ML20070T952
| ML20070T952 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/02/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070T943 | List: |
| References | |
| TAC-48295, TAC-49512, TAC-49513, NUDOCS 8302100060 | |
| Download: ML20070T952 (3) | |
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UNITED STATES
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SAFETY EVALUATION BY THE OFFICE OF flVCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.92 TO FACILITY OPERATIllG LICENSE fo. DPR-57 AND A'iEllDitENT NO. 30 TO FACILITY OPERATING LICENSE fo. NPF-5 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATIO!1 flVNICIPAL ELECTRIC AUTliORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN I. HATCH NUCLEAR PLANT, UNITS N05.1 & 2 DOCKETS NOS. 50-321 AND 50-366 By letter dated March 10, 1982 Georgia Power Company (GPC or the licensee) proposed revisions to the Technical Specifications (TSs) for both Hatch Unit No.1 and Unit No. 2 to change the numerical values of the reactor water levels and the setpoints for reactor water levels measured by the shroud water level instruments to make thetn consistent with a change in the zero reference level. The change is required in order to make the TSs consistent with previous modification of the reactor water level instruments to utilize a connon zero reference level pursuant to NUREG-0737, Item II.K.3.27.
In that modification, the zero water level references for the instrtments measuring the reactor shroud water level were changed from a level near the bottom head of the reactor vessel to that level near the bottom of the steam drier skirt that is used as the common zero level refercnce by all of the other reactor water level instruments. The changes to the TSs involve only numerical changes.
The actual levels at which protective
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functions or permissive signals are initiated. ave' unchanged. On the basis of the above discussion, we conclude that the proposed change is acceptable.
By' letters, as noted below, GPC proposed revisions to the Hatch Unit No.1 TSs as follows:
1.
Letter dated April 5,1982, proposed to establish an upper limit for the rod block monitor high flux trip setting.
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2.
Letter dated January 3,1983, proposed to increase the pressure setting of the Safety Relief Valve (SRV) tai) pipe pressure switches.
3.
Lette:r dated January 3,1983, proposed to add a newly installed torus access hatch to the list of testable penetrations.
Our evaluation of these proposals and our conclusions are discussed below.
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. Power Limit for Rod Block Monitor Trip The current Hatch Unit No.1 TS Table 3.2-7 requires a high flux trip setting for the rod block monitor at a percentage of actual themal power that is equal to 0.66 W + 41% where W is the loop recirculation flow. The current TS does not specify an upper limit for this trip setting. The licensee has requested in Attachment 3 to their letter dated April 5,1982, that an upper limit of 107% rated power be placed on this rod block monitor trip setting.
We have reviewed the proposal and found that the fuel thermal margins am not reduced by this change, none of the present rod block or reactor setpoints are affected and that the 107% power limit provides additional protection to the fuel for an above rated core flow condition. We conclude for the reasons stated above that this change is acceptable.
Safety Relief Valve Tailpipe Pressure Switch Setting By letter dated January 3,1983, GPC proposed a change to Section 4.6.H of the TSs. The change would increase the setpoint of the tailpipe pressure switch of each main steam SRV from 30 + 5 psig to 85, +15, -5 psig.
The licensee requested this change because it had been determined that the setpoint was not high enough to provide correct information on the "open" or " closed" status of the SRV to a reactor operator following a loss-of-coolant accident (LOCA). The revised setpoint accounts properly for an increase in primary containment pressure to the design.
value of 65 psig, combined with a static head pressure in the tailpipe of 10 psig caused by possible elevated water level in the torus. The setpoint also includes an allowance of 10 psi for inaccuracies and margin to prevent spurious SRV "open" indications during a LOCA.
On the basis of the above, we conclude that the 85 psig setpoint can provide more reliable information to the operator during a LOCA and is acceptable.
New Torus Access Hatch The Hatch Unit No.1 facility is currently in a torus modification /
refueling outage. During this outage, a new torus access hatch was added to containment. The licensee's letter of January 3,1983, proposed a revision to TS Table 3.7-2, " Testable Penetrations with 6
Double 0-Ring Seals". This revision would identify the new contain-ment penetration as being subject to local leak rate te' sting as required by Appendix J to 10 CFR Part 50.
The new to us access hatch has been designated penetration number X-200C, suppression chamber access manhole. This containment penetra-tion has been equipped with double 0-ring seals so it can be locally (Type B) leak tested in accordance with Appendix J to 10 CFR Part 50.
Table 3.7-2 of the Hatch TSs has been modified to include penetration X-200C. This table identifies those containment penetrations that will be locally (Type B) leak tested.
. The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety will not increase due to this change because the new torus hatch is constructed and sealed to meet the requirements of 10 CFR 50.55a as specified in ASME Code Case N236.
Modification to TS Table 3.7-2 is appropriate to identify the penetrations that will be locally leak tested in accordance with Appendix J.
Therefore, we conclude that the proposed change is acceptable.
Environmental Consideration We have determined that the amendments do 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. Itaving made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551,.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated, do not create the possibility of an accident of a type different from any evaluated previously, and do not involve a significant reduction in a margin of safety, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will'not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendnents will not be inimical to the common defense and security or to the health and safety of the public.
j Dated: February 2,1933 The following NRC personnel have contributed to this Safety Evaluation:
George W. Rivenbark, Charles C. Graves and Douglas Pickett.
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