ML20128G302
| ML20128G302 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/16/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20128G298 | List: |
| References | |
| NUDOCS 8505300116 | |
| Download: ML20128G302 (4) | |
Text
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UNITED STATES I
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 90 TO FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 s.
1.0 Introduction By submittal dated October 2, 1984, as supplemented October 22, 1984,
- the Power Authority of the State of New York (PASNY/ licensee) proposed a Technical Specification (TS) change to permit a temporary increase in the FitzPatrick main steam line high radiation scram and isolation setpoints to facilitate the testing of hydrogen addition water chemistry at FitzPatrick.- This proposed change is necessary to..the test, since it is anticipated that main steam line radiation levels may increase by a factor of five over the routinely experienced dose rates during maximum hydrogen addition rates. The purpose of this test is to study the feasibility of
, hydrogen addition to the coolant water to inhibit intergranular stress corrosion cracking (IGSCC) at FitzPatrick. PASNY has evaluated all other i
aspects of the proposed test under 10 CFR 50.59.
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2.0 Evaluation The main steam line radiation monitors (MSLRM) are used to detect gross failure of fuel cladding during nomal operation that may be caused by-any number of mechanisms (e.
manufacturing defects, etc.)g., pellet-cladding mechanical interaction, When high radiation in the main steam lines l-is detected by the MSLRM, a reactor trip is initiated to reduce the possibility of additional failure of fuel cladding and, at the same time, the main steam line isolation valves (MSIV) are closed to limit the release l
of: fission products. To perfom this reactor trip and main steam line isolation function, the MSLRM trip setting is set high enough above background radiation levels to prevent spurious trips yet low enough t
to detect gross failures'in fuel cladding.
For abnormal operational it occurrences (transients), the analyses that are perfomed determine.
i limiting conditions of operation (LCO) and Reactor Protection System trip settings to preclude any fuel-failures by meeting specified acceptable fuel design limits (SAFDL) as required by GDC 10. Therefore, credit is not required or taken for an MSLRM-initiated trip in the analysis of transients.
In addition, no credit is taken for an MSLRM trip in accident analyses (see below for its use in the control rod drop radiological dose calculation); however, the MSLRM trip will provide a backup trip to other t
primary Reactor Protection System trips if an accident were to result in significant fuel failures.
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. In the calculation of the radiological consequences of the control rod i
drop accident (CRDA), credit is taken for the MSLRM to provide a signal to close the MSIV upon the detection of high radiation in the main steam lines. The total time required to isolate the main steam lines, together with other assumptions, detemines the amount of fission product activity transported to the condenser before the main steam lines are isolated. The CRDA analysis, however, does not take credit for a reactor trip from the MSLRM in assuring that the fuel dispersal (enthalpy) criterion is met.
For a CRDA occurring at power levels above 20% of rated power, there is a significant margin to the fuel cladding failure threshold. The CRDA
.becomes a concern only at power levels below 10%. The licensee has stated that the hydrogen addition test will be conducted at power levels above 80%
of rated power and that MSLRM setpoint readjustments will be performed only above 20% of rated power.
In addition, the licensee has stated that, during controlled power reduction, restoration to pre-test setpoint values will be performed prior to going below 20% of rated power.
If, due to a recirculation pump trip or other unanticipated power reduction event, the reactor power drops below 20% rated power with the MSLRM setpoint at its test value, control rod withdrawal will be prohibited until the necessary setpoint readjustment can be made.
The capability for monitoring fuel defects and failures will be maintained through continued operability of the main steam radiation monitoring scram and isolation system, routine radiation surveys, the perfomance of daily i
primary coolant water analyses, and the continued operability of the Steam Jet-Air Ejector Off-Gas Monitor.
Furthermore, the licensee's existing quality assurance program and operating procedures, as applied to instrument adjustments, will minimize the potential for error associated with readjusting the MSLRM setpoint.
Based on the licensee's consnitment to increase the MSLRM setpoint to its test value only when the plant is operating at power levels greater than 20% rated power, to restore the setpoint to its pre-test value prior to reducing power below 20% of rated power and to prohibit control rod I
withdrawal in the event of an uncontrolled power reduction below 20% with i.
the MSLRM setpoint at its test value, as well as the licensee's capability l
to monitor fuel defects and failures during the test, we conclude that the l
proposed TS changes are acceptable.
We have also reviewed the proposed changes to assure that the licensee l-considered the radiological implications of the dose rate increase associated with nitrogen 16 (N-16) equilibrium changes during hydrogen I
addition at boiling water reactors (BWRs).
In addition, we have evaluated l-the submittals to determine whether the licensee 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 i
Ensuring That Occupational Radiation Exposure At Nuclear Power Stations Will Be As Low As Is Reasonably Achievable."
, The licensee has indicated that normal radiation protection /ALARA practices and procedures for FitzPatrick will be continued through the test.
Additionally, main steam system dose rates will be monitored by surveys on a routine basis, particularly in accessible areas. An overall objective of the test is to determine general in-plant and site boundary dose rate increases as a result of hydrogen addition. Additionally, specific locations where temporary shielding may be needed for long term implementation of hydrogen injection will be identified.
The licensee has taken additional measures to ensure that personnel exposure during the hydrogen addition testing are ALARA. These measures are:
(1)
In-plant surveys will be taken at various hydrogen flow rates (i.e.,radiationlevels),
(2) Area radiation monitors will be logged at specific increments of hydrogen addition, (3) Site boundary surveys will be conducted with Reuter-Stokes high pressure ionization chambers for measuring N-16 gamma; (4) Gamma isotopic surveys will be conducted in the environment during the CERT (Constant Extension Rate Test) test by an outside vendor. Surveys will be compared with normal operating data before/after the test.
The staff has discussed details of dose control measures and surveillance efforts planned for the test with licensee representatives. A similar test was conducted for the Dresden 2 and Peach Bottom facilities following a staff review and approval of a similar Technical Specification change.
The measures proposed for radiation protection /ALARA at FitzPatrick are consistent with those utilized at Dresden 2 during the successful tests at that unit, where no significant unanticipated radiological problems occurred.
The licensee has a radiation protection /ALARA program which has been recognized as adequate in overall NRC appraisals and includes the capability to conduct special tests and maintenance in accordance with 10 CFR Part 20 and is consistent with the criteria of Regulatory Guide 8.8.
An ALARA review of this program will be performed and submitted to NRC within 90 days of completion of testing / implementation.
Based on the adequacy of the licensee's radiation protection /ALARA program, utilization of special surveys to monitor dose rate changes at the site boandary, the success of the initial effort at Dresden 2 and the consistency of that effort with anticipated results, and the licensee's discussion of specific radiation protective /ALARA measures to be utilized, we find that the licensee has the capability to assure worker radiological protection and keep doses as low as is reasonably achievable. Based on these capabilities
... and the licensee's planned actions, we conclude that the proposed TS changes are acceptable.
3.0 Environmental Consideration This amendment involves a change in the 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 is 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 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed maaner, and (2) such activities will be conducted in compliance with the Comission'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.
Principal Contributors:
D. Fieno, F. Witt Dated:
May 16, 1985 i
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