ML20100Q504
| ML20100Q504 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/09/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20100Q502 | List: |
| References | |
| NUDOCS 9203170198 | |
| Download: ML20100Q504 (4) | |
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UNITED STATES r;
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NUCLEAR REGULATORY COMMISSION
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SAFEYY >ALVATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION fj,jf!"? 'TO AMENDMENT N0. --170 - TO FACILITY OPERATING LICENSE NO. OP.R-19 PQWER AUTHORITY OF T!iE STATE OF-NEW YORK JAMES A, FITZPATRICK NUCLEAR POWER PLANT D($1;ET NJ. 50-333 F
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- anuary.9, 1992, the Power Authority of the State of New York 7tne lict<
Snitted a request for changes to the James A. FitzPatrick Nuclear Poa f
., Technical. Specifications (TS).
The requested changes would :permf t hyWstatic pressure and leakage testing of the Reactor Coolant J
Sygem (RC5) as re.luired by Section XI-of the American Society of Mechanical Engin9ers (ASME) Boiler and-Pressure Vessel (B&PV) Code at RCS temperatures excee d,; 212: degrees F.--During'this testing, the High Pressure Coolant Injection (HPCI), Reactor Core -Isolation Cooling (RCIC),: and the Automatic -
Depressurization System (ADS)/ Safety Relief Valves (SRV) are.not'-required to be mpurable.:
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4 Hydrostatic testing and system 'akage testing of the Reactor Coolant System
- is required by Section XI of-'re A5ME B&PV code.
NRC Generic: Letter 68-11,
'NRC Position on Radiation Embr:ttlement of React r. Vessel Materials and Its Impact on Plant-Operations," is used to calculatt the reactor pressure vessel pressure and temperature;(P-T) limits required for;this test.
The P-T curves defining these limits are_pericdically recalculated to consider.the results of analyses of-irradiated surveillance-specimens to account for accumulateo reactor. fluence.
The current curves; require that these tests be conducted at RCS temperatures-approaching 190= degrees F.
Because decay: heat and mechanical heat used to H
-heat the~ reactor coolant do not allow exact control, the operators require p
margin to maintain the-test temperature between the ininimum temperature limit-and the maximut temperature linit of 212; degrees F.
Furthermore, in the future, these curves will be revised to require temperatures that. exceed 212 degrees F as. the' accumulated fluence increases. An extrapolation from the i
' minimum test temperature at 16 effective full power years (EFPY) indicates
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that minimum testing temperature will peak at about-250 degrees F at.32 EFPY.
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As reactor fluence increases, the minimum test temperature and pressure rises into ranges normally. associated with startup or hot shutdown.
RCS oressure and temperature are used throughout the TS as a basis for establishing system operability requirements.
However, sore Limiting Conditions for Operation (LCO) cannot be satisfied during hydrostatic and leak tests at elevated' temperatures.
Specifically, certain LCOs for HPCI, RCIC, ADS, and the SRVs cannot be satisfied during these tests for reasons discussed below:
1.
TS 3.5.C:
Requires that the HPCI systein be operable when irradiated fuel is in the vessel, the reactor pressure is greater than 150 psig, and the reactor coolant temperature is greater than 212 degrees F.
HPCI cannot be made operable during the test because piping r.ormally filled with steam is filled with water during the test.
2.
TS 3.5.D:
Requires that the ADS system be operable when irradiated fuel is in the vessel, reactor pressure is greater than 100 psig, and prior to startup from the cold, condition.
The ADS has not been evaluated for operability in the water-solid condition and may not-be operable.
- 3..TS 3.5.E:
Requires that the RCIC System be operable when irradiated fuel is in the vessel, the reactor pressure is greater than '50 ps19, and the reactor coolant temperature is greater than 212 degrees F.
RCIC cannot be made operable during the test because piping normally filled with steam is filled with water during the test.
4.
TS 3.6.E:
Requires the SRVs to be operable when the reactor coolant system exceeds atmospheric pressure and temperature is greater than 212 degrees F.
The SRVs will have to be gagged closed when test pressures exceed the SRV setpoints thus rendering them inoperable.
As stated above, the required hydrostatic pressure and inservice leak testing cannot be conducted without making HPCI, RCIC, ADS, and SRVs inoperable.. The proposed changes to the TS will-allow testing to be conducted at elevated temperatures with these systems inoperable.
'3.0 EVALUATION As outlined in Chapter 6 of the Updated Final Safety Analysis Report (UFSAR),
-" Emergency Core Cooling System -(LCCS)," in' the event. of a Loss-of Coolant Accident (LOCA), the ECCS is designed to remove residual heat including stored heat.and heat generated.due to radioactive decay, such that excessive fuel clad temperature is prevented.
The objective of the ECCS is to limit, in conjunction with primary'and secondary containments, the release of radioactive materials to.the environs following a LOCA so that resulting radiation exposures are kept within the guideline values given in 10 CFR Part.100.
In order to satisfy the Safety Design Bases, four systems are provided for emergency core cooling:
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HPCI System i
2.- Automatic Depressurization Syste.n (ADS) 3.
Core Spray System
-4.
.LPCI, an operating mode of the RCR System These are, in cddition to the other systems which supply core coolant, feedwater, control rod drive (CRD) hydraulic pumps, and RCIC.
The manner in which the ECLS operate to nrotect tha core is a function of.the rate at which coolant is lost from the break in the Reactor Coolant Pressure Boundary.
The HPCI System is designed to operate while the Reactor Coolant System is at high pressure.
The Core Spray System and LPCI are designed for operation at low pressures, if the break in the Reactor Coolant Pressure
-Boundary is of such isize that the loss-of-coolant exceeds the capacity of the HPCI System, Reactor Coolant System pressure drops at a rate fast enough to allow the Core Sp"1y System and LPCI to pump additional coolant into the reactor vessel in time to cool the cot e.
~Autone.ic depressurization is provided to automatically reduce Reactor Coolant System pressure if a break has occurred and vessel water level is not maintained by the HPCI System and other water addition systems.
Rapid depressurization of the Reactor Coolant System is desirable to permit flow from the Core Spray and LPCI Systems to enter the vessel, so that the temperature rise in the core is limited.
During hydrostatic testing and systen leakage testing of the RCS, the Recirculation pumps are in operation and a water-solid condition is maintained to control the necessary pressure and temperature.
Reactor water makeup, pressure, and level is controlled using the Control Rod Drive and Reactor Water Cleanup systems.
During the tests, all control rods are inserted to ensure the core remains subcritical and adequate subcriticality margins are maintained.
Furthermore, the decay heat level is minimized following the refueling or maintenance activities, and the reactor is maintained at.or near cold shutdown conditions.
During the hydrostatic pressure and leak test conditions, the postulated worst case accident is a LOCA.
The effects'of a small or large break LOCA are bounded by the existing plant analyses.
This is assured by the following test conditions: -the control rods are maintained fully inserted to maintain subcriticality margins, the reactor coolant inventory is large, the reactor coolant energy (enthalpy) is significantly less than that during power operation, and tiie decay neat ls low.
With a small break LOCA, the RCS will depressurize while the operator terminates the test and initiates RHR cooling and/or low pressure ECCS,- as necessary. With a large break LOCa, the reactor will rapidly deprer.surize and all low pressure ECCS with their initiating instrumentation will be available.
The operability of these low pressure ECLS
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Systems is assured by-the requirements of TS 3.5, " Core and Containment Cooling-Systems."
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- i., Primary containment integrity will be maintained during the hydrostatic testing and system leakage te; ting of the RCS.
Furthermore. other systcms.
designed to restrict radiological release (i.e.. Secondary Containment and the Standby Gas Treatment System) will also be available.
The availability of.
'these systems will assure that offsite releases remain within the guideline values of 10 CFR Part 100.
For the above reasons, the NRC-staff finds that the proposed amendment is acceptable.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 INVIRONMENTAL CONSIDERATION The 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 NRC staff has determined tnat 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 sign.i.ficant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant nazards consideration, and there has been no public comment on such finding (57 FR 4494). Accordingly, the amendment meets the slijibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(bl no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Commission has sancloded, based on the considerations discussed above,.
that:
(1) there is reasonable assurance that tha health and safety of the public will not be endangered by operation in the. proposed manner, (2) such activities will be conducted in compliance with the Commission's. regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
-Principal Contributor:
Brian McCabe Date: March 9, 1992
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Docket No. 50-333 Mr. Ralph E..Beedle Executive Vice President - Nuclear Generation Pocer Authority of the State of New York 123 Main Street White Plains, New York 10601 Daar Mr..Beedle:
SUBJECT-ISSUANCE OF AMENDMENT FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. M82630)
The Commis: ion has issued the enclosed Amendment No.179 to facility Operating
' License No. DPR-59 for the Jarnes A. FitzPatrick Nuclear Power Plant.
The amendment consists of changes to the Technicrl Specifications in response to your application transmitted by letter dated January 9, 1992.
The amendment revises the technical specifications to permit hydrostatic pressure and leakage testing of the Reactor Coolant System (RCS) as :? quired by Section XI of the American Society of Mechanical Engineers (ASME) Coller
. ar.0 Pressure Vessel (B&PV) Code at RCS temperatures exceeding 212 degrees F.
During this testing, the High Pressure Coolant Injection, Reactor Core Isolation Cooling, and the Automatic Depressurization System / Safety Relief Valves;are not required to be operable.
A copy of the related Safety Evaluation is enclosed. A Notice-of Issuance will be included in the Commission's next regular hiweekly Federal Reaister notice.
Sincerely, Original Signed By Brian C. Mr.Cabe, Project Manager Project Directorate I-l Division of Reactor Projects - !/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Amendment No.- 179to DPR-59
- 2.. Safety Evaluation cc w/ enclosures:
See next page OFFICE PDI-1:LA PDI-LPM SRXB M OGG n PDI-1:0 NAME csVogan W BCMcCabe:pch RJonds hN
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' DAN C2/19/92 -
02//i/92 02/lf/92 02_/$f92/
03/o9/92 OfflCIAL RECORD COPY I
FILENAME: C:\\WP\\WPDOCS\\M82630.AMD L
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DNTED March 9,:1992
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' AMENDMENT NO.179.T0 FACILITY OPERATING LICENSE NO. DPR-59-FITZPATRICK Dobket Files r NRC & Local 'PDRs PDI-l-Reading S. Varga, 14/E/4 J. Calvo, 14/A/4 R. Capra-C. Vogan B. McCabe C.'Cowgill OGC-WF D. Hagan, 3302 MNBB E. Jordan, 3302 MNBB B, Grimes, 9-A-1 R. Jones G. Hill (4), P-137 Wanda Jones, P-130A C. Grimes, 11/F/23 ACRS (10)
- OPA OC/LFMB
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- PD Plant-specific file
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Plant Service list f
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