ML19351F325
| ML19351F325 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/10/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19351F319 | List: |
| References | |
| NUDOCS 8101120235 | |
| Download: ML19351F325 (5) | |
Text
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f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 17 TO FACILITY CPERATING LICENSE 10. NPF-2 ALABAMA POWER COMPANY JOSEPH M.
FARLEY NUCLEAR PLANT, UNIT NO. 1 00CKET NO. 50.A8
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Introduction
\\labam3 Power Company (APCO), the licensee for the Joseph M. Farley Nuclear slant, Unit No.1, proposed changes to Operating License No. NPF-2.
These changes are included in APC0 letters dated October 15,1979 (modified by letter dated October 3,1980), and October 23, 1979. Our followup action to our May 1,1900 letter approving the Safeguards Contingency Plan is also discussed herein. The license changes included in this amendment and discussed below are as follows:
1.
Revised Administrative Controls Technical Specifications for entry into high radiation areas; 2.
Added feedwater control system bypass valves response times to Tech-nical Specifications; and r
3.
Added license condition relating to the approved " Joseph M. Farley Nuclear Contingency Plan." This condition is a followup action to our May 1,1980 letter which approved the plan.
ADMINISTRATIVE CONTROLS FOR HIGH RADITION AREAS (Specification 6.12 and 6.12.2)
Discussion and Evaluation By letter of October 15,1979, APC0 proposed changes to the Administrative Controls Technical Specification for entry into high radiation areas. Entry into high radiation areas requires positive control of personnel within those Conditions for each entry should be prepared in a manner which is areas.
both logical from the standpoint of good radiation orotection practice and unambiguous so that each of the alternative methods for control of entry will orovide reasonable protection of cersonnel. The current Standard Techni-cal Soecifications (STS) nas been written to clearly adcrass the manner in which radiation protection practice may be exercised for pt.sitive control for entry intt, high radiation areas. The APC0 submittal of October 15, 1979 falls short of this practice for the folicwing reasons:
120023 5 810
. (1) Specification 6.12.1.(a) as APC0 proposed would provide ffor a " control device" or "alann signal" required by paragraph 20.203(c)(2) of 10 CFR 20. The control device of caragraph 20.203(c)(2)(i) is not acplicable to most radiation sources in nuclear power reactors. Paragraph 20.203 (c)(2)(ii) requires a control device to energize a conspicuous visable or audible alarm signal. The APC0 preposal falls short of positive control of access into high radiation areas since the proposed system can either be de-energized by personnel, or, if used by itself with no other control device, could be ignored by personnel. Paragraph 20.203(c)
(2)(iii) is an alte.rnative addressed in our STS which has all the connota-tions of unambiguous positive access control. Therefore, Specification 6.12.1.(a) as procesed is' acceptable.
(2) Specification 6.12.1.(b) as prooosed by APC0 also does not provide positive access control. Pocket ioni:stion chambers are unaccept-able as survey meters since they are personnel dosimeters and should be used as such unless no survey meters are available. They are aftar-the-fact monitoring systems and, therefore cannot be considered positive control devices for determining stay time (i.e., their response is tou slow for mea:urement of dose rate in areas where the dose rate may be rapidly changing). Also audible warning devices (e.g., chirpers) reouf re some skill in interpretation of chirp rate as a function of dose rate and must also operate in a low noise area. The sum of the two instruments (i.e., cocket chamber clus chirper) is therefore not equal to or reliabile as a good radiation survey meter. Consequently, proposed Specification 6.12.1.(b) is unacceptable.
(3) Proposed Soecification 5.12.1.(c)(d) is acceptable since it conforns to the STS Section 6.12.1.(a) and (b).
(4) Proposed Specification 6.12.1.(e) is acceptable since it con-forms to the STS with the addition of "...by the Health Physics supeWisor."
The Saecification approved for Farley is 6.12.1.(c).
The !.coroved changes will provide a clear definitive condition of positive access control for entry into high radiation areas when t'le radiation levels are in excess of 1000 mR/hr. This action considers the case where it is not reasonable to provide locked enclosures for small areas having radiation levels in excess of 1000 mR/hr. Such areas may be located in nuch larger areas such as a pressuri:ed water reactor containment.
The conditions for entry into such areas require radiation level measure-ments in the area and delineation of maximum allowable stay-times in adcition to use of barricades, posting and flashing lights as the alterna-tive for locked encicsures. Positive exposurt control can also !;e made by continuous surveillance over the activities within the are_ by personnel cual1fied in radiation protection.
Conclusion The approved, modified Technical Specifications 6.12.1 and 6.12.2 replace the existing specifications and are acceptable. Changes to the APC3 orocosed revision were discussed with and agreed to by APC0 staff.
FEEDWATER CONTROL SYSTEM BYPASS VALVES RESPONSE TIMES (Specification Table 3.3-5)
Discussion and Evaluation By letter of October 23,1979, APC0 prooosed addition of three feedwater ccontrol system valves (FCV 479, FCV-489 and FCV-499) to Technical Specifi-cation Table 3.3-5.
These valves were installed as bypass control valves
'n part'lel with the main feedwater control valves (FCV-478, FCV 488 and 7G. 498) during the first refueling outage which was comoleted in late October 1979.
This system modification was accomplished by APC0 under 10 CFR 50.59.
Bypass control valves provide a means of operating the steam generator level control system at low reactor pcwer levels. The larger si:e of the main feedwater control valves preclude the use of the main valves at relatively low reactor power levels. Thus, the bypass control valves should result in improved system performance and should result in fewer reactor trips and system transients.
APC0 proposed changes to the Technical Specifications to add the bypass control valves to Table 3.3-5 wnere One main feedwater control valves are shown. This will assure that response time testing is accomolished in a manner consistent with the main feedwater control valves.
Conclusion Based on the discussion above, we conclude that the added recuirement to perform surveillance testing on the bypass control valves is acceptable.
FFurther, the testing is similar to testing creviously aporoved on the main feedwater control valves and gives added assurance of valve coerability as required by Technical Specification 3.3.2.1, a Limiting Condition for Operation.
CONTINGENCY ?LAN LICENSE CONDITION Ciscussion By letter dated March 23,1979 APC0 submitted a Safeguards Security Contingency Plan for the Josech M. Farley Nuclear Plant as reouired by t
4 10 CR 50.34(d) and 10 CFR 73.40. The plan was revised to meet the criteria established by Appendix C to 10 CR Part 73 and was fomulated per Regulatory Guide 5.54 as a self contained plan.
!n response to our letter dated August 31, 1979, APC0 provided a draft, nroposed amendment to the plan by letter of October 8,1979. Further, in res;:ense to our letter dated February 12, 1980, APC0 provided in their March 29, 1980 letter, a comoletely revised text incorporating all crevious changes. Our letter dated May 1,1980 approved the plan as revised.
Under provisions of 10 CR 2.790(d) the plan is being withheld from public disclosure.
Conclusicn Based on Our review of the revised Contingency Plan for the Joseph M.
Farley Nuclear Plant, we have concluded that the plan for this facility, when fully imolemented, will provide the protection needed to meet the general per#omance recuirements of 10 CFR 50.54(o) and 73.40(b) and the objectives of the specific requirements of 10 CFR 73.55(h) and Arpendix C to 10 CFR 73.
We, therefore, further conclude that your Safeguards Contingency Plan is acceptable.
Changes which would not decrease the effectiveness of your approved Safe.
guards Contingency Plan may be made without aporoval by the Comission pursuant to the authority of 10 CR 50.5a(p). A report containing a descriotion of each change shall be furnished to the Director, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Nashington, D. C. 20555, with a cc:y to the aporopriate NRC Regional Office within two months after the change is made. Records of changes nade without Ccmmission ac;:roval shall be maintained for a period of two years from the date of the change.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this detemination, we have further concluded that the amendment involves an action which is insignificant fro-the stand::oint of environmental imcact and, pursuant to 10 CR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
. Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendnent does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reason-able assurance that the health and safety of the public will not be endan-gered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Data: December 10, 1980 l
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