ML20236G490
| ML20236G490 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/27/1987 |
| From: | CONSOLIDATED EDISON CO. OF NEW YORK, INC. |
| To: | |
| Shared Package | |
| ML20236G473 | List: |
| References | |
| NUDOCS 8708040284 | |
| Download: ML20236G490 (8) | |
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ATTACHMENT A Technical Specification Page Revisions To Supplement the December 19, 1985 Application i
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Consolidated Edison Company.of New York, Inc.
Indian Point Unit No. 1 Docket No. 50-03 July, 1987 80 p P
3.0 Administrative and Procedural Safecuards 3.1 Organization 3.1.1 The organization for facility management and technical support shall be shown in Figure 3.1.
3.1.2 The Facility Organization shall be as shown in Figure 3.2. The Support Facilities Supervisor is responsible for operations at the Unit No. 1 facility.
3.1.3 The Vice President, Nuclear Power shall be responsible for over-all facility activities as shown in Figure 3.2 and shall delegate in writing the succession to this responsibility during his absence.
3.1.3.1 The General Manager, Nuclear Power Generation shall be responsi-ble for facility operations as shown in Figure 3.2 and shall delegate in writing the succession to this responsibility during his absence.
3.1.4 The operation of the facility, the operating organization, the procedures for operation, and modifications to the facility shall be subject to review by the Station Nuclear Safety Committee.
The committee shall report to the Vice President, Nuclear Power.
3.1.5 The Nuclear Facilities Safety Committee shall function to provide Independent review and audit of designated activities in areas of nuclear engineering, chemistry, radiochemistry, metallurgy and non-destructive testing, instrumentation and control, radiolo-gical safety, mechanical and electrical engineering, administra-tive controls and quality assurance practices, and radiological environmental effects.
3.1.6 All fuel handling shall be under the direct supervision of a licensed operator.*
3.2 Og rating Instructions and Procedures 3.2.1 No fuel will be loaded into the reactor core or moved into the reactor containment building without prior review and authoriza-tion by the Nuclear Regulatory Commission.
3.2.2 Detailed written instructions setting forth procedures used in connection with the operation and maintenance of the nuclear power plant shall conform to the Technical Specifications.
3.2.3 Operation and maintenance of equipment related to safety when there is no fuel in the reactor shall be in accordance with writ-ten instructions.
- Licensed operator for IP-1 or IP-2.
Amendment No.
4.0 Operating Limitations 4.1 General )
4.1.1 Whenever any operation is being performed that could result in the release of radioactivity or create a change in radiation levels, supporting facilities shall be maintained and operated as required in these Technical Specit'ications.
4.1.2 The concentration of radioactive materials released in liquid or gaseous form to unrestricted areas shall not exceed the limits specified in 10 CFR Part 20. Release of radioactive liquids and i l
gases shall also be consistent with the requirements of 10 CFR Part 50, Appendix I, as specified in Specifications 3.9 and 4.10 of Appendix A to the Indian Point Unit No. 2 Facility Operating License No. DPR-26. l 4.1.3 All radioactive waste material shall be handled in accordance with 10 CFR Part 20. In addition, solid radioactive waste shall be controlled as specified in Specifications 3.9.D and 4.10.D of ,
Appendix A to the Indian Point Unit No. 2 Facility Operating License No. DPR-26. l l
4.1.4 Radiation monitoring systems shall be maintained operational for: l (1) nuclear services building sewage, (2) chemical systems building sewage, (3) sphere foundation sump, (4) secondary purification blowdown cooling water, and (5) area radiation l monitors. If monitoring systems are not operational, effluent l l
sampling and/or local monitoring shall be accomplished to replace the non-operating system. In addition, Unit 1 radioactive effluent monitoring instrumentation shall be operable as spec-ified in Specification 3.9 of Appendix A to Indian Point Unit No. !
2 Facility Operating License No. DPR-26, 4.1.5 The Indian Point site meteorological monitoring system shall be maintained and operated as specified in Specifications 3.15 and 4.19 of Appendix A to the Indian Point Unit No. 2 Facility Oper-ating License No. DPR-26.
4.1.6 The Indian Point site Radiological Environmental Monitoring Program shall be conducted as specified in Specification 4.11 of Appendix A to the Indian Point Unit No. 2 Facility Operating License No. DPR-26.
4.1.7 Radiation Protection Program Procedure for personnel radiation protection shall be prepargd consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
Amendment No.
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4.1.8 High Radiation Area 4.1.8.1 As an acceptable alternate to the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20:
- a. Each High Radiation Area in which the intensity of radiation is i greater than 100 mrem /hr but less than 1000 mrem /hr shall be !
barricaded and conspicuously posted as a High Radiation Area and l entrance thereto shall be controlled by issuance of a Radiation Work Permit and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area. ]
- b. Each High Radiation Area in which the intensity of radiation is greater than 1000 mrem /hr shall be subject to the provisions of 4.1.8.1(a) above, and in addition locked doors shall be provided to prevent unauthorized entry to such areas and the keys shall be l maintained under the administrative control cf the Watch Supervi-sor on duty.
4.2 Operations Limits 4.2.10 Spent Fuel Storage and Handling 4.2.10.3 All irradiated fuel shall be stored in the racks provided in the Fuel Handling Building Storage pools.
Amendment No.
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ATTACHMENT B i
l Revised Safety Assessment To Supplement The December 19, 1985 Application I
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Consolidated Edison Company of New York, Inc.
Indian Point Unit No. 1 Docket No. 50-03 July, 1987
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Safety Assessment Description of Change and Safety Evaluation: .
t The proposed change revises the Indian Point Unit No. 1 (IP-1) Technical Specifications by deleting Section 3.2.6 and adding Sectfons 4.1.7 and 4.1.8 to assure that with respect to the Radiological Protection Program the IP-1 Technical Specifications are consistent with Indiar. Point Unit No. )
i 2 (IP-2) requirements.
l Currently, the IP-1 and ~.P-2 Technical Specifications requirements pertain-ing to personnel radiation protection are substantially similar. IP-1 currently has Technical Specifications which require that personnel radiation protection procedures be prepared consistent with the require- '
ments of 10 CPR Part 20. Similarly, the IP-2 Technical Specifications contain certain requirements for establishing personnel rediation protec- .
tion procedures pursuant to 10 CPR Part 20. However, in addition to these 5 requirements, the IP-2 Technical Specification allow alternative methods of controlling access to certain high radiation areas consistent with the NRC approved Standard Technical Specification (STS). Since the IP-1 Technical Specifications do not specify similar alternative high : radiation j access methods, this has led to different procedures and differing applica- ,,
tion of area access 2 requirements between the two units, soretimes re=;ulting in personnel confusion as to which requirement applies.
The proposed revisions to the IP-1 Technical Specifications contained in {
Attachment A to the application would allow the entrance to a high radia-tion area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr to be controlled by the issuance of a Radiation Work Permit (RWP). Such permits would require that individuals be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area. Under existing IP-1 Technical Specifications this is accomplished by maintaining these areas locked except during periods when access to the area is required. There are currently 20 areas in IP-1 which have radiation levels greater than 100 mrem /hr but less than 1000 mrem /hr. Those areas in IP-2 meeting the same radiation level criteria are unlocked. This difference in locking requirements presents a possibilities for confusion of station and contractor personnel, particularly when personnel are assigned to perform work at either unit. When an individual or group of individuals want to enter one of the IP-1 locked high radiation areas they must first contact an Health Physics (HP) technician to get the door unlocked. Because of the IP-1 physical layout and large distances between areas, the working individual or group experiences excessive time delays while waiting to obtain access to the desired areas. The HP technicians have other duties to perform and cannot devote a substantial amount of time to opening the doors. Also an entry into 9 of the 20 areas requires that an individual be posted at the entrance to provide positive access control. The station ALARA program stresses the need to minimize the number of individuals in a group working in the Radiologically Controlled Area (RCA). Because the
guard must remain at the entrance it is not possible for this individual to perform any other function in support of the work crew. Should the proposed change be approved, the need for this individual would be removed without diminishing the effectiveness of the station ALARA program.
Current estimates of the frequency of access to the locked areas are as follows:
4 areas are entered on a daily basis, 3 areas are entered on a weekly basis, and 13 areas are entered on a monthly basis or longer.
Should the proposed change become effective, all of the . entrances (28 doors) to these 20 locked areas would be unlocked. The radiological controls for entry into these areas would not be compromised by unlocking them, since entry would be controlled by the requirements of a RWP consistent with NRC approved Standard Technical Specifications and acceptable industry practices. One of the RWP requirements is the use of a l radiation monitoring device which continuously indicates the radiation dose rate in the area. In addition, these areas will be barricaded and conspicuously posted as a high radiation area. Individuals permitted to enter such areas are required to review radiation survey maps of the area before entry. The survey maps are also posted at the entrance of each area. " Hot Spots" or other radiological items of concern within the areas l are also identified locally. Thus, the alternative method of using an RWP for area access control is conservative in maintaining radiation protection for site personnel.
In summary, the proposed change would greatly improve the efficiency of work in the RCA of IP-1, would help reduce radiation exposure and would ;
minimize the potential for the confusion of workers who would otherwise I have to comply with two different sets of procedures in IP-1 and IP-2. j Bgis for No Significant flatards Consideration Determination The Commission has provided guidance concerning the application for the standards for determining whether a significant hazards consideration exists by providing certain criteria. Consistent with the Commission's criteria in 10 CFR 50.92 (48 FR 14871), we have determined that the pro-posed change does not involve a significant hazards consideration because the operation of Indian Point Unit No. 1 in accordance with this change j would noh #
- 1) involve a significant increase in the possibility or consequences of an accident previously evaluated. The proposed change does not entail any physical changes in plant equipment. The proposed change involves a revision to the site personnel access control requirements for IP-1 radiation areas in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr.
Thus, it does not in any way affect the dose consequences of offsite releases. The proposed change remains conservative based
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1 on consistency with NRC approved Standard Technical Specification (STS), acceptable industry practice and our own operating experi-ence. The proposed revision would also help to reduce site l radiation exposure and minimize the confusion of workers who have I to comply with two different procedures in IP-1 and IP-2.
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Therefore, this change will not increase the probability or j consequences of an accident previously evaluated. 1 a
- 2) create the probability of a new or different kind of accident from any previously evaluated, since the proposed change would not alter the configuration of any of the plant's equipment and remains conservative in providing assurance of maintaining adequate radiation protection for site personnel.
- 3) involve a significant reduction in a margin of safety, since the proposed change remains conservative for controlling access to high radiation areas. The proposed change involves revisions to the IP-1 access control requirements for radiation areas in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr. These areas will be barricaded and conspic-uosly posted as a high radiation area and entry will be con-trolled by the issuance of a Radiation Work Permit (RWP). In addition, a radiation monitoring device which continuously indicates the radiation dose rate in the area will be provided to the individuals who are permitted to enter such areas. These requirements provide a conservative measure of radiation protec-tion for site workers based on consistency with the NRC approved STS, acceptable industry practices and our own experience in applying similar requirements at IP-2. The proposed change will also reduce radiation exposure by eliminating the need for a guard to be posted at the entrance of the area and will minimize the confusion of workers who have to comply with two different sets of procedures in IP-1 and IP-2. m i
i Therefore, based on the above considerations, we conclude that the proposed j che.ge does not constitute a significant hazards consideration. ;
1 The proposed changes have been reviewed by Consolidated Edison's Station Nuclear Safety Committee and Nuclear Facilities Safety Committee. Both committees concur that this change does not represent a significant hazards j consideration and will not cause any change in the types or increase in the j amounts of effluents or any change in the authorized power level of the ]
facility.
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