ML20029B062

LER 91-001-00:on 910129,determined That Setpoints for Condensate Storage Tank Level Switches Not Set to Initiate Required Transfer When 10,000 Gallons Water Available.Caused by Inadequate Documentation.Setpoints raised.W/910226 Ltr

ML20029B062
Person / Time
Site:	Hatch
Issue date:	02/26/1991
From:	Hairston W, Tipps S GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-1497, LER-91-001-02, LER-91-1-2, NUDOCS 9103050317
Download: ML20029B062 (8)

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w. o. Hewston, m sew. n e m,2..e Noe,v onewo- HL-1497 001252 February 26, 1991 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 lLANT HATCH - UNIT 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 LICENSE EVENT REPORT INSTR'JMENT TRIP SETP0lNT3 DETERMfNED TO BE OUTSIDE TECHNICAL SPECIFICATIONS 1IMITS Gentlemen:

In accordance with the requirements of 10 CFR 50.73(a)(2)(1), Georgia Power t.ompany is submitting the enclosed License Event Report (LER) concers,ing incorrect setpoints for the Unit 2 Condensate Storage Tank level switches as a result of inadequate design documentation. This event occurred at Plant Hatch - Unit 2.

Sincerelv, hd W. G. Hairston, 111 JK0/cr

Enclosure:

LER 50-366/1991-001 c: (See next page.)

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U.S. Nuclear Regulatory Connission February 26, 1991 Page Two c: Georaia Power Company Mr. H. L. Sumner, Gent:ral Manager - Nuclear Plant Mr. J. D. Heidt, Manager Engineering and Licensing - Hatch NORMS U.S. Nuclear Rggylglory ConnisSign. Washtu9hnulLL.

Mr. K. Jabbour, Licensing Project Manager - Hatch U.S.. Nuclear Reculatory Commission. ReaJ.DD_11 Mr. S. D. Ebneter, Regional Administrator Mr. -L. D. Wert, Senior Resident inspector - Hatch l

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At that time, it was determined the setpoints for Unit 2 Condensate Storage Tank (CST, EIIS Code KA) level svitches 2E41-N002 and 2E41-N003 vere not in compliance with the requirements of Unit 2 Technical Specifications Table 3.3.3-2, item 3.c. Specifically, the switches, which cause liigh Pressure Coolant Injection (ilPCI. EIIS Code BJ) system suction source transfer from the CST to the Suppression Pool (EIIS Code lit) on lov CST vater level, vere not set to initiate the transfer when 10,000 uscable gallons of water vere available to the llPCI system as intended by the Unit 2 Technical Specifications. it the time of discovery of this event, the llPCI system was aligned to take suction from the Suppression Pool. It vas left in this alignment until the CST vater ic"cl svitch setpoints could be raised.

The cause of this event was less than adequate design documentation. Although the level svitch setpoints vere designed such that 10,000 gallons of vater remained in the CST at the time of the suction source transfer, the design documents did not require 10.000 gallons of vater to be available to the llPCI system.

Corrective actions include raising the CST level svitch setpoints, initiating revisions to appropriate portions of the Unit 2 Technical Specifications and Final Safety Analysis Report, and revieving additional setpoints.

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74ET PLJd[AND SYSTEM TDENTIFICATION General Electric - Boiling Vater Reactot Energy Industry Idtntification System codes are identified in the text as i (EIIS Code XX).

SUMMARY

OF EVENT 4 On 1/29/91 at approximate'ly 0915 CST, Unit 2 was in the Run mode at an

-approximate power level of 2436 CMVT (approximately 100% rated thermal pover).

At that time, it was determined the setpoints for Unit 2 Condensate Storage Tank (CST, EIIS Code KA) level switches 2E41-N002 and 2E41-N003 vere not in compliance with the requirements of Unit 2 Technical Speellications Table i 3.3.3-2, item 3.c. Specifically, the svitches, which cause High hessure Coolant Injection (flPCI, EIIS Code BJ) system suction source transfer from the CST to the Suppression Pool (EIIS Code BT) on lov CST vater level, vere not set to initiate the transfer when 10,000 uscable gallons of vater vere available to the HPCI system as intended by the Unit 2 Technical Specifications. At the time of discovery of this event, the llPCI system was aligned to take suction from the Suppression Pool. It was left in this alignment until the CST vnter level switch setpoints could be raised.

The cause of this event was less than adequate design documentation. The original design specification called for 10,000 gallons of water to be in the CST at the start of suction source transfer. Although the level switch setpoints were designed such that 10,000 gallons of water remained in the CST at i the time of the suction source transfer, the design documents did not require  !

10,000 gallons of water to be availabic to the HPCI system. Since the .

centerline of the HPCI system CST suction pipe is approximately 12 inches above  !

the bottom-of the CST, the setpoints were not adequate to ensure a successful transfer of the Unit 2 IIPCI aystem suction source to the suppression pool on lov CST vater level. A contributing factor to delaying identification of this issue i was Technical Specifications requirements based on a literal reading of the original design specification.

Corrective actions include raising the CST level svitch setpoints, initiating i . revisions to appropriate portions of the Unit 2 Technical Specifications and Final Safety Analysis Report, and reviewing additional setpoints.

. DESCRIPTION OF EVENT On 12/19/90,:the Plant' Review Board (PRB) van performing a routine review of a revision to procedure 575V-SUV-015-1S, "HPCI/RCIC Pump Suction Source Instrument Functional Test and Calibration." -In the course of revieving the revision, a-concern was raised by the PRB members that-the current-instrument setpoint (el 130 feet 11-1/2 inches MSL.) was inconsistent with other CST level setpoints and thus might not meet the intent of Unit 2 Technical Specifications Table 3.3.3-2, item 3.c. PRB Open Item 90-186-1 was issued to the plant's Nuclear Safety and Compliance (NSLC) Department to address this concern, i

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3 nn After reviev of this concern with Corporate Licensing and Architect /Engineet l (A/E) personnel, it was determined the trip setpoint of 130 feet 11 1/2 inches

, did not meet the intent of the reluirements of the Unit 2 Technical Specifications. Calculat. ions by A/E personnel indicated that CST vater level vould decrease approximately 13 1/2 inches following suction source transfet initiation as the llPCI system Suppression Pool suction valves opened and the CST

. suction valves closed. Since the centerline of the HPCI suction line from the CST is at elevation 131 feet, only 7 1/2 inches of the suction pipe vould be covered at the time the transfer initiated: thus, the CST suction pipe vould be substantially uncoveted and-the HPCI system would likely trip on lov suction pressure before the suction source transfer could be empleted. Therefore, the lov CST vater level trip setpoint of 130 feet 11 1/2 inches was not adequate to ensure a successful transfer of the Unit 2 HPCI system suction cource to the ,

Suppression Pool on lov CST vater level. Upon this dt. termination, a Deficiency Card was vritten to document this condition as required by plant procedures.

-It was also detetmined no probleas existed with the Unit 1 CST vater level  :

svitches'1041-N002 and 1E41-N003 because they are set to trip at 132 ieet 10 inches and 132 feet 8 inches, respectively. Additionally, the Unit 1 and Unit 2

- CST vater level svitches which in1tiate Reactor Core Isolation Cooling (RCIC,

• EIIS Code BN) suction source transfer vote also determined to be set correctly.

The llPCI system's suction source previously had been aligned to the Suppression

- Pool under Limiting Condition for Operation (LCO) 2-91-45. This was done on *

. 1/28/91 because Suppression Pool level instrument 2E41-H6620 had been temoved from service for calibre

• ion. CST level switches 2E41-N002 and 2E41-N003 vere declared inoperable and .dded to LCO 2-91-45 to ensure the HPCI system's suction source remained aligned to the Suppression Pool until the lov CST vater level tri; setpoint could be raised.

Design Change Request (DCR) 2H91-023 vas initiated to raise the lov CST vater level trip setpoint approximately 22 1/2 inches to 132 feet 10 inches. The DCR vas revieved and approved for implementation per plant administrative control procedures. Maintenance Vork Orders (MV0s) 2-91-419, 2-91-423, and 2-91-424 vere written to-perform the work required to implement the design change. ,

On 2/7/91, the work under the HV0s was completed. The CST level svitches vere

- then functionally' tested and calibrated per procedure 57SV-SUV-015-25, "IIPCI/RCIC Pump Suction Source Instrument Functional Test and Calibration." CST vater level svitches 2E41-N002 and 2E41-N003 vere declared operable and LCO 2-91-45 was closed on 2/8/91 at approximately 0400 CST. The HPCI system was aligned to take suction from the CST vhich is the system's normal lineup.

CAUSE OF THE EVENT The cause of this event was less than adequate design documentation. The original-setpoint specification, a design document, called for 10,000 gallons of water to be in the CST at the start of suction source transfer rather than

' 10,000 gallons of uscable vater. Because the centerline of the HPCI system CST suction pipe is approximately 12 inches above the bottom of the CST, the setpoints vete not adequate to ensure 10,000 useable gallons of vater vere y y,f te y- gw '

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Tt1T available to the flPCI system at the time of transfet initiation. A conttibuting factor to delaying identification of this issue was a Technical Specification requirement which was based on a litetal teading of the original design specification, i.e., 10,000 gallons of water in the CST instead of the intended 10,000 gallons of vater available to HPCI. Also several sections of the Unit 2 FSAR, including Table 7.3-1 and Sections 9.2.6.2 and 9.2.6.3, support this literal reading and, therefore, may have contributed to this event.

In addition, duting the investigation of this event, it was discoveted the CST vater level svitch setpoints vere determined in 1985 to be incottect. Proposed new setpoints vere transmitted to the site in April, 1985 as part of DCR 84-138.

At that time, the existing setpoints for these instruments vere identified as

" lover than the PSL (process safety limit)." This DCR vas initiated as patt of a program to establish an insttument setpoint index, establish consistency with Regulatory Guide 1.105 recommendations, and to implement setpoint changes calculated by the plant's A/Es. It is not clear why the setpoints fot these instruments were not raised in 1985. A review of the design modification transmittal packages associated with this prngram identified a potential foi a misunderstanding with regard to the necessity for implementing the new setpoints. For example, a previous setpoint index transmittal, dated 9/11/84, noted that the proposed setpoint changes do not constitute previous design deficiencies, but are upgrades of the design philosophy to present day standards. The proposed changes identified in the April, 1985 transmittal vere listed as more conservative than existing plant setpoints. Additionally, a subsequent transmittal stated that the proposed setpoint changes constitute an improved design. It appears that individuals involved in this development, reviev and implementation of the setpoint changes contained in DCR 84-138 may have failed to attach the approptiate level of significance to this concern since the setpoint vas in compliance with the literal vording of the existing Technical Specifications. Investigation of this aspect of the event vill continue.

REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This report is required by 10 CFR 50.73(a)(2)(1) because a condition existed which was prohibited by the plant's Technical Specifications. Specifically, it was determined the llPCI system's suction source vould not automatically ti nsfer from the CST to the Suppression Pool while 10,000 gallons of CSr water vote still available to the HPCI system. *!his is contrary to the tequitements of Unit 2 Technical Specifications Table 3.3.3-2, item 3.c.

The HPCI system is provided to assute the reactor is adequately cooled to limit fuel-clad temperatute in the event of a small break in the nuclear boiler system causing a loss of coolant which does not result in tapid depressurization of the reactor vessel. The HPCI system is provided with two suction sources: the CST, its normal source, and the Suppression Pool, its alternate source. On lov CST vater level or high Suppression Pool vater level, the liPCI system's suction source is designed to transfer automatically from the CST to the Supptession Pool.

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In this event, it was determined the trip setpoints for the CST vater level 4

svitches did not meet the intent of the Technical Specifications trip setpoint requirements. However, the plant's A/E has performed an analysis to show that

in all transients and accidents involving HPCI operation its suction source vill l transfer to the Suppression Pool on high Suppression Pool vater level before lov i CST vater level is reached. The analysis assumed the Suppression Pool is at the  ;

minimum Technical Specification alloved level of 12 feet 2 inches at the time an accident requiring the operation of HPCI occurs. Considering the Suppression Pool volume between the lov and high level setpoints, the water that could be held up in the dryvell and not reach the Suppression Pool, and the Reactor Vessel volume between the HPCI initiation setpoint (Level 2) and the !!PCI trip -

setpoint on high vater level (hevel 8), the maximum amount of makeup vater i

transferred by HPCI from the CST to the Reactor Vessel before suction source i transfer to'the Suppression Pool vas calculated to be less'than 62,000 gallons. i stand pipes installed in the CST for all non-essential equipment suction assure that-there is at least 100,000 gallons of CST _ vater available for the HPCI-system. The analysis assumed that the CST contained only the 100,000 gallons ,

assured by-the stand pipes; therefore, 38,000 gallons vould remain in the tank '

after the suction source transfer is accomplished. Calculations shoved that approxim=tely 10 1/2 inches of vater in the CST is equivalent to a volume of 10,000 gallons. Thus, 38,000 gallons of water would reach a height of over 39 inches in the CST. Since the bottom of the CST is at elevation 130 feet 0 inches, 30,000 gallons of water is at 133 feet 3 inches, vell above the required lov CST vater level trip setpoint of 132 feet 10 inches. Consequently, the analysis concluded the suction source transfer vill be caused by a high level in the Suppression Pool and not by a lov level in the CST. Under no identified transients does the CST vater level switch serve the primary function of facilitating the transfer of the !!PCI system's suction source from the CST to the Supprossion Pool.

Based on the above analysis, it is concluded that this event had no adverse

• Impact on. nuclear safety. The analysis is applicable to all power levels.- i CORRECTIVE ACTIONS Upon determination the trip setpoints for CST vater level switches 2E41-N002 and 2E41-N003 vere not in compliance with Technical Specifications requirements,-

they were declared inoperable and added to existing LCO 2-91-45. This ensured

.the llPCI system would remain aligned to take suction from the Suppression Pool until_the setpoints could be raised.

DCR 2H91-023 vas implemented which raised the water level svitch setpoints approximately 22 1/2 inches. -The new trip setpoint of 132 feet 10 inches ensures the CST vater level falls no lover than approximately 131 feet 8 1/2 inches before the suction source transfer is complete. A/E calculations shov the HPCI system' vill have adequate flow and suction pressure throughout the l entire valve opening and closing sequences at this trip setpoint.

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! Level svitches 2041-N002 and 2E41-N003 vete functionally tested and calibrated per procedure $7SV-SUV-015 25 iollowing implementation of DCR 21191-923 ano declared operable on 2/8/91 at approximately 0400 CST. The llPCI system was '

aligned to take suction from the CST vhich is its normal suction source.

i A request to tevise the Unit 1 and Unit 2 Technical Specifications vill be initiated. The requestod revision vill clearly state the lov CST vater level

trip setpoint is that equivalent to 10,000 gallons of wate, evallable to the IIPCI system. Additionally, applicable seetions of the Unit 1 and Unit 2 FSARs .

! vill be revised to indicate the trip setpoint is equivalent to 10.000 gallona of vater available to the itPCI syttem.

Additionally, all the setpoint changes covered under DCR 84-138 vill be revieved l by August 15, 1991 to ensure they have been implemented or the untr.odified l setpoints do not present any safety concerns. A supplement to this LPR vill be

issued by 9/1/911 it vill include the results of the above mentioned

' investigation and any additional cortective actions as vartanted.

c ADDITIONAL INFORMATION

1. Other Systems Affected:

No systems other than the Unit 2 IIPCI system vere affected by this event.

2. Failed Components Identiiicattons-No failed components caused or resulted from this event.

3. Previous Similat Events:

No previous similar events in the last tva years in which an inadequate design resulted in setpoints not in compliance with the plant's Technical Specifications vere noted.-

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