ML20216G952

Safety Evaluation Authorizing Request for Relief for IEEE 279-1971,Section 4.7.3 Requirements Concerning Steam Generator Water Level Control

ML20216G952
Person / Time
Site:	Farley
Issue date:	09/04/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20216G916	List: ML20216G913 ML20216G952
References
NUDOCS 9709160097
Download: ML20216G952 (3)

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. seeaHoot 3

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SAFETY EVALUATION SY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO REQUEST FOR RF1 IEF FOR lEEE 279-1971. SECTION 4.7.3 REQUIREMENTS - STEAM GENERATOR WATER LEVEL CONTROL UDB SOUTHERN NUCLEAR OPERATING COMPANY. INC.

JOSEPH M. FARI FY NUCI FAR PLANT. UNITS 1 AND 2 ,

DOCKET NOS. 50-348 AflD 50-364 -

1.0 INTRODUCHQS

- By letter dated April 29,1997, and supplemented by letter dated July 25,1997, the Southem .

Nuclear Operating Company, Inc. (SNC), proposed an altemative to the requirements of 10 CFR 50.55a(h).~ Pursuant to 10 CFR 50.55a(h), the protection system must meet the requirements of IEEE 279-1971, Section 4.7.3, which states, "Where a single random failure can cause a control system action that results in a generating station condition requiring ,

i protective action and can also prevent proper action of a protection system channel designed to L protect against the condition, the remaining redundant protection channels shall be capable of providing the protective action even when degraded by a second random failure." -

Section 50.55a(a)(3) states that altemativas to the requirements of paragraph (h) rnay be used, when authorized by the Director of the Office of Nuclear Reactor Regulation, if (i) the proposed

[4 attematives would provide an acceptable level of quality and safety or (ii) compliance with the

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specified requirements would result in hardship or unusual difficulty without a compeasat!ng '

i increase in the level of quality and safety.

SNC has proposed interim administrative controls of steam generator water level control -

(SGWLC) channels that provide reactor trip system actuation signals in conformance with the requirements of 10 CFR 50.55a(a)(3) to resolve the design deficiency idedified by the i j Westinghouse Electric Corporation (Westinghouse) in Nuclear Safety Ad.s,ory Letter _(NSAL)-

96-004, dated August 14,1996. In NSAL 96-004, Westinghouse stated that a common ~

instrument tap for the SGWLC steam generator level and steam flow channels can fall, thereby L ' resulting in_an inability _to meet the single failure criterion as required in 10 CFR 50.55a(h).

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in License Amendment Nos.104 (Unit 1) and 97 (Unit 2), SNC eliminated the low feedwater flow reactor trip based on the addition of a median signal selector (MSS) system in the SGWLC.

The MSS was installed to select the median of the steam generator water level instrument  !

channel input signals, so that the control system is prevented from acting on a single, failed .

1 protection system instrument channel. In a letter dated December 29,1993, NRC approved the

MSS change. _

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L i- In the present Farley plant design, each unit has three steam generator water level channels (1,11, and 111) per steam generator and two steam flow channels (lli and IV) per steam line. The i Channel ill level transmitter reference leg instrument tap is shared with the Channel lil steam j flow transmitter. A failure of that tap wiIl cause the associated instruments to fall high. Such a L - failure will result in the feedwater flow being shut-off and will prevent initiation of the required 1

- reactor trip when a random failure is postulated in a second steam generator water level '

l channel, per the requirements of IEEE 279-1971, Section 4.7.3. NSAL 96-004 notifed the

Westinghouse NSSS plants that because of this common tap, the existing SGWLC system did

!- ' not satisfy the requirements of IEEE 279-1971. Section 4.7.3. However, NSAL- 96-004 stated that the probability of a tap / impulse line failure coupled with a failure in another level channel of

- the redundant set (Channel I or ll) is extremely low and is, therefore, not considered to be a i substantial safety hazard. Furthermore, the NSAL stated that to meet the requirements of l IEEE 279-1971, plants with an analog MSS must *ther remove the steam flow control signal 3 3  : selector switch and connect it directly to the steam flow transmitter that does not share the tap ,

with a level transmitter used for protection functions, or include a statement in the TS Bases section stating that the steam flow selector switch must normelly select the steam flow i.

transmitter that does not share a tap connection with a narrow range steam generator water level transmitter. The Channel IV steam flow instrument tap is not shared with any water level protection channel.

2.0 EVALUATION NSAL 96-004 is applicable to the three-loop Farley units. As an interim measure,' SNC chose to

use Channel IV steam flow for normal plant operation as recommended in NSAL 96-004.

However, under this administrative control, the Channel 111 steam flow will be used (selected) i when Channel IV is unavailable during quarterly functional testing, which will be less than i 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> every quarter. The Farley Technical Specifications allow 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for testing of reactor n trip system and engineered safety feature actuation system analog channels. Further, when the Channel lli steam flow is selected, the associated Charne lil steam generator low-low level bistables will be placed in test within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to initiate a partial trip signal.

On December 19,1996, the SNC's Plant Onsite Review Committee approved this administrative control procedure and an associated 10 CFR 50.59 Safety Evaluation to-

' implement this interim measure.- The 10 CFR 50.59 Safety Evaluation stated that the Farley plant-specific probabilistic risir assessment (PRA) indicated that the probability of a simultaneous failure of the common tap selected by the flow control signal selector switch for steam flow and narrow range steam generator _ water level and an additional level channel during quarterly functional test of Channel IV steam pressure'and steam flow instruments is on the order of 8.0E-09/ year. This initiating event frequency in the PRA was based on reliance on Channel lli to provide the protection function when it is selected for an assumed total time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year when Channel IV is unavailable. Each Farley unit has a total of three Channel IV steam pressure channels and three Channel IV steam flow channels each of which .

is silotted 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for a quarterly functional test, for a total of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per year, during which the channels are unavailable. Thus, with the above administrative control, Channel ill would normally be used for only 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per year, which is below the 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year assumed in the event frequency determination in the PRA. . The calculated event frequency of 8.05-09/ year a

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is well below the guidelines of ANSI /ANS-51,1 1983, " Nuclear Safety Criteria for the Design of Stationefy Pressurized Water Reactor Plants," which states in Clause 3.2 that events with a  ;

. best-estimate frequency of occurrence of less than 10E-6 per reactor year need not be considered for design.-

- SNC stated that in the unlikely event of a common tap failure when Channel 111 steam flow is selected, control room indications, control system response, and operating procedures will provide sufficient information and instructions to ensure that the affected unit is safely

operated. Furthermore, SNC intends to rely on the interim administrative measures until -

l the Unit 1 16th refueling outage (spring 2000) and Unit 214th refueling outage (spring 2001).

Full compliance with IEEE 279-1971 requirements will be provided for the SGWLC channels before startup from these refueling outages. SNC stated that the interim measure will provide r ~ the necessary additional time to study design altematives prior to finalizing the required

'  ; modification to meet IEEE 279-1971 requirements. In addition SNC has made the decision to ,

replace the steam generators in Units 1 and 2 during the spring of 2000 and 2001 refueling

-outages, respectively. As a result, completing modifications twice, once for the current steam generators and once for the replacement steam generators, would result in hardship or unusual difficulty withcut a compensating increase in the level of quality and safety.

3.0 CONCLUSION

The staff has reviewed SNC's evaluation and justification to use administrative controls to

. correct the SGWLC design deficiency by selection of Channel IV as an interim measure for the SGWLC reactor trip function during normal plant operation, and use of Channel lli only for brief periods when Channel IV is unavailable due to surveillance tests. - Based on this review, the staff concludes that the proposed administrative controls are in accordance with Westinghouse NSAL g6-004 and provide sufficient plant safety measures due, to the low probability of event

... frequency for a limited time prior to restorir g the SGWLC to full compliance with the single failure requirements of IEEE 279-1971 during the 16th refueling outage (spring 2000) for Unit 1 and 14th refueling outage (spring 2001) for Unit 2. The staff authorizes the interim administrative controls as an attemative to the requirements of IEEE 279-1971 pursuant to 10 CFR 50.55a(s)(3)(ii), in that compliance would result in hardship or u, usual difficulty without

- a compensating increase in the level of quality and safety. The proposed interim administrative controls are authorized until the steam generator replacement outages, which are scheduled for spring 2000 for Unit 1 and spring 2001 for Unit 2.

Principal Contributors: S. Mazumdar J. Zimmerman Date:- September.4, 1997 1

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