ML20210U123
| ML20210U123 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/21/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210U113 | List: |
| References | |
| TAC-60895, NUDOCS 8606020176 | |
| Download: ML20210U123 (4) | |
Text
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NUCLEAR REGULATORY COMMISSION 5 (-
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.125 TO FACILITY OPERATING LICENSE N0. DPR-57 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORfTY M EORGIA CITYOFDALTON,G'E0RGB
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EDWIN I. HATCH NUCLEAR P_LANT, UNIT N0. 1 DOCKET NO. 50-321 1.0 I_NT_RODUCTION By submittal dated March 7, 1986 the Georgia Power Company has proposed a Technical Specification change to permit a temporary increase in the Edwin I. Hatch Nuclear Plant Unit No. 1 main steam line high radiation scram and isolation setpoints to facilitate the testing of hydrogen addition water chemistry at their Hatch plant. This proposed change is necessary since, on the basis of prior experience, it is anticipated that main stean line radiation levels may increase during the test by a factor of five over the routinely experienced dose rates.
In addition, in response to discussions with the staff, GPC, by letter dated April 22, 1986 provided details concerning the dose control measures and radiological surveillance efforts planned in support of the testing.
2.0 EVALUATION 2.1 HIGH RADIATION SCRAM AND ISOLATION SETPOINTS The Main Steam Line Radiation Monitoring (MSLRMs) provide reactor scram and reactor vessel and primary containment isolation signals upon detection of high activity levels in the main steam lines. Additionally, these monitors serve to limit radioactivity releases in the event of fuel failures. The proposed Technical Specification changes (to Tables 3.1-1, 3.2-1 and 3.2-8) would allow adjustments to the normal background radiation level and associated trip setpoints for the MSLRMs at reactor power levels greater than 20% rated power. The adjustments are needed to accommodate the expected increase in main steam activity levels as a result of hydrogen injection into the primary system. This is primarily due to increased nitrogen-16 (N-16) levels in the steam phase.
The licensee states that the only transient or postulated accident which takes credit for the main steam line high radiation scram and isolation signals is the control rod drop accident (CRDA).
The staff notes that for a CRDA, the MSLRMs' primary function is to limit the transport of f
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. activity released from failed fuel to the turbine and condensers by initiating closure of the main steam isolation valves and thus isolating the reactor vessel. Main steam line high radiation will also produce a reactor scram signal (reactor scram in the event of a CRDA, however, would be initiated by signals from the Neutron Monitoring System) and will isolate the mechanical vacuum pump and the gland seal steam exhaust system to reduce leakage of fission products to the atmosphere from the turbine and condensers.
Generic analyses of the consequences of a CRDA have shown that fuel failures are not expected to result from a CRDA occurring at greater than 10% power. As power increases, the severity of the rod accident rapidly decreases due to the effects of increased void formation and increased Doppler reactivity feedback. Since the setpoint adjustments will be restricted to power levels above 20% of rated power, the staff concludes that the currently approved CRDA analysis for Hatch 1 remains appropriately bounding.
2.2 RADIATION PROTECTION /ALARA The staff also has reviewed the proposed Technical Specification change to assure that the licensee has considered the radiological impl.ications of the dose rate increases associated with N-16 equilibrium changes daring hydrogen addition at BWRs. The review was also intended to determine that the licensee has adequately considered radiation protection /ALARA measures for the course of the test, in accordance with 10 CFR 20.1(c) and Regulatory Guide 8.8, "Information Relevant To Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be As Low As Is Reasonably Achievable."
An overall objective of the test is to determine general in-plant and site boundary dose rate increases due to hydrogen addition. The licensee has indicated that normal health physics /ALARA practices and procedures for Hatch will be continued throughout the test. Additionally, main steam system dose rates will be monitored by surveys on a routine basis. The licensee also indicated that specific locations will be identified where temporary shielding may be needed for long-term implementation of hydrogen injection.
The staff also has discussed with licensee representatives the details of the dose control measures and surveillance efforts planned for the hydrogen addition test. Tests of this type have been proposed and conducted at other operating BWRs following formal staff review and approval of similar Technical Specification changes. The test conditions, as identified by the vendor, as well as the measures proposed for radiation protection /ALARA at the Edwin I. Hatch Nuclear Plant Unit 1, are consistent with those utilized at the other BWRs during their successful hydrogen addition tests. None of these tests involved any significant, unanticipated, radiological exposures or releases.
On the basis of the adequacy of the licensee's radiation protection /ALARA program, utilization of special surveys to monitor dose rate increases on
. site and at the site boundary, the capability to monitor for fuel failures, as well as the success of similar efforts at other operating BWRs, the staff finds the licensee's request acceptable. Hence, the staff recommends that the Technical Specification change be approved as requested.
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COMPRESSED HYDR 0 GEN ST_0_ RAGE AND DISTRIBUTION SYSTEM The licensee's hydr. ogen addition system is designed to reduce the potential hazard to certain safety related systems from intergranular stress corrosion cracking.
Central storage of gaseous hydrogen is located outdocrs near the existing hydrogen storage facility for turbine generators.
Potential release of gaseous hydrogen outdoors is not expected to-be a significant hazard.
Past reviews of compressed hydrogen storage at other sites indicate that the release and ignition of pressurized hydrogen does not produce significant overpressures. The factors that determine this finding are the limited quantities of hydrogen that are available in compressed hydrogen storage containers and, given a release, the relatively high dispersion rate due to hydrogen bucyancy. With respect to the hydrogen distribution system, an excess flow valve is provided in the 1 inch flexible metallic hose connecting the hydrogen storage tanks with the injection system. The purpose of this valve is to limit the release rate of hydrogen in the event of a pipe break.
In the hydrogen injection area inside the plant, hydrogen monitors are provided at the booster pump and the hydrogen injection control valves.
These monitors are set to alarm and isolate the hydrogen injection system when hydrogen concentrations exceed 2%. On the basis of the above considerations, we conclude that the licensee's hydrogen addition system meets Section C.S.d of BTP CMEB 9.5.1 (NUREG 0800) and is, therefore, acceptable.
3.0 ENVIRONMENTA1. CONSIDERATION This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite; and that there should be no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 20 CFR 51.22(b), no environmental impact statement nor environmental assessment need be prepared in connection with the issuance of this amendment.
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4.0 CONCLUSION
On the basis of the considerations discussed above, the staff has concluded that-(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations.
Principal Contributor:
M. Lamastra Date: May 21, 1986
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