ML20087N400

Forwards Revised Leakage Reduction Program for Review. Leakage Reduction Test Results Will Be Submitted After Fuel Load Since Some Sys Cannot Be Tested Until Reactor Operating

ML20087N400
Person / Time
Site:	Fermi
Issue date:	03/27/1984
From:	Jens W DETROIT EDISON CO.
To:	Youngblood B Office of Nuclear Reactor Regulation
References
EF2---67-742, NUDOCS 8404030416
Download: ML20087N400 (7)

Text

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• W;yne H. Jens .

Vca Prisk$ent Nuclear Operations 2000 Gecond Ave Edison mwnue March 27, 1984 EF2 - 67,742 Director of Nuclear Reactor Regulation Attention: Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Reference:

Fermi-2 NRC Docket No. 50-341

Subject:

. Leakage Reduction Program A revised Leakage Reduction Program is attached for your review. It has been formatted for incorporation into Section H.III.D.l.1 of.the Fermi-2 FSAR in a forthcoming amendment. The program description has been revised to more clearly-define the program and its implementation.

It should be noted that Fermi will be submitting leakage reduction test results after fuel load. This is due to the fact that some systems cannot be tested until the reactor is operating. Consultation with other utilities indicates that this approach has been previously accepted by the'NRC.

Should you have any questions concerning the above, please

. contact Mr. 0.. Keener Earle, (313) 586-4211.

Sincerely .

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cc: Mr. P. M. Byron

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Mr. M. D. Lynch B404030416 840327 PDR ADOCK 05000341 PDR n A

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H.III.D.l.1 Primary Coolant Outside Containment H.III.D.l.l.1 Statement of Concern Parts 20 and 100 of Title 10 of the Code of Federal Regulations specify radiation limits and guidelines for licensed facilities to ensure the protection of public health and safety. In a power reactor, many systems that may or will contain significant radioactive liquid and/or gas inventories after a serious transient or accident have components located outside containment. At TMI-2, the major radioactive releases appear to

.have come from' leaks in such systems. Leakage from the systems must be maintained as low as practical to prevent releases of significant quantities of. radioactive material when the systems are operated. The plant operating staff should know the leakage rate of each system and have positive control over them to ensure the maximum availability of the equipment.

H.III.D.1.1.2 NRC Position H.III.D.1.1.2.1 Full Power License Requirement Applicants shall implement a program to reduce leakage. from systems outside containment. that would or could contain highly radioactive fluids during or af ter a serious transient or accident to as-low-as practical levels. This program.shall include the following:

(1) Immediate leak reduction (a) Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.

~ ( b) Measure actual leakage rates with system in operation and report them to the NRC.

(2) Continuing Leak Reduction - Establish and implement a program of preventive maintenance to reduce leakage to as-low-as practical levels. This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle.

H.III.D.l.1.2.2 ' Dated Requirement Applicants shall submit the information requested in the " Clarification" section of this position at least 4 months prior to issuance of a fuel-loading license.

This requirement shall be implemented by applicants for operating license prior to isreence of a full power license. ;(See Section III.D.l.1 of Ref. 4).

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H.III.D.l.1.2.3 Clarifici. tion _

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f"(Applicants shall provide a Cummary dedription, together with initial

-leak-test results, of theif.hogram, to reduce leakage froT systems outside

.,r containment that would orseculd contain primary coolant or'other highly radioactive fluids or gases during or following a serious transient or'

' ' w e-accident.

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( (1) Systems that shed,id b'e leak tested are as follows (any other plant system which has simiMc functions or post-accident characteristics even though not specified ,herein,,should be' included): .

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/- i. Residual heat removal'(RHR) .

'v ,b. Containment spfay recirculation ,, I s, ,

4" c .'. Hi W gh opressure injection recirculation ,'

d.4.$ Containment anet primary coolant samp1.ing t v, y e.- Reabtor core isolation cooling [ ..

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f. +- Waste /sas (including headers and cover gas sys:em outside of

_, Y chntainment in addit'lo$ to decay or storage system).

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.(2) Testidk ,of gaseous systhms sk uld include helium leak detection or

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4.quivalent testing methods.,

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(3)' Should consider program to 'r& duce potential release paths due to design and operator'deficiencias as' discussed in NRC letter to all 4

operating nuclear power plan'ts' regarding North' Anna and related incidenis, dated October '17, ;1979. O

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H.III.D.l.l.2.4 AppJicabiity[ '

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.This requirement applies to all operat-ing J cense applicants.

- H.III.D.1.1.3 Detroit Edison Position-f Detroit- Edison has developed a -Leakage Reduction Program to reduce and maintain-leakage to as-low-as practical from systems outside primary con-tainment ,that could or would contain 1-13hly radioactive fluids during

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and/or af ter a serious transient or ' accident. This program is based on Requirement 2.1.6a of NUREG-0578 (Reference 1) and the ~ requirements of item III.D.l.1 of NUREGs 0660, 0694 and 0737 (References 2, 3 and 4 respectively). ':

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H.III.D.l.l.3.1 Program Scope LTable H.III.D.l.1-1 identifies systems included in the Leakage Reduction Program. . Table H.III.D.l.1-2 lists systems to which the Leakage Reduction

. Program is not' applicable .and futher provides the justification for their exclusion. Only the systems listed in Table H.III.D.I.1-1 are included in the program.

H.III.D.l.l.3.2 Program Description The Detroit Edison Leakage Reduction Program includes the following features:

a. A combination of periodic visual inspections on accessible portions of the-systems and detailed system walkdowns to identify leakage into secondary containment out of components such as valve stems, pump seals, fittings, relief valve discharge lines, drains, vents and instrument loops. When possible, these inspections. are performed with the systems at approximately operating pressure in a normal or test condicion.

b. An aggressive maintenance program is utilized to correct identified leakage problems and assign a high priority to leakage related work requests for systems in this program. Essentially all leakage on concerned (i.e., those identified in Table H.III.D.l.1-1) systems will be addressed. These preventive and corrective maintenance measures ensure minimum leakage on a continuing basis.

c. Periodic leak rate testing of systems (those listed in Table H.III.D.l.1-1) and system components such as valves at intervals not to exceed each refueling outage. The general test methads used to. determine leakage from systems within the scope of this Leakage Reduction Program are provided in paragraph H.III.D.I.l.3.3.

-d. Records are maintained on inspections and tests performed and are used to identify chronic or-generic leakage problems in order to

implement modifications and/or corrective maintenance measures.

These records are also made available to the plant operators.

Approximately about the time full power is achieved, Detroit Edison will have collected the necessary data and will submit'to the NRC staff a report of the recorded leakage and preventive / corrective maintenance performed as the direct result of the evaluation of this leakage. The-report will also identify general-leakage criteria to be applied during the first fuel cycle as the basis for instituting corrective action in the form of~ preventive maintenance. Prior' to 'the start of the 'second fuel. cycle, Detroit Edison will revise the general criteria to the extent necessary based on the 100/LIC-6/5.2

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experience gained during the first operating cycle of Fermi 2. These revised criteria will be used as the basis for the long term leakage reduction / monitoring program for EF-2.

NOTE: ' In addition to this testing program, system leakage tests will be performed on many of these systems as part of the 10CFR50, Appendix J leakage testing program. The systems and components that are subject to this testing and which comprise the containment boundary are identified in' Table 6.2-2 of this FSAR.

H.III.D.1.1.3.3 Test Methods a) Liquid Systems - Systems or portions of systems that could coatain radioactive liquids during and/or after an accident are periodically placed into normal' operation or a testing mode.

'During these test conditions the systems are visually inspected for_ leakage with all results being recorded. Leakage detected

-during the paciodic visual inspections or the less frequent integrated leakrate test, will be measured where possible, and recorded. Techniques used for leakage measurement include collection into a graduated container and estimation by equating drops per unit of time to a standard volume.

b) Gaseous Systems - For systems or portions of systems that may contain radioactive gases during and/or after an accident, a pressure drop or make-up gas rate test is used. Clean air or nitrogen is used for these tests. When leakage is indicated by a pressure drop or excessive make-up, visual inspection techniques are applied to componenta during pressurization. The most common method of visual inspection will be the application of leak-detection fluid to suspected points of leakage (i.e., valve stem packings & air pump seals). The application of the helium leak detection method of inspection may be considered for some gaseous systems.

H.III.D.1.1.3.4 Test Procedures Each system identified in Table H.III.D.1.1-1 will have a surveillance testing procedure (s). These test procedure will contain the following elements as applicable:

a). A description of system and plant operating conditions necessary I to conduct each leak test. Test boundaries are identified and

' include only those portions of the system that could contain radioactive fluids during and/or after an accident. For example, the Core Spray suction piping from the condensate storage tank would not be inspected as this suction line is used for test purposes only and would not contain radioactive fluid during or after an accident.

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b) Elaboration.of special test ccthods necessary to supplement general test methods.

c) Data sheets listing .the specific areas to be inspected. These data sheets will identify isometric drawing numbers and provide spaces to record inspection results.

H.III.D.l.1.3.5 References 1.- U.S. Nuclear Regulatory Comission, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations, NUREG-0578, July 1979.

. 2. U.S. Nuclear Regulatory Commission, NRC Action Plan Developed as a Result of the THI-2 Accident, NUREG-0660, Vols.1 and 2, May 1980

3. U.S. Nuclear Regulatory, Comniselon, TMI-Related Requirements for New Operating Licenses, NUREG-0694, June 1980.

4. U.S. Nuclear Regulatory Commission, Clarification of TMI' Action Plan Reguirements, NUREG-0737, October 1980.

5.. ASME Boiler and Pressure Vessel Code, Section Kl.

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, *e EF-2-FSAR TABLE H.III.D.l.1-1 SYSTEMS OUTSIDE PRIMARY CONTAINMENT THAT COULD CONTAIN HIGHLY RADIOACTIVE FLUIDS Reactor core isolation coolinE Residual heat removal Containment Spray Suppression pool cooling

' Low pressure coolant injection Shutdown cooling Core spray Reactor water sample Reactor water cleanup Combustible gas control High pressure coolant injet ion Standby gas treatment Control rod drive discharge beaders Containment sampling system H.III.D.1.1-5 Amendment 48 - May 1983 022964

EF-2-FSAR TABLE H.III.D.l .1-2 SYSTEMS OUTSIDE PRIMARY CONTAINMENT TRAT WOULD NOT CONTAIN HIGHLY RADIOACTIVE FLUIDS System Comment RHR fuel pool cooling Not directly affected by accident.

Standby liquid control Injects fluid and does not circu-late reactor coolant.

General service water / emergency Does not circulate reactor coolant equipment service water and could become contaminated only due to system leaks.

Reactor building closed cooling Does not circulate reactor coolant water / emergency equipment and could become contaminated cooling water only due to system leaks.

Condensate storage Could become contaminated only due to isolation valve leakage.

Demineralized water makeup Could become contaminated only due to isolation valve leakage.

-Torus water management Isolated during LOCA and not required for accident mitigation.

. Control air / station air Would require system failure.

Fuel pool cooling and cleanup Not directly affected by accident.

Main steam lines Would require failure of MSIVs and failure of MSIV leakage control s ys tem.

Feedwater lines Would require failure of isolation valves.

Drywell cooling system Uses RBCCW or EECW and is not

- needed for safe shutdown of plant.

RHR steam condensing Not required for accident Mitiga-tion Reactor building floor / equipment Not required for accident mitiga-tion. Minimizing leakage from l systems in Table H.III.D.l.1-1 minimizes input to this system.

Radwaste Not required for accident mitigation.

H. III . D. l .1-6 Amendment 48 - May 1983 100 /LIC-6 /5.6 030284