ML20082K246

Proposed Decommissioning Tech Specs Administrative Control 5.3.1,reflecting Organizational Changes That Impact Membership of Decommissioning Safety Review Committee

ML20082K246
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/14/1995
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20082K220	List: ML20082K217 ML20082K227 ML20082K246
References
NUDOCS 9504190180
Download: ML20082K246 (14)

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l ATTACIIMENT 2 TO P-95037 PROPOSED CIIANGES 9504190180 950414 PDR ADOCK 05000267 W PDR

Fort St. Vrcin DTS Amendment No. _ - XX/XX/XX 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. j 2.7 EXCLUSION AREA BOUNDARY l 1 The EXCLUSION AREA BOUNDARY shall enclose the 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 l 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 l t recreational, occupational, or other' purposes not I associated with the plant. l

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! 2.9 OFFSITE 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 3 Program. The ODCM shall also contain (1) . the Radioactive  ;

Effluent Controls and Radiological Envircnmental i Monitoring Programs required by Specification 5.4.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental i l Operating and Annual Radioactive Effluent Release Reports l required by Specifications 5.5.1 and 5.5.2.

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Fort St. Vrain DTS j-

• Amendment No. _ - XX/XX/XX 3

Page 3.4-4 i

3.4 BASES (Continued)

I LC The LC establishes the maximum concentration tolerable in the PCRV shielding water to ensure 5l adequate protection to the MEMBERS OF THE PUBLIC.

i The LC requirements are consistent with the accident analysis assumptions. It should be noted l that the accident . analysis assumed 1 E+5 Curies i released. The.resulting tritium concentration of 1 62.4 #Ci/cc was chosen as the LC requirement

- because it is easier to determine 'a tritium l concentration for surveillance monitoring purposes.

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} APPLICABILITY This LC is applicable whenever there is shielding l l

l water within the PCRV.

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! ACTIONS A.1 or A.2 i When the PCRV shielding water tritium concentration j is greater than 62'.4.gCi/cc it is prudent to either i

reduce the concentration to less than or equal to 62.4 pCi/cc or perform an engineering evaluation to j verify that the total tritium content is less than i or equal to 1 E+5 Curies. A completion time of 72 j hours is a reasonable amount of time to change the j concentration of large water volumes and to perform

associated analyses.

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! When a Required Action cannot be completed within j the required Completion Time, a Special Report must j 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 j limits allowed by Reference 2. The Special Report 4 will be prepared as described in Specification 5.5.4.

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DTS

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Amendment No. _ - XX/XX/XX  :

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4.0 DESIGN FEATURES l

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l 4.1 Site )

j The Fort St. Vrain_ Nuclear Generatihg Station is located j approximately 35 miles north of Denver and 3.5 miles

' northwest of the town of Platteville, in Wald County, .

Colorado.

i The. site consists of 2798 acres. The EXCLUSION AREA

BOUNDARY encloses the decommissioning Emergency Planning

) Zone, as shown on Figure 4-1.

3 Points where radioactive gaseous and liquid effluents are released are shown on Figure 4-1.

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) 4.2 Reactor Buildina S The Reactor Building houses the prestressed concrete reactor vessel (PCRV), fuel handling area,. fuel storage l wells, fuel shipment preparation facilities, j_ decontamination and radioactive liquid and gas waste

processing equipment, and most reactor plant process and  ;

service systems. l Decommissioning will'not involve any major modifications j to the Reactor Building structural steel without j verification of the seismic qualification, as described in  ;

j Section 2.2.1 of the Decommissioning Plan. _

i i 4.3 PCRV Water Leakace Prevention

! The PCRV will be filled with water.to-provide shielding i for workers during initial PCRV internal dismantlement i activities. To prevent leakage from. the PCRV, all j penetrations which are below the PCRV water line and have i had their instrumentation removed are sealed. Sealing is i accomplished with either welded cover plates, welded-caps, or blind flanges.

Blind flanges for the seven outlet coolant thermocouple penetrations may be removed, one at a time, during underwater removal of the thermocouple assemblies. During 3

this time, PCRV shield water leakage will be prevented by

redundant seals on the thermocouple removal tools.

i i There are two independent trains in the PCRV shield water l

] system, to allow for maintenance and repair. . Each train t- has sufficient valves and drains to allow isolation as required.

Fort St. Vrain DTS Amendment No. __ - XX/XX/XX Page 5.0-1 5.0 ADMINISTRATIVE CONTROLS I

5.1 Resconsibility The Decommissioning Program Director shall have overall onsite responsibility for all Fort St. Vrain decommissioning l activities, for both PSC and contractor personnel. The Program j Director shall delegate in writing the succession to this responsibility during absences.

f The Vice President responsible for nuclear activities shall have l overall executive responsibility for all Fort St. Vrain !

decommissioning activities. j l

5.2 Oraanization

! The decommissioning organization, functional requirements, and

! qualification requirements for key decommissioning personnel, i for both PSC and contractor groups, shall be documented in the l FSV Decommissioning Plan.

The organization responsible for quality assurance shall report l to the Vice President responsible for nuclear activities 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 Decommissionina Safety Review Committee (DSRC) 5.3.1 The DSRC shall be comprised of the following:

Decommissioning Program Director (Chairman)

Radiation Protection Manager Engineering Manager Operations / Maintenance Manager Project Assurance Manager I

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.

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Fort St. Vrain DTS Amendment No. _ - XX/XX/XX Page 5.0-4 1 ADMINISTRATIVE CONTROLS (Continued) l 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 after completion of the audit.

I l 5.4 Procedures and Proarams t

l 5.4.1 Written administrative procedures, plans, manuals, i l and/or pr >qrams shall be established, implemented, and i maintained covering the activities referenced below:

a. Radiation Protection Program  :

b. Surveillance test activities of equipment required  ;

by these Decommi::sioning Technical Specifications l 1

c. Decommissioning Access Control Plan l

d. Decommissioning Emergency Response Plan l \

e. PROCESS CONTROL PROGRAM l 1

f. OFFSITE DOSE CALCULATION MANUAL l

g. Decommissioning Fire Protection Plan 5.4.2 Administrative procedures, plans, manuals, and/or programs of Specification 5.4.1 above, and permanent

! changes thereto, that affect nuclear safety, shall be l reviewed by the DSRC, or a subcommittee thereof, and approved by the appropriate management prior to implementation. Procedures shall be reviewed l

periodically as cet forth in Administrative Procedures.

i Changes to the OFFSITE DOSE CALCULATION MANUAL shall be i processed in accordance with Specification 5.10, and changes to the PROCESS CONTROL PROGRAM shall be 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 l

may be made provided the change is documented and approved by the appropriate management prior to implementation.

I Fort St. Vrain DTS Amendment No. _ - XX/XX/XX Page 5.0-12 l

ADMINISTRATIVE CONTROLS (Continued) l 5.8.2 In addition to the requirements of 5.8.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in) from the radiation source or from any surface which the radiation l penetrates shall be provided with locked enclosures to l 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 l

the immediate work area and the maximum allowable stay l 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 l

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 individua] areas, then that area shall be barricaded, conspicuously posted, and i

a flashing light shall be activated as a warning device l 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:

l 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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Fort St. Ircin

, DTS Amendment No. __ - XX/XX/XX Page 5.0-13 l

! ADMINISTRATIVE CONTROLS (Continued) 5.10 OFFSITE DOSE CALCULATION MANUAL Changes to the OFFSITE DOSE CALCULATION MANUAL:

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! 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 level of radioactive effluent control required by 10 CFR 20,

10 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 l CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint i calculations, j i

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 i complete, legible copy of the entire OFFSITE DOSE l l

CALCULATION MANUAL as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of  ;

l the report in which any change to the OFFSITE DOSE '

CALCULATION MANUAL was made.

5.11 Natural Gas Restrictio_D l As indicated in Specification 1.0, FSV may be converted to utilize a gas-fired boiler. The natural gas line supplying this boiler, or any other new natural gas source, shall not be 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 natural gas source demonstrating that the new source will not 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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1 ATTACIIMENT 3 1

TO P-95037 NO SIGNIFICANT IIAZARDS CONSIDERATION EVALUATION I

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DECOMMISSIONING OF THE FORT ST. VRAIN l NUCLEAR GENERATING STATION j NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION IN ACCORDANCE WITH 10 CFR 50.92 j INTRODUCTION The Decommissioning Order approving the Decommissioning Plan (DP) and authorizing the decommissioning of the Fort St. Vrain Nuclear Station was issued by the NRC on November 23,1992 (Reference 1). As identified in the Decommissioning Order, decommissioning of FSV is authorized in accordance with the DP subject to several conditions. Condition (c) states that "If the licensee desires (1) a change in the TS or (2) to make a change in the facility or procedures described in the Decommissioning Plan or to conduct tests or experiments not described in the Decommissioning Plan, which involve an unreviewed safety question or a change in the TS, it shall submit an application for amendment of its license pursuant to 10 CFR 50.90 or request approval of a revision to the Decommissioning Plan."

Pursuant to 10 CFR 50.92, each application for amendment to an operating license is 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 (10 CFR 50.92(c)]. A proposed amendment to an operating license for a facility involves no significant hazards consideration if the change 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 evaluated, or

3) involve a significant reduction in a margin of safety.

This amendment primarily addresses organizational changes within Public Service Company of Colorado (PSC) associated with recent corporate restructuring efforts and with the transition from a decommissioning organization to the organization that will manage the re-powered Fort St. Vrain facility. In addition, several minor editorial corrections are addressed. This revision will ensure that the Decommissioning Safety Review Committee membership requirements have the diversity of technical and administrative expertise necessary to monitor decommissioning operations, consistent with our corporate organization i planning.

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i BACKGROUND Administrative Control 5.3 defines the requirements for the Fort St. Vrain Decommissioning i Safety Review Committee (DSRC). The membership of the DSRC is currently defined as  !

I including the following positions-l Decommissioning Program Director (Chairman) I Deputy Director  !

Facility Support Manager (Radiation Protection Manager)  !

Decommissioning Engineering Manager .j Operations Manager . l Project Assurance Manager  ?

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Westinghouse Project Director j Plus consultants appointed as members, in writing, by the DSRC Chairman ]

Public Service Company of Colorado has recently undertaken a corporate restructuring in an effort to reduce costs and remain a viable provider of utility services. PSC's restructuring .

has eliminated numerous middle management positions, as is happening at similar- )

restructuring projects in many companies around the country. 'In the Fcit St. Vrain l organization, this effort has considered the organizational structure that is needed'to ensure l performance excellence, management competence, and high standards in every facet of the  ;

I decommissioning project. In addition, this effort is reviewing the organization that will be required for the FSV facility after it is repowered with natural gas-fired combustion turbines and heat recovery boilers, considering that the Independent Spent Fuel Storage Installation (ISFSI) will likely remain.

The impact that these efforts have had on the DSRC is that the Deputy Director and Project

! Controls Manager positions have been deleted; the Facility Support Manager has become the Radiation Protection Manager; the Operations Manager will be known as the Operations /

Maintenance Manager; and the Decommissioning Engineering Manager will be known as the Engineering Manager.

Another factor that will affect DSRC membership is the eventual relocation of individuals  ;

in the Westinghouse Project team. Most of the Westinghouse Team will be relocated when the final site survey has been completed, and some relocations may occur before that, such as upon completion of physical work activities on site. In anticipation of the possibility that  :

the Westinghouse Project Director will be relocated offsite before license termination, PSC is proposing to delete this position from the defined DSRC membership, and to retain this

, position as a consultant appointed in writing by the DSRC chairperson, for as long as the

! individual in this position is located on site.

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I With these changes, the revised DSRC membership described in this amendment consists of .

the following:

Decommissioning Program Director (Chairman) ,

Radiation Protection Manager Engineering Manager l Operations / Maintenance Manager Project Assurance Manager Plus consultants appointed as members, in writing, by the DSRC Chairman ,

The size of the DSRC will change from eight individuals to six, including the Westinghouse i Project Director, appointed as a DSRC consultant by the Chairman, t

EVALUATION With the exception of the change to the Decommissioning Safety _ Review Committee membership in.Section 5.3.1, all of the changes described in Attachments 1 and 2 are editorial in nature, correcting references and revising nomenclature for consistency. The following changes correct oversights in the original submittal of the Decommissioning Technical Specifications and involve No Significant Hazards Consideration:

• Definition 2.9. This revision of the definition of the Offsite Dose Calculation Manual to refer to the Annual Radioactive Effluent Release Report versus the Semiannual Radioactive Effluent Release Report provides consistency with the section that.

provides the frequency requirement for this report, Section 5.5.2.

• Bases 3.4. This revision to the Applicability discussion for the Technical Specification for tritium concentration in the PCRV shield water states that the LC is applicable whenever there is shielding water within the PCRV, not just until all Activated Graphite Blocks have been removedfrom the PCRV, to provide consistency with the Applicability section wording.

• Design Feature 4.3. This revision corrects the PCRV Water Irakage Prevention Design Features discussion to refer to the PCRV shield water system versus the  ;

PCRV water cleanup and clanfication system. l

• Administrative Control 5.3.7.b. This revision corrects the reference for audit reports from Specification 5.3.7 to Specification 5.3.6.

• Administrative Control 5.9.b. This revision corrects the reference for DSRC review of PCP changes from Specification 5.3.6 to Specification 5.3.4.

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• Administrative Control 5.10.b. This revision corrects the reference for DSRC review of ODCM changes from Specification 5.3.6 to Speci6 cation 5.3.4.

• Administrative Control 5.10.c. This revision corrects the reference from a Semiannual Radioactive Effluent Release Report to an Annual Radioactive Effluent Release Report, consistent with Section 5.5.2.

• Administrative Control 5.11. This revision changes the Natural Gas Restriction to state that FSV may be converted to utilize a gas-fired boiler instead of stating that FSV is being converted, to acknowledge PSC's plans to install natural gas-fired combustion turbines and to allow any other FSV repowering options.

The revision of the Decommissioning Safety Review Committee membership in l Administrative Control 5.3.1 also involves No Significant Hazards Consideration. This change reduces the size of the DSRC by two individuals by deleting the positions of Deputy Director and Project Controls Manager. This change also revises the titles of the Facility Support Manager (Radiation Protection Manager) to the Radiation Protection Manager, the Decommissioning Engineering Manager to the Engineering Manager, the Operations Manager to the Operations / Maintenance Manager, and the Westinghouse Project Director becomes a consultant appointed by the Chairman.

1 I These administrative changes have allowed the DSRC to retain a diverse mix of viewpoints l and perspectives, so that nuclear safety will be adequately considered and maintained during all facets of decommissioning activities. The position changes are addressed as follows:

• The duties of the Decommissioning Program Director and the Deputy Director have I been combined. This position continues to report directly to the Vice President and

! retains the responsibilities and background requirements described in Decommissioning Plan Section 2.4.4. Deleting the Deputy Director from the DSRC does not result in a loss of expertise or authority.

• The duties of the Project Controls Manager dealt mainly with costs, schedules, and contract issues. These duties have been assumed by the Decommissioning Engineering Manager, as described in Decommissioning Plan Section 2.4.9, and

deletion of this position from the DSRC has no affect on nuclear safety.

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• The Facility Support Manager (Radiation Protection Manager) has been redesignated as the Radiation Protection Manager, and this position continues to have overall responsibility for the Radiation Protection Program as described in Section 3.2 of the Decommissioning Plan. This position meets the qualifications of Regulatory Guide 1.8, " Qualification and Training of Personnel for Nuclear Power Plants" for the Radiation Protection Manager, and ensures that radiation protection perspectives are l adequately represented on the DSRC.

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• The-Decommissioning Engineering Manager title is being changed to the more encompassing Engineering Manager title, to acknowledge that this position will likely have technical responsibility for additional projects, such as the ISFSI and certain

! aspects of the repowering project. This position continues to meet the requirements of Decommissioning Plan Section 2.4.9 and ensures that technical perspectives are ,

adequately represented on the DSRC.

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• The Operations Manager title is being changed to Operations / Maintenance Manager, to reflect that this position is responsible for operations and maintenance, as described in Decommissioning Plan Section 2.4.8. This position ensures that both of these - ,

perspectives are adequately represented on the DSRC.

• The Westinghouse Project Director has been deleted from the defined DSRC membership, in anticipation that this position will eventually be relocated away from the FSV site. PSC commits to retain this position on the DSRC as a consultant appointed by the Chairman, as long as the individual in this position is located 'on site, so that the decommissioning contractor's perspective will be adequately i represented on the DSRC.

As described above, PSC considers that the proposed changes to the DSRC membership do ,

not result in a reduction of effectiveness of the committee, and that it continues to have the ,

diversity of technical and administrative expertise needed to monitor decommissioning operations to ensure that they are being performed safely, as described in Decommissioning Plan Section 2.4.10.

l CONCLUSIONS ,

i l Based upon the preceding evaluation, it has been determined that the proposed-l changes to the Decommissioning Technical Specifications at the Fort St. Vrain l Nuclear Station do not involve a significant increase in the probability or  ;

l consequences of an accident or malfunction previously evaluated, create the possibility of a new or different kind of accident or malfunction from any previously evaluated, or involve a significant reduction in a margin of safety. Therefore, it is concluded that the license amendment involves No Significant Hazards Consideration as defined in 10 CFR 50.92 (c).

REFERENCES

1. NRC Letter from Erickson to Crawford, dated November 23,1992;

Subject:

" Order to Authorize Decommissioning of Fort St. Vrain and Amendment No.

85 to Possession Only License No. DPR-34.

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