ML19338E101

Forwards Request for Addl Info & Power Sys & Mechanical Engineering Branch Positions Re Overload Protection Design & RCPB Safety Sys Respectively

ML19338E101
Person / Time
Site:	Farley
Issue date:	09/10/1980
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Clayton F ALABAMA POWER CO.
References
NUDOCS 8009240551
Download: ML19338E101 (9)

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September 10, 1980 Docket No. 50-364 Mr. F. L. Clayton, Jr.

Senior Vice President Alabama Power Company Post Office Box 2641 Birmingham, Alabama 35291

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR FARLEY 2 OPERATING LICENSE APPLICATION As a result of our continuing review of the operating license application for the Joseph M. Farley Nuclear Plant Unit 2, we have developed the enclosed request for additional information and position.

Please provide the information requested in the enclosure.' Our review schedule is based on the assumption that the additional information will be available for our review by September 22, 1980.

If you cannot meet this date, please i

inform us within 7 days after receipt of this letter so that we may revise our scheduling.

Sincerely, Robert L. Tedesco, Assistant Director for Licensing Division of Licensing i

Enclosura:

Request for Additional Informa tio' cc w/ enclosure:

See next page 9

800g240 5 5s p

-O Mr. F. L. Clayton, Jr., Senior' Vice.

President Alabama Power Company Post Office Box 2641 Birmingnam, Alabama 35291 cc.: Mr. Alan R. Barton Executive Vice President Alabama Power Company Post Office Box 2641 Birmingham, Alabama 35291 Mr. Ruble A. Thomas Vice President Southern Company Services, Inc.

Post Office Box 2625 Birmingham, Alabama 35202 Mr. George F. Trowbridge Shaw, Pittman, Potts and Trowbridge 1800 N Street, N. W.

Washington, D. C.

20036 Mr. W. Bradford NRC Resident Inspector

.t P. O. Box 1814 Dothan, Alabama 36302 1

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ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION JOSEPH fi. FARLEY NUCLEAR PLANT UNIT 2 DOCKET NO. 50-364 Requests from the following branch in NRC are included in this enclosure.

Requests and pages are numbered sequentially with respect to requests transmitted following issuunce of SER Supplement No. 3.

BRANCH PAGE NO.

Power Systems Branch 040-13 Mechanical Engineering Branch 110-31

040-13 040.0 POWER SYSTEMS BRANCH 040.15 Your design of overload protection for the penetration conductors is inadequate._ We require that the design of overload protection for electrical penetrations installed in Class lE as well as non-Class lE circuits must provide for independent primary and backup fault protective devices to preclude a single failure from impaired the integrity of a pecetration and that the design must meet the following requirements of IEEE-279:

A.

The system shall, with precision and reliability automatically i

disconnect power to the penetration conductors when currents i

through the conductors exceed the preset limits.

j B.

All primary and backup breaker overload and short circuit protection systems shall be qualified for the. service environ-ment including seismic. However, the seismic qualification for i

non-Class 1E circuit breaker protection systems should as a minimum assure that the protection systems remain operable during an operating basis earthquake.

In addition, the non-Class lE circuit breaker and protection system shall have a high pedigree.

C.

The circuit breaker protection ' system trip set points and breaker co-ordination between primary and backup protection shall have +.he capability for test and calibration.

Provisions for test under simulated fault conditions should be provided.

For designs where protection is provided by a combination of a breaker and a fuse or two fuses in series, provisions shall be provided for testing fuses.

D.

No single failure shall cause excessive currents in the penetra-tion conductors which will degrade the penetration seals.

E.

Where external control power is used for tripping breakers, signals for tripping primary and backup breakers shall be independent, physically separated and powered from separate sources.

Provide a modified design that includes the above requirements.

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110-31 110.0 MECHANICAL ENGINEERING BRANCH There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure. There are also some-systems which are rated at full reactor pressure.

In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems. The leak tight integrity of _these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA.

Periodic leak testing of pressure isolation valves is required to be performed after all disturbances to the valve are complete.

Pressure isclation valves are required to be Category A or AC per IWV-2000 and to meet the appropriate valve leak rate test require-ments of IWV-3420 of Section XI of the ASME Code and as discussed below.

Limiting Conditions for Operation (LCO) are required to added to the technical specifications which will require corrective action i.e.,

shutdown or system isolation when the final approved leakage limits are not met. Also serveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.

The staff's present position is that leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute per valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degradation over a finite period of time.

Significant increases over this limiting value would be an indication of valve degradation from one test to another.

Leak rates higher than 1 GPM will be considered if the leak rate changes are below 1 GPM above the previous test leak rate or system design precludes measuring 1 GPM with sufficient accruacy.

These items will be reviewed on a case by case basis.

The Class 1 to Class 2 boundary will be considered the isolatinn point which must be protected by redundant isolation valves.

In cases where pressure isolatica is provided by two valves, both will be independently leal tested. When three or more valves provide isolation, only two of the valves need to be leak tested.

Provide a list of all pressure isolation valves included in your testing program which will be categorized "A" or "AC" and discuss how your testing program will conform to the above staff position.

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EllCLOSURE REQUEST FOR ADDITIONAL INFORMATION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2 DOCKET NO. 50-364 Our review of your " Response to the TMI-2 Action Plan" submitted June 20, 1980 has resulted in the need for additional information. Requests and pages are numbered sequentially. The alpha numeric item designations correspond to the items in the TMI-2. Action Plan. The following requests are included in this enclosure.

Request No.

Page No.

9.

7 10.

7 11.

9 12.

9 L-

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9 9.

Your response to item 2.1.6.b/11.8.2 " plant shielding" is incomplete.

Provide the following information:

1.

Item 7b - last paragraph - Figures showing all radiation zones 4

i used in your plant shielding analyses to determine direct radiation i

and sources of indirect radiation. All figures should be clearly legible.

t 2.

Item 7c - the projected doses to individuals for necessary occupancy times in access-controlled areas in the auxiliary building.

3.

Item 7f - specify completion dates.

4.

For other modifications, which allow access to areas where access would be useful but not vital, you should specify the anticipated modifications and the scheduled completion date for modification.

4 It is our position that all items in Section [must be completed' prior to full power licensing except plant shielding modification which is January 1,1981, or full power operation, whichever is later.

4

10. provide a description of the two high range c'ontainment monitors required-and specify the location of these monitors inside containment. The description of the monitors should include:

1.

name of manufacturer and model number of the monitors; 2.

verification that the monitors meet the specifications of Table II.F.1-3; 3.

verification that the monitors are or will be operable on January 1, 1981, or prior to full power operation, whichever is later, and, 4.

a plant layout drawing showing the location of the monitors.

5.

The monitors should be located in a manner as to provide a reasonable assessment of area radiation conditions inside contain-ment. The monitors should be widely separated so as to provide independent measurements and should " view" a large fraction of the containment volume. Monitors should not be placed in areas which are protected by massive shielding and should be reasonably access-ible for replacement, maintenance, or calibration. Placement high in a reactor building dome is not recommended.

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i TABLE II.F.1-3 HIGH RANGE CONTAINMENT RADIATION MONITOR l

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REQUIREMENT The capability to detect and measure the radiation level l

within the reactor containment during and following an accident.

1 rad /hr to 108 rads /hr. (beta and gamma) or alternatively RANGE 1R/hr to 107 R/hr (gamma only) (overlap with normal radiation i

monitor (s) range by a factor of ten (10)).

60 kev to 3 MeV photons, with + 20% accuracy for photons of

RESPONSE

O.1 MeV to 3 MeV.

A minimum of two physically separated monitors (e.g.,

l REDUNDANT monitoring widely separated spaces within containment).

Per Regulatory Guide 1.97, Revision 2, Table 1, Instr. ment RELIABILITY Category 1.

SPECIAL In-situ calibration by electronics signal substitution is CALIBRATION acLaptable for all range decades above 10 R/hr.

In-situ calibration for deca e below 10 R/hr shall be by means of calibrated radiation source.

SPECIAL ENVIRONMENTAL QUALIFICATIONS-Vendors shall calibrate and type test representative specimens of detectors on at least one point in each deca '

range from 1 R/hr up to 106 R/hr. Vendors shall provit

.2rti fication of calibration of each detector for at least one psint per l

decase of range beis3en 1 R/hr and 103 R/hr.

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11. The Instrumentation and Control System Branch has reviewed the applicant's response to Item II.F.2 to determine the degree to which the proposed instrument system, a subcooling meter, meets the applicable criteria of the current draft version of Regulatory Guide 1.97, Revision 2.

Provide additional information on the Farley Unit 2 subcooling meter design to indicate how the design meets the requirements of Regulatory Guide 1.97, Revision 2, Table 1, Instrument Category 3.

Any deviations should be reported and either corrected or justified.

In addition, demonstrate the appropriate criteria have been developed to assure that th'e o'utput of the computer will be accurate within the overall uncertainty noted in Table 1 of the Farley 2 submittal. General criteria for safety computer soft-ware are being developed for the NitC at Oak Ridge National Laboratory.

Our consultants at ORNL will be available to discuss these criteria with the applicant if necessary. Provide also a description of periodic tests of the subcooling meter to confirm the accuracy of the calculator for temperatures and pressures anticipated during accidents.

12. Provide sample typeout sheets for the incore thermocouple data, including the core map display and the trend typewriter typeout.

Explain all synbols and numbers. For each display give the location of the printout (e.g. control room) and time to obtain typeout.