ML20141F448

Amend 73 to License DPR-61,modifying Tech Specs by Reducing Allowable Leakage Rate of Reactor Coolant Outside Containment & Changing Tag Number Designations for Two Check Valves

ML20141F448
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/14/1986
From:	Charemagne Grimes Office of Nuclear Reactor Regulation
To:
Shared Package
ML20141F440	List: ML20141F434 ML20141F448 ML20141F451
References
NUDOCS 8604230026
Download: ML20141F448 (6)

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CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET NO. 50-213 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No. DPR-61

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Connecticut Yankee Atomic Power Company (the licensee) dated January 30, 1986 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows:

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(2) Technical Specifications The technical specifications contained in Appendix A, as revised through Amendment No. 73, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the technical specifications.

3. This license amendment is effective as of the date of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Lw>dg e 1.

Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B

Attachment:

Changes to the Technical Specifications Date of Issuance: April 14,1986

ATTACHMENT TO LICENSE AMENDMENT NO. 73 FACILITY OPERATING LICENSE NO. DPR-61 DOCKET NO. 50-213 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT 3-25 3-25 3-26 3-26 3-26a

3.14 Primary System Leakage APPLICABILIIT: Applies to limiting operation of the plant under varying rates and conditions of primary plant leakage.

OBJECTIVE: To specify primary plant operability based upon primary plant leakage.

SPECIFICATIONS: A. Operation of the reactor coolant system shall be permitted by the following leakage criteria.

1. One CPM unidentified and uncontained leakage in the reactor coolant system.

2. Ten CPM in the reactor coolant system.

3. Three liters per hour combined leakage (outside containment) from:

(a), the normal makeup, seal injection, seal return, and loop fill portions of the chemical and volume control (CVCS) system and.

(b). the residual heat removal (RHR) system.

This includes thru-wall, mechanical seal, valve packing, and gasket leakage.

4. No pressure boundary leakage allowed in the reactor coolant system.

5. Primary - to - secondary leakage through the steam generator tubes shall be limited to 0.4 CPM total for all steam generators not isolated from the reactor coolant system and 150 gallons per day through any one steam generator not isolated from the reactor coolant system,

6. Leakage through each of the following ECCS valves shall not exceed one 'GPM: SI-CV-862A - SI-CV-862B -

SI-CV-862C - SI-CV-862D - SI-CV-872A - SI-CV-872B.

B. ACTION IS REQUIRED UNDER THE FOLLOWING CONDITIONS:

1. With any PRESSURE BOUNDARY LEAKAGE, be in COLD CHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> oy be in COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

3. If primary to secondary tube leakage (not including leaks originating from tube to tube sheet welds) in

- excess of specification 3.14.A.5 above, results in a cold shutdown, inservice inspections shall be per-formed in accordance with the first sample inspection in Table 4.10.1-2 during that cold shutdown.

BASIS: Uncontained leakage is flow to any open system. This includes leakage to the containment and primary auxiliary building sumps. Leakage which is attributed to a specific component is considered to be identified. Unidentified leakage would be that flow whose path is not known.

3-25 Amendment NO. X 73

4 Leakage from the primary plant can be detected in a nu=ber of ways including containment sump level, containment humidity and air particulate measurements, maintenance of water volui.e inventories and routine surveillance of charging header flow. Leakage that l is both uncontained and unidentified is undesirabic from the point of safety. A Icak rate of one CPM can be determined within a number of hours and without cause or definition the plant should be shutdown.

Uncontained but identified leakage does not constitute a safety hazard if it can be determined that operations l can safely continue. Knowledge of the source and path of the leak permits a sound judgement to be made regard-ing continued plant operation. Ten CPM is well within the minimum make-up capabilities but it is desirable to initiate plant shutdown.

Leakage that is both contained and identified does not constitute a safety hazard if it can be determined that ,

plant operations can be safely continued. 10 GPM leakage is well within the capacity of one charging pump (360 GPM) and ' makeup would be available even under the coincident loss of offsite power condition. Containing the leak within other' auxiliary systems permits control over disposition of the water volume and activity emanating from the primary system.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits,

( localized corrosion may likely result:in stress corrosion cracking. The extent of crdcking during plant operation would be limited by the limitation.of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 150 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation-will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated primary-to-secondary leakage of 150 gallons per day.per steam generator can ,readily be detected by radiation monitoring of steam generator blowdown. Leakage in excess of this limit will require plant shutdown during which the leaking tubes will be located and plugged.

Three 1,iters per hour is the amount of primary coolant I assumed in the offsite dose calculation to be released from systems outside of containment. The Residual Heat Removal (RHR) System and the listed portions of the Chemical and Volume Control (CVCS) System are likely 3-26 Amendment No. J C M 73

to be used in accident mitigation or recovery.

As such, the combined leakage from these systems, outside of containment must be maintained less than that assumed in the offsite dose calculation.

Excessive leakage through certain ECCS check valves could indicate that the valves are not performing their function of preventing reverse flow. The configuration of these valves is such that their failure to function could result in en intersystem loss-of-coolant accident.

Reference:

FDSA; Section 5.2.1 FDSA; Section 10.4.4 3-26a Amendment No. 73