ML20045G711

Forwards Writeups for GSI 73,113,120 & 151 Re Detached Thermal Sleeve,Dynamic Qualification Testing of Large Bore Hydraulic Snubbers,Online Testability of Protection Sys & Reliability of ATWS Recirculation Pump Trips,Respectively

ML20045G711
Person / Time
Site:	05200001
Issue date:	07/02/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
REF-GTECI-073, REF-GTECI-113, REF-GTECI-120, REF-GTECI-151, REF-GTECI-NI, TASK-073, TASK-113, TASK-120, TASK-151, TASK-73, TASK-OR NUDOCS 9307150083
Download: ML20045G711 (5)

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GENuclebrEnergy.

e Genera:Deanc Company 175 Cu@er Aanse San Jose, CA 35125 July 2,1993 Docket No. STN 52-001 Chet Poslusny, SenMr Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Schedule - GSI3 73,113,120 and 151

Dear Chet:

Encimed are writeups for GSIs 73,113,120 and 151 as identified in my May 18,1993 letter.

Ple,se provide a copy of this transmittal to Melinda Malloy.

Sincerely, 60 7

Jack Fox Advanced Reactor Programs cc: Alan Beard (GE)

Norman Fletcher (DOE)

Bernie Genetti (GE)

Carl Szybalski (GE)

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1 19.B.2.37.1 73 Detwhed Thermal Sleeve ISSUE During the period 1978 to 1980, there were reports of fatigue failure of thermal sleeve assemblies in the piping systems of both BWRs'and PWRs. The BWR problem was addressed by GE in NEDO-21821 (Reference 1) and was resolved with staff SER (Reference 2) and the publicadon of NUREG-0619 (Reference 3). Fatigue problems occurred subsequently in 1982 in some PWRs. In one instance, flow induced vibrations caused fadgue failures at the thermal sleeve attachment welds and subsequent cracking and tearing away of the thermal sleeves.

A CCEPTANCF CRITFRI A The feedwater nonle thermal sleeve may be subject to flow induced vibradon originating frrv wtex shedding from the feedwater sparger which is located in the downward coolant flow in the annulu' 'setween th e RPV wall and the shroud. The acceptance criterion for the feedwater nonle thermal sleeves is to demonstrate by realysis that vortex shedding from the feedwater sparger does not cause excessive vibration in the thermal sleeve. The objective of the design is to ensure a natural frequency of the thermal sleeve assembly that is three times the vortex shedding frequency from the sparger.

RFROI.imON In selecting the so-called welded thermal sleeve design,it was demonstrated by analysis that the natural frequency of the feedwater nozzle thermal sleeve is more than three times higher than the calculated vortex shedding frequency from the sparger. In calculadng the natural frequency it was conservatively assumed that the thermal sleeve is a cantilever.

Welded feedwater thermal sleeve designs (tuning fork design) has successfully been used in several domestic and overseas BWRs since 1977. There has never been a reported failure due to flow induced vibration which proves that the welded design used in the ABWR is acceptable, and, therefore, resolves this issue 73 for the ABWR.

REFERENCES 1.

NEDO 21821," Boiling Water Reactor FW Nozzle Sparger Final Report," GE,1978.

2.

letter to R. Gridley (GE) from D. Eisenhut, " Safety Evaluadon for the GE Topical Report NEDE-21821 02,'BWR Feedwater Nozzle /Sparger Final Report, Supplement 2," January 1980.

3.

NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Crachng", U. S.

Nuclear Regulatory Commission November 1980.

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19 B.2.64 11 % DYNAMIC Of f AIIFICATION TF_ STING OF L ARCF BORF HYDR Alfl_IC RN!!BRFRC.

ISSI'E issue 113 in NUREG-0933 (Reference 1), addresses the need for requirements for dynamic qualification testing -

oflarge bore hydraulic snubbers (>$0 kips load rating). Qualification tests oflarge bore hydraulic snubbers typically utilize a shutoff valve in place of the snubber control valve. To assure operability of the snubber control valves when subjected to dynamic loads, testing should be performed to determine the operational characteristics of the snubber control valve ACCFFF ANCF CRITFRI A i

ne acceptance criteria for the resolution ofIssue 113 for the ABWR design are the performance of dynamic -

tests in accordance with SS AR Section 3.9.3.4.1 (3).

RFSOLUTION For the ABWR design,large bore hydraulic snubbers will not be used as extensively as in some operating

_i plants. However, when they are used the required dynamic tests will be performed to confinn the operational characteristics of the snubber control valve.

RFFFRFNCES T

1.

NUREG-0933, "A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC, December 1992.

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19B.2A9 120 ON.LINE TESTABIt ITY OF PROTECTION SYSTEMS issus issue 120 was established to examine the on.line (at-power) testability of protection systems and the possibility that some plants might not provide complete testing capability. Protecdon systems consist of the reactor protection system (RPS) and the engineered safety features actuation system (ESFAS) (Reference 1).

ACCEI'TANCE CRITERI A De acceptance criteria for the resolution of Issue 120 is compliance with General Design Criterion (GDC) 21,

" Protection System Reliability and Testability" of Appendix A to 10 CFR 50 (Reference 4. Supplementary guidance is provided in Regulatory Guides 1.22 and 1.118 (Reference 2) and IEEE Standard 338(Reference 3) to ensure that protecdon systems (including logic, actuation devices, and associated actuated equipment) will be designed to permit tesdng while the plant is operating without adversely affecting the plant's operation. These requirements apply to both the RPS and the ESPAS. Existing Standard Technical Specification indicate that it is desirable to test all protection systems every 6 months.

RFcOLUTION in the ABWR design the RPS and ESFAS can be tested during reactor operation by six separate tests. The first five tests are primarily manual tests and, although each individually is a partial test, when combined with the sixth test they constitute a complete system test. The sixth test is a self test of the safety system logic and control which automadcally tests the complete system, excluding sensors and actuators. Online testability of protection systems is explained in Section 7.1.2.1.6. In the ABWR design, all actuation logic is solid state and in software.

Automatic system self-testing occurs during a portion of every periodic transmission period of the data communication network. Since exhaustive tests cannot be performed during any one transmission interval, the test software is written so that sufficient overlap coverage is provided to prove system performance during tests of portions of the circuitry, as allowed in IEEE 338 (Reference 3).

Herefore, this issue 120 is resolved for ABWR.

REFERENCES 1.

NUREG-0933,"A Prioridzation of Generic Safety Issues" (with supplements), U.S. NRC, July 1991.

2.

Code of Federal Regulations, Title 10 Part 50, Section 1.22 (Periodic Testing of Protection System Actuation Functions) and Section 1.118. (Periodic Testing of Electric Power and Protection Systems).

3.

IEEE Standards 3381977 " Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems".

4.

10CFR50 Appendix A," General Design Criteria for Nuclear Power Plants" Office of the Federal Register, National Archives and Records Administration.

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19 B.2.56 151 RFI I ABILITY OF ANTICIPATED TR ANSIENT WITHOliT SCR AM RFCIRClit. ATION PINP TRIP IN BWRs IS5t!E

-i Issue 151 in NUREG-0933 (Reference 1),' addresses the issue of the reliability of the ATWS RPT in BWRs.

i Issue 151 specifically identifies a reliability problem with GE's type AKF-25 circuit breaker and trip hardware,

~l (actually a type AKF-2-25 breaker, per NRC's IE Notice 87 12, Reference 2).

1 ACCEirTANCE CRITERIA He acceptance criterion for the resolution of Issue 151 is the use of reactor recirculation system pump trip -

i hardware or method that is more reliable then the previously used AKF-2 25 breaker hardware or method.

j RFROLtTTION De design for the ABWR reactor recirculation system and RPT method and hardware is completely different from the previously designed BWR reactor recirculation systems and RPT trip methods. De design is more diverse and redundantly reliable. Rather than using only two recirculation pumps and the associated single RPT breakers, the ABWR will use ten pumps and multiple pump and RPT trip logic, circuits and hardware. Adjustable speed drive (ASD), recirculation internal pumps (RIPS) are used. The ABWR RFT trip hardware (not yet specifically identified) will be completely different: Instead of using AKF 2-25 breaker switching hardware to provide a RPT, RFC controller switching and ASD gate inverter turn-off circuit hardware provides the RFT. See Subsection 7.7.1.3(7) and 7.7.1.3(8). Dus, by diversity and redundancy in design, the ABWR addresses and resolves hsue 151.

REFERENCES i

1.

NUREG-0933,"A Prioritization of Generic Safety Issues" (with supplements), U.S. NRC, July 1991.

2.

IE Information Notice 87 12 Potential Problems with Metal Clad Circuit Breakers, General Electric Type AKF-2-25", U.S. NRC, February l3,1987.

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