.7 Audits of decommissioning activities shall be performed under the cognizance of the OSRC.

These audits shall encompass:

i a.

A-decommissioning program audit to be performed at least - once per two

years, encompassing the following:

1) Decommissioning Technical Specifications

2) Radiation Protection Program

3) Training Program

4) -Decommissioning QA Plan

5) Decommissioning Access Control Plan l

6) Decommissioning Fire Protection Plan l

7) Decommissioning Emergency Response Plan b.

Any other area of faci!s ty activities considered appropriate by the OSRC.

5.3.8 Records of DSRC activities shall be prepared, approved, and distributed as indicated below:

a.

Minutes of each DSRC meeting and documentation of the reviews performed per Specification 5.3.5 above shall be approved and forwarded to the Vice President, Nuclear Operations within 30 days following the meeting.

Fort St. Vrain DTS Page 5.0-4 ADMINISTRATIVE CONTROLS (Continued) b.

Audit reports encompassed by Specification 5.3.7 above shall be forwarded to the Vice President, Nuclear Operations within 30 days after compietion of the audit.

5.4 Procedures and Programs 5.4.1 Written administrative procedures, plans, manuals, and/or programs shall be established, implemented, and maintained covering the activities referenced below:

a.

Radiation Protection Program b.

Surveillance test activities of equipment required by these Decommissioning Technical Specifications c.

Decommissioning Access Control Plan d.

Decommissioning Emergency Response Plan e.

PROCESS CONTROL PROGRAM f.

OFFSITE DOSE CALCULATION MANUAL g.

Decommissioning Fire Protection Plan 5,4.2 Administrative procedures,

plans, manuals, and/or programs of Specification 5.4.1 above, cnd permanent changes thereto, shall be reviewed by the DSRC, or a subcommittee thereof, and approved by the appropriate management prior to implementation.

Procedures shall be reviewed periodically as set forth in Administrative Procedures.

Changes to the 0FFSITE DOSE CALCULATION MANUAL shall be processed in accordance with Specification 5.10, and changes to the PROCESS CONTROL PROGRAM shall be processed in accordance with Specification 5.9.

5.4.3 Temporary changes to administrative procedures, plans, manuals, and/or programs of Specification 5.4.1 above may be made provided the change is documented and approved by the appropriate management prior to implementation.

o Fort St. Vrain OTS Page 5.0 5 ADMINISTRATIVE CONTROLS (Continued) 5.4.4 The following programs shall be e stabl i shed, implemented and maintained:

a.

Radioactive Ef fluent Controls Program A program shall be provided conforming with 10 CFi<

50.36a for the control of radioactive effluents and for maintr.ining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable.

The program (1) shall be contained in the OFFSITE DOSE CALCULATION MAN!!AL, (2) shall be implemented by procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements:

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the OFFSITE DOSE CALCULATION MANUAL,

2) Limitations on the concentrations of radioactive material released in liquid ef fluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,

3) Monitorin'g,

sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20,106 and with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL,

4) Limitations on the annual and quarterly doses or dose commitment to a-MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,

5) Determination of cumulative and projected dose contributions from radioactive ef fluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL at least every 31 days,

i

=

Fort St. Vrain DTS Page 5.0-6 ADMINISTRATIVE CONTROLS (Continued)

6) Limitations on the operability and use of the liquid and gaseous ef fluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting frorr radioactive material released in gaseous of fluents to areas beyond the EXCLUSION AREA BOUNDARY conforming to the doses associated with 10 CFR part 20 Appendix B, Table II, Column 1, 4

8) Limitations on the a n.nua l and quarterly air doses resulting fec.T.

noble gases released in gaseous ef fluents to areas beyond ti.

EXCLUSION AREA BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on-the annual and quarterly doses to a MEMBER OF THE PUBLIC 'from tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous ef fluents released to areas beyond the EXCLUSION AREA BOUNDARY conforming to Appendix I to 10 CFR' Part 50,

10) Limitations on the annual

~ dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle. sources conforming to 40 CFR Part 190.

l

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Fort St. Vrain DTS Page 5.0-7 ADMINISTRATIVE CONTROLS (Continued) b.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.

The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and i

modeling of environmental exposure pathways. The program shall (1) be contained in the OFFSITE DOSE CALCULATION MANUAL, (2) conform to the guidance of Appendix 1 to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION

MANUAL,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the EXCLUSION AREA BOUNDARY are identified and that modifications

'to the monitoring program are made if required by the results of this' census, and

3) Participation in a Interlaboratory Corparison Program to ensure that independent checks on the l

precision and accuracy of the measurements of radioactive materials in environmental sample matrices.~are performed as part of the quality assurance program for environmental monitoring.

1 L

l' l:

Fort St. Vrain DTS Page 5.0-8 ADMINISTRATIVECONTROLS(ContinuedL__

5.5 Reporting Requirements In-addition to the applicable reporting requirements of 10 CFR, the following reports shall be submitted to the Regional Administrator of the NRC's Region IV office unless otherwile noted:

5.5.1 Annual Radiological Reports Annual reports covering the activities described below, for the previous calendar year shall be submitted as follows:

a.

Annual Radiation Exposure Report The Annual Radiation Exposure Report for the previous calendar year shall be submitted to the Commission within the first calendar quarter of each calendar.y' ear in compliance with 10 CFR 20.407 and in accordance with the guidance contained in Regulatory Guide 1,16.

b.

Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the activities of cne unit during the previous calendar year shall be submitted before May 1 of each year.

The report shall include summaries, interpretations. -and analysis of trends of the_results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent-with the objectives outlined in (1) the OFFSITE DOSE CALCULATION MANUAL and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

l l

l 4

Fort St. Vrain DTS Page 5,0-9 ADMINISTRATIVE CONTROLS (Continued)_____

5.5.2 Semiannual Radioactive Effluent Release Report The Semiannual Radioactive Effluent Release Report covering activities during the previous 6 months shall be submitted within 60 days af ter January 1 and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The report shall also include a copy of the OFFSITE DOSE CALCULATION MANUAL, if any changes were made during the report period, as required by Specification 5.10.

The material provided shall be (1) consistent with the objectives outlined in the OFFSITE 00SE CALCULATION MANUAL and PROCESS CONTROL PROGRAM, and (2) in conformance with 10 CFR 50.36a and Section IV.B,1 of Appendix I to 10 CFR Part 50.

5.5.3 Nontoutine Reports a.

The NRC Operations Center shall be notified of emergency and nonemergency events in accordance with 10 CFR 50.72.

b.

Reportable events shall be reported in accordance with 10 CFR 50.73.

5.S.4 Special Reports Special Reports required by Specification 3.4 shall be submitted to the NRC Regional Administrator within the time period specified.

5.6 Re.ord Retention c

5.6.1 The following records shall be retained for at least three years:

a.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to RADIATION SAFETY.

b.

Licensee Event Reports (LERs).

c.

Records of surveillance activities, inspections, and calibrations required by the Decommissioning Technical Specifications.

Fort St. Vrain DTS Page 5.0-10 ADMINISTRATIVE CONTROLS (Continued)-

d.

Records of changes made to procedures related to RADIATION SAFETY.

e.

Records of radioactive shipments, f,

Records of sealed source leak tests and results.

5.6.2 The following records shall be retained for the duration of the Facility License:

a.

01smantlement records for systems and equipment related to RADIATION SAFETY.

b.

Records of facility radiation and contamination surveys, including final site release records, c.

Records of radiation exposure for all individuals entering radiation control areas, d.

Records of gaseous and liquid radioactive material released to the environs, e.

Records of training and qualification for current members of the decommissioning staff, f.

Records of activities required by the Decommissioning QA Plan.

g.

Records of reviews performed pursuant to 10 CFR 50.59, h.

Records of meetings of the OSRC.

i.

Records and logs pertaining to the Radiological Environmental Monitoring Program.

J.

Records of changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

5.7 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be

approved, maintained, and adhered to for all activities involving personnel radiation exposure.

"A Fort St. Vrain DTS Page 5.0-11

' ADMINISTRATIVE CONTROLS;(Continued) t 5~.8 - High' Radia'ti'on Area L5.8.1 Pursuant _=to _ paragraph 20.203 (c)(5) of 10 CFR 20, in lieu of the " control device" or " alarm signal" required by -paragraph 20.203(c),- each high' radiation area, as defined in 10 CFR Part 20, in which the ! intensity of radiation is greater than-100 mR/h but equal to or less than 1000 mR/h at 45 cm -(18 in.).from the radiation source or_

from any_ surface _which the radiation penetrates shall:be barricaded and conspicuously posted as: a' high radiation area and' entrance thereto shall be controlled by requiring issuance of a Radiation Work

~

. Permit

_( RWP). -

Individuals qualified in radiation protectionprocedures(e.g., Health Physics personnel) or: personnel _ continuously escorted by such individuals may be exempt.from the RWP issuance requirement -during L

the performance of their assigned duties in high h

radiation areas with exposure rates equal to or less:

-th'an; 1000- mR/h, provided they are otherwise following plant radiation protection-procedures' for entry into

'i L_.

Osuch _high radiation areas. Any individual or group of individuals permitted to enter =such areas shall be provided -with or-accompanied by. one or more of the following:

a.

A radiation monitoring devic'e; which continuously indicates the radiation dose rate in the area, or-u t

L b..

A -radiation monitoring device which continuously integrates the radiation dose rate'in'the area-and

-alarms when na = preset' integrated dosetis received.

L

-Entry into such areas with: this monitoring device L

.may; be.-made after the dose rate level in the area

'has.been: established ~and personnel. have been made knowledgeable of them, or L

c.- A _ health physics

-qualified individual' (i.e.,-

K qualified:in radiation protection procedures) with a-g radiation dose rate monitoring ' device who _is_

h responsible.for-providing positive. control over the.

activities.

within the area -and shall perform periodic radiation surveillance' at the frequency specified by the f acility-Health Physics staff in

.the1RWP.

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a Fort Sti Vrain DTS Page 5,0-12 ADMINISTRATIVE-CONTROLS (Continued) 5.8.2 In' addition to the requirements of 5.8.1.-

areas accessible te m m el with radiation levels greater

.than-- 1000 mR/h at 45 cm (18 in) from the radiation source or from any surface which the radiation penetrates shall be.provided with locked enclosures to prevent' unauthorized entry, and.the keys.shall be maintained under the administrative control of health physics supervision.

Enclosures shall remain locked except during periods of access by. personnel under an approved RWP which shall specify the dose rate levels in the' immediate work area and the maximum allowable stay time for individuals in the area.

In lieu of the stay

-time specification of the RWP, direct or remote (such as

'use.of -

closed.

circuit TV cameras) continuous surveillance -may be made by -personnel qualified in.

radiation-- protection procedures to provide -positive exposure control over the activities within the area.

For-individual areas accessible to -personnel with radiation _ levels of greater than_ 1000 mR/h that are-

-located within large areas,~where no enclosure exists for purposes _of locking,_ and no enclosure can~ be reasonably constructed around the individual areas, then

-that--area shall be roped off, conspicuously posted,- and a flashing light shall be activated as a warning device.

=5!9 PROCESS CONTROL PROGRAM (pCP)

Permanent changes to the-PROCESS CONTROL PROGRAM':

a

a..Shall befdocumented'and records of reviews performed shall.

be retained.as-;part. of the 0SRC meeting 're c.o rd s, as-required by Specification 5.6;2. LThis documentation shall contain:

1) Sufficient 'information.to support the change together

~

i with the. appropriate analyses or evaluations justifying thechange(s),and

-2) A determination-that. the change will maintain the-overall conformance of-the solidified waste product to existing: requirements of Federal',

State,t or other applicable-regulations.

b.

Shall become effective ~after review and acceptance by the OSRC in accordance with-Specification 5.3.6.

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i Fort St. Vrain 3

OTS Page 5.0-13 i

i ADMINISTRATIVE CONTROLS (Continued) 5.10 0FFSITE DOSE CALCULATION MANUAL Changes to the OFFSITE DOSE CALCULATION MANUAL:

a.

Sha'l be documented and records of reviews performed shall

.)

be retained as part of the OSRC meeting records, as required by Specification 5.6.2.

This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 10 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of

effluent, dose, or setpoint calcula tions.'

b.

Shall become effective after review and acceptance by DSRC in accordance with Specification 5.3.6.

c.

Shall be submitted to the Commission in the form of a

complete, legible copy of the entire OFFSITE DOSE CALCULATION MANUAL as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the OFFSITE 00SE CALCULATION MANUAL was made.

5.11 Natural Gas Restriction As indicated in Specification 1.0, FSV is being converted to utilize a gas-fired boiler.

The natural gas line supplying this boiler shall not be charged within the EXCLUSION AREA-BOUNDARY, during any handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building.

ATTACHMENT 4 TO P-90367 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS

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DECOMMISSIONIND OF THE FORT ST, VRAIN NUCLEAR GENERATING STATION NO SIGNIFICANT HAZARDS CpNSIDERATION ANALYSIS INTRODUCTION Pursuant to 10 CFR 50.92, each application for amendment to an operating license must be reviewed to determine if the proposed change involves a

significant hazards consideration.

The Commission has provided standards for determining whether a

significant hazards consideration exists (10CFR50.92(c)].

A proposed amendment to an operating. license for a facility involves no significant hazards consideration if the change to the facility in accordance with the-proposed amendment would not:

1) involve-a significant increase in the probability or consequences of an accident previously evaluated, or 2) create-the possibility of a new or_ different kind of accident from any accident previously.evaluhted, or 3) involve a significant_ reduction in a margin of safety.

.The - amendment, as defined below,_ describing the replacina of the existing Technical-Specifications, in. their entirety, witn the Decommissioning Technical Specifications has been reviewed and deemed not to-involve a_significant hazards consideration.-

The basis for this determination:follows.

BACKGROUND Public Service. Company of Colorado (PSC) is _ proposing to decommission - the Fort St, Vrain Nuclear Generating. Station.

2 Pursuant to 10 CFR 50,82 PSC has prepared and submitted a Proposed.

Decommissioning Plan (PDP) to the NRC for review and approval. The decommissioning of nuclear _ facilities is a regulated process whereby radioactive material is. removed from the plant site, the site --is decontaminated to. established limits, and NRC licenses are terminated.

LThe-current requirements with regard to occupational or public doses or effluents to the environment continue to apply throughout the decommissioning. period until the' license is terminated by the Commission.

The _ decommissioning planning requirements are considered appropriate means of-assuring that the decommissioning will be carried out - in accordance with 10CFR Part 20,

-and specifically that the doses will be kept as low as reasonably achievable (ALARA).

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4-DECOMMISSIONING OF THE FORT ST. VRAIN

. NUCLEAR GENERATING STATION-PSC has selected the DECON option for decommissioning Fort St.

Vrain.

PSC is proposing the immediate dismantlement and decommissioning.of Fort St. Vrain to release all site areas for unrestricted use.

To accomplish this, the following activities will be undertaken:

1.

Removal of the Prestressed Concrete Reactor Vessel (PCRV) internal. radioactive components remaining after the defueling of-the reactor.

-2.

Decontaminate and/or dismantle those portions of the PCRV structure and radioactive balance-of-plant systems which exceed limits for unrestricted release of residual radioactive materials.

3.

Ship all radioactive waste offsite for disposal.

4.

Perform a final site radiation survey _to confirm that all site areas can be released for unrestricted use.

5.

Terminate the 10CFR50 operating license.

The major dismantling and decontamination activities that will be performed during decommissioning are described in detail in Section 2.3 of the PDP and are summarized below:

Decontamination ~ and Dismantlement of the Prectressed Concret_e Reactor Vessel-The major decommissioning task is the dismantlement and decontamination of the radioactive portions of the PCRV.

Initial dismantlement of the PCRV'will include removal of selected PCRV internal components and removal of portions of the steam generators.

Simultaneously, the non-contaminated portion of the steam generators will be removed from the lower portion of the _PCRV to provide access for detachment of the contaminated steam generator upper assemblies.

After the steam generator. lower assemblies are removed from the bottom of the PCRV, the PCRV bottom head and side wall penetrations will be sealed, a water cleanup and clarification system will' be connected, and the PCRV will be flooded.

Flooding of the PCRV will provide shielding for the workers associated with PCRV-dismantlement activities.

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DECOMMISSIONING OF THE FORT ST0 VRAIN NUCLEAR GENERATING STATION A plug of concrete will be removed from the top head of the PCRV to gain access to the PCRV cavity for removing the core components, including the defueling elements, hexagonal reflector blocks, large side reflector blocks, side spacer blocks, core support blocks, and core support posts.

After the core internals have been removed, the core barrel will be removed by cutting it into pieces sized for disposal in radwaste containers.

Following removal of the core barrel, the PCRV water level will be lowered and the core support floor (CSF) insulation will be removed.

The CSF will then be segmented into pieces and removed using the Reactor Building crane.

Once the CSF is removed, the helium circulator diffusers and steam generator upper modules will be removed.

The remaining radioactive components, which include the activated " beltline concrete" of the PCRV around the reactor core region, the PCRV liner, liner insulation and insulation cover plates, and the PCRV lower floor with its supports will also be removed.

Decontamination and Dismantlement of Contaminated Balance of Plant (BOP) Systems The decontamination and dismantlement of contaminated or potentially contaminated balance of plant systems will be performed by either decontamination in place, removal and decontamination, or removal and disposal as radioactive waste.

In

general, contaminated or potentially contaminated

piping, components, structures, wall s and ductwork will be surveyed to determine acceptability for unrestricted release or to determine the cleanup required for release.

EVALUATIONS Accidents that could result in radiation exposure at the site boundary during the dismantling and disposal activities have been postulated and analyzed.

However, since the reactor will be defueled prior to the commencement of decommissioning operations and all fuel will be removed from the Reactor Building, the risk of accidents resulting in a

radiological release during decommissioning activities is considerably less than during plant operation.

Furthermore, the accident analyses have determined that the radiation doses to the general public from postulated accidents would be a small fraction of the Protective Action Guide levels recommended by the U.S.

Environmental Protection Agency (Reference 1).

Page 3 of 10

_._.___._____.-______.---.-..---_.m

.}

1 DECOMMISSIONING OF THE FORT STo VRAIN

NUCLEAR-GENERATING STATION

-Section 3.4 of-the Proposed Decommissioning Plan describes the analyses for the following postulated accidents:

-Dropping of contaminated concrete rubble.

Conversion construction near-PCRV-dismantlement.

Heavy load drop.

Fire

- Loss-of PCRV shielding water.

Loss of power.

Natural disasters.

The components with the highest potential radiation release were considered in the accident analyses. Therefore, accidents that were analyzed bound.the radiological consequences from other postulated accident -scenarios.

In evaluating the postulated accidents,

. conservative assumptions were made when -data orknowledge to

- support-more-realistic analyses were lacking.

Conservatism in this context means -that _the radiological-- consequences from the postulated accidents will be overestimated rather than underestimated.

A capsule summary of, the accident scenarios is given in Table 3.4-1 of the PDP.

A distance of 100 meters was used as the' exclusion area boundary __(EAB), the distance from the Reactor Building walla

~

to the-nearest fenced - site boundary,.

A worst case atmospheric dispersion-factorL of '3.53 E-2 sec/m has been calculated and-is used-in the accidentc analyses, with the--exception of the-Tornado accident,3 which utilizes.: an = atmospheric dispersion factor of 4.53 E-4'sec/m.-

The atmospheric dispersion factor of 4.59 E-4.sec/m represents an-annual._ average. dispersion factor for Fort St. Vrain, andEis believed to be conservative'in the event of a-tornado.

L These atmospheric Edispersion factors were calculated using_ the -

. guidelines -presented in _ Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident _Conseque. ace Assessments at Nuclear _ Power Plants, and' conservatively assumed a wind speed of 1 mph (1 mph for-the worst: case-and 5_ mph for annual' average).

The analyses also~ assumed that all releases =-to the environment'were ground: level releases.

The major exposure-pathway.was assumed to

-be air inhalation, with the lung representing the critical organ in--

most cases.

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DECOMMISSIONING OF THE FORT ST6 VRAIN NUCLEAR GENERATING STATION 1

The doses to an offsite individual from the postulated accident scenarios are prosented in Table 3.4-2 of the PDP.

An inspection of this table reveals that the limiting accident is a fire with a whole body dose of 121 millirem and a 215 millirem dose to the organ (lung).

The results of the accident analyses indicate that the radiation exposures to the general public will be very low.

In all

cases, the radiological consequences from the postulated decommissioning accident scenarios are well within the 25 rem whole body dose and-300 rem to any specific organ guidelines established by 10CFR100.

The radiological consequences are also a small fraction of the one rem whole body dose and five rem to any specific organ guidelines cited in the EPA Protective Action Guidelines.-

For comparison with previously analyzed accidents, Section 14.11 of the Fort St. Vrain FSAR (Reference 2) provides a discussion of Design Basis Accident Number 2 (Rapid Depressurization/ Blowdown).

As stated in the FSAR, this accident postulates a hypothetical sudden failure of both closures in a PCRV penetration resulting in the release of the fission product inventory circulating within the primary-coolant and a fraction of the plateout activity. For this accident sconario it was conservatively assumed that the coolant escapes directly from the building into the atmosphere at ground level without any credit for holdup or filtration by the ventilation system.

The resultant doses at the EAD (590 meter boundary) from this analysis are 2.5 rem whole body gamma,17.4 rem thyroid (the highest inhalation organ dose) and 4.8 rom bone.

The analysis determined that the radiological consequences of the rapid depressurization would be within the guidelines of 25 rem whole body 'and 300 rem to any specific organ prescribed by 10CFR100. As presented in the PDP, the worst case decommissioning accident is-a postulated fire occurring at ground level iemediately outside of the Reactor Building truck loading bay resulting in EAD (100 meter boundary) doses of 121 millirem to the whole body and 215-millirem to the lung ~(the highest inhalation organ dose).

No credit was taken for holdup or filtration. by the ventilation system..-

The radiological consequences from this postulated

-accident are significantly less than those from the Design Basic Accident -Number 2.

These accident scenarios are judged -to be comparable and the same type of accidents since both involve a

. sudden release of radioactivity into the atmosphere. Due to the low consequences (a small' fraction of' EPA Protective Action G'lidelines)-

of postulated decommissioning accidents, it can be concluded that the Fort St.-Vrain decommissioning activities do not pose an undue risk to the health and safety of the general public.

Page 3 of 10

DECOMMISSIONING OF THE FORT STo VRAIN NUCLEAR GENERATING STATION CONCLUSIONS Since the reactor will be defueled prior to the commencement of decommissioning operations and all irradiated fuel will be removed from the Reactor

Building, there will be no need for shutdown /cooldown systems such as decay heat removal nor is there need for safety systems portaining to reactivity control.

All plant systems whin will be relied upon during decommissioning are described in the PDP and are governed by the Decommissioning Technical Specifications.

Based on the information presented above, the following conclusions can be reached with respect to 10 CFR 50.92.

1.

Superseding the existing Technical Specifications by the Decommissioning Technical Specifications does not increase the probability or consequence of an accident previously evaluated in the FSAR.

Ceasing plant operations and removing all irradiated fuel from the Reactor Building eliminates the probability of power operations and refueling accidents that are evaluated in the Fort St. Vrain FSAR.

The probability of occurrence of the accidents analyzed in the PDP are generally quite low.

An accident which could reasonably be expected to occur over the course of decommissioning is the loss of power.

However, the probability of such an occurrence is not significantly different from the loss of outsido electric power events analyzed in FSAR Section 10.3.

Heavy load drops of concrete rubble or activated graphite blocks are not anticipated during decommissioning, and are not considered to be significantly more probable than the load drops assessed in FSAR Section 9.2.11.

Should such load drops

occur, Decommissioning Technical Specifications 3.1 and 3.2 provide assurance that activity releases will be

filtered, as necessary, such that dose conseque. aces do not exceed a small fraction of EPA Protective Action Guidelines.

The probability of a fire which could result in the release of significant quantities of activity, such as analyzed in the PDP, is considered to be extremely low.

A rupture which would result in PCRV shielding water flowing into the Reactor Building is not expected to occur over the course of decommissioning, although the consequences of such an event are negligible since the water inventory is contained by the Reactor Building and only small amounts of tritium are released due to evaporation.

Decommissioning Technical Specification 3.4 assures tritium concentrations will not exceed those assumed in the PDP accident analyses.

The probability of natural disasters, such as earthquakes or tornadoes, is unchanged.

Page 6 of 10

DECOMMISSIONING OF THE FORT STo VRAIN NUCLEAR GENERATING STATION The radiological effluent disposal system specifications, Section 8 of the existing Technical Specifications, are not included in the Decommissioning Technical Specifications.

However, Specifications 5.9,

" Process Control Program" and 5.10, "Of fsite Dose Calculation and Radiological Environmental Monitoring Program Manuals",

in conjunction with Decommissioning Technical Specifications 2.9 and 2.11, require controls associated with radiological effluents to be maintained to assure compliance with 10CFR20 and 10CFR50 requirements.

Removal of detailed requirements for radiological effluents from the Technical Specifications, and incorporation of those requirements in programs referenced by the Technical Specifications, was approved by the NRC in NRC Generic Letter 89-01, dated January 31, 1989.

Since process controls on radiological effluents will remain, this activity does not increase the probability of uncontrolled effluent releases which could exceed 10CFR20 limits.

Based on the above, the probability of occurrence of accidents and malfunctions which could result in release of radioactivity is not significantly different from similar accidents and malfunctions evaluated in the FSAR.

With respect to the consequences of accident analyses, all fissionable material will be removed from the Reactor Building prior to commencing decommissioning activities.

As

such, accidents involving fissionable material are no longer credible.

Therefore, the risk of accidents resulting in a radiological release during decommissioning activities is considerably less than during plant operation.

Detailed analyses of postulated decommissioning accidents have been performed for the Proposed Decommissioning Plan.

These analyses have determined that the worst case accident would result in a whole body dose of 121 millirem and a 215 mil.14 rem dose to the organ (lung) at the EAB (100 meter boundary).

The results of the accident analyses indicate that the radiation exposures to the general public will be very low.

In all

cases, the radiological conscquences from the postulated decommissioning accident scenarios are well within the 25 rem whole body dose and 300 rem to any specific organ guidelines established by 10CFR 100 and also lower than accidents involving radioactive releases previously evaluated in the FSAR.

'I here fore, the decommissioning accident scenarios are bounded by the current design basis accident analyses, such that no increase in radiological consequences will result.

Page 7 of 10

DECOMMISSIONING OF THE FORT STo VRAIN j

NUCLEAR CENERATING STATION 1

2.

The issusnce of the Decommissioning Technical Specifications to superso.ie the existing Technical Specifications does not croato the porsibility of dif ferent typos of accidents or malfunctions than tnoso ovaluated previously in the FSAR.

Issuanco of the Dacommissior m Technical Specifications will provide usurance that structt.*or, systems and equipment rolled on to provent or nitigato the consequences of accidents postulated to occur during decommissioning will perform their intended safety function.

The majority of the existing Technical Specifications apply to components and systems relied upon to mitigate the consequences of accidents analyzed in the FSAR, such as reactivity excursions, loss of cooling accidents, etc.,

whoso occiarronco are no longer credible during decommissioning conditions, with all fuel removed.

Dolotion of those obscleto specification.9, and inclusion of the specifications required to mitigato the consequences of postulated decommissioning i

accidents is warranted.

The propocod Decommissioning Technical Specifications do r.nt placo plant systems in configurations conducivo to the occurrence of accidents or malfunctions not previously ovaluated.

The integrity of the Reactor Building, in conjuction with operation of the ventilation exhaust system, limits the off-sito dosos under normal and abnormal conditions during decommissioning activities.

The controls will assure the Reactor Building confinomont and the Reactor Building vontilation exhaust system are availablo to mitigato an activity release whenever decommissioning operations give riso to the possibility of a rolcase which could exceed a small fraction of the EPA Protectivo Action Guidelinos.

Controls also provido assurance that tritium activity concentrations in the PCRV shiclu water will not exceed those postulated in the decommisoloning accident analysos.

These Section 3 Conoral Requirements, as well as additional Administrative Controls delineated in Section 5 of the PDP, servo to promote radiation safety during decommiccioning, and do not contribute to the possibility of different types of accidents or malfunctions.

Many of the categories of accidents, or accident

types, ovaluated in the PSAR are not possible during decommissioning.

Sinco all irradiated fuel will be removed from the Reactor

Building, accidents related to reactivity excursions, power-to-flow mismatch, and loss of decay heat removal are no longer a concern.

An accident type ovaluated in the FSAR which is applicable during decommissioning operations is a broach or malfunction of containment permitting releaso of radioactivity.

Failures of the reactor coolant pressure boundary resulting in radioactivo roloases are evaluated in FSAR Sections 10. 2. 3. 4,

" Radiation Monitoring of Hot Roheat Piping,"

14.7,

" Primary Coolant Leakago",

14.8,

" Maximum Credible Accident", and 14.11, " Design Basis Accident No.

2-Rapid Depressurization/ Blowdown".

Malfunctions of auxiliary systems resulting in itskage of radioactivity are evaluated in Page 8 of 10

--_-~_-------.--

DECOMMISSIONING OF THE FORT ST. VRAIN NUCLEAR GENERATING STATION FSAR Section 14.6, which includes failures of the holium purification system, accidental release of the contents of a gas wasto surge tank, fuel handling and fuel storage accidents and drop of a fully loaded spent fuel shipping cask.

While specific decommissioning accidents differ from those containment failure accidents, in that fuel is not present during decommissioning and the PCRV is open to Reactor '

iding atmosphere and flooded with water, they generally f.1

.o the same accident

category, rolcaso of radioacti

.mo to containment failure.

Design Basis Accident No.

2 tapid Depressurization/ Blowdown, is the most sovero of this type of accident analyzed in the PSAR.

FSAR Section 14.11 identifics the of fsito dose consequences of this accident to an individual at the current 590 motor EAD as 2.5 rom whole body gamma, 17.4 rom thyroid and 4.8 rem bono.

The consequences of this accident are much more sovoro than thoso of the postulated decommissioning accidents analyzod in the PDP.

In that the possiblo range of accidents and malfunctions which could occur i

over the courso of decommissioning are the samo type of accident as previously evaluated in the FSAR under the category of primary coolant leaks and auxiliary system leaks, bounded by Design Basis Accident No.

2, no now types of accidents or malfunctions are being croatou.

3.

The postulated accident analysos havo verified that decommissioning activitics will be maintained within the bounds of safo, analyzed conditions as defined in the Proposed Decommissioning Plan and governed by the Docommissioning Technical Specifications.

Tho evaluation has taken into account the applicable Decommissioning Technical Specifications and has bounded the conditions under which the specifications permit decommissioning. The results as presented in the FSAR are bounding. As such, the margin of safoty, as defined in the bases to the Decommissioning Technical Specifications, is not reduced by replacing the current Technical Specifications, in I

their

entirety, with the Decommissioning Technical Specifications.

Based upon the proceding ovaluation, it has boon determined that the proposed chango to replaco the current Technical Specifications with the Decommissioning Technical Specifications at the Fort St.

Vrain Nuclear Gonorating Station does not involvo a significant t

l increase in the probability or consequences of an accident ar malfunction previously evaluated, croato the possibility of a now or different kind of accident or malfunction from any previously evaluated or involve a significant reduction in a margin of safety in the basis of any Technical Specification.

Thorofore, it is concluded that the licensing amendment does not involvo a

Significant Hazards Considoration as defined in 10 CPR 50.92 (c).

l Page 9 of 10

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DECOMMISSIONING OF THE FORT ST. VRAIN NUCLEAR GENERATING STATION FEPEREllCES i

1.

Manual af Protective Action Guidos and Protective Actions for Nuclear Incidents, EPA-520/1-75-001-A, January

1990, U.S.

Environmental Protection Agency.

2.

Fort St. Vrain Updated Final Safety A.lalysis Report, Revision 8,

Public Service company of Colorado.

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ATTACHMENT 5 TO P-90367

-i CONFORMANCE OF FSV DECOMMISS10lilNG TECHld1 CAL SPECIFICATIONS l

TO RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION (RETS)

GUIDANCE IN GENERIC LETTER 89-01 l

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4 CONFORMANCE OF FSV DECOMMISSIONING TECHNICAL SPEClf! CAT 13NS TO RAbl6EOML EFFLTIBIFTETHNICAL SPECIFTCATION (UT&T-GITfDANCFIN GtMNTCEETTYk 89-01 i

In Generic Letter 89-01, the NRC provided guidance to all power i

reactor licensees for implementing programmatic controls for the Radiological Effluent Technical Specifications (RETS) in the Administrative Controls section of the Technical Specifications, and for including procedural details of RETS in the Offsite Dose Calculation manual (ODCM) or in the Process Control Program (PCP).

The Fort St.

Vrain Deccmmissioning Technical Specifications (OTS),

which are provided as a separate attachment to this letter, conform to the Generic Letter 89-01 Model Technical Specification (MTS) guidance, with the following exceptions:

1.

The PCP definition was revised to add 49 CFR 100 to the list of applicable regulations.

This regulation provides requirements related to the transportation of radioactive material.

2.

Administrative Controls on the Radioactive Effluent Controls Program, paragraph 6.8.4.g in the MTS (paragraph 5.4.4.a in the FSV DTS):

a.

The introductory paragraph was revised from "shall be implemented by operating procedures" to "shall be implemented by procedures".

FSV operating procedures are currently controlled by the Onerations organization, which may not be radi' active effluent the organization responsible for o

controls procedures in the FSV decommissioning organization.

This revi' Ion provides organizational flexibility for the required procedures, b.

Program elements 7, 8, and 9 were changed from limiting dose rat 6s due to gaseous effluents released to areas "beyond the SITE BOUNDARY" to "beyond the EXCLUSION AREA BOUNDARY",

This change is conservative, since the Exclusion Area Boundary is L

within the Site Boundary, and will be consistent with PSC's Decommissioning Emergency Response Plan, which will determine-doses at the Exclusion Area Boundary.

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c.

Program element 9 was changed to delete limits on doses from lodine-131 and Iodine-133.

The UTS will not be applicable until all nuclear fuel has been removed from the FSV Reactor Building, so there will be no source of Iodine-131 or lodine-133.

3.

Administrative Controls on the Radiological Environmental Monitoring Program, paragraph 6.8.4.h in the MTS (paragraph 5.4.4.b in tne FSV DTS):

a.

Program element 5 was changed from requiring a Land Use Census for areas "beyond the SITE BOUNDARY" to "beyond the EXCLUSION AREA BOUNDARY", similar to the change described in paragraph 2.b above.

4.

The Annual Radiological Environmental Operating Report was changed from " covering the operation of the u r.i t" to " covering the activities of the unit", to reflect the non-operating status of the plant during decommissioning.

5.

The Semiannual Radioactive Effluent Release Report was also changed to refer to " activities" rather than " operation", similar to paragraph 4 above.

In addition, a sentence was added to require that a copy of the ODCM be provided if changes were made, for consistency with the requirements of MTS Specification 6.14 (Specification 5.10 in the FSV DTS).

6.

The Records Retention section was revised from requiring " records of reviews performed" for changes to the ODCM and PCP, to requiring " records of changes made" to the ODCM and PCP.

The actual reviews will be documented in the minutes of the Decommissioning Safety Review Comm". tee (DSRC), and Wese minutes will be retained for the duration of the Facility License.

7.

The PCP requirements in MTS Specification 6.13 (Section 5.9 in the FSV DTS) were revised as follows:

a.

The requirement was changed to apply to " permanent" changes, consistent with the temporary change provisions of FSV DTS Specification 5.4.3.

b.

The documentation equirement was revised to require that i

records of change rt<tews be retained as part of the DSRC meeting records, as described in paragraph 6 above.

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c.

The statement that PCP changes shall become effective "after rb.iew and acceptance by the [URG) and the approval of the Plant Manager" was revised to "after review and acceptance by the OSRC in accordance with Specification 5.3.6".

DSRC acceptance includes FSV's equivalent to the Plant Manager, because the Program Manager for Decommi n ning is the OSRC chairman.

DTS Specification 5.3.6 requires the DSRC to recommend approval of the PCP and changes thereto.

DTS Specification 5.4.2 requires that the PCP be reviewed by the OSRC, and approved by appropriate management.

8.

The ODCM requirements in MTS Specification 6.14 (Specification 5.10 in the FSV DTS) were revised as follows:

a.

The same changes described in paragraphs 7.a, b, and c above for the PCP were made for the ODCM.

b.

PSC deleted the requirement that changed ODCM pages "be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and i

shall indicate the date (e.g.,

month / year) the change was i mpl emian tr.d".

PSC considers that this requirement is a procedural detail that may be a good practice, but is not appropriate for a Technical Specification requirement.

PSC considers that the above changes from the guidance in Generic Letter 89-01 are not changes in substance and are consistent with the intent of providing programatic controls in the Technical Specifications, l

PSC is in the process of revising the ODCM and/or PCP, as appropriate, to include the procedural details covered in the current FSV RETS, including the limiting conditions for operation, the'.;

applicability, remedial actions, surveillance requirements, and bases.

This revision is expected to be completed by April 1, N91.

As requested in Generic l.e tter 89-01, the revised ODCM vil be provided for NRC reference, but not approval or concurrence, at that time.

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