ML20098E985

Forwards New &/Or Revised Pages to 840724 Equipment Qualification Program Justifications for Continued Operation Transmittal.Revs Due to Changes in Qualification Status of Certain Equipment

ML20098E985
Person / Time
Site:	Hatch
Issue date:	09/26/1984
From:	Gucwa L GEORGIA POWER CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
NED-84-508, TAC-42505, TAC-42506, TAC-55488, TAC-55489, NUDOCS 8410020213
Download: ML20098E985 (9)

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. Georgia Power Company 333 Piedmont Avenue Atlanta, Georg4 30308

. Telephone 404 526-6526 Maling Address:

Post Office Box 4545 Atlanta, Georgia 30302 Georgia Power L T. Gucwa tre southern e:ectnc system Manager Nuc%ar Engineering and chief Nuclear Eng neer Septenber 26, 1984 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S.' Nuclear Regulatory Cmunission Washington, D. C. 20555 NRC D00 7S 50-321, 50-366 OPERATING LIGNSES DPR-57, NPF-5 EDWIN I. HA701 NUCLEAR PIANT UNITS 1, 2

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UPDATE OF EQUI 1 MENT GUALIFICATION PROGRAM JUSTIFICATIONS FOR CONTINUED OPERATION ,

. Gentlenen:

On July 24, 1984, Georgia Power Company (GPC) subnitted as letter NED-84-395 the 10 GR 50.49 equipnent qualification progran Justifications for Continued Operation (JOOs) which were still effective for Plant Hatch.

Since that subnittal GPC has determined that revisions to seven of the Hatch Unit 1 JOOs and addition of one new JCO are required due to changes which have occurred in the qualification status of certain equipnent.

Enclosed are the eight new or revised Unit 1 JOOs. These pages should be used to replace the pages with identical attachnent and page ntsnbers which were transnitted by our July 24, 1984 letter. It should be noted that pages 2 and 3 of Attachnents 1 and 2 of the July 24, 1984 subnittal are deleted since the equipnent covered by those JCOs is now fully qualified.

.In addition, page 46 of Attachnent 2, enclosed with this letter, is a new

-JOO to cover a limit switch which was recently added to the scope of the Hatch equipnent qualification progran.

Attachnent 3 to our July 24, 1984 letter has not been revised since there have been no changes to the Unit 2 JOOs since that date. The three attachroents to our July 24, 1984 letter, along with the revisions now being

- subnitted, justify -continued operation of Plant Hatch equipnent for which complete environnental qualification per 10 ' CFR 50.49 has not yet been

demonstrated.

Very truly yours, 8410020213 840926

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L. T. Gucwa

. CBS Enclosures xc: H. C. Nix, Jr.

J. P. O'Reilly go Senior Resident Inspector i l.

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REV. I ATTACHMENT 1 b

T47-N003, T47-N009 and Cable N3-04 These are RTD instruments and associated cable which monitor the drywell temperature in the area of the reference legs on the RPV water level instruments. The instruments are used by the operators to density compensate the level instruments based on drywell temperatures.

The level instruments do not require the density compensation in order for all automatic system actuations to take place to assure plant safety. The temperature instruments in conjunction with the level instrument will be used by the operator to confirm proper injection system operations. The maximum expected error in the level instruments is 25 inches which corresponds to a drywell temperature of 340 F. The operator is very aware of this error and if in doubt will apply the maximum error in the safe direction. If the instrument provided an incorrect reading the operatcr could be misled but the automatic system actuations which occur on high and low RPV level will protect the core and will alert the operator to the fact that he is being misled by the temperature elements. Therefore, plant safety is assured.

Based.on the above, continued operation is justified.

24.

REV. 1 ATTACHMENT 1 CABLES N1-4, H1-16, M5-20, N3-03, and N3-19 /i\

The specific cables noted above are of the PVC type which have not been tested to the standards required by the D0R guidelines which are an attachment to I.

E.Bulletin 79-018. Justification for continued plant operation until replacement of the cables is achieved is principally based on IEEE paper entitled" Insulation and Jackets for Control and Power Cables in Thermal Reactor Nuclear Generating Stations, (IEEE Transactions on Power Apparatus and Systems Vol. PAS-88, No. 5, May 1969)."

A key consideration in the evaluation of this report relative to the acceptability of the cable installed at_HNP-1 is that under no circumstances will the accident dose of radiation occur concurrent with the harsh environment created by a high energy line break since all the subject cables are located exterior to the drywell. With this in mind radiation can be evaluated independently of tne high energy line break.

ihe IEEE paper in Table X1 concludes that PVC has an "Overall threshold of damage" of 5 X 510 Rads. Were it not for the failure of the test cable in a simulated steam environment after being irradiated to 5 X 106 Rads the overall 6

threshold of damage would have been 5 X 10 Rads. Since we are not considering a steam environment concurrent with the maximum radiation, the 5 X 106Rads threshold of damage compares favorably with the highest actual accident dose of 6

1.86 X 10 Rads in the area of the NE and SE corner rooms.

The IEEE paper in Table II indicates that the useful life of PVC cable in terms of elongation loss is greater than 115 years even after being irradiated to 5 X 106 Rads. Although further evaluations of cable life are possible, these evaluations are not necessary because the cable is being replaced.

The maximum accident temperature of 214 F presents no significant problem because PVC cable, as pointed out in the IEEE paper on page 534, has a useful life of 200 hous at 136 C (276 F). Additionally, a steam environment presents no signi-ficant problem because PVC cable, as pointed out in the IEEE paper on page 534 l

and Table IX, was found suitable in a steam environment greater than 9 days.

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REV. 1 ATTACHMENT 2 E51-N023A&B, E51-N026A&B, E51-N025A&B, E51-N027A&B fii These instruments are temperature sensors which monitor temperatures in areas where the RCIC steam line is located and initiate an isolation signal in the event of a steam leak in the RCIC steam line.

The instruments could be subjected to a harsh environment due to radiation in the unlikely event of a LOCA. No credit is taken in the accident analysis for the. operation of these switches for the mitigation of a LOCA.

In the event of a HELB the instruments could be subjected to a harsh environment due to temperature and pressure. A high degree of confidence exists that the instruments will operate to isolate the HPCI steam line break since test experience shows that the switches will operate up to a test temperature of 212'F. The set point of these switches is 170*F which is well below the temperature for which test experience exists.

This piece of equipment has been used during normal operation in plants similar to Hatch I for the past eleven (11) years. To date, no age related common mode failures have been reported. These devices have experienced relatively limited service in Hatch 1 as they have been used during normal operation for only five (5) years. With this limited service, this equipment is not expected to fail before replacement.

Based on the above, continued operation is justified.

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• REV. 1 ATTACHMENT 2 E41-C002(LS-4)

This limit' switch is attached to the HPCI Turbine stop valve and controls the

- signal which opens the HPCI system discharge valve as well as activates the turbine regulator ramp generator on HPCI turbine start.

The limit switch is not required for a HELB in the HPCI room which is a rupture of the HPCI steam supply line. No credit is taken for the operation of the HPCI turbine to mitigate the consequences of a HPCI steam line break.

The limit switch could be subjected to a harsh environment due to radiation after a LOCA. The limit switch is a NAMC0 Model D200G-ST-2 and does not have any documented radiation exposure test data. The manufacturer has provided a jdhi list of non-metallic parts which are u:ed in the subject switch. Each part has been reviewed to determine its radiation damage threshold value. The results of the evaluation has demonstrated that the most susceptible part has a threshold value 5 times higher than the maximum expected total integrated dose in the HPCI room; therefore, the switch is expected to operate as designed.

Based on the above, continued operation is justified.

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