ML20100L358
| ML20100L358 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 02/22/1996 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20100L353 | List: |
| References | |
| NUDOCS 9603040182 | |
| Download: ML20100L358 (44) | |
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l DECOMMISSIONING j
l TECHNICAL SPECIFICATIONS
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For FORT ST. VRAIN i
Unit No. 1 a
i Docket No. 50-267 i
Appendix A to Facility Licensa No. DPR-34 1
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FORT ST. VRAIN f
DECOMMISSIONING TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page Number 1-1
1.0 INTRODUCTION
2-1 2.0 DEFINITIONS 3.0-1 3.0 GENERAL REQUIREMENTS 2
3.1 Reactor Building Confinement Integrity 3.1-1 3.2 Reactor Building Ventilation Exhaust System 3.2-1 3.3 Radiation Monitoring Instrumentation 3.3-1 3.4 PCRV Shielding Water Tritium Concentration 3.4-1 4.0-1 4.0 DESIGN FEATURES 5.0-1 5.0 ADMINISTRATIVE CONTROLS 5.0-1 5.1 Responsibility 5.0-1 5.2 Organization 5.3 Decommissioning Safety Review Committee 5.0-1 5.0-4 5.4 Procedures and Programs 5.0-8 5.5 Reporting Requirements 5.0-9 1
5.6 Record Retention 5.0-10 5.7 Radiation Protection Program 5.0-11 5.8 High Radiation Area 5.0-12 5.9 Process Control Program 5.10 Offsite Dose Calculation and Radiological Environmental Monitoring Program Manuals 5.0-13 5.0-13 5.11 Natural Gas Restriction 1
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a Fort St.-Vrain DTS l
l Amendment No. 85 - 11/23/92 Page 1-1
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' 1.O INTRODUCTION r
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These Decommissioning Technical Specifications are applicable l
during the decommissioning of the Fort St. Vrain (FSV) reactor.
Decommissioning is considered to begin after all of the nuclear
. fuel has been removed from the FSV Reactor Building and after the NRC has approved the Decommissioning Plan.
i The Fort St.
Vrain Nuclear. Generating Station originally i
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 l
Specifications.
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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.
i There are two categories of FSV Technical Specifications:
" Decommissioning Technical Specifications (DTS)" include
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Amendment 85 and all subsequent amendments.
" Operating Technical Specifications" refers to the historical Technical Specifications included in all i
previous amendments.
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Fort St. Vrain DTS Amendment No. 86 - 11/29/93 Page 2-1 i
2.0 DEFINITIONS The defined terms in this section appear in capitalized type and are applicable throughout these Technical Specifications.
2.1 ACTIONS ACTIONS shal.1 be that part of a specification which prescribes hequired Actions under designated conditions, which shall be completed within specified Completion Times.
2.2 ACTIVATED GRAPHITE BLOCKS ACTIVATED GRAPHITE BLOCKS shall include all activated graphite components that were inside the PCRV when there was irradiated fuel in the core.
Defueling elements 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, l
including the sensors and alarm, interlock and/or trip functions, and may be performed by any series of l
sequential, overlapping, or total channel steps such that
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the entire channel is calibrated.
2.5 CHANNEL CHECK l
A CHANNEL CHECK shall be the qualitative assessment of
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channel behavior during its operation by observation.
This determination shall
- include, where
- possible, comparison of the channel indication and/or status with 4
other' indications and/or status derived from independent instrument channels measuring the same parameter.
Fcrt St. Vrain DTS Amendment No. 88 -
6/30/95 Page 2-2 DEFINITIONS (Continued)
I 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 the i
decommissioning Emergency Planning Zone (EPZ), as shown on Figure 4.1.
The EXCLUSION AREA BOUNDARY is a minimum of 100 meters from the Reactor Building, Fuel Storage Building, and Radioactive Waste Compactor Building.
2.8 MEMBER (S) OF THE PUBLIC MEMBER (S) OF 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) OF THE PUBLIC also includes persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
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2.9 OFFSITE DOSE CALCULATION MANUAL (ODCM) i The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain i
the methodology and parameters used in the calculation of j
offsite doses resulting from radioactive gaseous and
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liquid affluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in j
the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Specification 5.4.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental l
l Operating and Annual Radioactive Effluent Release Reports required by Specifications 5.5.1 and 5.5.2.
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Fort St. Vrain f
DTS Amendment No. 85 - 11/23/92 Page 2-3 DEFINITIONS (Continued) 2.10 OPERABLE - OPERABILITY A
component or system shall be OPERABLE or have i
OPERABILITY when it is capable of performing its intended safety function within the required range. The component 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 with the Surveillance Requirements defined in these Decommissioning Technical Specifications.
2.11 PROCESS CONTROL PROGRAM (PCP)
The PROCESS CONTROL PROGRAM (PCP) shall contain the i
procedure, 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 to assure compliance with 10 CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
2.12 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area inside or outside the EXCLUSION AREA BOUNDARY (or Emergency Planning Zone) 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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3 Fort St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.0-1 3.0 GENERAL REOUIREMENTS 3.0.1 Compliance with the Limiting Conditions (LC) contained in Section 3 of the Specifications is required during Fort St.
Vrain Decommissioning; except -that upon discovery of a failure to meet the LC, the associated
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Required Actions shall be het 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 l
need not be completed.
3.0.3 Surveillance Requirements shall be met as specified in' the Applicability for individual LCs unless otherwise stated in an individual Surveillance Requirement.
Failure to meet a Surveillance Requirement, 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.
j 3.0.4 Each Surveillance Requirement, any Required Actions which 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.
F 3.0.5 For a Surveillance Requirement not performed within the Frequency defined by 3.0.4, the ACTIONS are applicable at the time it is identified that the Surveillance has not been performed.
The Required l
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 the 24-hour j
allowance and the surveillance Requirements are not
- met, the Completion Times of the ACTIONS are applicable at that time.
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I Fort St. Vrain i
DTS I".
Amendment No. 85 - 11/23/92 Page 3.0-2 3.0 BASES 4
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3.0.1 and 3.0.2 3.0.1 and 3.0.2 establish the general requirements applicable to LCs.
These requirements are based on j
j the requirements consistent with operating plants' 1
Limiting Conditions for Operation per the Code of Federal Regulations, 10 CFR 50.36 (c) (2).
3.0.1 j
establishes the Applicability statement within
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individual specifications as the requirement for when conformance to the LC is required for safe i
decommissioning of the unit.
The Required Actions establish those remedial measures that must be taken within specified Completion Times when d
requirements of a LC are not met.
j 3.0.2 establishes that noncompliance with a
l specification exists when the requirements of the j
LC are not met and the associated Required Actions 2
have not been met within the specified Completion i
Times.
The purpose of this general requirement is j
to clarify that:
(1) completion of the Required Actions within the specified Completion Times l
constitutes compliance with a specification, and l
(2) completion of the remedial measures of the Required Actions is not required when compliance with an LC is restored within the completion Time 7
specified in the associated
- ACTIONS, unless 1
otherwise specified.
f 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 operating plants' j
j Surveillance Requirements stated in the Code of l
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 otherwise stated in an individual Surveillance Requirement.
The purpose of this general requirement is to ensure that Surveillances are performed to verify i
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.
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l Fort St. Vrain DTS t
Amendment No. 85 - 11/23/92 Page 3.0-3 r
3.0 BASES (Continued) 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 any Required Action with a
Completion Time requiring the periodic performance of an action on 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 sufficient 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
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t condition that constitutes a failure to meet the l
OPERABILITY requirements for an LC.
Under the i
provisions of this general requirement, systems and components are assumed to be OPERABLE when the associated Surveillance Requirements have not been I
met.
However, nothing in this provision is to be construed as implying that systems or components i
are OPERABLE when they are found or known to be inoperable although still meeting the Surveillance Requirement frequency.
This general requirement i;
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
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the point in time it is identified that a
Surveillance has not been performed and not at the c
time that the allowed Surveillance Frequency was exceeded.
1-s Fcrt St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.0-4 6
5 3.0 BASES (Continued) t I
If the Completion Times of the ACTIONS are less l
than 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 1
j performed.
If a Surveillance is not completed within the 24-hour allowance, the Completion Times i
of the ACTIONS are applicable at that time.
t For the purpose of making the transition from the operating Technical Specifications to the Decommissioning Technical Specifications, l
surveillances performed under the operating Technical Specifications may be utilized to satisfy i
the applicable surveillance requirements of the j
Decommissioning Technical Specifications.
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Fort St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.1-1 1
1 3.1 REACTOR BUILDING' CONFINEMENT INTEGRITY r
LC 3.1 Reactor Building confinement integrity shall be maintained with:
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The' Reactor Building overpressure protection system louvers closed *, and t
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Either:
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The outer truck bay closures closed, or 2.
The inner truck bay closures closed.
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' APPLICABILITY:
Whenever ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV and remain inside the Reactor 4
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Building
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l ACTION l
CONDITION REQUIRED ACTION COMPLETION TIME i.
A.
Do not have A.1 Suspend activities 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> j-Reactor Building involving physical confinement handling of integrity ACTIVATED GRAPHITE l
BLOCKS within the Reactor Building f
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i The Reactor Building overpressure protection system louvers may be open provided there are no activities j
in progress involving the physical handling of any ACTIVATED GRAPHITE BLOCKS.
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Fort St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.1-2 J
SURVEIr.TANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1 Verify all Reactor Building Daily, when physical overpressure protection system handling of louvers are in the closed ACTIVATFD GRAPHITE position, except as permitted by BLOCKS is in LC 3.1.
progress SR 3.1.2 Verify inner truck bay closures Prior to opening are closed outer truck bay closures i
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Fort St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.1-3 3.1 BASES BACKGROUND The integrity of the Reactor
- Building, in 2
conjunction with operation of the ventilation exhaust system, limits the off-site doses under 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 1
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 mair.tained subatmospheric at all times including normal accees (see LC 3.2).
Subatmospheric conditions can be maintained with several louver banks open.
Tha 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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a Fort St. Vrain DTS Amendment No. 86 - 11/29/93 Page 3.1-4 i,
j 3.1 BASES (Continued) j l
Reactor building confinement integrity is taken credit for in the Heavy Load Drop and the Loss of AC Power accident analyses, as described in Section 3.4 of'the Decommissioning Plan. (Reference 1)
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LC The LC establishes the minimum conditions required I
to ensure that Reactor Building confinement integrity is maintained during applicable accident j
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 s
that the Reactor Building overpressure protection system louvers may be-open provided there are no activities in progress involving the physical 4
l handling of any ACTIVATED GRAPHITE BLOCKS..
For example, the louvers may be open while ACTIVATED GRAPHITE BLOCKS are being dried or are in temporary i
storage within the Reactor Building, as long as they are not being
- moved, cut, or otherwise physically handled.
i APPLICABILITY The Reactor Building confinement integrity 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 l
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.
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Consistent with the Accident Analyses, ACTIVATED GRAPHITE BLOCKS include all graphite components inside the PCRV, except defueling elements.
The defueling elements are not activated.
In the' event of a
load drop accident involving defueling elements, the resultant doses are low enough that j
confinement integrity or ventilation are not required.
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Fcrt St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.1-5 i
1 3.1 BASES (Continued)
ACTIONS A,1 When Reactor Building confinement integrity is
- reached, suspend activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time to suspend physical handling of the ACTIVATED GRAPHITE BLOCKS allows an orderly suspension of activities.
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SURVEILLANCE REQUIREMENTS SR 3.1.1 The Reactor Building overpressure protection system louvers are verified in their closed position daily during activities when they are required to be closed, that is, during, although not necessarily contemporaneously
- with, physical handling of ACTIVATED GRAPHITE BLOCKS.
SR 3.1.2 Prior to opening the outer truck bay closures, the inner truck bay closures are verified closed.
While the outer truck bay closures are open, locks or signs are posted on the inner truck bay closures to prevent them from being opened.
This ensures Reactor Building confinement integrity.
i 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 4
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A Fort St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.2-1 3.2 REACTOR BUILDING VENTILATION EXHAUST SYSTEM i
LC 3.2 The Reactor Building ventilation exhaust system shall be l
OPERABLE with:
a.
Reactor Building internal pressure subatmospheric, and b.
At least one of the three ventilation exhaust trains OPERABLE, with each train consisting of one exhaust fan 4
(C-7301, C-7302, or C-7302S) and the HEPA filter section of the associated filter assembly (F-7301, F-7302, or F-7302S).
i APPLICABILITY:
Whenever ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV and remain inside the Reactor Building ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Reactor Building A.1 Suspend activities 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> pressure is involving physical atomspheric or handling of greater ACTIVATED GRAPHITE BLOCKS within the Reactor Building l
B.
All exhaust B.1 Restore at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> trains one ventilation inoperable exhaust train to OPERABLE status C.
Required Action C.1 Suspend activities 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B.1 not met involving physical within handling of Completion Time ACTIVATED GRAPHITE BLOCKS within the Reactor Building i
6 Fort St. Vrcin DTS Amendment No. 86 - 11/29/93 Page 3.2-2 i
i SURVEITTANCE REOUIREMENTS i
SURVEILLANCE FREQUENCY d
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SR 3.2.1 Verify Reactor Building pressure 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is subatmospheric i
SR 3.2.2 Verify pressure drop across each Weekly i
HEPA filter is less than 6 inches i
of water, with a flow rate of at least 17,100 cfm SR 3.2.3 Verify HEPA filter bank satisfies 18 months, after in-place penetration and bypass structural leakage test acceptance criteria maintenance on the l
l of less than 0.05 percent, using HEPA filter housing, test procedure guidance in or after each 3
Regulatory Positions C.5.a and complete or partial l
C.5.c of Regulatory Guide 1.52, replacement of a Rev.
2, March 1978, with a flow HEPA filter bank i
rate of at least 17,100 cfm i:'
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Fcrt St. Vrain l
DTS Amendment No. 85 - 11/23/92 Page 3.2-3 l
I 3.2 BASES l
BACKGROUND The Reactor Building ventilation exhaust filter system is designed to filter the Reactor Building i
atmosphere prior to release to the vent stack l
during both normal and most accident conditions during decommissioning.
The system consists of three trains, one of which i
is normally in continuous operation.
The design flow rate for each train is 19,000 cfa.
Allowing 10% for degradation, the minimum flow rate is
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1 17,100 cfm.
One train is sufficient to maintain the Reactor Building subatmospheric and thereby l
minimize unfiltered fission product release fron the building.
With only one exhaust fan operating, the ventilation system-controls will throttle fresh l
air supply to the air handler in order to reduce the pressure.
The Reactor Building is maintained in a
subatmospheric condition to ensure that all air leakage will be inward and to minimize unfiltered fission product release from the building.
The
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ventilation system was designed to maintain a subatmospheric condition approximately 1/4 inch l
water gauge negative.
In actual practice, the l
Reactor Building pressure is normally 0.15 to 0.20 inches water gauge negative, depending on building l
l activities and ventilation system configuration.
There is an alarm at approximately 0.08 inches l
l water gauge negative, and the outside air supply will fully close if the building pressure increases 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).
i LCs The LC establishes the minimum conditions required i
to ensure the Reactor Building ventilation exhaust system is maintained while the potential exists for a drop of an ACTIVATED GRAPHITE BLOCK.
One train 4
i is sufficient to maintain the Reactor Building subatmospheric and thereby minimize unfiltered fission product release from the building.
HEPA filters provide the required particulate
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filtration.
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Fcrt St. Vrain DTS Amendment No. 85 - 11/23/92 j~
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Page 3.2-4 3
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3.2 RASES (Continued) i APPLICABILITY The Reactor Building ventilation exhaust system 2
will remain
- OPERABLE, providing filtration of effluents to the environment, while the potential exists for dropping an ACTIVATED GRAPHITE BLOCK.
ACTIONS M
i When the Reactor Building pressure is atmospheric I
or greater, suspend activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building.
The one hour completion time to j
suspend activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building minimizes the-time exposure of the Reactor e
j Building to atmospheric or greater conditions and j
is a conservative time frame.
The suspension of physical handling activities is acceptable because l
all analyzed accidents assume something active is t
i happening - no passive postulated accidents will l
result in radiological conditions where the need for ventilation and confinement exists.
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The ability of the Reactor Building ventilation i
exhaust system to perform its filtering function during a Heavy Load (ACTIVATED GRAPHITE BLOCK) Drop i
is dependent on at least one exhaust train being l
With all the exhaust trains inoperable, i
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 i
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 will still be maintained at subatmospheric conditions.
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9.21 l-When Required Action B.1 cannot be completed within thw required completion
- Time, all -activities involving physical handling of ACTIVATED GRAPHITE BLOCKS-within the Reactor Building are suspended.
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Twelve hours is reasonable to suspend handling activities.
The suspension of physical handling activities is acceptable 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.
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Fcrt St. Vrain l
DTS Amendment No. 86 - 11/29/93 Page 3.2-5 l
3.2 BASES (Continued) i-SURVEILLANCE REQUIREMENTS SR 3.2.1 Verification that Reactor Building pressure is subatmospheric ensures that the confinement integrity is intact.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance i
frequency is more frequent than the operating 1
i technical specification requirements.
SR 3.2 1 i
A pressure drop across the HEPA filter of less.than 6 inches of water gauge at 90% of the filter design i
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 l
Particulate Air (HEPA) filters are determined by i
dioctyl phthalate (DOP) testing.
The filter j
penetration and bypass acceptance limits in the
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j surveillance are applicable based on a HEPA filter i
efficiency of 99
- percent, as assumed in the 1
decommissioning accident analysis.
The surveillance frequencies specified establish system j
performance capabilities.
Verification of the HEPA filter functions ensures system performance capabilities.
The surveillance frequency is the same as the operating technical specifications.
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REFERENCES 1.
FSV Decommissioning Plan i
2.
Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, EPA-
$20/1-75-001-A, January
- 1990, U.S.
Environmental Protection Agency a
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Fort St. Vrcin
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DTS Amendment No. 85 - 11/23/92 Page 3.3-1 3.3 RADIATION MONITORING INSTRUMENTATION s
LC 3.3 The area radiation monitoring instrumentation channels shown in Table 3.3-1 shall be OPERABLE with their alarm setpoints within the limits specified for the activities in progress, depending on whether Radiation Work Permit (RWP) controls l
are in effect.
l APPLICABILITY:
At all times, until all significantly contaminated or activated items that could exceed alarm setpoints have been removed from the Reactor Building.
ACTIONS
~
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more A.1 Adjust alarm 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l
radiation monitor setpoint within channel alarm limit setpoint exceeds i
value in Table OR 3.3-1 3
A.2 Declare the channel inoperable B.
One or more B.1 Place a portable 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> i
radiation monitor monitor (with 1
channels alarm) in the area j
4 SURVEILLANCE REQUIREMENTS i
I SR 3.3.1 Perform the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION surveillances as shown in Table 3.3-2.
Fort St. Vrain DTS Amendment No. 85 - 11/23/92 Page 3.3-3 TABLE 3.3-1 RADIATION MONITORING INSTRUMENTATION INSTRUMENT ALARM SETPOINTS DURING ACTIVITIES DURING ACTIVITIES i
NOT CONTROLLED CONTROLLED i
a.
Refueling Floor
< 15 mR/hr
< 100 mR/hr*
i 4
b.
Truck Bay
< 15 mR/hr
< 100 mR/hr*
1 4
Monitors may be reset to alarm at a radiation level within a l
factor of 2 of the expected radiation level.
i 4
TABLE 3.3-2 SURVEILLANCE REOUIREMENTS l
CHANNEL i
CHANNEL FUNCTIONAL CHANNEL l
INSTRUMENT CHECK TEST CALIBRATION a.
Refueling Floor Daily Monthly 6 months i
b.
Truck Bay Daily Monthly 6 months j
i i
..s w
Fort St. Vrain 5
DTS J
Amendment No. 85 - 11/23/92 Page 3.3-3 3.3 BASES i
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 i
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 i
investigated and appropriately resolved.
l Decommissioning 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 l
radiation levels will normally be anticipated and i
planned for, with monitoring provided as required.
l-Individual work activities will be performed under Radiation Work Permits (RWFs), which will include monitoring provisions.
- Also, gaseous effluent releases will be monitored and controlled by the i
Offsite Dose Calculation Manual (ODCM) program.
Liquid releases will also be monitored and controlled, in accordance with the ODCM 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.
4 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 setpoint, 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 z ll_ tion levels.
Fort St. Vrain DTS i
l l
Amendment No. 85 - 11/23/92 Page 3.3-4 i
}
3.3 BASES (Continued)
APPLICABILITY This LC is applicable at all times.
ACTIONS A.1 or A.2 When one or more radiation monitor channel alarm / trip setpoint exceeds the values in Table 3.3-1 either adjust the alarm / trip setpoint within its limits or declare ~the channel inoperable.
The 4
j 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.
M When one or more radiation monitor channels is i
inoperable, place a portable monitor with an alarm l
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
)
i-hour Completion Time is reasonable to complete the Required Action.
i SURVEILLANCE REQUIREMENTS SR 3.3.1 l
The surveillance requirements frequencies specified for CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION conform to industry practice i
and the surveillance frequencies given in Standard Technical Specifications and are adequate to ensure l~
the proper operation of these detectors.
j REFERENCES 1.
FSV Decommissioning Plan 2.
Offsite Dose Calculation Manual Program i
s
.-e
__.. _ -. _ ~.. _.
Fort St. Vrain 4
DTS r
Amendment No. 85 - 11/23/92 Page 3.4-1 3.4~
PCRV SHIRT.nING WATER TRITIUM CONCENTRATION LC 3.4 Tritium concentration in PCRV shielding water shall not exceed 62.4 #Ci/cc.
APPLICABILITY:
Whenever there is shielding water within the PCRV.
l ACTIONS.
3 CONDITION REQUIRED ACTION COMPLETION TIME l
l-A.
PCRV shielding A.1 Reduce tritium 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> water tritium concentration to concentration s 62.4 pCi/ c
> 62.4 gCi/cc OR
=
i A.2 Perfora engineering evaluation to verify total i
tritium content
$ 1 E+5 Ci B.
Required Actions B.1 Prepara and submit The next l
not met within to the NRC a 30 days i
required Special Report Completion Time describing the safety concerns and the plans for restoring tritium concentration to within the limit i
i 1
d 4
r
1 7
'Fcrt St. Vrain DTS i
Amendment No. 85 - 11/23/92 Page 3.4-2
[
SURVEIT T14CE REQUIREMENTS l
SURVEILLANCE FREQUENCY j
SR 3.4.1 Verify PCRV shielding water Daily, during tritium concentration within initial filling i
limits of the PCRV with shielding water, until tritium concentration is less than 0.1 pCi/cc for three consecutive samples.
SR 3.4.2 Verify PCRV shielding water Weekly, after i
tritium concentration within~
tritium limits concentration is less than 0.1 i
pCi/cc, until tritium l
concentration is less than 0.01 pCi/cc for three consecutive samples.
SR 3.4.3 Verify PCRV shielding' water Monthly, after tritium concentration within tritium limits concentration is less than 0.01 pCi/cc e
a l
I l
Fcrt St. Vrain i
DTS Amendment No. 85 - 11/23/92 Page.3.4-3 l
3'.4 BASES I
BACKGROUND During Decommissioning of Fort St.
- Vrain, the i
Prestressed Concrete Reactor Vessel (PCRV) cavity will be. flooded with water to facilitate the removal of the reactor core components.
PCRV i
dismantlement activities will begin only after all spent fuel has been removed from the reactor building.
The water will be circulated, and 3
purified by the PCRV water circulation system to gradually decrease
-the radioactivity, except i
tritium, in the water.
Thus, the flooding of the PCRV will provide shielding-for the workers associated with PCRV dismantlement activities.
t 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 domineralization for controlling dissolved solids, and " Feed and Bleed" for adding clean makeup water and for removing contaminated (primarily tritium) water.
The initial fluctuating increase in the tritium 1
concentration during the flooding of the PCRV will be controlled by the " Feed and Bleed" dilution i
process.
In accordance with the ODCM, released i
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 Loss of PCRV Shielding Water accident analysis, as described in Section 3.4 of the Decommissioning Plan (Reference l
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.
i n
l Fort St. Vrain DTS-
)
Amendment No. 88 -
6/30/95 Page 3.4-4 i '
t 1
3.4 BASES (Continued) i LC The LC establishes the maximum concentration i
j-tolerable in the PCRV shielding water to ensure j
adequate protection to the MEMBERS OF THE PUBLIC.
l The LC requirements are consistent with the accident analysis assumptions.
It should be noted that the accident analysis assumed 1 E+5 Curies 4
i released.
The resulting tritium concentration of j
62.4 pCi/cc was chosen as the LC requirement because it is easier to determine a
]
concentration for surveillance monitoring purposes.
i i
j APPLICABILITY This LC is applicable whenever there is shielding l
water within the PCRV.
j i
i i
ACTIONS A.1 or A.2 j
4 l
When the PCRV shielding water tritium concentration is greater than 62.4 pCi/cc it is prudent to either i
reduce the concentration to less than or equal to p
62.4 #Ci/cc or perform an engineering evaluation to verify 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 i
concentration of large water volumes and to perform i
associated analyses.
l M
l When a Required Action cannot be completed within l
the required Completion Time, a Special Report must l
be prepared and submitted to the NRC describing the safety concerns and the plans for restoring tritium j
concentration to within its safety analysis limit.
j 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 l
limits allowed by Reference 2.
The Special Report will be prepared as described in Specification i
5.5.4.
l 4
I l
I i
I J
i w-
Fcrt St. Vrain 1
DTS Amendment No. 85 - 11/23/92 Page 3.4-5 I
)
4 r
3.4 BASES (Continued) 1 i
SURVEILLANCE i
4 REQUIREMENTS SR 3.4.1 and 3.4.2
]
1 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-l shielding
- water, until tritium concentration I
i decreases below 0.1 pCi/cc, will detect any j
fluctuations in the tritium concentration during non-steady state conditions.
- Also, performing j
daily sampling until tritium concentration is less than 0.1 pCi/cc for three consecutive samples ensures that equilibrium conditions are achieved i
l before the surveillance frequency is decreased.
The 7 day surveillance frequency will ensure that a i
fluctuation in the tritium concentration during i
subsequent material handling activities will be detected.
After equilibrium tritium concentration has decreased below 0.01 pCi/cc, monthly sampling j.
will be performed.
This is conservative with l
~
respect to Regulatory Guide 8.32 sampling requirements.
i REFERENCES 1.
FSV Decommissioning Plan 3
2.
Manual of Protective Action Guides and l
Protective Actions for Nuclear Incidents, EPA-520/1-75-001-A, January
- 1990, U.S.
]
Environmental Protection Agency j-3.
Regulatory Guide 8.32, Criteria for Establishing a Tritium Bioassay Program.
i 1
a
-e er :
4
+
Fcrt St. Vrcin DTS 1
j-Amendment No. 88 -
6/30/95 f
Page 4.0-1 2
f
'4.O~
DESIGN FEATUDH 1
4.1 Site The Fort St. Vrain Nuclear Generating Station is located approximately 35 miles north of Denver and 3.5 miles i
northwest of the town of Platteville, in ' Wald County, l
i l
The site consists of 2798 acres.
The EXCLUSION AREA BOUNDARY encloses the decommissioning Emergency Planning j
Zone, as shown on Figure 4-1.
Points where radioactive gaseous and liquid effluents are released are shown on Figure 4-1.
j 4.2 Reactor Buildina 4
1 The Reactor Building houses the prestressed concrete reactor vessel (PCRV), fuel handling area, fuel storage j
- wells, fuel shipment preparation facilities, decontamination and radioactive liquid and gas waste l.
processing equipment, and most reactor plant process and l
service systems, i
l Decommissioning will not involve any major modifications j
to
-the Reactor Building structural steel without verification of the seismic qualification, as described in i
Section 2.2.1 of the Decommissioning Plan.
5 4.3 PCRV Water Leakaae Prevention The PCRV will be filled with water to provide shielding l
for workers during initial PCRV internal dismantlement l
activities.
To prevent leakage from the
- PCRV, all penetrations which are below the PCRV water line and have d
had their instrumentation removed are sealed.
Sealing is l'
accomplished with either welded cover plates, welded caps, i
or blind flanges.
Blind flanges for the seven outlet coolant thermocouple penetrations may be removed, one at a
- time, during i'
underwater removal of the thermocouple assemblies.
During this time, PCRV shield water leakage will be prevented by redundant seals on the thermocouple removal tools.
l l
There are two independent trains in the PCRV shield water system, to allow for maintenance and repair.
Each train has sufficient valves and drains to allow isolation as required.
l
1 e
Fort St. Vrain OTS m.
Amendment No. 85 - 11/23/92
'k'1\\ '; f__- -
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_. _ - ~. _
i i
Fort St. Vrain DTS Amendment No. 88 -
6/30/95 i
Page 5.0-1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility The Decommissioning Program Director shall have overall onsite responsibility for.all Fort St.
Vrain decommissioning activities, for both PSC and contractor. personnel.
The Program Director shall delegate in writing the succession to this responsibility during absences.
The Vice President responsible for nuclear activities shall have i
overall executive responsibility for all Fort St.
Vrain l
decommissioning activities.
l l
1
)
i 5.2 Organization The decommissioning organization, functional requirements, and i
qualification requirements for key decommissioning personnel, for both PSC and contractor groups, shall be documented in the f
FSV Decommissioning Plan.
J;'
The organization responsible for quality assurance shall report to the Vice President responsible for nuclear activities on j
quality assurance matters, to ensure independence.
An individual qualified in radiation protection procedures shall 4
be present at the facility at all times during physical decommissioning activities.
i l
5.3 Decommissionino Safety Review Committee (DSRC) l 5.3.1 The DSRC shall be comprised of the following:
l Decommissioning Program Director (Chairman)
{
Radiation Protection Manager Engineering Manager Operations / Maintenance Manager Project Assurance Manager l
consultants may be appointed as members, in writing, by the DSRC Chairman An alternate Chairman and alternate
- members, if required, shall be appointed in writing by the DSRC Chairman.
i 1
i
~
Fort St. Vrain DTS Amendment No. 85 - 11/23/92 Page 5.0-2
~
ADMINISTRATIVE CONTROLS (Continued)
]
4 i
5.3.2 The DSRC shall meet at least once per calendar quarter, or more frequently as convened by the DSRC Chairman or the Vice President responsible for nuclear activities.
[!
]
5.3.3 A quorum of the DSRC shall consist of-the Chairman or alternate
- Chairman, and a simple majority of the
]
- members, including alternates.
No more than two alternate members shall participate as voting members in DSRC activities at any one time.
a 5.3.4 The DSRC shall be responsible for review of:
i a.
Administrative procedures,
- plans, manuals, and i
programs required by Specifications 5.4.1 through I
5.4.4, 5.7, and permanent changes thereto, that
[
affect nuclear safety.._
J j
b.
Proposed tests and experiments that affect nuclear i
safety.
i c.
The following items which involve an unreviewed safety question as defined in 10 CFR 50.59:
i l
1)
Administrative procedures, plans, manuals, and programs required by specifications 5.4.1 i.
through 5.4.4, 5.7, and permanent changes
- thereto, l
2)
Proposed changes or modifications to plant i
systems or equipment, and 3)
Proposed tests and experiments.
I d.
Proposed changes to the decommissioning work specifications that affect nuclear safety, and any new decommissioning work specifications that affect nuclear safety.
i e.
Proposed changes to the Decommissioning Technical Specifications or Facility License.
f.
Investigations of violations of Decommissioning Technical Specifications, and of regulations or
-license requirements.
t l
g.
Reportable events as defined by 10 CFR 50.73.
l h.
Unplanned release of radioactive material to the environs.
i,
)
Fort St. Vrain DTS i
Amendment No. 85 - 11/23/92 Page 5.0-3 j.
ADMINISTRATIVE CONTROLS (Continued)
~
5.3.5 The DSRC shall:
5 a.
Advise the Decommissioning Program Director on matters that affect nuclear safety.
I b.
Recommend to the Decommissioning Program Director in 1
writing, approval or disapproval of items considered i
under Specifications 5.3.4.a through 5.3.4.d above.
i i
c.
Render determinations in writing with regard to whether or not each item considered under Specification 5.3.4.c constitutes an unreviewed safety question.
j i
j.
d.
Recommend to the Decommissioning Program Director j
other areas of facility activities where additional L
oversight is prudent and/or where independent 1
l auditing is needed.
5.3.6 Audits of decommissioning activities shall be performed
)
i under the cognizance of the DSRC.
These audits shall l
encompass:
A decommissioning program audit to be performed at a.
i least once per year, encompassing the following:
1)
Decommissioning Technical Specifications 2)
Radiation Protection Program 3)
Training Program j
4)
Decommissioning QA Plan 5)
Decommissioning Access Control Plan t
6)
Decommissio.unes Fire Protection Plan 7)
Decommiss!.oning Emergency Response Plan l
l b.
-Any other atea of facility activities considered appropriate by the DSRC.
?
5.3.7 Records of DSRC activities shall be prepared, approved, and distributed as indicated below:-
l.
Minutes of each DSRC meeting and documentation of l
a.
l the reviews performed per Specification 5.3.4 above shall be approved and forwarded to the Vice President responsible for nuclear activities within i
i 30 days following the meeting.
1
N*,
I E
, i.
Fort St. Vrain DTS i
Amendment No. 88 -
6/30/95 Page 5.0-4 i
e i
q ADMINISTRATIVE CONTROLS (Continued)
{
i e
'l b.
Audit reports encompassed by - Specification 5.3.6 above shall be forwarded to the Vice President responsible for nuclear activities within 30 days j
after completion of the audit.
~
4:
5.4 Procedures and Proarm==
5.4.1 Written administrative procedures,
- plans, manuals, t
i and/or programs shall be established, implemented, and maintained covering the activities referenced below:
i 1
l a.
Radiation Protection Program j
b.
Surveillance test activities of equipment required j
by these Decommissioning Technical Specifications i
i c.
Decommissioning Access Control Plan i
d.
Decommissioning Emergency Response Plan i
i e.
f f.
OFFSITE DOSE CALCULATION MANUAL g.
Decommissioning Fire Protection Plan l
5.4.2 Administrative procedures,
- plans, manuals, and/or f
programs of Specification 5.4.1 above, and permanent changes thereto, that affect nuclear safety, shall be a
i reviewed by - the DSRC, or a subcommittee thereof, and j
approved by the appropriate management prior to implementation.
Procedures shall be reviewed l
3 periodically as set forth in Administrative Procedures.
i i
~
Changes to the OFFSITE DOSE CALCULATION MANUAL shall be j
processed in accordance with Specification 5.10, and j
changes to the PROCESS CONTROL PROGRAM shall be I
processed in accordance with Specification 5.9.
l 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 tihe appropriate management prior to i
implementation.
s t
f
l 4
Fort St. Vrain i
DTS Amendment No. 85 - 11/23/92 i
i, Page 5.0-5 i
- ADMINISTRATIVE CONTROLS (Continued) i 5.4.4 The following programs shall be established, implemented, and maintained:
I a.
Radioactive Effluent Controls Procram l
A program shall be provided conforming with 10 CFR l
50.36a for the control of radioactive affluents and i
for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive affluents as low as reasonably i
achievable.
The program (1) shall be contained in j
j the OFFSITE DOSE CALCULATION MANUAL, (2) shall be implemented by procedures, and (3) shall include remedial actions to be taken whenever the program i.
limits are exceeded.
The program shall include the following elements:
t 1)
Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation i
including surveillance tests and setpoint determination in accordance with the methodology in the OFFSITE DOSE CALCULATION MANUAL, i
[
2)
Limitations on the concentrations of radioactive material released in liquid affluents to i
UNRESTRICTED AREAS conforming to 10 CFR Part 20 4
- limits, l
3)
Monitoring,
- sampling, and analysis of
}
radioactive liquid and gaseous affluents in accordance with 10 CFR 20 and with the j.
methodology and parameters in the OFFSITE DOSE l
CALCULATION MANUAL, t
l 4)
Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid affluents 4
released from each unit to UNRESTRICTED AREAS
)
conforming to Appendix I to 10 CFR Part 50, 1
5)
Determination of cumulative and projected dose contributions from radioactive affluents for the current calendar quarter and current calendar 4
year in accordance with the methodology and i
parameters in the OFFSITE DOSE CALCULATION MANUAL at least every 31 days, i
1 1
.. ~ -
= _.. -..
1 g
i Fort St. Vrain DTS Amendment No. 85 - 11/23/92 i
Page 5.0-6 i
i 1
ADMINISTRATIVE CONTROLS (Continued)'
6)
Limitations on the operability and use of the liquid and gaseous effluent treatment systems to l
ensure that the appropriate portions of these j
systems are used to reduce releases of j
radioactivity when the projected doses in a 31-day period would exceed 2
percent of the l
guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, j
7)
Limitations on the dose rate resulting from radioactive material released in gaseous affluents to areas beyond the EXCLUSION AREA BOUNDARY conforming to the doses associated with 10 CFR part 20, 8)
Limitations on 'the annual and quarterly air i
doses resulting from noble gases released in i
gaseous effluents to areas beyond the EXCLUSION l
AREA BOUNDARY conforming to Appendix I to 10 CFR Part 50, t
i-9)
Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from tritium and all radionuclides in particulate form with half-l lives greater than 8 days in gaseous affluents i
released to areas beyond the EXCLUSION AREA BOUNDARY conforming to Appendix I to 10 CFR Part 50, l
i
- 10) Limitations on the annual dose or dose j
commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR l
Part 190.
I
(
=
s e
Fcrt St. Vrain DTS Amendment No. 85 - 11/23/92 Page 5.0-7 ADMINISTRATIVE CONTROLS (Continued) b.
Radioloaical Environmental Monitorina Procram 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 affluent monitoring program and modeling of environmental exposure pathways.
The program shall (1) be contained in the OFFSITE DOSE CALCULATION MANUAL, (2) conform to the guidance of Appendix I 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 Comparison Program to ensure that independent checks on the 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 4
1 l
i l
r a
4
, +
Fort St. Vrain DTS
[
Amendment No. 85 - 11/23/92 i
Page 5.0-8 4
ADMINISTRATIVE CONTROLS (Continued) j 5.5 Reportina Requirements J
In addition to the applicable reporting requirements of 10.CFR, j
the following reports shall be submitted to the Regional j
Administrator of the NRC's Region IV office unless otherwise
{
.noted:
5.'5.1 Annual Radioloaical Reports i
Annual reports covering the activities described below, for the previous calendar year shall be submitted as i
follows:
l a.
Annual Radiation Exposure Report
)
The Annual Radiation. Exposure Report for the previous calendar year shall be submitted to the j
commission within the first calendar quarter of each calendar year in compliance with 10 CFR 20 and in accordance with the guidance contained in Regulatory Guide 1.16.
t I
b.
Annual Radiological Environmental Operating Report
}
The Annual Radiological Environmental Operating Report covering the activities of the unit during j
the previous ' calendar year shall be submitted before i
May 1 of each year.
The report shall include j
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 l
l objectives outlined in (1) the OFFSITE DOSE
)
CALCULATION MANUAL and (2) Sections IV.B.2, IV.B.3, j
j and IV.C of Appendix I to 10 CFR Part 50.
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Fcrt St. Vrain DTS Amendment No. 85 - 11/23/92 Page 5.0-9 ADMINISTRATIVE CONTROLS (Continued) 5.5.2 Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report covering activities during the previous 12 months shall be submitted within 90 days after January 1 of each year.
The report shall include a summary of the quantities of radioactive liquid and gaseous affluents 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
(1) consistent with the objectives outlined in the OFFSITE DOSE 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 Nonroutine ReDorts 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.
5.5.4 Soecial Recorts Special Reports required by Specification 3.4 shall be submitted to the NRC Regional Administrator within the time period specified.
1 i
5.6 Record Retention 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 nuclear safety.
b.
Licensee Event Reports (LERs).
c.
Records of surveillance activities, inspections, and calibrations required by the Decommissioning Technical Specifications.
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Fcrt St. Vrnin DTS Amendment No. 85 - 11/23/92 f
Page 5.0-10 ADMINISTRATIVE CONTROLS (Continued) d.
Records of changes made to procedures related to c
l nuclear safety.
e.
Records of radioactive shipments.
f.
Records of sealed source leak tests and results.
i 5.6.2 The following records shall be retained for the duration i
of the Facility License:
a.
Dismantlement records for systems and equipment related to nuclear 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.
1 j-e.
Records of training and qualification for current l
members of the decommissioning staff.
f.
Records of activities required by the i
Decommissioning QA Plan.
l' g.
Records of reviews performed pursuant to 10 CFR 50.59.
l h.
Records of meetings of the DSRC.
l i.
Records and logs pertaining to the Radiological Environmental Monitoring Program.
I j.
Records of changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
5.7 Radiation Protection Procram 4
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.
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Fort St. Vrain j
DTS Amendment No. 85 - 11/23/92 Page 5.0-11 ADMINISTRATIVE CONTROLS (Continued) 5.8 Hiah Radiation Area 5.8.1 Pursuant to 10 CFR 20, in lieu of the " control device" or " alarm signal", each high radiation area, as defined in 10 CFR Part 20, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation 1
Work Permit (RWP).
Individuals qualified in radiation protection procedures (e.g., Health Physics personnel) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such 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 device which continuously indicates the radiation dose rate in the area, or j
b.
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device i.
may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them, or c.
A health physics qualified individual (i.e.,
qualified in radiation protection procedures) with a i
radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics staff in the RWP.
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Fort St. Vrain DTS l.
Amendment No. 88 -
6/30/95 Page 5.0-12 I
l ADMINISTRATIVE CONTROLS (Continued) 5.8.2 In addition to the requirements of 5.8.1, areas 3
accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in) from the radiation l;
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 i
l physics supervision.
Enclosures shall remain locked except during periods of access by personnel under an i
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 remoto (such as l
use of closed circuit TV cameras) continuous i
surveillance may be made by personnel qualified in l
radiation protection procedures to provide positive exposure control over the activities within the area.
l For individual areas accessible to personnel with radiation levels of greater than 1000 mR/h that are i
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 barricaded, conspicuously posted, and a flashing light shall be activated as a warning device i
whenever the dose rate in the area exceeds or will i
shortly exceed 1000 mR/hr.
5.9 PROCESS CONTROL PROGRAM (PCP)
Permanent changes to the PROCESS CONTROL PROGRAM:
a.
Shall be documented and records of reviews performed shall be retained as part of the DSRC 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 overall conformance of the solidified waste product to existing requirements of
- Federal, State, or other applicable regulations.
b.
Shall become effective after review and acceptance by the l_
DSRC in accordance with Specification 5.3.4.
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6/30/95 Page 5.0-13 l
ADMINISTRATIVE CONTROLS (Continued) 5.10 OFFSITE DOSE CALCULATION MANUAL Changes to the OFFSITE DOSE CALCULATION MANUAL:
a.
Shall be documented and records of reviews performed shall be retained as part of the DSRC 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 l
the change (s) and i
2)
A determination that the change will maintain the level of radioactive affluent control required by 10 CFR 20, 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 I
calculations.
b.
Shall become effective after review and acceptance by DSRC l
in accordance with Specification 5.3.4.
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 l
Annual Radioactive Effluent Release Report for the period of the report in which any change to the OFFSITE DOSE CALCULATION MANUAL was made.
5.11 Natural Gas Restriction l
As indicated in Specification 1.0, FSV may be converted to j.
utilize a gas-fired boiler.
The natural gas line supplying this boiler, or any other new natural gas source, shall not be i
introduced within 0.5 miles of the location where ACTIVATED GRAPHITE BLOCKS are stored, for any purpose, without prior NRC approval.
PSC shall submit an analysis of any proposed new i
natural gas source demonstrating that the new source will not L
present an unacceptable hazard to the ACTIVATED GRAPHITE BLOCKS or to the equipment or systems needed to protect the ACTIVATED GRAPHITE BLOCKS.
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