ML20127F569
| ML20127F569 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/03/1992 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| SECY-92-374, NUDOCS 9211100442 | |
| Download: ML20127F569 (102) | |
Text
MMMMMMMMMMMMMMA
- ,,,,,,......o.e,eeoeoeooo
- RELEASEDTOmE POR .
c6 i
/""%. Io - n /25/*
[' '
g' g
'j dge ird$ls
,ggoooooo000000000000000 q
\ ""' /
'n. 2a POLICY ISSUE November 3, 1992 (NEGATIVE CONSENT) SECY-92-374 f_ot: The Commitsioners from: James M. Taylor Executive Director for Operations Sub.iec t : FORT ST. VRAIN NUCLEAR GENERATING STATION (FSV) -
PUBLIC SERVICE COMPANY OF COLORADO (PSC) - APPROVAL 0F DECOMMISSIONING PLAN AND AMENDMENT OF LICENSE
Purpose:
To inform the Commission of the staff's intent to issue an Order approving the FSV Decommissioning Plan and corresponding amendment to FSV License DPR-34.
Backaround: While operating, Fort St. Vrain Nuclear Generating Station (FSV) was a 842 MW thermal, high temperature gas (helium) cooled reactor (HTGR) that operated from January 31, 1974, to August 18, 1989. The Public Service Company of Colorado (the licensee or PSC) shut down FSV because of control rod drive failures and later made the shutdown permanent upon discovering degradation of the steam generator ring headers.
PSC requested a possession only license on November 21, 1989, and submitted a proposed Decommissioning Plan on -
November 5, 1990. On May 21, 1991, the NRC issued a possession only license to PSC. On August 31, 1991, PSC submitted proposed decommissioning technical specifications (TS).
PSC removed all spent fuel from the reactor protected area and transferred it to an independent spent fuel storage installation (ISFSI) that is separately licensed to PSC under 10 CFR Part 72. Proposed decommissioning of FSV involves promptly dismantling, decontaminating and disposing of radioactively activated and contaminated material and components. PSC will dismantle and remove much of the plant. However, the reactor building, turbine building, CONTACT: Peter Erickson, ONDD/NRR, 504-1101 Seymour Weiss, ONDD/NRR, 504-2170
_ NOTE: TO BE MADE PUBLICLY AVAILABLE cd~103 y waEn Tae rinAt sas 1s pinnE AVAILABLE J
z { ) 10 mq 7 $ [fbk mmmmmmmmmmmmmmmmm
a Commissioners l and other structures that are not radioactive above limits acceptable for unrestricted access will remain.
Discussion: The staff has enclosed (1) an order approving the FSV Decommissioning Plan, (2) an amendment to establish the -l Decommissioning TS, (3) the staff's supporting safety evaluation,-(4) an environmental assessment, and (5) a Notice of Issuance of Environmental Assessment and Finding of No Significant Impact.
The order approves the FSV Decommissioning Plan and adds a condition to apply 10 CFR 50.59 principles to changes in the decommissioning plan by PSC.
The amendment replaces the existing FSV TS with.the Decommissioning TS and allows the licensee to receive FSV byproduct and special nuclear material that may have to be returned to FSV during the offsite shipment or the disposal process (License Condition 2.C.5).
The enclosed safety evaluation (SE) summarizes the proposed dismantling and decontamination (DECON) of FSV. Major tasks covered are (1) decontamination and dismantlement of the prestressed concrete reactor vessel (PCRV),-(2) decontamination of the balance of the-plant, and (3) final radiation survey plan. In the SE, the staff also evaluates -
the PSC cost estimates and funding' plan, radiation exposure estimates, the PSC decommissioning organization, training
-programs, protection of occupational and public health and safety, radioactive waste processing and disposal and waste volume estimates. The staff concludes that there is-reasonable assurance that the health and safety of the public will be adequately protected.
In the Environmental Assessment the staff compares environmental aspects of decommissioning alternatives (SAFSTOR, ENTOMB and DECON) and evaluates in' detail the environmental aspects of the DECON alternative selected by PSC. The major issue evaluated is the tritium that will be released from the graphite reflector blocks to the PCRV shield water. The staff reviewed the tritium dilution and release and the possible radiation exposure to the public.
PSC estimated that data from a British reactor indicates that no more than 500 curies of tritium will be released to the shield water during FSV decommissioning.
Comrissioners Even if tritium release to the shield water exceeds 500 curies, PSC has proposed adequate controls of total tritium release, release rate, dilution and monitoring to ensure that any exposures to the public would remain below the as low as reasonably achievable (ALARA) guidelines of 3.0 millirem per year as specified in Appendix ! to 10 CFR Part
- 50. The State of Colorado has also reviewed PSC's proposed handling of tritium releases and has concluded that the issue is adequately addressed, in the EA, the staff also evaluated exposure to workers, radioactive waste impact, and other environmental issues as outlined in 10 CFR Part 51. The stnff concluded that the proposed FSV decommissioning will not significantly affect the qua'.ity of the human environment and that an environmental impact statement was not required. Therefore, the staff will publish a Notice of Issuance of Environmental Assessment and Finding of No Significant Impact (enclosed) in the FEDERAL REGISTER.
Timina: To satisfy financial assurance requirements of 10 CFR 50.75(e) for the decommissioning of FSV, PSC has obtained a commitment for a $125 million Letter of Credit from the Bank of New York in cooperation with several other banks. The Letter of Credit is subject to NRC approval of the FSV Decommissioning Plan and the commitment from the banks expires on November 25, 1992 unless NRC approval of the plan is granted before that date. PSC has applied to the Bank of New York for a 120 day extension to the commitment for the Letter of Credit and the Bank of New York is consulting with the other banks regarding the extension. If the extension is not granted, another financial assurance mechanism for decommissioning FSV would have to be established by PSC and approved by the NRC. Significant delay in decommissioning would likely occur if the November 25th date passes without either an extension of the commitment for the Letter cf Credit or NRC approval of the FSV Decommissioning Plan. If the extension is granted, PSC indicates that it alone would cost about $90,000 in financial and legal fees.
Coordination: The Office of the General Counsel has no legal objection to the paper.
Commissioners Recommendation: Unless the Commission otherwise directs within 10 days from the date of this paper, the staff will issue the order approving the FSV Decommissioning Plan and the associated amendment that revises the FSV TS and authorizes the temporary return of byproduct and special nuclear material to FSV.
/
R4-~
a es M. TayJbr
.ecutive Director for Operations SECY NOTE: In the absence of instructions to the contrary, SECY will notify the staff on Tuesday, November 17, 1992, that the Commission, by negative consent, assents to the action proposed in this paper.
DISTRIBUTION:
Commissioners OGC OCAA OIG OPA OPP REGION IV EDO ACRS ACNW ASLBP SECY I
i i
1
/p mg'o, UNITED STATES
['
- ~,
g NUCLEAR REGULATORY COMMISSION wAssisoton o, c.20sss g ,p a., .
%, .....f Docket No. 50-267 Mr. A. Clegg Crawford Vice President, Electric Production Public Service Company of Colorado Post Office Box 840 Denver, Colorado 80201-0840
Dear Mr. Crawford:
SUBJECT:
ORDER TO AUTHORIZE DECOMMISSIONING 0F FORT ST. VRAIN AND AMENDMENT pn 85 TO POSSESSION ONLY LICENSE NO. DPR-34 (TAC N0. M82592)
The Commission has issued the enclosed Order to authorize decommissioning of the Fort St. Vrain Nuclear Generating Station and has also issued Amendment No. 85 to Possession Only License No. DPR 34. The order and amendment responds to your application dated November 5, 1990, as revised December 17 and 21, 1990, January 14, 1991, April 15 and 26, 1991, May 15, 1991, June 6 and 17, 1991, July 1 and 10, 1991, August 28 and 30, 1991, November 15, 1991, December 6, 1991, January 9, 1992, March 19 and 20, 1992, April 17 and 30, 1992, June 24, 1992, and September 1, 18 and 25, 1992. The Order has been forwarded to the Office of the Federal Register for publication.
A Notice of Consideration of Issuance of Orders Authorizing Decommissioning and Termination of Possession Only License was published in the FEDERAL P,EGISTER on March 13, 1992 (57 FR 8940). No comments or requests for iiearing were received.
A copy of the related Safety Evaluation and Environmental Assessment supporting Amendment No. 85 is enclosed. Also enclosed is a copy of the Notice of Issuance of Environmental Assessment which was published in the FEDERAL REGISTER on Sincerely, ,
hPeter B. Erickson,
/ kh -z o s Senior Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation
Enclosures:
- 1. Order Authorizing Decommissioning
- 2. Amendment No, 85
- 3. Safety Evaluation
- 4. Environmental Assessment
- 5. Notice of EA cc w/ enclosures: See next page
Mr. A. Clegg Crawford Fort St. Vrain Public Service Company of Colorado Docket No.-50-267 cc:
Mr. David Alberstein, Manager Mr. D. D. Hock Fort St. Vrain Services President and Chief Executive GA International Services Corporation Officer -
P. O. Box 85608 P. O. Box 840 San Diego, California 92138 Denver, Colorado 80201-0840 Senior Resident inspector U.S. Nuclear Regulatory Commission P. O. Box 640 Platteville, Colorado 80651 Kelley, Standfield, & 0'Donnell 1225 17th Street, Suite 2600 Denver, Colorado 80251 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 800631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413 Radiation Control Division (RCD-D0-B1) colorado Department of Health 4300 Cherry Creek Drive South Denver, Colorado 80222-1530 Commitment Control Program Coordinator Public Service Company of Colorado 16805 Weld County Road 19-1/2 Platteville, Colorado 80651-Mr. M. H. Holmes Nuclear licensing Manager Public Service Company of Colorado 16805 Weld County Road 19-1/2 Platteville, Colorado. 80651
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )
)
PUBLIC SERVICE COMPANY OF COLORADO )- Docket No. 50-267-
)
(Fort St. Vrain Nuclear )
Generating Station) )
ORDER APPROVING DECOMMISSIONING PLAN AND AUTHORIZING DECOMMISSIONING OF FACILITY By application dated November 5, 1990,'as revised December 17 and 21, 1990, January 14, 1991, April 15 and 26, 1991, May 15, 1991, June 6 and 17, 1991, July 1, 1991, August 28 and 30, 1991, November 15, 1991, December 6, 1991, January 9, 1992, March 19, 1992, April 17, 1992, and September 25, 1992, Public Service Company of Colorado (PSC) requested approval of its proposed Decommissioning Plan for the Fort St. Vrain Nuclear Generating Station (FSV).
In support of the Decommissioning Plan, PSC submitted an Environmental Report (ER) Supplement on July 10, 1991, as revised March 20, 1992, April 30, 1992, June 24, 1992, and September 1 and 18, 1992. A Notice.of Consideration of Issuance of Amendment and Opportunity for Hearing was published in the FEDERAL REGISTER on March 13, 1992'(57 FR 8940). No request for hearing or petition for leave to intervene was filed following notice of the proposed action.
The U.S. Nuclear Regulatory Commission (the Commission) has reviewed the application with respect to the provisions of the Commission's rules and regulations and has found that decommissioning as stated in the licensee's -
Decommissioning Plan will be consistent with the regulations in 10 CFR Chapter I, and will not be inimical to the common defense and security or to
the health and safety of the public. The basis for these findings is set forth in the concurrently issued Safety Evaluation by the Office of Nuclear Reactor Regulation.
l The Decommissioning Plan replaces the licensee's Updated Safety Analysis '
Report. Accordingly, a license condition has been added allowing the licensee i
to make changes to the Decommissioning Plan after performing a review based upon criteria similar to the criteria of 10 CFR 50.59 to ensure that such changes do not involve an unreviewed safety question.
The Commission has prepared an Environmental Assessment and Finding of No Significant Impact for the proposed action. Based on that Assessment, the Commission has determined that the proposed action will not result in any significant environmental impact and that an environmental impact statement need not be prepared. The Notice of Issuance of Environmental Assessment was published in the FEDERAL REGISTER on Accordingly, pursuant to Section 103, 161b, 1611, and 161o of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.82, the licensee's Decommis-sioning Plan dated November 5, 1990 as revised, is approved and decommissioning of the Fort St. Vrain facility is authorized in accordance with the Decommissioning Plan and the Commission's rules and regulations, subject to the following conditions:
(a)(1) The approved Decommissioning Plan replaces the Safety Analysis Report in its entirety and the licensee may (i) make changes in the facility or procedures as described in the Decommissioning Plan and (ii) conduct tests or experiments not described in the Decommissioning Plan, without prior Commission approval, unless the proposed changes, tests or experiments involve a change in the Technical Specifications (TS) incorporated in the license or an unreviewed safety question.
(2) A proposed change, test or experiment shall be deemed to-involve an unreviewed safety question (1)-if the probability of occurrer.ce or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Decommissioning Plan may be increased or (ii) if a possibility for an accident or malfunction of a different type _than evaluated previously in the Decommissioning Plan may be created; or (iii) if the margin of safety as defined in the basis for any TS is reduced.
(b)(1) The licensee shall maintain records of changes in the fac lity and of changes in procedures made pursuant to-this section, to the extent that these changes constitute changes in the facility or proceduros as described in the Decorraissioning Plan. The licensee shall also maintain records of tests and experiments carried out pursuant to paragraph (a) of this section. These records must include a written safety evaluation which provides the basis for the determination that the rhanges, tests or experiments do not involve an unreviewed safety question.
(2) The licensee shall submit, as specified in 10 CFR 50.4, a report containing a brief description of any changes, tests and- experiments, including a summary.of the safety evaluation of each. The report must be submitted querterly.
(3) The records of. changes in the facility shall be maintained untii the date of termination of the license and records of changes in procedures and records of tests-and experiments shall be maintained for a period of.three years.
(c) 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.
For further details with respect to this action, see: (1) the licensee's application for authorization to decommission the facility, dated November 5,1990, as revised; (2) the licensee's Environmental Report Supplement dated July 10, 1991, as revised; (3) Amendment No. 85 to License No. DPR-34; (4) the Commission's Safety Evaluation; and (5) the Commission's i Environmental Assessment and Finding of No Significant Impact. These
_4_
doc'1ments are available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street N.W., Washington, D.C.
20555, and at the Greeley Public Library, City Complex Building, Greeley, Colorado 80631.
Dated at Rockville, Maryland this FOR THE NUCLEAR REGULATORY COMMISSION 7WE Thomas E. Murley, Director Office of Nuclear Reactor Regulation
~ __ .
je # " ** !
UNITED STATFS e- r . r ';,g NUCLEAR REGULATORY COMMISSION WASWNGTON D. C,20555 ,
PUBLIC SERVICE r,0MPANY OF COLORADQ FORT ST. VRAIN NUCLEA'l GENERATING STATION DOCKET NO. 50-262 AMENOMENT TO POSSESSION ONLY LICENSE Amendment No. 85 License No. DPR-34
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to Possession Only License No.
DPR-34 by Public Service Company of Colorado (the licensee), dated November 5, 1990, as revised December 17 and 21, 1990, January 14, 1991, April 15 and 26,1991, May 15,1991, June 6 and 17,1991, July 1 and 10,1991, August 28 and 30, 1991, November 15, 1991, December 6, 1991, January 9, 1992, March 19 and 20, 1992, April 17 and 30, 1992, June 24, 1992, and September 1, 18 and 25, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will be maintained in conformity with the application, the-provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that-the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuanta of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's_ regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attechment to this license amendment, and paragraphs 2.C.(5) and 2.0.(2) of Fossession Only License No. DPR-34 is hereby amended to read as follows:
2.C.(5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive and possess, but not separate, byproduct and special nuclear materials that were produced by the previous operation of the facility or during decommissioning activities.
2.D.(2) Technical Snecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 85, are hereby incorporated in the license. The licensee shall maintain the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
Technical Specifications must be implemented within 14 days of the date of license amendment issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Seymour H. Weiss, Director Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation
Attachment:
Appendix A Technical Specifications Changes Date of Issuance:
~ _ - _ -~ .- . ._- _-__ _. _ _ _ __ _ _ . . .
. . _ _ _ ._ .. ._._ .- - . . . . . ~ . . . . _ _ . . . _ . . _ . . . _ . - - . . . _ . _ . . . . . - ~ . .
~
1<
ATTACHMENT TO LICENSE AMENDMENT NO. POSSESS 10N'ONLY LICENSE NO. DPR-34 DOCKET NO. 50-267 I
Replace all of the pages of the Appendix A Technical Specifications with 'th'e enclosed pages.
k k
,e ,, , , . , - - . . _ , , - , . . , , , . , - - - , - , , ~ ~ , , -
DECOMMISSIONING TECHNICAL SPECIFICATIONS For FORT ST. VRAIN Vnit No. 1 Docket No. 50-267 Appendix A to Facilit) License No. OPR-34
__m__ _2- . _ _ _ _ . _ _ . _
t FORT ST. VRAIN DECOMMISSIONING TECHNICAL 1 SPECIFICATIONS ,
TABLE OF CONTENTS Page Number
1.0 INTRODUCTION
1 2.0 DEFINITIONS 2-1 3.0 GENERAL REQUIREMENTS 3.0-1 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 Watt Tritium Concentration 3.4-1 4.0 DESIGN FEATURES 4.0-1 5.0 ADMINISTRATIVE CONTROLS 5.0-1 5.1 Responsibility. 5.0-1 5.2 Organization 5.0 '
5.3 Decommissioning Safety Review Committee 5.0-1 5.4 Procedures and Programs 5,0-4 5,5 Reporting Requirements 5.0-8 5.6 Record Retention- 5.0-9 5.7 Radiation Protection Program 5.0-10 5.8 High Radiation Area 5.0-11 5.9 Process Control Program _ 5.0-12 5.10 Offsite. Dose Calculation and Radiological Environmental Monitoring Program Manuals- 5.0 5.11 Natural Gas Restriction 5.0-13 l
. . ~ - -- ,4 , , . , ...-#
1 i
Fort St. Vrain l DTS i' Amendment No. 85 Page 1-1
1.0 INTRODUCTION
These Decommissioning Technical Specificationi are applicable during the decommissioning of the Fort St. Vrain (FSV) reactor.
Decommissioning is considered to begin af ter all of the nuclear fuel has been removed frem the FSV Reactor Building and after the NRC has approved the Decommissioning Plan.
The Fort St. Vrain Nuclear Generating Station originally ,
operated as a High Temperature Gas-Cooled Reactor, which supplied steam to a turbine generator. The facility may be converted to utilize a gas-fi.ed boilcr. 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 Decommissioling Technical Specifications.
Activities that will 5e undertaKec in accordance with these ,
Decommissioning Technical Snceirications include the dismantlement and decommissioning (DECON) of the radiologically-activated and contaminated portions of th? facility to release all site areas for unrestricted use. ,
There are two categories of FSV Technical Specifications:
- " Decommissioning Technical Specifications-(DTS)" include-Amendment 85 and all subsequent amendments.
+ " Operating Technical Specifications" refers to the historical -Technical Specifications included in all previous amendments.
1 4
-, -. . __ _, _ . , . _ , , . _ _ . . _ ~ . _ - . , _ _ - . , _ _ . . _ _ _ . _ . _ _ _ . , - ~ , _ . . , -. ~ - _ - . .-- . . . _ . , _
Fort St. Vrain DTS
- Amendment No. 85 Pago 2-1 2.0 DEFINITIONS The defined terms in this section appear in capitalized type and are applicable throughout these Technical Specifications.
2.1 ACTIONS ACTIONS shall be that part of a specification which prescribes Required Actions under designated conditions, which shall be completed within specified Completion Times.
2.2 ACTIVATED GRAPHITE BLOCKS ACTIVATED GRAPHITE BLOCKt shall include the reflector blocks and spacer blocks. Other graphite items, such as defueling elements, core support blocks, and core support posts, are not considered ACTIVATED GRAPHITE BLOCKS.
2.3 BASES The BASES shall summarize the reasons for the Limiting Conditions, Applicabilities, ACTIONS, and Surveillance Requirements. In accordance with 10 CFR S0.36, the BASES are not considered part of the Decommissioning Technical Specifications.
2.4 CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and with the required accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel, considering system design, including the sensors and alarm, interlock and/or trip functions, and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
2.5 CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during its operation by observation.
This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
a Fert St. Vrain DTS Amendment No. 85 - -
Page 2-2 DEFINITIONS (Continued)__ _
2.6 CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable, considering system design, to verify OPERABILITY including alarm, interlock, and/or trip functions.
2.7 EXCLUSION AREA BOUNDARY The EXCLUSION AREA BOUNDARY shall enclose the decommissioning Emergency Planning Zone (EPZ), as snown 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) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with decommissioning the plant. Individuals who are occupationally associated with the conversion of the plant, and persons who enter the site to service equipment or make deliveries, are included in this category. MEMBER (S) 0F THE PJBLIC also includes persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
2.9 0FFSITE DOSE CALCULATION MANUAL (ODCM)
The OFFSITE DOSE CALCULATION MANUAL (02CM) .shall contain the methodology and parameters used in the calculation of offsite doses resulting frem 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 Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiole11 cal Environmental Monitoring Programs required by ...ecification 5.4.4 and (2) descriptions of the information that should be included- in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 5.5.1 and 5.5.2.
y=,y9- vi y9-e -, g- ,y y ,y y ge1, ,-,e, yyy .w ep.
1 Fort St.'Vrain DTS Amendment No.-85 Page 2 DEFINITIONS (Continued) _
2.10 OPERABLE - OPERABILITY A component or system shall be OPERABLE or have 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 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 others requirements governing the disposal of solid radioactive:
waste.
2.12 UNRESTRICTED AREA An UfrESTRICTED 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.
I
Fort St. Vrain DTS Amendment ~No. 85-Page 3.0-1 i
3.0 GENERAL REQUIREMENTS 1 ___ ___ __
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 Required Actions shall be met within the specified Completion T'.mes.
3.0.2 Noncompliance with a specificat on i 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 i the specified Completion Time, the Required Actions need not be completed.
3.0.3 Surveillance Requirements shall be met as specified in the Applicability for individual LCs unless otherwise stated -in an individual Surveillance Requirement.
Failure to meet a Surveillance Require. ment, except as provided in 3.0.4, shall constitute failure to meet the LC. Surveillance Requirements do not have to be performed-on inoperable equipment.
3.0.4 Each Surveillance Requirement, any Required Actions 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.
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-
-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 allowance. and the Surveillance Requirements are not
-met, the Completion Times of the ACTIONS are appitcable at that time.
w- -at , ,p - ~
v -
-g-vy, y w- +
._ _ . _ __ _ _ ._~ ._ _ . _ _ _ _ _
Fort:St. Vrain DTS Amendment No. 85 '
Page'3.0-2 3.0 BASES 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 the requirements consistent with operating plants' Limiting Conditions for Operation per the Code of Federal Regulations, 10 CFR 50.36 (c)(2). 3.0.1 establishes the Applicability statement within individual specificatiors as the requirement for when conformance to the .C is required for safe decommissioning of the unit. The Required Actions establish those reme al measures that must be taken within spet1fied Completion Times when requirements of a_LC are not met. l 3.0.2 establishes that- noncompliance with a specification exists when the requirements of the-LC are not met _and the associated Required Actions have not been met within the specified Completion Times. The purpose of this general requirement is to clarify that: (1) completion of the Required Actions within the specified Completion Times constitutes compliance with a specification, and I (2) completion of the remedial measures of the j Required Actions is not required when compliance with an LC is restored within the Completion Time specified in the associated ACTIONS, unless otherwise specified.
3.0.3 - 3.0.5 3.0.3, 3.0.4, and 3.0.5 establish the general i requirements applicable Surveillance to-Requirements. These requirements are based on the requirements consistent with operating- plants' Surveillance Requirements stated in the Code of Federal Regulations, 10 CFR 50.36 (c)(3).
3.0.3 establishes _the requirements that Surveillance Requirements must be met during the conditions specified in the Applicability.for which- ;
the requirements of the LC apply unless otherwise stated in an individual Surveillarice Requirement.
The purpov' of this- general- requirement is to ensure that Surveillances are performed to verify the status of systems and _ components' and that parameters are within specified- limits.
Surveillance. Requirements do not have , to be performed when outside of the Applicability of-the LC unless otherwise specified.
E o
Fort St. Vrain DTS Amendment No. 85 Page 3.0-3 3.0 BASES _(Continued) 3.0.4 establishes the conditions under which tha 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 Surveillence 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 confe mance 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 Surveillaf.ce within the allowed Surveillance Frequency, defined by the provisions of 3.0.4, is a condition that constitutes a failure to meet the OPERABILITY requirements for an LC. Under the provisions-of this general requirement, systems and components are assumed to be OPERABLE when the associated Surveillance Requirements have not- been met. However, nothing in this provision is to be construed as implying that systems or omponents are OPERABLE -when they _are found or known to be inoperable although still meeting the -Surveillance Requirement frequency. This general requirement.
also clarifies that the ACTIONS are applicable when Surveillances have not been completed within_the allowed Surveillance Frequency- and that the Completion Times of the Required Actions apply from the point in time it is identified that a Surveillance has not been performed and not at the time-that'the allowed Surveillance Frequency was' exceeded.
Fort St. Vrain DTS Amendment No. 05 -
Page 3,0-4 i 3.0 BASES (Continued)
If the Completion Times of the ACTIONS are less 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 performed. If a Surveillance is not completed within tho 24-hour allowance, tF) Completion Times of the AC IONS are applicable at that time.
For the purpose of making _the transition from the operating Technical Specifications to the Decommissioning Technical Specifications, surveillances performed under the operating Technical Specifications may be utilized to satisfy the applicable surveillance requirements of the Decommissioning Tecnnical ;pecifications.
Fort .. Vrain 4 DTS Amendment No. 85 Page 3,1-1
- 3,1 REACTOR BUILDING CONFINEMENT INTEGRITY. _ __
LC 3.1 Reactor Building confiner it integrity shall be maintained with:
i a, The Reactor Building overpressure protection system louvers closed *, and i
- b. Either:
- 1. The outer truck bay closures closed, or ;
, 2. The inner truck bay closures closed. I APPLICABILITY: Whenever ACTIVATED GRAPHITE BLOCKS have been removed ;
from the PCRV and remain inside the Reactor ;
Building
- 2 ACTION ,
I I I I l CONDITION l REQUIRED ACTION l COMPLETION TIME l l 1 l l l l 1 l lA. Do not have lA.1 Suspend activities I hour l 1 l Reactor Building l involving physical l l confinement l hand.ing of l l i l integrity l ACTIVATED GRAPHITE l l i l l BLOCKS within the l l l l Re,.: tor Building l l l l . .I l
'he ' Reactor Building overpressure protection system-louvers may-be open provided there are no activities i in- progress involving the physical handling of any ACTIVATED GRAPHITE BLOCKS.
.e_.- ..-_____ .. ,_,.-..,,-,-.,_.,,_.g,__,__ . . _ , , , . , , , , . , , , . . _ . .._,.,p,._ . .___,,__y . , , , _ . , , - . . .
r Fort St. Vrain DTS i Amendment No. 85 -
Page 3.1-2 ,
SURVEILLANCE REQUIREMENTS _ ,_ ,,_ _ .__- . mm._ .
w== m,--...- . _ .. . .. -.
l SURVE!LLANCE l FREQUENCY l l -. _ l I I I l l 54 3.1.1 Verify all Reactor Building l Daily, when physica1l overpressure protection system handling of
[ l i l louvers are in the closed l ACTIVATED GRAPHITE l position, except as permitted l BLOCKS is in j by LC 3.1. l progress l-l l l l l l l SR 3.1.2 Verify inner truck bay closures l Prior to opening l l are closed l outer truck bay l l l closures !
I u==c _ .
.-l _. .l 1
+
Fort St. Vrain DTS i Amendment No. 85 Page 3.1-3 ;
a 3.1 BASE L ,___ m _
BACKGROUND The integrity of the Reactor Building, in 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 i of a major release of activity from the Prestressed
> Concrete Reactor Vessel (PCRV) dismantlement (i.e., i 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 buy includes two redundant sets of closures. The outer closures have historically included a truck door and the personnel access door 1 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 t uck doors, in an airlock-type configuration.
The Reactor Guilding shall be- maintained subatmospheric at all times-including normal access (see LC 3.2). Subatmospheric conditions can be maintained with several louver banks open. The overpressure protection system louvers may bi 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 at- alosed to ensure integrity of the- Reactor- Building confinement prior to the opening of the outer truck 3 doors to the truck bay. i 1
l l
- - - - - -_ . - ..-.- - . - . = _ = . . - -, . - . -
+ i Fort St. Vrain ;
DTS i Amendment No. 85 -
Page 3.1-4 6
I 3.1 PASES (Continued).,, _ , ._
Reactor building confinement integrity is taken credit for in the Heavy Load Drop and the Loss of AC Power accident analys => described in Section 3.4 of the Decommission ; Plan. (Reference 1)
LC The LC establishes the minimum conditions required ,
to ensure that Reactor Building confinement integrity is maintained during applicable accident ,
scenarios (i.e., Heavy Load Drop and/or Loss of AC Power). The LC requirements are consistent with the accident analysis assumptions, and the criteria used during plant operation. It should be noted that the Reactor Building overpressure protection system louvers may be open provided tnere are no activities in progress involving the physical handling of any ACTIVATED GRAPHITE BLOCKS. For example, the louvers may be open whilt. ACTIVATED 4 GRAPHITE BLOCKS are being dried or are in temporary '
storage within the Reacto* Building, as long as they are not being moved, cut, or otherwise physically handled.
APPLICABILITY The Reactor Building confinement integrity applicability is based on complying with the off-site dose requirements established in the 10.CFR 100 guidelt t s and the EPA Protective Action-Guidelines 'e the event of a Heavy Load Drop accident and/or _uss of AC Power.- However, the Reactor Building overpressure protection system.
louvers may be open provided there are no activities in progress involving the physical handling of any ACTIVATED GRAPHITE BLOCKS.
Consistent with the Accident Analyses, ACTIVATED .'
GRAPHITE BLOCKS include reflector blocks and spacer blocks, as defined in Specification 2.2. The activation level of other graphite materials is significantly less than the reflector blocks and spacer blocks. In the event of a load drop accident involving other graphite materials, the resultant doses are low enough that confinement integrity or ventilation are not required.
4 I
-.._-.--.y._c.... m..y , ,,[..y, c., -
.,_.,___.-.l,.,
t Fort St. Vrain ,
OTS Amendment No. 85 Page 3.1-5 r 3.1 BASES (Continued) ._,,_____.. .
ACTIONS Aj When Reactor Building confinement integrity is ;
breached, s* spend activities involving physical ,
handling of ACTIVATED GRAPHITE BLOCKS within the 1 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.
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 contemporaneous 1y 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.
REFERENCES 1. FSV Decommissioning P.lan ;
- 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
-t i
t a y .,--,r _-m-.
, .,--,a.,,., n - . . em., ,....,.,;,.g. .,n.,w r. , ,,y ,.m,,.,w .e,,,,,., .,..r,...pe .w -,m,,-, , n n.,-. - - -. . , .
Fort St. Vrain DTS Amendment No. BS Page 3.2-1 3.2 REACTOR BUILDING VENTILATION EXHAUST SYSTEM LC 3.2 The Reactor Building ventilation exhaust system shall be OPERABLE with:
- a. Reactor Building internal pressure subatmospheric, and
- b. At least one of the th ee ventilation exhaust trains OPERABLE. with each train consisting of one exhaust fan (C-7301, C-7302, or C-7302S) and the HEPA filter section of the associated filter assembly ( F-7301, F-
'302, or F-73025).
APPLICAB, CITY: Whenever ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV and remain inside the Reactor Du11 ding ACTIONS
~
l l l l l CONDITION l 3EQUIRED ACTION l COMPLETION TIME l l 1 l l l l 1 l lA. Reactor Building lA.1 Suspend activities l 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> l l pressure is l involving physical l l l atmospheric or l handling of l l l greater l ACTIVATED GRAPHITE l l l l ELOCKS within the l l l l Reactor Building l l l l __ l ___I l l l l lB. .All e_xhaust lB.1 Restore at least l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l l trains l one ventilation l l l inoperable i exhaust train to l l l l OPERABLE status l l l l l l l l l l lC. Requitard Action 10.1 Suspend activitiet l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l l B.1 not met l involving physical l l.
l within l handling of l l l Completion Time l ACTIVATED GRAPHITE l l l l BLOCKS within the l l-l l Reactor Building l l' l l 1 I
Fort St. Vrain DT S Amendment No. 85 Page 3.2-2 SURVEIL _ LANCE REQ!J1REMENTS_ = = ._ _ _ = = = m _ .___ ,
. _ _ -. . _ _ _ _ _ _.. _ _ _ _ _ . _ ___ _ .__..._ _._ _ ..___ _-_ _ _..-._ _ _____._._r=2 l SURVEILLANCE i FREQUENCY l l l l 1 1 I l SR 3.2.1 Verify Reactor Building pressure l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l l is subatmospheric l l l_ l l I I l l SR 3.2.2 Verify pressure drop across each l Weekly l l HEPA filter is less than 6 inches l l l of water, with a flow rate of at l l l least 17,100 cfm l l l- l 1 I I I l SR 3.2.3 Verify HEPA filter bank satisfies l 18 months, after l l in place penetration and bypass I structural l I leakage test acceptance criteria i maintenance on the l I of less than 1 percent, using i HEPA filter hcusitj,1 l test procedure guidance in I or after each I l Regulatory Positions C.S.a and l complete or partial l l C.S.c of Regulatory Guide 1.52, I replacement of a l l Rev. 2, March 1978, with a flow l HEPA filter bank l l rate of at least 17,100 cfm I l I _ ___ _ _ _ . . _ _ _ - _- _.l -_ . _l
Fort St. Vrain 075 Amendment No. 85 Page 3.2-3 3.2 BASES BACKGROUND The Reactor Building ventilation exhaust filter system is designed to filter the Reactor Building attrosphere prior to release to the vent stack during both normal and most accident conditions during decommissioning.
The system consists of.three trains, one of which is normally in continuous operation. The design flow rate for each train is 19,000 cfm. Allowing 10*4 for degradation, the minimum flow rate is 17,100 cfm. One train is sufficient to maintain the Reactor Building subatmospheric and thereby minimize unfiltered fission product release from the building. With only one exhaust fan operating, the ventilation system controls will throttle fresh air supply to the air handler in order to reduce the pressure.
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 ventilation system was designed to maintain a subatmospheric condition approximately 1/4 inch water gauge negative. In actual practice, the Reactor Building pressure is normally 0.15 to 0.20 inches water gauge negative, depending on building activities and ventilation system configuration.
There is an alarm at approximately 0.08 inches water gauge negative, and the outside air supply will fully close if the building pressure-increases to atmospheric.
The Reaccor Building ventilation exhaust system is s
" taken credit for in the-Heavy Load Drop accident analysis, as described in Section 3.4 of the Decommissioning Plan (Reference 1). ,
LCs The -LC estabi'shes the minimum conditions-.. required to ensure the Reactor Building ventilation exhaust system is maintained while the potontial exists fer a drop of an ACTIVATED GRAPHITE BLOCK. One train is sufficient to maintain the Reactor Building subatmospheric and thereby minimize unfiltered fission' product release from the building.
HEPA filters provide the required part 1 :ulate
. filtration.
o
, , ~ , , , , , , , , .
Fort St. Vrain D3 Amendment tio. 85 Page 3.2-4 3.2 BASES (Continued) _
APPLICADILITY The Reactor Building ventilation exhaust system will remain OPERABLE, providing filtration of effluents to the environment, while the potential exists for dropping an ACTIVATED GRAPHITE BLOCK.
ACTIONS M When the Reactor Building pressure is atmospheric of greater, suspend activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building. The one hour completion time to suspend activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building minim 17es the time exposure of the Reactor Building to atmospheric or greater conditions and is a conservative time frame. The suspension of physical handling activities is acceptable because all analyzed accidents assume something active is happening - no passive postulated accidents will result in radiological conditions where the need for ventilation anu confinement exists.
.Bd The ability of the Reactor Building ventilation exhaust system to perform its filtering function during a Heavy Load ( ACTIVATED GRAPHITE BLOCK) Drop is dependent on at least one exhaust train being OPERABLE. With all the exhaust trains inoperable, restore at least one ventilation exhaust train to OPERABLE status. A ventilation train may be operating but not OPERABLE, e.g., in the event a required Surveillance is not ccmpleted 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.
CJ W.,s E quired Action B.1 cannot be completed within the required Completion Time, all activities involving physical handling of ACTIVATED GRAPHITE BLOCKS within the Reactor Building are suspended.
Twelve hours is reasonable to suspend ti.ndling 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 existi.
Fort St. Vrain DTS Amendment No. 85 Page 3.2-5 3.2 BASES (Centinued) _
SURVEILLANCE REQUIREMENTS SR 3.2.1 Verifica': ion 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 frequency is more frequent than the operating technical specification requirements.
SR 3.2.2 ~
A pressure drop across the HEPA filter of less than 6 inches of water gauge at 90% of the filter design -
flow rate will indicate that the filters are not clogged by excessive amounts of foreign matter.
SR 3.2.3 Bypass leakage and penetration for High Efficiency-Particulate Air (HEPA) filters are - determined by dioctyi phthalate (DOP) te s t.i n g . The filter penetration and bypass acceptance limits -in the surveillances are applicable based on a HEPA filter-efficiency of 95%. The surveillance frequencies specified establish system performance capabilities.
Verification of the HEPA filter functions ensures system performance capabilities. The surveillance frequency is the same as the operating technical specifications.
REFERENCES 1. FSV Decommissioning Plan
- 2. Manual of Protective Action Guides and Protective Actions for Nuclear Incidents EPA-520/1-75-001-A, January 1990, U.S.
Environmental Protection Agency J
. . . . - . . . - - - . _ - . - - - - . - - . - . _ . . - - . - . - - - ~ . .
Fort St. Vrain i DTS i Amendment No. 85 Page 3.3-1 3.3 RADI ATION MONITORING INSTRUMENTATIOL f I
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 are in effect. 2 APPLICABILITY: At all times, until all significantly contaminated
- or activated items that could exceed alarm setpoints have been removed from the Reactor Building.
ACTIONS l 1 l- l l CONDITION l REQUIRED ACTION l COMPLETION TIME l 1 l l l l '
I I -
l l lA. One or more lA.1 Adjust alarm l 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l l radiation monitor l setpoint within l l l channel alarm l limit l l l setpoint exceeds l l l l value in Table l OR l l l 3.3-1 l l l l lA.2 Declare the l l l l channel inoperable l l l l l l i I I l lB.1 Place a portable
- 18. One or more l 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> l l radiation monitor l monitor (with l l l channels l alarm) in the area l l -r
.l_ inoperable l l l I; l l I t SURVEILLANCE REQUIREMENTS 2 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 Page 3.3-2 TABLE 3.3-1 RADIATION MONITORING INSTRUMENIATION INSTRUMENT ALARM SETPOINTS 4
DURING ACTIVITIES DURING ACTIVITIES NOT CONTROLLED CONTROLLED BY AN RWP BY AN RWP
- a. Refueling Floor < 15 mR/hr < 100 mR/hr* l
- b. Truck Bay < 15 mR/hr < 100 mR/hr*
-- r
. m Monitors may be reset to alarm at a radiation level within a factor of ? of the expected radiation level.
I-1ABLE 3.3-2 SURVEILLANCEREQUJREM'lNTS CHANNEL CHANNEL FUNCTIONAL CHANNEL INSTRUMENT CHECK TEST CALIBRATION
- a. Refueling Floor Daily Monthly 6 months .
- b. Truck Bay. Daily Monthly 6 months-I f .y
-. ,.,,w-- _.-.,y.es 9 - - ,- ., , , .,,..y+,--,,_y ,._-,#._--.c.- . . . . . . . - , #
Fort St. Vrain DTS Amendment !!o, 85 Page 3.3-3 3.3 BASES _
BACKGROUND The radiation monitoring instrumentation required by this specification at all times during decommissioning activities, until all significantly contaminated or activated items that could exceed alarm setpoints have been removed from the Reactor Building, includes two area radiation monitors, one on the refueling floor and one in the truck bay of the Reactor Building. These monitors serve as accident monitors to detect unplanned radiation levels in the Reactor Building, that should be investigated and appropriately resolved.
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 radiation levels will normally be anticipated and planned for, with monitoring provided as required.
Individual work activities will be performed under R6uiation Work Permits (RWPs), which will include nonitoring . provisions. Also, gaseous effluent releases will be monitored and controlled by the 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.
LC The LC establishes the minimum conditions required to ensure the radiat'on 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 radiation levels.
l
Fort St. Vrain DTS Amendment No. 85 Page 3.3-4 3.3 BASES (Continued)._ _ ___ .
APPLICABILITY This LC is applicable at all times.
ACTIONS A.1 or A.2 When one or more radiation menitor channel alarm / trip setpoint exceeds the valuet in Table 3.3-1 either adjust the alarm / trip setpoint within The its limits or declare the channel inoperable.
Required Action and Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is consistent and comparable with Standard Technical Specifications.
B.d When one or more radiation monitor channels is inoperable, place a portable monitor with an alarm in the area. The OPERABILITY of the radiation monitoring channels ensures that the radiation levels are measured in the areas served by the individual channels and an alarm is initiated when the radiation level trip setpoint is exceeded. A6 hour Completion Time is reasonable to complete the Required Action.
SURVEILLANCE REQUIREMENTS SR 3.3.1 -
The surveillance requirements frequencies specified for CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CAllBRATION conform to industry practice and the surveillance frequencies given in Standard Technical Specifications and are adequate to ensure the proper operation of these detectors.
REFERENCES 1. FSV Decommissioning Plan
- 2. Offsite Dose Calculation Manual Program
G Fort St. Vrain DTS Amendment No. 85 Page 3.4-1 3.4 PCRV SHIELDING WATER TRITIUM CONCENTRATION 1
l LC 3.4 Tritium concentration in PCRV shielding water shall n ;
exceed 62.4 pCi/cc. j i
APPLICABILITY: Whenever there is shielding water within the PCRV.
f ACTIONS l l l l l CONDITION i REQUIRED ACTION l COMPLETION TIME l l l- l l l l l l I lA, PCRV shielding lA.1 Reduce tritium l 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l water tritium concentration to I l l l l l concentration l 5 62.4 pCi/cc l l l > 62.4 pC1/cc l l l l l OR l l l l l l )
l lA.2 Perform engineering l l l l l evaluation to l l l l Verify total l l l l tritium content l l ,
l l s 1 E+5 Ci l l l 1 l l l 1 -
-1 - l-lB. Required Actions lB.1 Prepare and submit l The next -l l not met within l to the NRC a -l - 30 days l l required l Special Report l l Completion Time l I describing the l l- - -
4 l l safety concerns l l ',
l l and'the-plans _for l l l l' restoring tritium- l _l l l concentration to 1 -l l l- within the limit l l l l_ l -l b
l I
- , , - r-wr--vm. +-v- :p, iv ,gw,--- ,- --,,y-,- ,--~--<,-m-=*-9-9.*qww , p 9 y- - , - -r,e vev-+ -v w- . is ,...h,-weym-<.-w,- -w we. ..e rtr'c-'+ =%ew --+r- -e6-1e9-e-6-s+-*we ?
Fort St. Vrain i DTS Amendment No. 85 -
l Page 3.4-2 ,
j
. SURVEILLANCE REQUIREMENTS l SURVEILLANCE l FREQUENCY l I I I ;
I I I l SR 3.4.1 Verify PCRV shielding water l Daily, during l l tritium concentration within l initial filling l l limits I of the PCRV with. l l l shielding water. l ;
l l until_ tritium l l l l concentration is a i l l 1e", than 0.1 pC1/cc l l for three l l consecutive samples.1 '
l- l I I l SR 3.4.2 Verify PCRV shielding water l Weekly, after ,
l tritium concentration within l tritium l l limits l concentration is l l less than 0.1 l t I ; pCi/ce, l ,
l l until tritium l l l concentration is l ;
I l 1ess than -l l l 0.01 pC1/cc l l l for three -l l l consecutive samples.l ,
I- I I I -
_ _1 I l SR 3.4.3 Verify PCRV shielding water l Monthly, after l l tritium concentration within l tritium l_
l limits- l- concentration is
-l l 1ess than 0.01
- l l _-p Ci/cc l:
l- _. --
I l l _
w.e- e w~rw -,m r y--w-
- w r va- e ww +,ww. vew p + , e., .. w w w , ,v, i
Fort St. Vrain DTS
+
Amendment No. 85 Page 3.4-3 3.4 BASES - ,_ _ __, ., _
BACKGROUND Ouring Decommissioning of Fort $t. Vrain, the Prestressed Concrete Reactor Vessel (PCRV) cavity will be flooded with water to facilitate the removal of the reactor core components. PCRV dismantlement activities will begin only after all spent fuel has been removed from the reactor building. The water will be circulated, and purified by the PCRV water circulation system to gradually decrease the radioactivity, except tritium, in the water. Thus, the-flooding of the PCRV will provide shielding for the workers associated with PCRV dismantlement activities.
There are a number of systems associated with the flooding of the PCRV to control radioactive material. Their functions include-filtration of the PCRV water inventory, partial demineralization for controlling dissolved solids, and " Feed and Bleed" for adding clean makeup water and -for removing contaminated (primarily tritium)' water.
The initial fluctuating increase in the tritium concentration during the flooding of the PCRV will be controlled by the " Feed and Bleed" dilution process. In accordance with the ODCM, released tritiated water will normally be treated as normal liquid radwaste, diluted- and released at a controlled rate.
A maximum PCRV shielding water tritium concentration is assumed in the Loss of PCRV Shielding Water accident analysis, as described-in Section 3,4 of the Decommissioning Plan (Reference 1).
For this analysis, it is conservatively assumed that the theoretical maximum amount of tritium is transferred to graphite blocks,-the PCRV which shielding water is approximately _1from the E+5-Curies. The tritium concentration in the spilled water is calculated to he 62.4 pC1/cc.
Fort St. Vrain DTS Amendment No. 85 Page 3.4-4 1
3.4 BASES _(S ntinued)m = . _ =.,m.m=m _ _,_ m m_
LC The LC establishes the maximum concentration toler 31e in the PCRV shielding water to ensure adequate protection to the MEMBERS OF THE PUBLIC.
The LC requirements are consistent with the accident analysis ass'_mptions. It should be noted that the accident analysis assumed 1 E45 Curies released. The resulting tritium concentration of 62.4 pCi/cc was chosen as the LC requirement because it is easier to determine a tritium concentration for surveillance monitoring purposes.
APPLICABILITY This LC is applicable whenever there is shielding water within the PCRV, until all ACTIVATED GRAPHITE BLOCKS have been removed from the PCRV, after which there is no credible source of additional tritium.
ACTIONS A.1 or A.2 When the PCRV shielding water tritium concentration is greater than 62.4 pCi/cc it is prudent to either reduce the concentration to less than or equal to 62.4 pCi/cc cr 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 me to change the concentration of large water volumes ano to perform associated analyses.
Ed When a Rec 31 red Action cannot be completed within the required Completion Time, a Special Report must be prepared and submitted to the NRC describing the safety concerns and the plans for restoring tritium concentration to within its safety analysis limit.
The preparation and submittal of a Special Report is an acceptable action because the 1 E+5 Curie analysis value results in doses far below the limits allowed by Reference 2. The Special Report will be prepared as described in Specification 5.5.4.
b Fort St. Vrain i 075-Amendment No. 85 i Page 3.4-5 !
3,.,4_ BASES 1 Continued)__ s_ _ . . ,, !
i i SURVEILLANCE l REQUIREMENTS SR 3.4.1 and 3.4.2 ,
Verification of PCRV shielding water- tritium ;
concentration limits ensures adequate protection to '
the MEMBERS OF THE PUBLIC. The daily surveillance-frequency during the filling of the PCRV with the .
shielding water, until tritium concentration i decreases below 0.1 pC1/cc, will detect any ,
fluctuations in the tritium concentration-during i non-steady state conditions.. Also, performing daily sampling until tritium concentration it less !
than 0.1 uCi/cc for three consecutive samples [
ensures that equilibrium conditions are achieved '
before the surveillance frequency is decreased.
- The 7 day surveillance frequency will ensure-that a ,
fluctuation in the tritium concentration during subsequent material handling activities will be detected. After equilibrium tritium . concentration :
has decreased below 0.01 pCi/ce, monthly sampling will be performed. This is conservative with respect to Regulatory Guide 8.32 sampling :;
requirements.
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 '
- 3. Regulatory. Guide- 8-32,
. Criteria for Establishing a Tritium Bioassay Program,
+
b b
b t
t y
,,,n v, mw r y r ,- ,, n n ew- -v.,..-,nn,- ...,,,r,l- vn r m e .L . . . . r ---n ,,,-~-wa,m,,~,--..,nn-- ---~-L,-,.-----v
Fort St. Vrain i DTS
+
Amendment No. 85 :
Page 4.0-1 !
4.0 DESIGN FEATURES _ _ _ _ _ _ _
4.1 Site I The Fort St. Vrain Nuclear Generating Station is located approximately 35 miles north of Denver and 3.5 miles :
northwest of the town of Platteville, in Weld County, Colorado.
l The site consists of 2798 acres. The EXCLUSION AREA BOUNDARY ent:Osts the decommissioning Emergency Planning '
2one, as shown on Figure 4-1.
Points where radf"--tive gaseous and liquid effluents are released are shown on Figure 4-1.
t 4.2 Reactor Building :
The Reactor Building houses the prestressed concrete reactor vessel (PCRV), fuel handling area, fuel storage wells, fuel _ shipment preparation facilities, ,
decontamination and radioactive liquid and gas waste processing equipment, and most reactor plant process and i service systems.
f Decommissioning will involve any major modifications to the Reactor Building structural _ steel without 'l verification of the seismic qualification, as described in Section 2.2.1_of the Decommissioning plan.
4.3 pCRV Water Leakage Prevention The PCRV will be filled with water to provide' shielding i for workers during initial- PCRV internal dismantlement-activities. To prevent _ leakage from the PCPV, all-penetrations which are below the PCRV water line and have
- had their instrumentation removed are sealed. Sealing _is I
accomplished with either welded cover plates, welded caps, or blind flanges.
There are two independent. trains-in the PCRV water _ cleanup and clarification. system, to allow for maintenance and '
repair. Each train has s'r 'cient valves and drains to allow isolation as required.
1r y +- - ,y g ycqq m ,-m-w7mg- -v-sy,y*sw- +y-ii-y w- g y yv wire,yyy-www-yw--p*-- Vmeye dr- r- gy 'ye p-gr+wsve , uW re de->+7mes-v-dr--pTPw-+T*wre :We w p T' + v'a v e &re w-e-ru*ve-w&m--***M'"-"'l-* * <"
--- ~u u., . . . - . . -..u .s ws s. -.w -u.w..-.n- u --u-w.~ . , - -==x.._~ +w~s-.1.u.- .n._+ax -s .- a x. - ----wa.-a ~..s--~_a1-.
Fort St. Vrain
- m. 075 Amendment No. 85 -
1
.,\ \ \ , ..-w________.---------,3 Page 4.0-2 l
-[
_ - .- ~l j j --- l :
\ l1 \. , l _ - -- , .-. - .- --- .
~ - -
\_. 6,. , -... - - 3., y 4, ;
),,
I
}
s...+...... s 1 e
....... .. al -
i l
k l l I'i n i
y.j ,
i
't i i . l <
j .
6 I i
i4i g i 1
,' i . . ... . a $ . . ..
~ .d h. , . ..._. d ..M- .
- 4. {.. k. ; "' "s'iX4" "-
- l 8
~
s, e ;
': ; V. . ,; . . . ". . .".". ." s ' li l ,
, , 3 g o !! I . ! < > ' : ., ., 5 \
u '* -
ei ,,- ! . '\
d.
[bm-i
.L _ _ __ ; / :. . . . . . .' .:
g
, p.
__ ._.. _ * '- -e._ .
,, , . , it -
, ia-f M _.s. sa.
.. ., (
/ t' _ _ . . . .1 , ..u. .t. ut:ss2*. . . . . .P.I :..M.; s t-... t r i cqg ' l j i
, j ; .. , , , .
- g - _ ,, ,4 je$ ,f * * \- ;i <
j}
g
! 7 g.s .5 ) 6 fJ,' :.
POINT OF LIQUID *
'l I
I i (O
4
} *** *"* .!.
l lf-N ~ * '
'I ErrLUENT ll {,t.....
l t, I}j RELEASE
.i
.... O(xf , ld,i il*.,, ,l V (- L R ,
i 1, ,, n ::::
.... j
.::.:: :L . . . . . . . . .::::.::::::n) 1 i,
!i
!j H j
"8 g L n._
g .. j i e ,[ . - - - - 7 '/ j; g t n tu ! rm <- :- u > ( I. e / %. L
*ma.
@ v{) A Nf 6
- i. h x
16! "*
.e -i .
1 R ( )b4.Me Iifar.ar 29;, 9w .. , r-- e
'a p avur p I g
*=D '" B d f! -
.- c--
lf :==-4 _- r-~i".lE-- 3w w[.l n n,=
. N)m@._- .-_ mFAdIb*
n -
; r-
)s .
i bh [h b tra ~ '
-! *N r r' " y, g b g s v= ar p s &
f t an, i --.;
. . .. . . .v..-- a ,
1 l, \ auqw" t% A $ ,, 4 ,7 tcro o g, pet p .. . 7 stoo ], l j g wo
** 4
'N
\
P6M-.
""'#N U M%{t{[e{IM[#[fh,,,
9 ! %D PCINT OF
-f-sa-hih, 9 Mj 2
== . L. Q! t i s s4 cAstous-w ij g
- '[i w
.. - '[ -**'C EP [ TURBINE;
- ar-* ^
,BLDo . .
EFFLVENT RELEASE- I
.....ee.... ..
..I,_. ,
l- (@4 -,- r .@r$4%!d6Was;-iI ,1*M j E
. . . . . . . . . . . . /F.
z / n yb e-NIN~d
.__ ._ ~1_ pm. h... . _ . . .' . .q . ...,d
., - - - , . . . . . . _ _-~_ _
. _ _ _. l
.. et m _- 6 .
iR
- 3. . .;,;,,.j 4-
! - 1 a y -- im ,p l .,ly"g,- ,
wm
,,, .t
~
- m &, ru i 33
.... ,, .J -o me w I, s g. s w. m z
me x = 9 i p* g'M 3 =_ g,yu, -
+ --
a r ll ,)., s, s -
> < x - r .
3 , 4x Ir m .4p- ,.i..
- - 4,,... ,, .a . '
g
-I 6
'\ 'g
\,s ,u. g
,mnunnnna m:.. .-
- l
-//
g;. / ./ h litii.11I'ilILLj -
, ......~,' n-o M --......L.s :; ____..__ L___.....c ./ :f FSV SITE Figure 4-1
e Fort St. Vrain DTS Amendment No. 85 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 overall executive responsibility for all Fort St. Vrain , decommissioning activities. ' 1 5.2 Organization The decommissioning organization, functional requirements, and qualification requirements for key decommissioning personnel, for both PSC and contractur groups, shall be documented in the FSV Decommissioning Plan. The organization responsible for quality assurance shall report 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 Decommissioning Safety Review Jommittee (DSRC) 5.3.1 The DSRC shall be comprised of the following: Decommissioning Program Director (Chairman) Deputy Director-Facility Support Manager (Radiation-Protection Manager) Decommissioning Engineering Manager Operations Manager Project Assurance Manager Project Controls Manager Vestinghouse Project Director 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.
- . . , ,,m.. .,
- t Fort St. Vrain DTS ,
Amendment No. 85 + i Page 5.0-2 - ADMINISTRATIVE CONTROLS (Continued) 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 : alterno e Chairman, and a sirrple majority of the i members, including alternates. No more than two alternate members shall participate as voting members in DSRC activities at any one time. 5.3.4 The DSRC shall be responsible for review of: [
- a. Administrative procedures, plans, manuals, and prC+ ams required by Specifications 5.4,1 through 5.4.4, 5.7, and permanent changes thereto, that ;
- affect nuclear safety. !
- b. Preposed tests and experiments that affect nuclear s lety. ,
- c. The following items which involve an unreviewed safety question as defined in 10 CFR 50.59:
- 1) Administrative procedures, plans, manuals, and-programs required by Specifications 5.4.1 through. 5.4.4, 5.7, and permanent _: changes thereto,
- 2) Proposed changes or modifications to plant systems or equipment, and
- 3) Proposed tests and experiments.
- d. Proposed changes to the decommissioning work specifications that affect nuclear safety, and any new decommissioning work specifications that affect nuclear safety. ;
- e. Proposed changes -to the Decommissioning Technical 4 Specifications or Facility License. !
- f. Investigations of violations of DecommissioningL Technical Specifications,. and of regulations or >
license requirements.
- g. Reportable events.as defined by 10 CFR 50.73. '-i
. h. Unplanned release of radioactive material _to_the environs.-
T "#'7 ^9 1 D'*TeVFrWytP-9'te---PWtirY' - ev'19 4M"'T V'C*' 7 NFC'-'N 'r' P'-"'" -r%W9'1'mbfp t'-'+4# v e's '-'dP- P W F'w='W4- N-'WWw~'1' **"-"M-"-'Fl- Ne"- 64" l N'""-YMW
l I Fort St. Vrain DTS l
+
Amendment No. 85 l Page 5.0-3 I ADMINISTRATIVE CONTR0f.S (Continued) _ 5.3.5 The DSRC shall:
- a. Advise the Decommissioning Program Director on matters that affect nuclear safety.
- b. Recommend to the Decommissioning Program Director in- )
writing, approval or disapproval of items considered j under Specifications 5.3.4.a through 5.3.4.d above. '
- c. Render determinations in writing witi. regard to whether or not each item considered under Specificattun 5.3.4.c constitutes cn unreviewed safety question,
- d. Recommend to the Decommissioning Program Director other areas of facility activities where additional oversight is prudent and/or where independent auditing is needed.
5.3.6 Audits of decommissioning activities shall be performed under the cognizance of the DSRC. These audits shall i encompass: l
- a. A decommissioning program audit to be performed at least once per year, encompassing the f ollowing:
- 1) Decommissioning Technical Specifications
- 2) Radiation Protection Program
- 3) Training Program ,
- 4) Decommissioning QA Plan
- 5) Decommissioning Access Control Plan
- 6) Decommissioning Fire Protection Plan
- 7) Decommissioning Emergency Response Plan b, Any other area of- facility activities considered-appropriate by the DSRC.
5.3.7 Records of DSRC activities shall be prepared, approved, and distributed as indicated-below:
- a. Minutes of each DSRC meeting and documentation of the. reviews performed per Specification 5.3.4_ above ,
shall -_ be approved- and forwarded to the Vice President responsible for_ nuclear: activities within 30 days following the meeting.
Fort St. Vrait, i' DTS Amendment No. Si : Page 5.0-4 l AkilNISTRATIVECONTR0; (Continued) ___ _
- b. Audit reports encompassed by Specification 5.3.7 -
, above shall be forwarded to the Vice President responsible for nuclear activities within 30 days - , after completion of the audit. 5.4 procedures and programs t 5.4.1 Written administrative procedures, plans, manuals, and/or programs shall be established, implemented, and maintained covering the activities referenced below:- Radiation Protection Program a.
- b. Surveillance test activities of equipment required by these Decommissioning Technical Specifications
- c. Decommissioning Access Control Plan
'd. Decommissioning Emergency Response Plan
- e. PROCESS CONTROL PROGRAM
- f. OFFSITE DOSE CALCULATION MANUAL
- g. Decommissioning Fire Protection Plan 5.4.2 Administrative procedures, plans, manuals, and/or programs of Specification 5.4.1 above, and permanent changes thereto.- that affect nuclear safety, shall be reviewed by the DSRC, or a -subcommittee thereof, and approved by the appropriate management prior to implementation. Procedures shall be reviewed periodically as set forth in Administrative Procedures.
Changes to the OFFSITE DOSE CALCULATIC;1 MANUAL shall be processed in accordance with Specification 5.10, and changes to the PROCESS CONTROL PROGRAM shall be processed in accordance with Specification 5.9. l 5.4.3 Tempora ry changes to administrative procedures, plans, manuals, and/or programs of Specification 5.4.1 above
- may be made- provided the change is documented and '
l approved:'by the appropriate management prior to implamentation, 1 c -- - -- - - , - - . - - . . - . . .- , - , - - - - - .-, --
Fe t St. Vrain DTS Amendment No. 85 Page 5.0-5 ADMINISTRATIVE CONTROLS (Continued) 5.4.4 The following programs shall be established, implemented, and maintained:
- a. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the OFFSITE DOSE CALCULATION KANUAL, (2) shall be implemented by procedures, and (3) shall include remedial actions to be taken whenever the program ,
limits are exceeded. The program shall include the following elements: 1} Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the OFFSITE DOSE CALCULATION MANUAL,
- 2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conformir.g to 10 CFR Part 20 limits,
- 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with'- -10 CFR 20 and -with -the methodology and parameters in the 0FFSITE DOSE CALCULATION HANUAL, 4} Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix 1 to 10 CFS Part 50,
- 5) Determination of cumulative and projected dese I
contributions from radioactive effluents for the current calendar quarter. and current calendar year in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL at least every 31 days,
a Fort St. Vrain DTS Amendment No. 85 i Page 5.0-6 j ADMINISTRATIVE CONTROLS (Continued)_
- 6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to 1 ensure that the appropriate portions of these .
systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the : guidelines for the annual ' dose or dose ' commitment conforming to Appendix 1 to 10 CFR Part 50,
- 7) Limitations on the dose rate resulting from radioactive material released in gaseous.
effluents 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 doses resulting from noble gases released in gaseous effluents to areas beyond the EXCLUSION AREA BOUNDARY conforming to Appevix I to 10 CFR.
Part 50,
- 9) Limitation; on the annual and quarterly doses to a MEMBER OF THE PUBLIC from tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas beyond ~ the EXCLUSION AREA BOUNDARY conforming to Appendix I-to 10 CFR Part ,
50,
- 10) Limitations on the annual dose or dose -
commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation frc.d uranium fue'l cycle sources conforming to 40 CFR Part 190. i L
?
r
- v. - - . ...e,.,,...,.++ . . , ,,._r5 ..#.+r,,.-_-.,.%e~ , ,.,~.r r..y ,,_.,,,,r-,m-..c.',e.--,,,,-y v.',,,,.r , w:m , , b,.,.w- yv.g+
Fort St. Vrain DTS Amendment No. 85 Page 5.0-7 ADMINISTRATIVE CONTROLS (Continued) _ _ _
- b. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposuc pathways, and (2) verification of the accuracy of the effluent monitoring program snd modeling of environmental exposure pathways. The program shall (1) be contatoJ 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 a nt' parameters in the OFFSITE DOSE CALCULATI'r 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 inoependent checks on the precision _ and accuracy of the measurements of radioactive materials in environmental sample matrices -are performed as part of the quality a assurance program for environmental monitoring.
Fort St. Vrain DTS Amendment No. 85 - Page 5.0-8 ADMINISTRATIVE CONTROLS (Continued) _ 5.5 Reporting R quirements In addi ti ot, to the applicable reporting requi-ements of 10 CFR, the 'ollowing reports shall be submitted to the Regional Adm'. .. strator of the NRC's Region IV office unless otherwise noted: 5.5.1 Annual Radiological Reports Annual report: covering the activities described below, for the previous calendar year shall be submitted as follows:
- a. Annual Radiation Exposure Report The Annual Radiation Exposure Report for the previous calendar year thall be submitted to the 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,
- b. Annual Radiological Environmental Operating Report The Annual Radioloc ed Environmental Operating Report covering the _11 ties of the unit during the previous calendar ye.r shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental -
Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the 0FFSITE '75E CALCULATION MANUAL and (2) Sections IV.B.2, IV. 3, and IV.C of Appendix I to 10 CFR Part 50.
6 Fort St. Vrain DTS Amendment No. 85 Page 5.0-9 2 ADMINISTRATIVECONTROLS(Continuedl_ , 5.5.2 Annual Radioactive Effluent Release Report The Annual Radioac ;> F.ffluent Release Report covering activ sies during c< previous. 12 months shall be submitted 'within r /s af ter January 1 of each year. , The report shall inciese a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The report shall also include a copy of the 0FFSITE DOSE CALCULATION MANUAL, if any , chan;es were made during the report period, as required by Specification 5.10. The material provided shall be (1) consistent with the objectives outlined in the OFFSITE 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 Reports
- a. The .NRC Operations Center shall be notified of emergency and nonemergency events in accordance with 10 CFR 50.72.
- b. Reportable events shall be reported in accordance with 10 CFR 50.73.
5.5.4 Special Reports Special Reports required by Specification 3.4-shall be submitted to the NRC Regional Administrator within the time period specified. 1 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.
., . . . ~ , . - - . - - . - - . .- - - . _ - . ~ . . - . . -. . . .
Fort-St. Vrain DTS Amendment No. 85-Fage 5.0-10 ADMINISTRATIVE CONTROLS (Continued)
- d. 4ecords of changes made to procedures related to nuclear safety.
- e. Records of radioactive shipments,
- f. Records of sealed source leak tests and results.
5.6.2 The following records shall be retained for the duration of the Facil'ty 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 radiatic axposure for all individuals entering radiation contrui areas.
- d. Records of gaseous and liquid radioactive material released to the environs.
- e. Records of training and qualification for current members of the decommissioning staff,
- f. Records of activities required by the Decommissioning QA Plan,
- g. Records of reviews performed pursuant to 10 CFR 50.59.
- h. Records of meetings of the DSRC.
- i. Records and logs pertaining to- the Radiological Environmental Monitoring Program.
J. Records of changes maoe to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM. 5.7 Radiation Protection Program Procedures .for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall -be approved, maintained, and adhered to for all activities involving personnel radiation exposure.
.~ _
' Fort St. Vrain DTS' - -
Amendment No. 85- , Page 5.0-11 ADMINISTRATIVE CONTROLS'(Continued) 5.8 High 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 ' 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
- 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 may be made after the dose rate level in the_ area has been established and personnel -have been .made-knowledgeable of thum, or
- c. A . health physics- qualified individual (i.e...
qualified in radiation protection procedures) with a 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.
*w r o sn~= w M2 w M 9-Hm>'-k , - W ev M 4 v
, - . ~- . --
b
-Fort St. Vrain DTS Amendment No. 85 Page~5.0-12 ADMINISTRATIVE CONTROLS (Continued) 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 penetrates shall be provided with locked enclosures to prevent unauthorized entry, and the keys _shall be maintained under the administrative control of health physics supervision. Enclosures _ shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate-levels in the immediate work area and the maximum allowable stay time for individuals in the area. In lieu of the stay time specification of the RWP, direct'or remote (such as-use of closed _ circuit TV cameras) continuous surveillance m6y be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.
For individual areas accessible to personnel with radiation levels of greater than 1000 mR/h thatf are located within large areas, where.no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device whenever the dose rate in the area exceeds or will 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 - tosupportfthechange-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 DSRC in accordance with Specification 5.3.6.
1
Fort St. Vrain DTS Amendment No. 85 Page 5."-13 ADMINISTRATIVE CONTROLS (Continued)__ _ 5.10 0FFSITE DOSE CALCULATION MANUAL Changes to the OFFSITE 00SE 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 the change (s) and
- 2) A determination that t..e 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 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- b. Shall become effective after review and acceptance by DSRC in accordance with Specification 5.3.6.
- c. Shall -be submitted to the Commission in the form of a complete, legible copy of the entire OFFSITE DOSE CALCULATION MANUAL as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the OFFSITE DOSE CALCULATION MANUAL was made. -
5.11 Natural Gas Restriction As indicated in Specification 1.0, FSV is being converted to utilize a gas-fired boiler. The natural gas line supplying this boiler, 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 unaccepttble hazard to the ACTIVATED GRAPHITE BLOCKS or to the equipment or systems needed to protect the ACTIVATED GRAPHITE BLOCKS. __~
9 SAFETY fVALUATION BY THE U.S. NUCLEAR REGULATORY COMMISSION RELATED TO AN ORDER AUTHORIZING DECOMMISSIONING AND AMENDMENT NO 8S TO POSSESSION ONLY LICENSE NO. DPR-34 PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 3 l I
CONTENTS Page 1
SUMMARY
OF PLAN ......................... I 1.1 Description of Decommissioning Plan and Decommissioning Alternative ....................... I 1.1.1 Introduction ................... I 1.1.2 Background .................... I 1.1.3 Proposed Action . . . . . . . , . . . . . . . . . I 1.2 Major Tasks, Schedules, and Activities . . . . . . . . . . 2 l.2.1 Description of Major Activities . . . . . . . . . . 2 1.2.2 Schedule for Decommissioning Activities . . . . . . 4 1.3 Cost Estimate and Availability of Funds ......... 6 1.3.1 Decommissioning Cost .....,......... 6 1.3.2 Decommissioning Funding Plan ........... 6 2 GENERAL DESCRIPTION OF FORT ST. VRAIN ,.......... .. 6 2.1 Decommissioning Activities, Planning, and Exposure Estimates .............. .......... 7 2.2 Decommissioning Organization cnd Responsibilities .... 9 2.3 Training Program . . . . . . . . . ........... 9 2.4 Contractor Assistance .................. 10 3 PROTECTION OF OCCUPATION AND PUBLIC HEALTH AND SAFETY ,..... 10 3.1 Facility Radiological Status . . . . . . . . . . . . . .. . 10 3.1.1 Facility Operating History ........,... 10 3.1.2 Radiological Status of Plant ........... 11 3.2 Radiation Protection . . . . . . . . . . . . . . . . . . 11 3.3 Radioactive Waste Management . . . . . . . . . . . . . . . 13 3.3.1 Spent Fuel Disposal . . . . . . . . . . . . . . . . 13 3.3.2 Radioactive Waste Processing ........... 14 3.3.3 Radioactive Waste Disposal ...,........ 16 3.4 Accident Analysis .................... 19 3.5 Industrial Safety .................... -20 3.6 Asbestos . . . . . . . . . . . . . . . . . . . . . . . . . 20 4 FINAL RADIAT!0N SURVEY PLAN ................... 21 i i ... ._ . . . . . - . .
CONTENTS (cont'd) Page 5 UPDATED COST ESTIMATE FOR DECOMMISSIONING ..., ....... 21 6 TECHNICAL SPECIFICATIONS IN PL ACE DURING DECOMMISSIONING . . . . . 23 7 QUAli1Y ASSURANCE ........................ 23 8 DECOMMISSIONING ACCESS CONTROL PLAN ............... 23 9 DECOMMISSIONING EMERGENCY RESPONSE PLAN ............. 23 10 DECOMMISSIONING FIRE PROTECT 10h PLAN . . . . . . . . . . . . . . . 23 - 11 LICENSE CONDITION 2.C.(5) .................... 23 12 STATE CONSULTATION . . . . . . . . . . . . . . . . . . . . . . . . 24 13 ENVIRONMENTAL CONSIDERATIONS . . . . . . . . . . ........ 24 14 CONCLUSIONS ........................... 24 15 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 ii
9 1
SUMMARY
OF PLAN 1.1 Descriotion of Decommissionina Plan and D.3commissionina Alternative 1.1.1 Introduction On December 5,1988, the Public Service Company of Colorado (PSC) notified the Nuclear Regulatory Commission (NRC) that it had elected to terminate Fort St. Vrain operations early because of economic considerations associated with the ongoing operating costs at the plant. PSC submitted its proposed decommissioning plan (PDP) in accordance with Section 50.82(a), Title 10 of the Code of Federal Reaulations,10 CFR 50.62(a), which requires the PDP be submitted "within two years following permanent cessation of operations." The PDP was submitted on November 5, - 1990, with the DECON option as the selected decommissioning alternative. The NRC staff submitted several requests for additional information (RAls) to PSC. In response to the staff's RAls, PSC revised the PDP on December 17 and 21, 1990, January 14, April 15 and 26, May 15, June 6 and 17 July 1, August 28 and 30, November 15, and December 6, 1991; and January 9, March 19, April 17, and September 25, 1992. PSC submitted an Environmental Report Supplement on July 10, 1991, and revisions on March 20, April 30, June 24, 1992, and September 1 and 18, 1992. Subsequent PSC submittals have addressed all outstanding NRC RAls. The NRC staff evaluated the adequacy of the licensee's proposal on the basis of applicable NRC regulations and regulatory guidance and in accordance with applicable sections of the Standard Review Plan (SRP, NUREG-0800). The results of its evaluation are provided below. 1.1.2 Background Fort St. Vrain (FSV) was shut down on August 18, 1989. On August 29, 1989, the PSC Board of Directors confirmed the decision that FSV would not be restarted and that PSC would pursue the decommissioning. PSC identified problems with the control rod drive assemblies and the sNam generator steam ring headers that presented significant technical obstacles that could-be overcome, but at significant cost and time to PSC. Additional-ly, the uniqueness of the one-of-a-kind high-temperature gas-cooled reactor fuel cycle made the cost to purchase new fuel prohibitive. This, in-conjunc-tion with low plant availability and correspondingly high operating costs, was the basis for the PSC's decision to discontinue operation of FSV. 1.1.3 Proposed Action The PSC selected the DECON option as'the decommissioning alternative and intends to decontaminate and dismantle the prestressed concrete' reactor vessel (PCRV) and supporting systems to the extent necessary to ensure removal of radioactive materials and to allow release of the facility and site for unrestricted use. The contamination and activation levels are low at FSV because the plant had a relatively short operating history of approximately 447 full-power days since
2 1979 when commercial operation was initiated. The licensee elected the DECON alternative to: (1) allow maximum flexibility in use of the site and facility;
-(2) decommission the facility without significant radiation exposure; (3) eli-minate the need for long-term monitoring, surveillance, and maintenance; ,
(4) avoid any significant effects to the environment; and (5) support the agreement between the Colorado Public Utilities Commission and PSC regarding funding for the DECON alternative. The proposed decommissioning is necessary to terminate the FSV license in accordance with the requirements of 10 CFR 50.82. Dismantlement and decontam-ination of the plant systems and the PCRV to conditions suitable for unre-stricted release are the required results of the decommissioning action that the licensee will be undertaking. The licensee's selected decommissioning alternative meets the requirements of 10 CFR 50.82(b)(1) and is acceptable because the decommissioning will be completed prouptly, within the term of the present license and adequate procedures and controls to protect occupational and public health have been developed. The licensee has provided an adequate description of the final radiation survey and developed an updated cost estimate for the DECON alternative. 1.2 Ma_ior Tasks. Schedules. a.nd Activities 1.2.1 Description of Major Activities The major dismantlement and decontamination activities to be performed during decommissioning are divided into three major work areas: (1) decontamination-and dismantlement of the PCRV; (2) decontamination and dismantlement of the contaminated balance-of-plant (80P) systems; and (3) site cleanup and final site radiation survey. (1) Decontamination and DismantlemeD1 of the PCRV The major decommissioning task is the dismantlement and decontamination of the reactor internal components and the radioactive portions of the PCRV. Activi-ties to dismantle the PCRV will begin'after all irradiated fuel has been removed from the reactor building and transferred to the independent spent fuel storage installation (ISFSI) or to the Idaho National Engineering Laboratory. Section 2.3 of the PDP provides a detailed description of the steps necessary to dismantle and decontaminate the PCRV. -All-dismantlement and decontamination activities must-be accomplished in accordance with Technical Specification (TS) 5.7, " Radiation Protection Program."- After it evaluated several technical options for dismantling radioactive portions of the PCRV, PSC decided to flood the PCRV so that a majority of dismantlement activities could be performed under water. This approach allows direct access to highly radioactive portions of the PCRV, while affording the maximum shielding benefit, which provides significantly lower estimated worker exposure than other approaches. However,.this approach has raised concern re-garding the sealing of the PCRV penetrations and about the possible releases of large amounts of tritium from the graphite reflector blocks. l
3 The PCRV will be dismantled using a diamond wire cutting technique. This is a standard construction method for cutting large volumes of concrete. In summary, the diamond wire cutting system consists of a wire with collars containing a diamond-matrix, made to length for each individual cut, and a hydraulic puiley to drive the system to circulate the wire. The diamond wire is routed to envelop the cut area. Chapter 2 of the PDP provides a detailed discussion on the use of the diamond wire cutting to dismantle the PCRV. The PCRV top head will be cut in several sections using the diamond wire and removed. The activated concrete in the PCRV walls will be cut with the diamond wires. This will be accomplished by removing vertical and circumferential tendons for access for the diamond wires. In cases where the tendon tubes are not useable, new vertical holes will be core drilled to allow a complete cut. After the concrete has been removed by the diamond wire method, additional decontamination by scabbling, vacuum sand blast, or wiping may be required in some areas to met the release criteria. The licensee has demonstrated that the use of these methods to dismantle the PCRV provides adequate protection of tha worker and maintains occupational doses as low as is reasonably achievable (ALARA). The sealing of the PCRV, and the possible release Of large amounts cf tritium are addressed in Chapters 2 and 3 of the PDP respectively, and in Chapter 4 of the Environmental Report Supplement. TS 5.7 provides adequate radiation protection requirements with regard to tritium in the PCRV. The licensee has committed to the application of Regulatory Guide 1.143 for activities related to flooding the PCRV. Before flooding the PCRV, all penetrations through the PCRV will be sealed. The penetrations will be sealed by either cutting and capping the penetration outside the PCRV or by installing bolted and gasketed fl anges . In addition, where welding is required, all welds will be nonde-structively tested in accordance with applicable codes. All leakage resulting from flooding the PCRV will be treated by means of the disposal demineraliza-tion and filtration system that is part of the PCRV water cleanup and clarifi-cation system. The leakage will be detected by visual inspection. Section 2.2 of this report provides a detailed discussion regarding the dismantlement of the PCRV, and Section 3.3.2 of this report provides a detailed analysis and evaluation of the tritium concern. (2) D_econtamination and Dismantlement of Contaminated Balance of Plant (80P) Systems For the purposes of the PDP BAP systems refer to those contaminated plant systems outside the PCRV. PSC will decontaminate and dismantle contaminated B0P systems as described in Chapter 2 of the PDP. The BCP systems are listed below. In' summary, the BOP systems will either be-decontaminated in place by conventional methods or removed and disposed of as low-level radioactive waste. The licensee will use conventional methods such as shears, scabbling, - mechanical cutting and flame cutting to remove the 80P. These methods minimize worker exposure and maintain occupational doses ALARA. B0P SYSTEMS System 13 - Fuel Handling Equipment System 14 - Fuel Storage Facility
l 9 4 System 16 - Auxiliary Equipment System 21 - Helium Circulatory Auxiliary Equipment System 23 - Helium Purification System System 24 - Helium Storage System System 46 - Reactor Plant Cooling Water System System 47 - Purification Cooling Water System System 61 - Decontamination System System 62 - Radioactive Liquid Waste System System 63 - Radioactive Gas Waste System System 72 - Reactor Building Drain System System 73 - Reactor Building Ventilation System System 93 - Inetrumentation & Controls (3) Final Radiation Survey Plan and Site Cleanuo Chapter 4 of the PDP specifies the release criteria that PSC will use in decommissioning FSV and provides a detailed description of the final radiation survey. The PDP release criteria are consistent with criteria provio'ed to PSC in an NRC letter dated October 4, 1989, SECY-92-106, and confirmed by NRC letter dated April 27, 1992. Therefore, decontamination of FSV to levels that meet the release criteria specified in the PDP will allow termination of License DPR-34. The proposed firal radiation survey must demonstrate the effectiveness of the decommissioning and provide documentation that all contaminated materials, structures, areas, and components have been successfully removed or decontam-inated to acceptable levels to permit release for unrestricted use. This final radiation survey to release the FSV site, facilities, and installed equipment for unrestricted use will be performed following the completion of the decontamination and dismantlement activities. The proposed survey plan is acceptable for proceeding with decommissioning, but the staff will reevaluate the adequacy of the survey and the survey results when FSV decommissioning is complete. An independent survey by the NRC or an independent contractor will be used to confirm that the residual radioactivity at FSV meets the NRC criteria discussed above. Section 4 of the SER < raluates the final radiation survey methodology and criteria and concludes it . the final survey plan meets the requirements of 10 CFR 50.82(b)(3). 1.2.2 Schedule for Decommissioning Activities PSC has completed the decommissioning planning phase. It consisted of preparation of work scope planning, work specificaticas and procedures, and equipment and material staging. PSC esti.nated that decontamination and dismantlement (i.e., actual dismantlement, decontamination, and physical decommissioning activities) will take about 39 months. Section 2.3.5 of the PDP provides a detailed schedule of FSV decommissioning and decontamination activities. Figure 1 of this report provides a time line of FSV decommission-ing. The NRC staff concludes that PSC has addressed all major activities and the schedule for completion of decontamination of FSV is reascnable on the basis of NUREG/CR-0130 as well as a comparison with other facilities. i
5 g ................................................
...................................g...............
r 3 ......................................................... !
............................I.5...... .I.
s
.................................................... . . . . . . . . . . . . . . . . . E..........
y .... . . . . . . . . . . . . . . . -..
..=...4 - . . . . . . .
.....,lE.l. m gI l*_-g 5 _
c a .
..................-....E =e _u s
- I. ....... .
5
.~.lI=!.s e
- I og" . _$
m a a 1s, Ira 5 g- .E w R .......-....-~~.w..-~~.~~cc = .y
- 5 :! $.
3 -5 a
- l'
- l. .m
- . . . - - - ~ . . . g g=y~~"g f,, ip II I E I
g .............. g .s s l'g I.a; a g.= i
- s. I.
- w. . . .. . .
......l. .
I 353 - e EEf f a !8,s5= I E W cla= = =, s - 1 E e
= ..... ..
I w=.I e,...e. E,1I )
..x. . . .= ..,..._$........
gaf ., f.r. g 3 lii!; a EliS
!!Ple E!E E ..I s -
t i ii e . r gpm[.Lr sw
.= , luL . .. . . .. . . ..N .. h._ . -
I h Ew _a 1 ' S
. ill s f ,'
> g -l!
l a 5
. t g I E
.c--.-.....
=I .I e If is'.
55 x a r - 3'
.. g -
d = ir i z Fig"ce 1 Sunaary Schedule of Major Decommissioning Tasks
6 1.3 Cost Estimate and Availability of Funds 1.3.1 Decommissioning Cost PSC estimated a total decommissioning cost of $157,472,700 for FSV. Assump-tions used as the basis for these costs are identified in the PDP. Section 5 of this report provides a detailed evaluation of the' cost for decommissioning FSV. The NRC staff concludes this cost estimate is reasonable and satisfies the requirement of 10 CF" 50.82(b)(4). 1.3.2 Decommissioning Funding Plan The First Interest Bank of Denver, N.A., entered into a standby trust agree-ment with PSC for the purpose of receiving payment under an irrevocable letter of credit issued to the PSC account. The letter of credit provides financial assurance for the decommissioning of FSV. The term of the standby trust agreement commences when the letter of credit is issued and expires when the decommissicning activities at the facility are completed, or as otherwise provided in the standby trust agreement. The bank acts as trustee and administers any funds received to fund costs for decom-missioning in accordance with the terms of the standby trust agreement. PSC proposes to use a $28 million external trust fund plus a $125 mmicn Icar of credit that will decline as DECON operations are completed. The combir M total of $153 million exceeds the estimated decommissioning cost of $157.5 million, less expenses to date of $10.5 million. The NRC staff concludes that PSC's proposed decommissioning funding assurance mechanism is acceptable and in compliance with 10 CFR 50.75(e). 2 GENERAL DESCRIPTION OF FORT ST. VRAIN FSV is a high-temperature gas-cooled reactor (HTGR) owned and operated by PSC. FSV is located approximately 35 miles north of Denver and 3.5 miles northwest of the town of Platteville in Weld County, Colorado. FSV had a capacity of 330 MWe. The PSC-owned site consists of 2798 acres. Approximately 1 mi2 within the site area is designated as the exclusion area; PSC maintains complete control over this area. The closest distance from the reactor building to the nearest exclusion area boundary is about 1935 feet, but the reactor building is about 3500 feet from the nearest site boundary. The Atomic Energy Commission -(AEC) issued a provisional construction permit to PSC on September 17, 1968 (AEC Docket No. 50-267). Fort St. Vrain was initi-ally scheduled for commercial operation in 1972. Although PSC received a full-power operating license in 1973, NRC mandated extensive pre-operational testing and the resulting engineering modifications delayed commercial operation until 1979. Chapter 2 of the PDP provides a complete description of the Fort St. Vrain facility. The major events and milestones that occurreo at FSV are listed below.
4 ' 7 1973 December Plant construction completed, facility .perating License DPR-34 issued to PSC 1974 January- Initial criticality achieved, startup testing, low-power operation, and required plant modifications implemented (1974-1979) 1979 Jul;e Commercial operation began 1981 November 100 percent full-power operation achieved 1984 June Six control rod drives (CRDs) failed to automatically scram causing shutdown 1986 February Plant restarted following CRD refurbishment outage May Environmental qualification outage September FSV removed from the rate base October Safe shutdown cooling reanalysis performed, reducing maximum power level to 82 percent of rated power (270 of 330 Mwe) 1987 July Plant shut down following helium circulator _ bolt failure October Hydraulic fire during plant restart 1988 June Plant record achieved for MWe generated for 1-month period July Plant shut down to refurbish helium circulators December Decision approved by PSC Board of Directors to shut down and decommission (operations-will cease on or before June 30, 1990.) 1989 August Plant permanently shut down because of control rod failures and subsequent discoveries of failure of steam generator ring headers 1990 November Decommissioning plan (DECON option) submitted 1991 May Possession only license issued December Movement of the fuel to ISFSI initiated 1992 May Coro defueling completed 2.1 Decommissionina Activities. Plannina. and Exposure Estimates Decommissioning of FSV includes the dismantlement, decontamination, and disposal of radioactively contaminated material and components within the PCRV, contaminated R0P systems, and contamination on the remaining site, followed by the final radiation survey. The activated and contaminated portions of FSV will be decontaminated, dismantled, and removed during the decommissioning process in compliance with TS 5.7 and TS 5.8, "High Radiation Area," of Appendix A to Docket 50-267. The activities are divided into decon-tamination of the PCRV and decontamination and dismantlement of the contami-nated B0P. The licensee has provided a detailed analysis in Section 3.1.2 of
8 the PDP regarding the radiation levels at the facility. The radiation levels were determined on the basis of an August 1990 survey and supporting activa-tion analysis. Table 1 of this report provides a summary of the estimated exposures for decommissioning the facility. While FSV is an HTGR and is different from the typical boiling-water and prc surized-water reactors, many activities and pro- - cedures are similar. A comparison of FSV estimated exposure rates to those of Pathfinder, Shoreham, and the generic estimates of NUREG/CR-0130 show that the exposure rates are reasonable and the estimated exposure for decommissioning ' FSV is considerably less than the generic estimates. NUREG/rR-0130 provides total exposure estimates of 1400 man-rem compared to 433 man-rem for FSV. A detailed evaluation of these exposure estimates is provided in Chapter 3 of this report. The staff concludes that although personnel conducting the dis-mantling activities will be exposed to radiation during the dismantling and decentamination, PSC has developed activities and procedures to limit exposure and control radioactive material in order to maintain occupational doses as ALARA. Exposure estimates to accomplish the individual tasks and overall project are reasonable. Table 1 Projected Person-Rem Exposure for the fort St. Vrain Decommissioning Project Person Person Work Activity Hours
- Exposure **
PCRV Dismantlement and Decontamination (D/D):
- Initial preparation / disassembly- 23,733 7.4 Remove PCRV concrete top head 20,578 20.4 Dismantle PCRV core and core barrel 49,368 157.3 Remove core support floor, barrel, and insulation 9,213 103.4 D/D PCRV lower plenum 16,103 59.9 Final PCRV dismantlement, decontamination
- and cleanup 15.047 17.7 Subtotal
- 134,042 366.1 Contaminated B0P D/D and Waste Packaaina:
Initial preparation / characterization 7,279 0.25 Dismantle /decon operations 58.684 1.4 Subtotal: 65,963 1.65-Waste preparation, packaging, shipping, and disposal 33.055 65.4 Total: 233,060 433.15
- Person-hours only for those tasks where the potential for measuring radia-tion exposures exists.
** Exposure time (worker efficiency) is ..timated to be 50 percent of scheduled 1 work time for PCRV tasks where the potential for radiation' exposure exits.
< r w m ,e -
,-.y , ,-- , nn - . , . . -~ - .w~,..
s To accomplish decommissioning, substantial portions of the existing plant will be dismantled and removed. However, the reactor and turbine buildings and structures that are not radioactive above limits suitable for unrestricted use will remain. The radiation program provides adecuate requirements for radia-tion protection of workers and the public. Floocing of the PCRV with water to provide shielding will reduce worker exposure by a significant factor. PSC estimates an exposure reduction factor of 2 to 11 based on decommissioning proposals by other vendors. Site cleanup involves pre- and post-decommissioning surveys of the site and the radiological decontamination necessary to meet the regulatory guidelines to allow release for unrestricted use. These activities are discussed in detail in Section 4 of this report. 2.2 Decommissionina Oraanization and ResAgnsibilities In Section 2.4 of the PDP and in TS 5.0, " Administrative Control," PSC ident-ified the key positions in the decommissioning organization and described their functions. The lines of authority to the corporate level are indicated in figure 2.4.1 of the PDP. The education, training, and experience require-ments are described for all positions important to decommissioning safety. The person with ultimate onsite authority for various functional areas is the PSC Decommissioning Program Director who has overall responsibility for all decommissioning activities conducted by PSC and contractors. The decommis-sioning organization also includes a Decommissioning Safety Review Committee (DSRC) to monitor the decommissioning operation to ensure that it is being performed safely. The DSRC will review and audit mcjor decommissioning operations dealing with radioactive material, radiologi;al controls, review procedures, records, reportable occurrences under 10 CFR Parts 20 and 50, and changes made in accordance with 10 CFR 50.59. The responsibilities and function of the DSRC are defined in TS 5.3, " Decommissioning Safety Review Committee " The committee reports to the Vice President Nuclear Operations. The staff concludes that the licensee's proposed decommissioning organiza-tional structure is acceptable and is accordance with the provisions of NUREG-0800, Sections 13.1.1, " Management and Technical Support," and 13.1.2 and 13.1.3, " Operating Organization." 2.3 Trainina Proaram The licensee's training program is described in Section 2.6 of the PDP and provides general employee training for all decommissioning personnel. The radiation worker training will incorporate the requirements of 10 CFR 20.103 and the guidance of Regulatory Guide 8.15. The training and qualifications of the health physics technicians and supervisors will be conducted in accordance with American Nuclear Society /American National Standards Institute (ANS/ ANSI) Standard 3.1-1981. PSC stated that specifit job training will be provided for decommissioning personnel on the basis of specific job requirements. Records of all training will be maintained. Because the training program for the decommissioning personnel is in accordance with the provisions of NUREG-0800 Section 13. " Training," it is acceptable and ensures that the licensee will be able to maintain ALARA.
10 2.4 Contractor Ass %'aEt PSC will retain ovt al responsibility for the decommissioning of FSV. PSC selected Westinghoust and its support contractors to perform the decommissicn-ing of FSV. Westinghouse is tne prime contractor and will provide engineering and licensing support to PSC. H. K. ferguson will provide site labor, labor management, and support to Westinghouse. Chapter 2.5 of the PDP describes the scope of work to be accomplished, the administrative controls to be ust.d to ensure adequate healtn and safety protection, and the qualifications and experience of the contractors. -The staff concludes that PSC provided adequate information on.its contractors and is capable of retaining overall responsibility for decommissioning. 3 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 3.1 Dtcility Radiolooical Status The staf f reviewed the oparating history and radiological conditions in-the plant and evaluated the activities and tasks to be carried out in contaminated dreas. The staff relied on Regulatory Guide DG-1005, " Standard Format and Content for Decommissioning Plans for Nuclear Reactor," and applicable sect-ions of 10 CFR Parts 20 and 50 for review guidance. 3.1.1 Facility Operating History in Section 3.1.1 of the PDP, PSC addressed conditions in the plant that could affect decommissioning, such as radioactive spills, potential contamination in inaccessible areas, and operating events that had the potential to spread radiation. During the operational history of the plant there have been no spills or releases of radioactive effluents resulting in significant residual radioactive contamination either on site or off site. However, there have been a few routine plant operatio" that may have resulted in residual radioactive contamination in areas that are inaccessible. Specifically, the fuel storage wells and equipment storage wells on the re-fueling floor were used to store spent fuel and highly radioactive components. Over the years of transferring various components and spent fuel, it is anticipated that high levels (e.g., 5,000,000 dpm/100 cm2) of loose surface contamination have accumulated on horizontal surfaces. The lower portions of these wells are inaccessible. At various times throughout plant history, the hot service facility also has had levels of loose surface contamination measuring greater than-5,000,000 dpm/100 cm2 Periodic decontamination was typically performed using water; as a result, crud traps may have been created in inaccessible areas. To date, no crud traps have been identified-in accessible areas containing drain piping from the hot surface facility. The results of the August 1990 radiation survey demonstrated that greater than 95 percent of the plant areas have radiation levels corresponding to back-ground. Table 3.1-1 of the PDP identifies those areas with radiation levels above background.
11 The staff finds that PSC provided sufficient information and that the informa-tion is acceptable and meets the requirements of 10-CFR 50.75(g)(1). 3.1.2 Radiological Status of Plant The radiological conditions of the plant are described in Section 3.1.2 of the PDP. The staff evaluated the radiation hazards at the plant on the basis off the activation analyses performed and the August 1990 radiation and contamina-tion survey performed on the reactor building and turbine building. Table ' 3.1-1 of the PDP provides a summary of the contamination levels at the facility with a description of major contributors. General area radiation levels throughout the turbine building are primarily due to natural background. Contamination levels both fixed and loose are less than 1000 dpm/100 cmr at all locations and generally less than 100 dpm/100 cmt. External radiation levels outside both the turbine building and reactor buildings are typically less than 2 mrem /hr. The PCRV served as the containment of the nuclear steam supply system and all the internal PCRV components will either require decontamination or will be removed and disposed of as radioactive waste. PSC conducted activation analyses on the PCRV. Table 3-1.2 of the PDP provides a summary of the radiation levels for the PCRV and its components. PSC has identified several additional surveys and samplings to be conducted once the fuel is removed to confirm the radiation levels. In addition, the radiological status of the site and surrounding areas has been monitored during the entire life of the facility-through the radiologic,1 environmental monitoring program. The results are included in the PDP. The staff concludes that PSC has provided sufficient information on the radio-logical status of the plant to meet the requirements of 10 CFR 20.201 to survey the facility for radiological hazards. 3.2 Radiation Protection The staff reviewed the licensee's radiation protection program and the licensee's commitment to the protection of the workers and public during decommissioning and evaluated the task and activities that_would be required to support decontamination. The staff relied on Regulatory Guide DG-1005 and applicable sections of 10 CFR Parts 20 and 50 for review guidance. Section 3.2 of the PDP was prepared to be consistent with NUREG-0761, which provides-guidance for the content of a radiation protection plan. It also in ' corporates the guidance contained in Regulatory Guides 8.8 and 8.10. Section 3.2 of the PDP also stated that the radiation protection program for FSV will incorporate the requi sments of the 1991 revision to 10 CFR Part 20 no later than January 1,1994, in accordance with the current schedule for implementa-tion. However, depending.on the decommissioning status at that time, PSC may apply for an exemption to complete decommissier' ; under the current Part 20. a e
~y a w e- w v e- ,- - an - .-+e-.n--- - - , -mem ,e - -
4 12 PSC's radiation exposure estimates are discussed in Section 2.1 of this-report. In support of the ALARA goals, PSC will develop and implement the respiratory protection program in accordance with-10 CFR Part 20, Regu;atory Guide 8.15, and NUREG-0041. In addition, the Radiation Protection Managers (RPMs) and radiation protection staff will be qualified in accordance with Regulatory Guide 1.8 and ANS/ ANSI 3.1. The administrative organization and functional responsibilities for implemen-tation of the radiation protection program are described in Section 3.2.3 of the PDP and in TS 5.0. The PSC (RPM) is assigned primary responsibility for implementation of the program with administration of the Westinghouse team radiation protection activities under the Westinghouse Project Radiation Protection Manager (PRPM). The PSC RPM will have direct communication inter-face with the PRPM and maintain overall responsibility for the radiation protection program. Adequate radiation protection staffing will be maintained consistent with the decommissioning activities in progress. The radiation protection staff and the RPMs will be qualified and trained in accordance with Regulatory Guide 1.8 and ANS/ ANSI 3.1. Section 3.2.4 of the PDP describes the radiation protection ini.ial training qualification, and retraining program. Appropriate training will be provideJ for non-radiation workers, radiation workers, and radiation protection person-nel in accordance with 10 CFR Part 19 and applicable guidance contained in Regulatory Guides 8.13, 8.27, and 8.29. The content of radiation worker training also will be consistent with Appendix A of NUREG-0761. Section 3.2.3 of the DDP describes the radiation dose control elements to be incorporated in decommissioning radiation protection procedures. These in-clude controlling sources of radiation, controlling access to areas containing radioactive materials, using radiation work permits for administrative control of personnel entering and working in radiological areas, measuring radiation exposures of workers, controlling and monitoring internal doses, and adminis-tering a program to maintain occupational doses ALARA. Appropriate caution signs and labels will be provided in accordance with 10 CFR 20.203 and 20,204. All project workers entering radiologically controlled areas will be required to wear external radiation monitoring devices consisting of thermoluminescent dosimeters (TLDs) and self-reading or digital alarming dosimeters as described in Section 3.2.5.6 of the PDP. The TLDs will be processed at an appropriate frequency by an outside vendor accredited by the National Voluntary Laboratory Accreditation Program. Whole-body counts of all radiation workers will be conducted on a scheduled basis and indirect bioassay mer.urements will be made-as necessary to access the intake of radioactive materials in accordance with 10 CFR 20.103. . The health physics instruments and equipment used in the radiation protection program are described in Chapter '. of the PDP and include portable radiation survey instruments, personnel monitoring equipment, air samplers, respiratory protection equipment, and protective. clothing. Table 3.2-2 of the PDP provides a summary of the types and models of equipment used in the radiation protection program. Section 3.2.8.2 of the PDP identifies the procedures for
13 calibration and response checks of the radiation monitoring equipment and air sampling equipment. Radiological surveys will be conducted with appropriate instruments in accordance with 10 CFR Part 20. All radioactive material entering nr leaving the radiologically restricted-areas will be coatrslieu as described in Section 3.2.6 of the PDP. Interim storage of radioactive materials ano orocessing of liquids containing radio-active materials will require a safety evaluation and will be in compliance with NRC Generic Letter 81-38. Materials and equipment released from radio-logically controlled areas for unrestricted use will not contain detectable amounts of radioactive material as determined in accordance with the guidance of NRC Circular 81-07 and NRC information Notice 85-92. Radioactive liquid and gaseous effluent releases will be monitored'and con-trolled using installed plant equipment in accordance with TS 5.4.4 and the methodology contained in the Offsite Dose Calculation Manual (0DCM), in con-formance with 10 CFR Part 20, Appendix A and Appendix I to 10 CFR Part 50, and 40 CFR Part 190. The ODCM will be used to establish set points for FSV effluent radiation monitors so that the concentration limits at the site boundary are within limits of 10 CFR Part 20. The ODCM also will be used to establish the methods for periodic assessment of doses to individuals from routine gasoous and liquid effluents to demonstrate compliance with Appendix 1 to 10 CFR Part 50. This will limit the average concentrations of radioactive materials released in effluents to unrestricted areas to a small fraction of the limits in Appendix B, Table II, Columns I and 2 to 10 CFR Part 20. The resulting individual doses per year for gaseous effluents will not exceed 5 mrem to the total body and 15 mrem to the skin or 15 mrem to any organ from
~
particulate radioactivity. For liquid effluents, including tritium, the resulting doses per year will not exceed 3 mrem to the total body or 10 mrem to any organ. Compliance with Appendix I to 10 CFR Part 50 will ensure that the dose from tritium in the drinking water pathway will not exceed the standard in 40 CFR 141. The effect on radiological conditions in the environ-ment as a result of decommissionir.g activities will be determined by continua-tion of specific parts of the existing radiological environmental monitoring program that will monitor area radiation, water samples, air samples, and vegetation samples. The staff concludes that the radiation protection program provides sufficient control of radioactive materials during decommissioning and meets the require-ments of 10 CFR 50.82(b)(2) regarding the description of th:e controls and limits on procedures and equipment to protect occupational and public health and safety. 3.3 Radioactive Waste Manaaement Section 3.3 of the PDP provides detailed information on the technologies, equipment, and procedures to be implemented for the management of radioactive waste during the decommissioning of FSV. i
-~
14 3.3.1 Spent Fuel Disposal Although not directly related to proposed decommissioning plans, the following information is provided on the disposition of the FSV spent fuel. PSC's pre-ferred plan to manage spent fuel is to ship the spent fuel to the DOE facility in Idaho. A three-party agreement between PSC, General Atomic, and the Atomic Energy Commission provided storage for eight segments of FSV spent fuel at the Idaho National Engineering Laboratory (INEL) facility. To date, PSC has shipped three segments of spent fuel to INEL as a result of three previous refuelings. Because of the uncertain schedule for shipping of spent fuel to Idaho or other DOE f acilities, PSC pursued an alternate plan end licensed, constt ucted, and is operating an ISFSI that is separately licensed under 10 CFR 72. The ISFSI facility is located immediately adjacent to the current site and the location is outside the plant's existing protected area and is approximately 1500 feet northeast of the reactor building. The ISFSI, using the modular vault dry store system, is designed to store all of the remaining iSV spent fuel, up to 37 metal-clad reflector blocks (MCRBs) and up to 6 neutron sources. FSV is currently utilizing the ISFSI alternative and all remaining spent fuel has been transferred to the ISFSI. 3.3.2 Radioacti.e Waste Processing During the FSV decommissioning project, the PCRV cavity will be flooded with water to provide sh.elding and contamination control. Flooding the PCRV will result in the release of radionuclides (that exist in the PCRV as a result of activation and plateout) into the water. The radionuclide of primary concern is tritium. Part of the tritium inventory is expected to leach out of the graphite blocks into the wcter and tritium cannot be removed by conventional processing means employed by the PCRV shield water system. The amount of tritium to be handled by the PCRV shield water system and potential exposure to personnel will depend on both the total amount of tritium present in the graphite and other components inside the PCRV and the fraction that is released to the water. Chapter 4 of the Environmental Report Supplement provides a detailed discus-sion of the effect resulting from release of tritium. Because the tritium concentrations in the PCRV graphite components and the rate at which the tritium leaches into the water from the graphite cannot be easily measured, the amount of tritium that enters the PCRV w:tter has been estimated, based on a conservative calculation of the total amount of tritium produced during power operation (i.e., 100,000 curies [Ci]) and actual measurements of tritium leach rates from British Magnox reactor graphite. PSC estimates that approxi-mately 500 Ci (or 0.5 percent of the total tritium inventory) will enter the water. The PCRV shield water system will process this tritium inventory for discharge using the existing liquid effluent discharge path and dilution. 1
l l 15 The maximum tritium inventory in the graphite that could exist in the PCRV when it is flooded is: 39EGA Curies large permanent side reflectors 82,588 Boronated side spacer blocks 11,532 Removable hexagonal reflector blocks 3,500 Core support blocks and bottom reflectors with hastelloy cans 48 Total: 97,638 for the purposes of estimating the amount of tritium in the graphite, a tri- - tium inventory of 100,000 Ci is assumed. The 100,000 Ci inventory was based on the March 1992 Actuation Analysis, EE-DEC-0010, for FSV. Data on tritium leaching fro. graphite obtained by the British was tne basis for the estimate of the fraction of the tritium inventory likely to be leached from the FSV graphite after the PCRV is flooded. These British measurements . were made in support of decommissioning of the Magnox plants by P. B. WoMam ! and I. G. Pugh. For the two samples of British graphite that were tested in demineralized water, the leach rate of the tritium was measured to decrease with time starting at about 0.1 percent per day and declining to below 0.0001 percent per day after several months. Applying these values to FSV, a curve of tritium release rate versus time was prepared with a cumulative tritium release rate of 0.5 percent of the tritium inventory in the graphite released in about the first month after flooding the PCRV. Use of this release rate results in a release of 500 Ci from the graphite and absorbed by-the water, based on an assumed initial tritium inventory of 100,000 Ci in the core graphite. The large side reflector blocks and the boronated side reflector blocks at FSV are made of commercial grade "lM graphite and have about 50 times as much lithium per unit volume as the British reactor grade graphite. Therefore, the j production rate for tritium during reactor operations at FSV was also 50 times e as much. The HLM graphite at FSV differs from the reactor-grade British graphite in other aspects also and no confirming tritium release tests have been done on the FSV HLM graphite or any other similar, commercial grade - graphite. The entire analysis is based on the assumption that the tritium leaching properties of the FSV HLM graphite are expected to result in a more conservative behavior than the British Magnox test samples. HLM surface-to-volume ratios are significantly lower, indicating that HLM graphite water ingress will not occur as rapidly and tritium migration to the graphite surface will take significantly longer. The densities of the irradiated HLM are greater than the British graphite samples, indicating. lower porosity and a lower leach rate in the HLM graphite as a result of density. In addition, effect on increased porosity should be greater in the British samples than in the HLM because the effects of reactor power history favor the HLM graphite. Therefore, with no applicable confirming tests for the 0.5 percent (500 Ci) release, PSC has established additional administrative limits on the amount and the rate that tritium could be released through dilution of the shield water.
I 16 PSC established administrative controls to limit the total tritium release to 8000 Ci and to restrict the tritium release rate, following dilution, to 20,000 picoCi per liter in the South Platte River (the U.S. Environmental Protection Agency's { EPA's] average annual concentration limit for tritium in drinking water, 40 CFR 141.16). Technical Specification 5.4.4 establishes 10 CFR Part 50 Appendix 1, ALARA guidelines for tritium release of 3.0 mrem per year to any member of the public. If more than 8000 Ci of tritium are released to the shield water, PSC has stated that it would solidify the water and dispose of it as a solid waste. PSC estimated that filling the PCRV will require approximately 325,000 gallons of water. Filling of the PCRV will be stopped at predetermined levels (1/4 core incret ets) to allow tritium sampling and analysis. No discharge will be made until tae trend of tritium concentration is determined. The initial concentration of tritium in the PCRV (approximately 5 days after fill) is estimated to be less than 0.40 pCi/ml, based on 500 Ci of tritium diluted in 325,000 gallons of water. The decommissioning technical specifications require that the PCRV water be sampled and analyzeJ daily for tritium concentrations during the initial fill of the PCRV. Sample frequency may be reduced to weekly after the tritium concentration has decreased to less than 0.1 pCi/cc. Limits have been estab-lished in the decommissioning technical specifications to ensure that tritium activity concentrations in the PCRV shield water system will not exceed those postulated in the decommissioning accident analyses. TS 3.4, "PCRV Shielding Water Tritium Concentration," established specific requirements regarding tritium concentration and frequency of the analysis of the PCRV shield water. Because the entire estimate of the release is based on theoretical analysis, PSC assessed what the effects might be if the maximum theoretical amount of tritium (100,000 Ci) is released into the PCRV shield water, including effects on air handling, tritiated water disposal, contamination, and personnel pro-tection. If the 100,000 Ci is released, the licensee has allowed sufficient funding to solidify the tritiated water, and ship it to a low-level waste disposal site, Allowing for this case, decommissioning can proceed and will be accomplished within the decommissioning cost estimate previously submitted to the NRC. In addition, with considerations for the worst credible accident and this extreme case, the staff finds that decommissioning can be accom-plished without undue risk to the safety of the public. In conclusion the assumptions regarding the amount of tritium released into the water are reasonable, and the worst-case scenario was analyzed for the entire 100,000 Ci of tritium released into the PCRV. The staff concludes the requirements of 10 CFR 50.75(f)(2) have been adequately addressed. The environmental effects of the volume of tritium are addressed in detail in the Environmental Report Supplement and the staff's Environmental Assessment. 3.3.3 Radioactive Waste Disposal PSC initially estimated the processed and volume-reduced radioactively contaminated saste for disposal as 100,072 ft), with 99,219 ft3 from the PCRV 1
17 and associated operations, and 853 ft3 from B0P. PSC stated in Section 3.3.3 of the PDP that it is negotiating a contract between the Rocky Mountain Compact (RMC) Board and the Northwest Compact Board to allow access for the waste generated from RMC States to the existing Northwest Compact disposal facility beginning in January 1993. In support of this effort, PSC has added an additional 512,441,000 to the standby trust agreement to cover the addi-tional cost of disposal. The waste from the PCRV consists of activated concrete, graphite blocks, other activated components, miscellaneous equipment and piping, and concrete rubble. The PCRV waste is contaminated principally with Fe-55, tritium, and Co-60. The waste from the BOP consists of tanks; ((1 pumps; heating, ventilation, and air conditioning (HVAC) filters; and miscel-laneous equipment and piping. There also may be radioactively contaminated asbestos. After processing and volume reduction, PSC estimated that the _ volume of radioactive waste will be segregated into the following categories: Class Volume-(cubic-feet) A 84,000 B 15,000 C 1,000 PSC stated that, because of high concentrations of cobalt-60 in the keyed-top reflettor control rod elements, as much as 400 ft3 of Class C wastes may be reclassified as greater than Class C (GTCC). The PDP has stated that waste - volume estimates may change as decommissioning operations proceed. Tables 2 and 3 of this report provide summaries of the estimated volume of wastes and the classification, number, and type of containers necessary for shipping and disposal. PSC also stated that, if mixed wastes are generated, they will be managed according to Subtitle C of the Resource Conservation and Recovery Act (RCRA). PSC also stated that it did not intend to petition the EPA to delist any mixed waste. Section 3.3.3.6 of the PDP addresses the storage of the waste at FSV. The wastes will be stored at various plant locations depending on the classifica-tion of the wastes. As an example, the ISFSI may be used for storage of the keyed-top reflector control rod elements if they are GTCC. Storage of these elements at the ISFSI is permitted by the ISFSI license (SNM-2504). In addition the fuel ttorage building, compactor building, ;he reactor building, as well as additional areas, will be available for storage. The waste storage will be based on guidelines in NRC Generic Letter 81-38 and Appendix 11.4-A to NUREG-0800. The staff finds that pac's analyses and estimates of the volumes of waste generated during decommissioning as well as the waste classification of the proposed waste and practices and methods for meeting the transportation requirements are reasonable and consistent with the applicable requirements of 10 CFR Parts 20, 61, 71, and the requirements of 10 CFR 50.82(b)(1)(iii).
18 , Table 2 FSV PCRV Vs.te Volume' Estimates Volume Con-Item / System Class LSA Number (ft3)* tainers Region constraint device and pin C No 84 200 2 Metal control rod reflectors - C No 37 400 3 Metal block, non control rod B No 276 2000 13 Defueling blocks A Yes 1482 7200 75 Top reflector graphite blocks A No 1215 1500 8 Bottom reflector graphite blocks A No 1215 1400 8 Radial reflector (perm. and rmvble) A No 480 1900 9 Large reflector blocks B No 312 12600 50 Half-size reflector blocks A No 312 2100 8 Upper reflector keys (carbon steel) A No 24 200 '2 Side spacer blocks with boron rods B No 197 500 66 rods removed A No 1152 2400 25 Bottom reflector blocks with cans (Hastelloy) C No 20061 375 50 cans removed A 276 800 8 Lower reficctor keys (Hastelloy) B No 24 200 1 Core support blocks A Yes 61 1500 15 Core support posts A Yes 183 200 2-Core support floor columns A Yes 12 600 7 Misc steel from beneath CSF A Yes 1000 10 Metal on large side reflector A Yes 24 100 1 Core barrel A Yes 1 1400 31 Lower plenum insulation A Yes 900 10 Silica blocks (25,000 lbs) A Yes 500 12 Concrete - top A Yes 3700 9 Concrete - CSF A Yes 6200 15 Concrete - side A Yes 19000 45 Concrete rubble - Jackhammer A Yes 700 16 Hist Inconel parts on CSF A No 400 - 5 Concrete cutting debris - top A Yes 200 Concrete cutting debris - CSF A Yes 200 8 Concrete cutting debris - side A Yes- 300 Helium purifiers in PCKV head A Yes 10 500 5 Helium diffusers A Yes 4 1750 4 Helium circ shutoff -valve assembly A Yes 4 200 2 Helium bellows A Yes 12 1600 12 Steam generators A Yes 12 21000 12 Themocouples and guide tubes B No 100 1 Lower floor / appurtenances A Yes 1200 42 Platform / handling tools / jib cranes A Yes- 576 10 Crane cable / drum /3 bucket inverters A Yes- 500 5 Misc containers A Yes 300 3 PCRV water system A Yes 2100 2 Resins - solidify, ship, bury A** No 20 2/30 20 Misc soft waste A Yes 13000 125 PCRV Totals: 114,500 740
- Estimated pre-volume reduced quantity.
** Estimated Burial Class - Specific burial class ident.Tication may require additional analysis with 10 CFR 61.
19 Table 3 FSV B0P Waste Volume Estimates Volume item / System Class LSA Number (FT3)* Reaction isolation valves A Yes 5 1000 Refueling sleeves A Yes 2 200 Sand from fuel storage wells A Yes 800 Sand from equipment storage wells A Yes 200 Sand from helium regeneration pit A Yes 100 Auxiliary transfer casw sand A Yes 100 Hot cell facility A Yes 400 Sand from hot cell facility A Yes 500 Core support vent filters A Yes 1C Gaseous waste surge tanks A Yes 1 1000 Gaseous waste compressors A Yes 2 2100 Liquid waste monitor tank A Yes 1 600 Liquid waste demineralizers A Yes 2 200 Liquid waste receivers A Yes 2 1100 Liquid waste sump (sand) A Yes 20 Liquid waste transfer pumps A Yes 2 100 Liquid waste sump pumps ' A Yes 2 10 Liquid waste filters A "s 2 -10 Decon solution tank A Yes 1 400 Decon recycle pump A Yes 1 2 Decon chem supply pump A Yes 1 2 Purified helium filters A Yes 2 10 Helium removal filter A Yes 1 100 Helium getter units A Yes 2 10 HVAC filters A Yes 1000 Fuel handling machine A Yes 200 Fuel handling machine components A Yes 400 Small and large bore piping A Yes 600 Reactor building drain system A Yes 100 Instrumentation and controls A Yes 200 B0P Totals 11,500
- Estimated pre-volume reduced quantities.
3.4 Accident Analysis In Section 3.4 of the PDP, PSC evaluated the effect of potential decommis-sioning accidents at FSV on the health and safety of the public. The activi-ties, equipment, and circumstances associated with decommissioning are differ-ent from those evaluated in the FSV Final Safety Analysis Report for power operations and refueling.
-The risk of accidents resulting in a radiological release during decommission-ing_ activities was considerably less than during plant operation be- se all spent fuel will be f moved from the reactor building. Therefore, v non-operations accident scenarios will be evaluated in this section.
20 The type of postulated accident and the resultant doses to an individual at the emergency planning zone (EPZ.100 Meter Minimum) are given below. l 2-lipur Dose -(mrem)_ kqidr_n.1 Whole-B.21Y Organ Droppin.) of concrete rubble 4.92 58.9 (bone) Heavy load drop 7.10 202 (lung) Fire 121 215 (lung) Loss of PCRV shielding water 34.8 34.8 (lung) Loss of power 1.54 40.0 (lung) , Natural dis ster Itornado) 0.58 16.8 (lung) Dropping of steat generator -. primary module 8.3 40.7 (lung) The results of the accident scenarios postulated for FSV decommissioning indi-cate radiation exposures to the general public are very low. The resulting analysis show that the radiological consequences at the EPZ are within the 10 CFR Part 100 guidelines and are only a small fraction of the EPA Protective ! Action Guidelines (EPA-520/1-75-001-A) and would therefore require no offsite response to the accident. The staff compared the accident scenarios and releases to accidents in NUREG/CR-0130 and concludes the scenarios analyzed are representstive of accidents that could occur at FSV during decommissioning and that none of the accidents has potential consequences (radiation doses) in excess of the EPA Protective Action Guidelines. 3.5 Industrial Safety The proposed decommissioning activities involve a number of routine industrial safety hazards that are subject to regulation by other Federal agencies, in ese areas, the NRC staff has not reviewed the licensee's decommissioning pian for regulaDry compliance, limiting its review to radiological aspects only. Nevertheless, the staff has noted the presence of these hazards. 3.6 Asbestos PSC's asbestos removal procedures will follow procedures for the safe removal and disposal of materials containing asbestos required by EPA regulations pro-mulgated as " National Emission Standards for Hazardous Air Pollutants" (40 CFR Part 61) and Occupational Safety and Health Administration (OSHA) safe work practices required under 29 CFR 1926.58. PSC will provide respiratory protec-tion in accordance with OSHA regulation 29 CFR 1910.134. The BOP systems were surveyed and material containing asbestos was found on some piping in the helium purification system and radioactive waste aas system, Approximately 1500 linear feet of metal Jacketed material will be required to be removed, packaged, and disposed. Two industrial hygienists and necessary industrial services will be on site to support this operation. The asbestos removal is addressed in detail in Section 3.5 o' the Environmental Report Supplement and in the FSV decommissioning cost estimate (WBS 2.4.1).
21 The staf f concludes that the removal of asbestos is adequately addressed and follows the procedures required by 29 CFR 1926.58. 4 flNAL RADIATION SURVEY PLAN Chapter 4 of the PDP describes the methodology and criteria that will be used in performing the final surveys at FSV. It included a definition of the residual radioactivity limits, radiation survey methods, materials release criteria, and the site release criteria. The final radiation survey plan is based on the guidsnce provided on NUREG/CR-2082, in addition to the criteria discussed for unrestricted release. PSC will follow the guidance in Regulatory Guide 1.86 for both loose and fixed-surface contamination, adopting NRC's guidance of 5 microR/hr above background. In addition, equipment and materials will be released according to NRC Circular 81-07 and NRC Information Notice 85-92. PSC stated that the , effective dose equivalent for an individual will be less that 10 mrem /yr for residual contamination in groundwater and soil. The staff considers these criteria to be reasonable and acceptable. The staff concludes that the final survey plan meets the requirements of 10 CFR 50.82(b)(3) and is reasonable and acceptable. 5 bPDATED COST ESTlHATE FOR DECOMMISSIONING lt is the responsibility of the NRC to determine if the cost estimate provided in the PDP provides a reasonable basis for sufficient funding to complete decommissioning of the facility. The review of the cost estimate for decom-missioning the FSV facility was based on independent estimates and comparison of several cost activities to be conducted at this facility to similar activities conducted at other facilities. The review included an evaluation ot the cost assumptions used, major decommissioning activities and tasks, dismantlement and decontamination costs, volumes of waste to be removed, disposal costs, transportation costs, equipment costs, and labor rates. The basis for the evaluation was similar information provided in the Pathfinder decommissioning cost estimate, the Shoreham decommissioning cost estimate, the "1992 Heans Building Construction Cost Data," the " Dodge Manual for Building Construction Cost Data 1984," and in NUREG/CR-0130. All cost information was escalated to 1991 dollars using an inflation rate of 5 percent. The estimated cost of $157,472,700 represents a reasonable estimate of decommissioning the FSV facility. While FSV is an HTGR, many activities that will be conducted to decontaminate and dismantle this facility are similar to activities conducted at other re-actor facilities that are or have been decommissioned. In addition, several activities that support decommissioning are standard construction practices. The staff reviewed several areas to ensure the estimated cost to DECON the FSV facility are reasonable, for example, the cost of removal of contaminated pumps (1,000-10,000 lbs) was compared to similar activity that was conducted at the Pathfinder facility. The removal of similar pumps (1,000-10,000 lbs) at Pathfinder, cost approximately $1900.00. The removal of similar pumps for
m _ __ _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ _ - _ _ _ __ ._ _ _ _ _ _ _ i 22 the FSV liquid waste system is estimated to be $3055.00. Even after adjust-ments for regional differences and inflation, the FSV costs wc.a greater than the estimated cost at Pathfinder. To date, the actual costs for decommission-ing at Pathfinder have been consistent with the initial estimate and, there-fore, represents an example cost for comparison. The staff compared the labor rates summarized in Table 3.1-1 of the PDP cost estimate to the labor rates for Pathfinder, which were escalated at 5 percent per year to 1991 dollars, and the city cost indexes in "1992 Means Building Construction Cost Data" and found them reasonable. In addition, the staff compared labor rates for FSV to Shoreham, using the city cost indexes listed : in the "1992 Heans Building Construction Cost Data" and found ther reasonable. The staff compared PSC's estimated equipment rental costs to the cost for equipment rentals listed in "1992 Heans Building Construction Cost Data" snd adjusted the costs using the city cost index. It examined the rental costs of-many different types and sizes of equipment ranging from small air compressors to Sp-70 ton cranes. For example, the cost to rent and run a crane (RT 15-24 T) ivr PCRV work was estimated to be about $30 per hour for r$v compared to the industry estimate of $100 per hour. After adjustment for regional dif-forences, the FSV cost was considerably less. The estimated cost for rental of a 750-cfm compressor used for blasting and for running tools was $100 for 40 hours of use. Typical industry rate fcr a similar compressor for. a weeks rental (40 hours) was estimated to be $865. Even after adjustmen' for reg-ional differences, the FSV estimate war less. M. K. Fergerson/We. inghouse stated that equipment costs used in t', cost estimate included depreciation costs on company-owned equipment and these costs were considerably less'than actual rental costs. if the company is required to rent equipment because they do not own a particular piece of equipment,-the additional costs will be taken from the $23 million contingency included in the cost estimate. Estimated operating costs were consistent with industry standards. The estimated costs of removing many of the BOP systems at FSV were compared to those at Pathfinder. For example, the cost to remove piping up to 5 inches in diameter for the helium purification system is $74.83 per foot compared to the cost or removing 2- to 8-inch piping at Pathfinder of $30.54 per foot. The estimated cost for removir.g the FSV cooling water system piping, which , consisted of over 47,000 feet of piping ranging in diameter from 0.5 to 20 inches, averaged $108.79 per-foot. The estimated cost to remove piping greater than 8 inches in diameter at Pathfinder was $60.36 per foot. After adjustments for inflation and regional differences, the FSV estimates were considered conservative. In addition, the staff compared the estimated cost of removing contaminated concrete from the PCRV at FSV to the actual cost of removing c.ontaminated reinforced concrete at Pathfinder. Although the methods of removing the concrete were different, the cost of removal should be similar. PSC estimated
- the cost for cutting the core support floor at approximately $1120.00 per cubic yard compared to the cost of removing contaminated concrete at Pathfind-er of $650.00 per cubic yard. Therefore, the estimate to remove the contami-ncted conr. rete at FSV is conservative. T!e staff also reviewed the cost-for disposal of the 100,000 ft3 of radioactive materials and finds it reasonable
23 on the basis of the proposed contractual agreement with Richland site, and current disposal costs. The staff concludes that this cost estimate for decommissioning the FSV facil-ity meets the requirement of 10 CFR 50.82(b)(4). 6 TECHNICAL SPECIFICATIONS IN PLACE DURING DECOMMISSIONING The staff reviewed PSC's proposed TSs that will be in place for decommission-ing and found them acceptable. The TSs are incorporated in Appendix _A to the license and provide requirements for reactor building integrity, radiation monitoring instrumentation, PCRV tritium concentration and surveillance, administrative organization, radiation protection program, access control, effluent control, radiological environmentai monitoring and reporting. The staff has determined that the TSs adequi..ely address decommissioning activi-ties and meet th9 requirements of 10 CfR 50.82(b)(5).
) QUAllTY ASSURANCE PSC's quality assurance program (QAP) is described in Chapter 7 of the PDP and is designed to meet the requirements of Appendix-B to 10 CFR Part 50.
The Corporate Vice-President, Nuclear 0)erations of PSC is the corporate offi-cer responsible for implementation of tie QAP. The Vice President has direct access to the President of PSC. The Project Quality Assurance Manager reports directly to the Vice-President. The staff concludes that the Quality Assur-ance Plan is adequate and meets the requirements of 10 CFR 50.82(b)(5) and the guidelines of SRP Section 17.2. 8 DECOMMISSIONING ACCESS CONTF PLAN PSC's access control plan is described in Chapter 8 of the PDP. It is designed to meet the requirements of 10 CFR 20.105 and follows the guidance in NRC Regulatory Ct.ido 1.86. The staff concludes this access control plan is < reasonable and acceptable. 9 DECOMMISS!0NING EMERGENCY RESPONSE PLAN PSC's emergency response plan has been reviewed and approved by the NRC staff on March 3, 1992, separately from the PDP. 10 DECOMMISSIONING FIRE PROTECTION PLAN PSC's fire protection program has been revised and revisions related to decom-missioning were approved on March 28, 1992 and June 5, 1992. 11 LICENSE CONDITION 2.C.(5) In its submittal PSC also proposed to revise license condition 2.C.(5) to- _ allow the temporary return of its own byproduct and special nuclear material that is shipped offsite during decommissioning for compaction, testing or disposal but has to be returned to the site for repackaging before final
24 disposal. We have reviewed the proposed change to license condition 2.C.(5) and have determined that it is consistent with the intent of a proposed rule change to 10 CFR Part 50.54(eo) which will soon be finalized. Thus the change is an acceptable and appropriate modification of the existing license condition. 12 STATE C0f45ULTA110!1 in accordance with the Commissions regulations, the Colorado State official was notified of the proposed issuance of the order and amendn. ant. Questions on tritium rele,'ne from the State of Colorado Water Quality Control Division were satisfatttifly resolved (Environmental Assessment, Section 6). The state official had no additional comments. 13 Ef4VIRONMENTAL C0!iSIDERA110ft This action involves an Order that authorizes Public Service Company of Colorado to decommission FSV in accordance with it's Proposed Decommissioning plan and an amendment of License No. DPR-34 to establish TS requirements for the decommissioning activities. An Environmental Assessment has berr completed by the staff. A flotice of Issuance of Environmental Assessment and finding of No Significant impact has beer published in the FEDERAL REGISTER (57 FR ). Based on the Environmental Assessment the Commission concluded that the proposed order and amendment will not have a significant affect on the quality of the human environment. 14 00!4CLUS10flS The Commission has concluded, based on considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by decommissioning FSV in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the Order and amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors: C.L. Pittiglio Jr. J.B. Baird P.B. Erickson Date:
25 13 REFERENCES American Nuclear Society /American National Standards Institute, Standard ANS/ ANSI 3.1-1981, " Selection, Qualification and Training of Personnel for Nuclear Power Plants."
"1992 Means Building Construction Data," 50th Annual, R.S. Means Company Inc.
*0ffsite Dose Calculation Manual," Procedure SUSMAT-2, Issue 19, April ?2, 1992.
" Pathfinder Decommissioning Plan Cost Estimates Calculation Notebook,"
Northern States Power Company, February 28, 1990. Public Service Company of Colorado, " Fort St. Vrain Decommissioning Cost Estimate," June 1991.
-- , Fort St. Vrain Environmental Report Supplement, " Post Operating License Stage - Decommissioning," April 1992.
--- , " Fort St. Vrain Proposed Decommissioning Plan," June 28, 1991.
-- , Decommissioning Technical Specification for Fort St. Vrain Unit 1, Docket NO. 50-267, Appendix A.
"Shoreham Decommissioning Book for Cost Estimates," Shoreham Decommissioning Plan, Long Island Lighting Company, October 1991.
U.S. Environmental Protection Agency, EPA-520/1-75-001-A, " Manual of Protec-tion Action Guidelines and Protective Action for Nuclear Incidents," January 1990. U.S. Government Printing Office, Code of Federal Reaulations, Title 10,
" Energy," U.S. Nuclear Regulatory Commission, Washington, DC.
-- , Title 29, " Labor," Occupational Safety and Health Administration, Washington, DC.
-- , Title 40, " Protection of the Environment," U.S. Environmental Protection Agency, Washington, DC.
U.S. Nuclear Regulatory Commission, Circular 81-07, " Control of Radioactive Contaminated Material."
-- , Draf t Regulatory Guide DG-1005, " Standard Format and Content for Decom-missioning Plans for Nuclear Reactors," September 1989.
-- , Generic Letter 81-38, " Storage of low-Level Radioactive Waste at Power Reactor Sites," November 1981.
l 1
26 -- , Information Notice 85-92, " Surveys of Waste Before Disposal for Nuclear Facilities." -- , Regulatory Guide 3.143, " Design Guidance for Radioactive Waste Management Systems, Structures and Components Installed in Light-Water-Coolad Nuclear ! Power Plants," Rev. 1, October 1979. , -- , Regulatory Guide 1.8, " Qualification and Training of Personnel for : Nuclear Power Plants," Rev. 2, April 1987. ! -- , Regulatory Guide 1.86, " Termination of Operating License for Nuclear Reactors," June 1974. -- , Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Reasonably Achievable," Rev. 3, June 1978. -- , Regulatory Guide 8.10. "0)erating Philosophy for Maintaining Occupation Exposures As Low As is Reasona)1y Achievable," Rev. 1-R, May 1977. -- , Regulatory Guide 8.13. " Instruction Concerning Prenatal Radiation Exposure," Rev. 1, November 1975.
-- , Regulatory Guide 8.15, " Acceptable Programs for Regulatory Protection,"
October 1975. -- , Regulatory Guide 8.27, " Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants," March .,81. -- , Regulatory Guide 8.29, " Instruction Concerning Risks from Occupational Radiation Exposure," July 1981. -- , NUREG-0761, " Radiation Protection Plans for Nuclear Power Reactors licansee," March 1981. -- , NUREG-0800, " Standard Review Plans for the Review Plans for the Review of Safety Analysis Reports for Nuclear Power Plants," July 1981. -- , NUREG/CR-0130, " Technology, Safety and Cost of Decommissioning a Refer-enced Pressurized Water Reactor Power Station," Addendum 4, July 1988. -- , NUREG/CR-2082, " Monitoring Compliance With Decommissioning Termination Survey Criteria," June 1981. -- , SECY-92-106,
Subject:
Action Plan To Ensure Timely Remediation of Sites Listed In the Site Decommissioning Management Plan, March 24,-1992. Woolam, P. B. and I. G. Pugh,1978, "The Radioactive Inventory of a Decommis-- stoned Magnox Power Station Structure."
.. _.- . . . _ _ _ - - . . . . ~ . ~ . - - - _ . - - _ _ _ ~ . - _ - . ~ . ~ ~ - . . . ~ . - . . - - - - - ~ _ . _ _
l l 00 0 c
+ jo, UNITED STATES
' . t,p, NUCLEAR REGULATORY COMMISSION i 8 }-
; W A$HINGToN, o. C4 20566 !
t,4....$ i ENVIRONMENTAL ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING ORDER AUTHORIZING DECOMMISSIONING OF ; FORT ST, VRAIN NUCLEAR GENERATING STATION i PUBLIC SERVICE COMPANY OF COLORADO t DOCKET NO. 50-267 , 1.0 IDENilflCATION OF PROPOSED-ACTION AND
SUMMARY
Fort St. v. ;in Nuclear Generating Station (fSV) is a 842 MW-thermal, High Temperature Gas (helium) Cooled Reactor (HTGR) that operated frca January 31, 1974 to August 18, 1989. Public Service Company of Colorado (the licensee or PSC) shut down FSV due to control rod drive failures and subsequently made the shutdown permanent due to a discovery of degradation of the steam generator ring headers. Pursuant to 10 CFR 50.82, PSC submitted a Proposed Decommis-4 sioning Plan (PDP) on November 5, 1990. The NRC staff submitted several Requests for Additional Information (RAls) to PSC. In response to the RAls, PSC revised the PDP on December 17 and 21, 1990, January 14, 1991, April 15 and 26, 1991, May 15, 1991, June 6 and 17, 1991, July 1, 1991, August 28 and ._ 30,-1991, November 15, 1991, December 6, 1991, January 9, 1992, March 19, 1992, April 17, 1992, and September 25, 1992. In support.of.the PDP, PSC - submitted an Environmental Report (ER)-Supplement on July 10, 1991, as revised March 20, 1992, April 30, 1992, June 24, 1992, and September I and 18, 1992. Environmental review requirements for decommissioning licensed reactors were clarified by the NRC when it amended the regulations on June 27, 1988 (53 FR 24018), " General Requirements for Decommissioning Nuclear f acilities." In so doing, the NRC eliminated the mandatory requirement for an' environmental-
' impact statement (EIS) after considering decommissioning impacts in its Final Generic Environmental Impact Statement on Decommissioning of Nuclear-Facilities (FGEIS). The FGEIS included both' fuel cycle and non-fuel-cycle nuclear facilities, including pressurized water reactors-(PWRs) and boiling-water reactors (BWRs) but did not *specifically address-the HTGR technology of FSV.
The environmental aspects of decommissioning an HTGR were, however, considered for Peach Bottom Unit I decommissioning in the-Commission's environmental. assessment of April 24,1990, (55 FR;17320), _with respect to a renewal of the plant's possession _only license.. In that environmental assessment the Commis-sion concluded that the proposed action would not have a significant.effect on the quality of the' human environment and determined that an environmental impact statement would not be prepared. The staff used the FGEIS and the Peach Bottom 1 Environmental Assessment as basis for this assessment.
~
lhe new regulations still require an ER Supplement to update the licensee's original ER and reflec, any new information or significant environmental change associated with the proposed decommissioning. The Commission's final Environmental Statement (fES) for fSV of August 1972 evaluated the environ-mental effects of operating the facility. The licensee's ER Supplement references the licensee's original ER, the Commission's FES and the FLEIS as appropriate. Environmental assessments by the NRC staff are required by 10 CFR 51.95(b) for power reactor decommissioning, and as presented in the statement of consid-erations for the amendment to the decommissioning regulations (53 FR 24039),
"If the impacts for a particular plant are significantly different from those studied generically because of site-specific considerations, the environmental assessment would discover those and lay the foundation for the preparation of an EIS. If the impacts for a particular plant are not significantly differ-ent, a finding of No Significant Impact would be prepared.'
PSC selected prompt dismantling (DECON) as the decommissioning alternative for FSV and the FSV ER Supplement evaluates the environmental impact of that alternative. This environmental assessment reviews PSC's ER supplement and the selected DECON alternative in accordance with 10 CFR 51.95(b). 1 Decommissioning of FSV includes the dismantirment, decontamination and disposal of radioactively activated and contaminated material, and components produced by TSV operations. Substantial partions of the plant will be dismantled and removed. However, the Reactor Building, Turbine Building, and other structures that are not radioactive _above limits acceptable for unrestricted access will remain, following completion of dismantling and decontamination activities, PSC will conduct extensive radiation serveys to verify that the plant and the FSV site meet NRC release criteria and can be released to unrestricted access. The FSV Part 50 license will be terminated following satisfacto,y completion of deconnissioning actions by PSC and verification surveys by the NRC. All spent fuel has now been removed from the reactor protected area and transferred to a recently completed Independent Spent fuel Storage Installation (ISFSI). The ISFSI is licensed separately (SNM-2505 dated November 4, 1991) under 10 CFR Part 72. The Part 72 licensing action includes an Environmental Assessment relating to tne construction and operation of the ISFSI which supports issuance of the SNM license and storage of the spent fuel at the ISFSI. 20- PURPOSE OF AND NEED TOR PROPOSED ACTION , The licensee's purpose of decommissioning as proposed is to dismantle and decontaminate FSV such that the possession only license can be terminated and the site released for unrestricted access. l Since FSV is permanently shut down, PSC is required by 10 CFR 50.82 to submit a PDP for NRC review and approval. The approval of PSC's PDP'would allow PSC l to dismantle FSV in accordance with that plan. Also, completion of the l
dismantlement would allow the FSV site to be released for unrestricted access and the license to be terminated following NRC verification of the removal of residual radioactivity to acceptable levels, i 3.0 ALTERNATIVES INCLUDING pR0p0 SED ACTION l As discussed in the fGE15 the three decommissioning alternatives for reactor f acilities are Stf STOR, [N10MB, and DECON. The environmental impacts of these alternatives as well as a "No Action" alternative are compared in Table 1 and discussed below. 3.1 SAFSTOR - Deferred Dismantlement
" SAfS'iOR is the alternative / option in which a facility is placed and maintained i in a condition that allows it to Ls safely stored and subsequently decon-taminated to levels that permit release of the property to unrestricted use.
The f acility may be lef t intact, except that all fuel must be removed from the reactor core and radioactive fluids and certain radioactive wastes removed from the site. PSC states that it did not select this alternative for FSV because of its uncertainties about future escalation of waste disposal costs, I uncertainties about decommissioning costs in 50 or 60 yr .rs and the continued maintenante, inspections and nuclear insurance costs required during the SAf510R period. 3.2 ENTOMB - Entorks nt The EN10MB option involves encasing radioactive contaminants in a strncturally long-lived material, such as concrete. The entombed structure must be appropriately maintained, and continued surveillance is required until the radioactivity is removed from the site or decays to a level that permits unrestricted use of the property. Waiting for decay is impractical at power reactors because of the long-lived radionuclides such as nickel-63 and " niobium-94 that are present in reactor components near the core. PSC has performed an activation analysis of FSV and has determined that even after 100 years of decay, sufficient radioactivity from these long-lived radionuclides will be present in reactor structures to preclude the release of FSV for unrestricted use. Entombment would make later removal of long-lived radionuclides more difficult than would other alternatives. 3.3 DECON - Prompt Dismantlement The DECON option involves prompt removal of equipment, structures, and other portions of the facility containing radioactive contaminants, or decontami-nating them to a level that permits the f acility to be released for unrestricted access. PSC has selected DECON because it allows termination of the FSV license shortly after defueling is ccmplete and thereby eliminates continued insurance, maintenance and monitoring costs. The DECON option at FSV reduces uncertainties over the cost for delayed dismantling and radioactive waste disposal and ailow; the use of the existing, experienced FSV operating staff to assist with the dismantlement. An
1 additional consideration given by PSC in selecting DEC0ll was that it was more equitable to future generations in that the current generation that received the power generated by FSV pays the dismantling cost. A disadvantage of DECON is the higher occupational radiation dose as compared to the other alternatives. Also, the DECON option at FSV will involve flooding of the Prestressed Concrete Reactor Vessel (PCRV) for radiation shielding during removal of reactor components. This flooding is projected to involve the release of some of the tritium that is present in the graphite reflector blocks (tritium is produced when trace lithium in graphite absorbs neutrons). The environmental impact of this tritium release is evaluated in Section 5.1 of this assessment. 3.4 [Lo _ Act ion The objective of decommissioning is to restore a radioactive facility to a , condition such that there is reasonable assurance that the decommissioned facility does not adversely impact public health and safety. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, some action is required. Thus, independent of the type of f acility and its level of contamination, no action, implying that the licensee would simply abandon or leave the facility after ceasing operations, is not permitted by f4RC regulations and thus is not a viable alternative for any nuclear facility that is permanently shutdown (FGEIS Section 2.4.1). 4.0 AffECTED ENVIRONMENT The area surrounding FSV is agricultural with a low population density. The major farm products are grain, feed corn, sugar beets, beef cattle, sheep and turkeys. There is also a limited amount of dairy farming in the area. fourteen oil and gas wells are located within one mile of the reactor building. The nearest town, Platteville, has a population of approximately 1500 and is located 3.5 miles from FSV. 5.0 ENVIRONMENTAL CONSEQUENCES AND MITIGATING ACTION Our purpose is to analyze the proposed decommissioning of FSV, to determine whether the potential environmental impacts are significantly different from those discussed in the FGEIS and whether any such different impacts significantly affect the human environment. Environmental aspects of the DECON alternative are evaluated in the sections that follow. 5.1 Tritium. and Other Radionuclides in Shield Water During the fort St. Vrain decommissioning p:oject, the_PCRV cavity will be flooded with water to provide shieldin,, and contamination control. Flooding _ _ _ of the PCRV will result in the release of some of the radionuc' ides (that exist in the PCRV as a result of activation and plateout) into the water. Of primary concern is tritium which, after dilution of the shield water, would be released to the aquatic envirormcht during decommissioning over a period of up to two years. ! PSC proposes to_ reduce tritium levels in the PCRV shield water by dilution and discharge as liquid effluent. PSC has provided an analysis of the tritium
l present in the graphite reflector blocks and estimated that 0.5 percent (500 Curies) of the tritium present would be released from the FSV graphite to the , shield water during the PCRV flooding operations. This estimate is based on results from tests on reactor grade graphite from a British HTGR (Magnox reactor) which demonstrated that less than 0.5 percent of the tritium present , would be released. The large side reflector blocks and the boronated side reflector blocks used at FSV are made of commercial grade, (HLM) graphite and have about 50 times as t much lithium per unit volume as does the British, reactor grade graphite. Therefore, the production rate for tritium durirg reactor operations with HLH graphite is also 50 times as much as with the British graphite, assuming the same neutron flux. The HLH graphite at FSV differs from the reactor grade British graphite in other aspects also; no confirmatory tritium release tests have been performed on the FSV HLH graphite or other similar, commercial grade graphite. We have determined that PSC's 500 curie estimate for potential tritium ri. lease represents a credible estimate for the following reasons:
- 1) PSC's estimate properly considers the higher production rate for tritium in the HLH graphite;
- 2) The British graphite samples have higher porosity than the HLH graphite and, therefore, greater potential for tritium migration; consequently, since fewer pores would reduce egress of tritium, the movement of tritium to the surface in the HLH graphite should be slower;
- 3) The British tests were conducted with dimensionally smaller samales by comparison with the HLH graphite blocks. The HLH graphite bloc (s have longer diffusion pathways and smaller surface to volume ratios which would result in a slower rate of tritium outgassing than the British tests; and,
- 4) Water has leaked into the PCRV many times during FSV operations such that the water and high humidity in contact with the reflector graphite has likely caused a significant amount of the easily removable tritium to transfer to the water vapor and subsequently to have been removed from the PCRV during a helium drying process. PSC liquid effluent reports document tritium releases of about 200 curies per year during FSV operations that may have originated in this manner.
The licensee has stated that the upper limit for tritium release is 8000 curies with a maximum of 4000 curies per year and has established administrative requirements to limit the releases to these values. The staff is using these limits as wil as other established (concentration) limits to determine the potential environmental impacts of tritium releases. Technical Specification (TS) 5.4.4 limits tritium discharged to unrestricted areas in accordance with 10 CFR Part 20, Appendix B, Table II, Column 2 (0.003 microcuries/ml). TS 5.4.4 also limits tritium release in accordance with 10 CFR Part 50, Appendix I guidance to assure that the releases of radioactive material to unrestricted areas will be kept "as low as reasonable achievable" (ALARA), For tritium in liquid effluent, the ALARA gu.deline is 3.0 millirem per year to the any member of the public. I
I The trittated shield water will be diluted with blowdown water from the cooling tower system and emptied into a 25 acre farm pond via the discharge canal (Goosequill ditch). Dilution and discharge of tritium will occur during a period cf up to two years after PCRV flooding. The estimated tritium concentration was based on a 5-year average dilution flow rate of 5 E+11 liters per year including plant process water and aaditional dilution between fort St. Vrain and the Gilcrest well (the closest downstream public water supply). Since drinking water is not taken directly from the South Platte River the concentration of tritium in the river water conservatively estimates the possible tritium concentration in drinking water wells. Assuming the maximum possible total release of 8000 curies of tritium at a rate of 4000 curies per year, the average concentration of tritium in the river near the Gilcrest well would be 8000 picocuries per liter with the - 5 E+11 liters per year dilution rate. The EPA Safe Drinking Water Standard of 40 CFR Part 141.16 is 20,000 picocuries per liter average concentration. PSC proposes to release no more than 4000 curies of tritium yearly to the environment, to release no more than 8000 curies total, to remain within the EPA Safe Drinking Water Standard, to meet 10 CfR Part 20 requirements and to remain within 10 CFR Part 50 Appendix ! ALARA guidance. Measurements will verify that offsite concentrations remain within these limits. PSC l as evaluated the following potential exposure pathways to members of the general public for a release of 4000 curies of tritium per year; (1) drinking water consumption, (2) milk consumption, (3) vegetable consumption, (4) meat and waterfowl consumption, and (5) fish consumption. The staff has reviewed the licensee's calculation methods and assumptions and finds that they are consistent with NRC's Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Evaluating Compliances with 10 CfR Part 50, Appendix 1," Revision 1, October 1977. The potential exposures ft each pathway and total exposure are listed in Table 2. The 1 millirem per year dose to the maximally exposed individual (Table 2) is a small fraction of the avarage 300 millirem received annually by members of the puolic in the United States and Canada from sources of natural background radiation. Also, PSC estimated (hat the inttgrated public exposure out to a radius of 50 miles from FSV (conservatively 17,000 persons) would be 2.5 person-rem per year if the release were 4000 curies of tritium per year or a total of 5.0 person-rem for 8000 curies released over a 2 year perica. PSC' estimated public dose due to waste transportation is 9.0 person-rem. Therefore, with the addition of 5.0 person-rem dose from a release of 8000 curies of tritium the total public dose is estimated at 14 person-rem. Public dose from o+her radionuclides in the shield water will t be significant due to their removal by the FSV demineralizer system to less tnan 1 percent of 10 CfR Part 20 limits and subsequent dilution of shield water by blowdown and river water. The 14 person-rem dose estimate is consistent with the fGEIS estimated public doses for decommissioning BWR and PWR facilities of 10 and 22 person-rem respectively. If more than 8000 curies of tritium are released to the PCRV shield water, PSC has committed to solidify the water and dispose of the resulting material in a licensed low level radioactive waste burial f acility. l
r PSC has committed in TS 5.4.4 to: 1) incorporate requirements in the FSV Offsite Site Dcse Calculation manual (ODCH) for monitoring and control of tritium concentrations; and 2) report future modifications of these requirements to the NRC in accordance with applicable ODCM change protocol. j Based on the above considerations, the staff concludes that: 1) PSC's estimate of 500 curies of tritium release is credible, (2) potential public exposures are within the estimates given in the FGEIS for light water reactors; and (3) adequate procedural controls will be implemented in the ODCM to ensure that the EPA Safe Drinking Water Standard in 40 CFR Part 141.16, 10 CfR Part 20 and Appendix 1 ALARA guidance will be met and that the total release will not exceed 8000 curies. Other radionuclides such as cobalt-60, cesium-137, strontium 90, and iron-55 will be removed by the PCRV water cleanup system domineralizer to an average concentration of less than 1.0 percent of 10 CFR Part 20, Appendix fs, Table !! values prior to their dilution in the receiving water. The staff has determined that such releases will have no significant impact on the human environment. 5.2 Status of federal m State. Regional and local Environmental Permits By letter dated September 18, 1992, PSC provided the status of required permits. State of Colorado permits have been obtained fc,r asbestos removal and disposal, service water wells, fuel oil storage, and auxiliary boilers, a A building permit was obtained from Weld County for oil storage tank. PSC has submitted applications to the State for new ground water monitoring wells and permit modification for existing monitoring wells. PSC has reached agreement with the Colorado Department of Health, Water Quality Control Division (WOCD) regarding the dilution and discharge of tritium during decommissioning. . The Colorado Wastewater Discharge Permit for FSV is in the process of renewal with the WQCD and liquid discharge will continue under the existing permit until the revised permit is issued. All required FCC licenses for radio and microwave communications are current. Currently no air emission permits are applicable to FSV activities. No permits or approvals are required for hazardous waste generated during decommissioning. A radioactive material disposal site user's permit for the Beatty Nevada Site is valid through the end of 1992. PSC has applied for approval to use the disposal site in Richland, Washington after January 1,. 1993. We have determined that PSC has adequately considered and has provided the status of federal, State and local permits needed for FSV decommissioning. , 5.3 Lpeal Short-Term Uses VersgLlona-Term Productivity Af ter decommissioning is complete PSC intends to continue to use the FSV site for power production. PSC plans to build a gas fired boiler to provide steam to the existing TSV turbine. Therefore, there is no conflict between short-term usas vs long-term productivity of the site. 5" 1, reversible and Irretrievable Commitments of Resources The proposed DECON of FSV will involve the use of about 2000 gpm of water for dilution of the PCRV shield water before its release to the environment. The
t Commission's FES for FSV considered flow rates of 2650 gpm for FSV with no adverse environmental impact and the 2000 gpm falls well within that rate. In , addition, the dilution water 1: returned to the South Platte River and is therefore not lost. FSV facilities will continue to occupy the same space that was previously evaluated in the Commission's FES for FSV. Consistent with the impacts evaluated in the FGEls approximately 2 acres of land elsewhere, at a licensed waste burial site will be used for FSV waste. After decommissioning is complete, the FSV site will continue to be used for electric power production with the use of a new gas fired boiler and the existing FSV turbine. 5.5 Potential Exaq1gre to Workers PSC has estimated that radiation exposure for workers during the proposed FSV decommissioning would be approximately 500 person-rem. This exposure estimate is based on a work breakdown for each task and that all work associated with radioactive components is accomplished in accordance with Radiation Work Permits that are issued for each activity. The projected exposure is adequately define / and is significantly less than the 1874 person-rem occupational exposare estimated for the reference BWR tr 1215 person-rem estimated for the reference PWR enploying the DECON alternative as evaluated in the FGEIS. PSC estimates that the proposed use of the PCRV water shield will result in a projected personnel exposure that is 2 to 11 times less than the personnel exposure estimated by other vendors for decommissioning without the PCRV water shield. The NRC staff has reviewed the licensee's data, calculations and results and has determined that potential exposures to workers are adequately estimated and are acceptable with respect to ALARA considerations. 5.6 Radioactive Waste PSC estimated that FSV decommissioning would produce 100,000 cubic-feet of radioactive waste after processing and volume reduction. PSC estimates 71,000 cubic feet of Class A, 28,000 cubic feet of Class B, and 1000 cubic feet of Class C waste. The waste consists of activated and contaminated concrete, graphite blocks, equipment and piping. Radionuclides of concern are cobalt-60, iron-55, europium-152, cesium-137 and tritium. Chapter 3.3 of PSC's PDP provides a detailed discussion of the waste generated and Section 3.3.3 of the staff's Safety Evaluation Report (SER) provides an evaluation of the waste and it's disposal. The FGEIS indicates that the DECON option at large light water reactors would produce about 670,000 cubic feet of low level radioactive waste. PSC indicates that there may be as much as 400 cubic feet of radioactive -te that is
- Greater Than Class C" (GTCC) because of high concentrations of cobalt-60 in the keyed-top reflector control rod elementr. If these elements are GTCC, they would be stored at the ISFSI as permitted by the existing ISFSI license (SNM-2504, November 4, 1991, pages 1 and 2). Storage of these '
reflector elements would continue at the ISFSI until the cobalt-60 (5.2 year half-life) decayed to levels that would allow disposal at a low level waste burial facility or the Department of Energy established a facility for GTCC
-- . - , - - - .- . . , - - - - - - . - - _ - _ _ - . w
waste. if these elements are not found to be GTCC they would be promptly lisposed of at a licensed low level waste burial facility. The staff cnneludes that the volume of waste generated from dismantlement of FSV is well within FGLIS estimates and that waste disposal from FSV will not impose a significant environmental impact. 5.7 Air Quality Dust will be generated during diamond wire cutting of the PCRV concrete. However, the water used as a lubricant for diamond wire will retain most of the dust. In addition any dust generated by this cutting operation or similar operations will be filtered by the existing resctor building ventilation system consisting of moisture separators, HEPA filters and charcoal absorbers. General dust control will be accomplished with HEPA filtered vacuum cleaners. The staff concludes that the filtered ventilation system and dust control procedures will assure that dust generated during FSV decommissioning will not impose a significant environmental impact. 5.8 Exhnst Emissions Exhaust emissions from diesel and gasoline powered equipment and vehicles may have a slight impact on air quality. PSC estimates that approximately 350 truck shipments of radioactive waste will be made. in addition several hundred shipments of non radioactive waste will be made to local landfills. However, the total number of vehicles at the site will be significantly reduced from the levels previously needed to support FSV operations. Also, the exhaust from the plant's auxiliary boilers will be reduced during decommissioning because their maximum capacity has been reduced by a factor of ten to accommodate building heating alone. 5.9 Asbesto1 Asbestos removal will be performed within controlled ventilation areas equipped with exhaust filtration to ninimize its release. The removal of asbestos is discussed in section 3.5 of the licensee's ER Supplement and evaluated in the staff's SER. Removal will be accomplished in accordance with Federal (OSHA 29 CFR 1910/1926 and EPA 40 CFR 61. H) and State regulations. The staff has determined that the removal of asbutos dill not have a g significant environmental impact. S.10 ffhsts of Chemical and Biocide Discharoe During plant operations, waste water was produced by demineralizer regenera- ' tion and blowdown from the main and service water cooling towers. During decommissioning, the main cooling tower will be out of service and dominer-alizer regeneration and service water tower use will be reduced. Therefore, with a greatly reduced use of chemicals / biocides the resulting environmental impact during decommissioning will also be reduced to less than that considered in the FSV FES.
~
l l
5.11 Effects of Sanitary Discharae Sanitary wastes from approximately 320 PSC workers and contractors have been previously treated by the site sewage system. The maximum number of people required on site during decommissioning is not expected to exceed 300. Therefore, the environmental impact would not increase during decommissioning. 5.12 Endanaered Specjn No significant impact on the flora and fauna of the site or its surrounding areas have been observed during construction or operation of FSV and no change in such impact is expected during decommissioning. In addition, there is no known natural habitat for endangered species on land areas directly involved with decommissioning, i 5.13 Rajn There will be very little noise impact during decommissioning. Demolition activities will normally be conducted during the day with the exception of diamond wire cutting of the PCRV concrete. Diamond wire cutting is inside the reactor building and is relatively quiet. Therefore, the cutting is not expected to be heard offsite. The areas in the vicinity of the plant are i sparsely populated and the nearest population center of approximately 1500 residents (Platteville, Colorado) is about 3.5 miles distant. 5.14 Social Economic Impac1 Decommissioning is expected to be completed over a 57 month period and to employ a peak work force of approximately 300 personnel. The temporary nature of the project and limited work force supports the conclusion that no significant demographic shifts or significant effects on the regional economy will result. 5.15 Accident Analysis Section 3.4 of the PDP and also 3.4 of the staff's SER evaluated potential decommissioning accidents with respect- to the health and safety of the public. The risk and consequences of accidents during decommissioning is considerably less than during plant operations due to the removal of all fuel from the reactor protected area. Whole body and organ doses are shown in the PDP to be l a small fraction of EPA Protective Action Guidelines at the Emergency Planning I Zone boundary (a square with sides approximately 100 meters from the reactor building)'for all accident scenarios. Section 5.0 of the ER Supplement calculates potential exposures using the EPA code "AIRDOSE." Again the potential dosea were shown to be a small fraction of EPA Protective Action Guidelines, significantly less than the potential doses considered in the FSV FES and will have no significant impact on the human environment. 5.16 Ground Water FSV is between the South Platte River and the St. Vrain Creek about 2 miles south of the intersection of these two streams.- The contours of the water table indicate the flow of the ground water is predominant 1y'toward the South Platte River to the east. The local water district (Central Weld County Wate'r l
District) is supplied fron .arter Lake that is located about 20 miles west of FSV. The closest downstream public well is the Gilcrest well that is incated about 7 miles downstream (north) near the South Platte River. The only radionuclide that has a potential for getting to the groundwater during dismantling is tritium from the shield water dilution process (EA Section 5.1). Other radionuclides such as cobalt-60, cesium-137, iron-55 and strontium-90 in shield water are removed to concentrations less than 1.0 percent of 10 CFR Part 20 limits by demineralizers. PSC has calculated the potential exposure from Gilcrest well water assuming that-the maximum possi-ble, 8000 curies of tritium were released by the dilution process to the Sr'th Platte River. The potential annual exposure from the Gilcrest well water , ,th that conservative assumption is 0.047 millirem which is well below the EPA drinking water limit of 4.0 millirem per year. Following decommissioning, the potential exposures from groundwater will be significar.tly less as tritium will no longer L : released and residual radionuclides at the site will be in the form of low-level activation or contamination of solid reactor structures. The site must be shown to meet the 4.0 millirem annual criteria for drinking water (40 CFR 141.16) as well as the NRC criteria for residual surface contamination and direct radiation in reactor structures (SER Section 1.2.l(3)) before the license-is terminated. 6.0 AGENCIES AND PERSONS CONSULTED The NRC staff reviewed PSC's application and it's Environmental Report Supplement and communicated frequently with officials of the State of Colorado but did not consult other agencies with respect to this Environmental Assessment. Questions on tritium release from the State of Colorado Water Quality Control Division (letters dated July 16 and 24,1992) were directed to PSC and the NRC. Responses from PSC and the NRC dated August 6 and September 17, 1992, respectively were provided to the State. PSC has agrad to comply with the State of Colorado requirements on tritium concentrations and monitoring in the South Platte River. 7.0 flNDING Of NO SIGNIFICANT IMPACT Based upon the foregoing environmental assessment, the Commission has concluded that the environmental impacts of decommissioning FSV are adequately bounded by the environmental impacts of decommissioning light water reactors as anslyzed in the FGEls and that the proposed action will not have a significant impact on the quality of the human environment. Accordingly, the Commission has determined not to prepare an environmental impact statement for this proposed action. Attachments:' As stated (Tables 1 and 2) Date:
1&ELLI FSV ENVIRONMENTAL ASSESSMEN1 ALTERNATIVES COMPARISON
&llernative fotential Advantanes Potential Disadvantanes EafslDB Lower waste curies & volume; Requires surveillance & access lower radiation exposure & control; site not available for dose; least commitment of unrestricted access but FSV 16nd to radioactive waste turbine could be repowered disposal; PCRV shield water with natural gas; new workers wo.ld probably not be needed would require more training; for SAFSTOR, resulting in residual radioactivity must be-a reduction in potential removed at end of SAFSTOR tritium release, period (60 years maximum).
ENTOMB Lower waste curies & volume; Long-lived radionuclides (Ni 63 1 lower radiation exposure to & Hb-94) must be removed in the workers; entombment future; site not available structure controls access to for unrestricted access; more radioactivity. difficult to remove residual radioactivity at the end of ENTOMB period due to the entombment structure, pICON Fastest release of land Higher waste curies, waste for unrestricted use; no volume and radiation dose; requirement for continued tritium release to shield surveillance or access water and the environment control; prompt removal increased over other options. of radioactivity; PSC workers more familiar with FSV construction; FSV equipment in newer / better working condition. NO ACTION Not a viable alternative Would not ensure that risk to the public from the facility will be within acceptable bounds, p
,- ~-
m,,e- --
-a wy ,,vm.,.-rt,- w-- , . -..w,--4.m , w g ,,.e.r...<r ,me.,we,, , -wwow,,- .,,-m,-_.r.g-+,,mw,,,n,r- - - + , e-e--g,w +, w
( 181Ll_2 i WHOLE BODY DOSE TO THE MAXIMALLY EXPOSED INDIVIDUAL T PATHWAY (mrem /vear) Ground Water Ingestion 0.047 Milk Consumption 0.512 Vegetable Consumption 0.326 Heat and Waterfowl Consumption 0.084 Fish Consumption 0.024 TOTAL: 0.993 mr/yr 7 i -,-4 y, _ . - . . --y,,,,,_-,-.,,,-..y.-,.,,c._.,,,..,e.y,.... .~ ,. ,__ ,r,.,,-- . , , - . . . ,,.%-..e,, .,y .- , . , , . . , , , , , , af,, , . , , . , - . , , _ y ...,.m... .
7590-01 VnllED STATFS NUCLEAR REGULATORY COMMISSION PUBLIC SERVICE COMPANY OF COLORADO f_0RT ST. VRAlti NVCLEAR CENERATING ST ATION DOCKET NO. 50-267 NOTICE OF ISSUANCE OF ENVIRONMENTAL ASSESSMENI AND FINDING OF NO SIGNIFICANT IMPACT The U.S. Nuclear Regulatory Commission (the Commission) is considering the issuance of an ordtr authorizing the decommissioning of the Fort St. Vrain Nuclear Generating Station (FSV) that is licensed to Public Service Company of Colorado (PSC). The Commission is also considering the issuance of an amendment to revise the Technical Specifications to be consistent with the Decommissioning Plan. Identification of Proposed Action FSV has been shut down since August 18, 1988, and all spent fuel has been removed from the reactor protected area and transferred to an Independent Spent Fuel Storage Installation (ISFSI) that is separately licensed to PSC under 10 CFR Part 72. Decommissioning of FSV includes the dismantlement, decontamination and disposal of radioactively activated and contaminated material and components produced by FSV operations. Substantial portions of the plant will be dismantled and removed. However, the reactor building, turbine building, and other structures that are not radioactive above limits acceptable for unrestricted access will remain. Following completion of dismantling and decontamination activities, PSC will conduct extensive radiation surveys to verify that the plant and the FSV site meet NRC release criteria and can be released to it. restricted access. The FSV Part 50 license will be terminated following satisfactory completion of decommissioning i
i l actions by PSC and verification surveys by the NRC. Approval of the Decommissioning Plan will allow PSC to dismantle and decontaminate FSV in accordance with the approved plan. Environmental Impacts The NRC staff reviewed the proposed Decommissioning Plan and the related Environmental Report Supplement with respect to 10 CFR SI 53(b) and to document its review prepared an Environmental Assessment. Decommissioning FSV in accordance with the plan will allow termination of License No. DPR-34 and the continued use of the site for electric power production using the FSV turbine with a natural gas fired boiler. Findino of No Sianificant Impact The staff has reviewed the proposed decommissioning relative to the requirements set forth in 10 CFR Part 51. Based upon the Environmental Assessment, the staff concluded that there are no significant environmental impacts associated with the proposed decommissioning and that the aroposed decommissioning will not have a significant effect on the quality of the human environment. Therefore, the Commission has determined, pursuant to 10 CFR 51.31, not to prepare an environmcntal impact statement for the proposed decommissioning of FSV. For further details with respect to this action, see: (1) the licensee's l application for authorization to decommission FSV, dated November 5, 1990, as l- revised December 17 and 21, 1990, January 14, 1991, April 15 and 26, 1991, ! May 15, 1991, June 6 and 17, 1991, July 1, 1991, August 28 and 30. 1991, l
- j. November 15, 1991, December 6, 1991, January 9, 1992, March 19, 1992, i
l
1 April 17, 1992, and September 25, 1992; (2) the licensee's Environmental Report Supplement dated July 10, 1991, as revised March 20, 1992, April 30, 1992, June 24, 1992, and September 1 and 18, 1992; (3) Amendment No. 85 to License No. DPR-34; (4) the Commission's related Safety Evaluation; and (5) the Commission's Environmental Assessment and Finding of No Significant lepact. These documents are available for inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, N.W., Washington, _ 0.C. 20555, and at the Greeley Public Library, City Complex Building, Greeley, Colorado 80631. Dated at Rockville, Maryland this FOR THE NUCLEAR REGULATORY COMMISSION Seymour H. Weiss, Director Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation i _--------w. - _ _ - _ . . _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _}}