ML20043B032

Provides Details of Resolution Util Will Pursue Re Reactor Bldg Flooding Detailed in Encl LER 90-005.Mod Will Be Installed to Add Alarm in Main Control Room to Indicate When Flood Level Reaches Point & Operator Action Begins

ML20043B032
Person / Time
Site:	Crystal River
Issue date:	05/17/1990
From:	Beard P FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9005240103
Download: ML20043B032 (3)

Text

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Power C O R P O R AT 10 N '

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.May 17,.1990; l 3F0590-06 U.S.-Nuclear Regul'atoryfCommission [

Attn: Document Control Deskt Washington, D.C. 20555~

Subject:

Crystal River. Unit 3 >

Docket No. 50-302 '

Operating License No. DPR-72-'

' Reactor Building Flooding i

Reference:

Licensee Event' Report No.-90-005, dated April.29, 1990 '

Dear Sir:

l On March 29, 1990, Florida Power Corporation-(FPC)' determined that '

Crystal River Unit 3-(CR-3) was operating outside the plant design.

basis duet to procedures which were based upon a non-conservative '

Reactor Building (RB) maximum accident flood level r calculation.

L This problem is described in the reference LER (copy enclosed) .

The purpose of this letter'is .to describe 'the resolution..which. FPC

(

will pursue. ,

The Borated Water Storage Tank (BWST) is the primary source of ECCS water to mitigate design basis accidents. During the initial phase of LOCA mitigation, the BWST provides the source of water for the' ,

High Pressure Inj ection (UPI) , , Low : Pressure - Injection (LPI) 'and Building Spray (BS) pumps to inject water.into.the reactor vessel' and spray into the:RB atmosphere.- .Thel water inventory from the e reactor coolant system, the BWST, and the other tanks (Core. Flood Tank, Makeup Tank, and Sodium Hydroxide' Tank) . fall to the RB floor, eventually. reaching the RB sump' ' located in the- 95 ft floor elevation (plant' datum). Flooding of.this elevation will occur.

-In 'accordance with the current emergency -operating procedures, when  ;

the BWST. level reaches 2.5 ft, the operator manually; transfers ~the.  ;

LPI and BS1 pump suctions from the BWST to the RB sump, initiating the recirculation phase. -This action would result in a~RB flood l 90052401g[hyh02 ADO

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May 17, 1990 3F0590-06

~Page 2 level of epproximately 102 ft elevation with maximum Technical' specification volumes in the various tanks at the beginning of the )

accident. A number of instruments necessary to mitigate the design ,

basis accidents are installed on the walls below the 102 - f t l elevation, but above the 99.85 ft elevation (the maximum flood level described . in FSAR Section 6.2.2.1) . Thus, a design basis -

accident could cause these instruments to' be submerged. It is,_

therefore, necessary to initiate operator _ action to limit the flood ,

level to the FSAR value of 99.85.ft elevation. l l

To limit the flood level, there are three considerations which must be addressed. These are:

a. Water volumes and' boron concentrations for core' cooling '

and shutdown requirements must be satisfied, l

b. Minimum- required NPSH of the LPI and BS pumps must be satisfied before switchover to sump, and j Materials inside the RB must be' compatible with the pH c.

of the RB spray or sump fluid. l The licensing bases LOCALmodels consider the first item and-limit I the minimum volume of water and boron concentration in the1BWST. l The required levels-for the'LP1 pumps-and the'BS' pumps NPSH are 95.6 ft and 97.0 ft, respectively. Maintenance of the expected!pH  ;

in the range of 7.2 to 11.0 by establishing-a proper. balance of '

sodium hydroxide and borated water will assure . material ,

compatibility. A limited duration excursion-above a pH of.11.0 is verified to be acceptable in our EQ' program documentation.

The approach which best satisfies these considerations and resolves the RB flooding issue is to limit.the' volume of water contributed by the BWST. This.will'be accomplished by a procedural change in operator action. The operator will ensure ~the~RB flood-level will not exceed the 99.85 ft elevation. Following a LOCA,-the operator I will begin manual transfer of.the ECCS pump suctions from the BWST to the RB sump when.he receives an alarm-indicating-that the RB flood level has reached approximately 97.6 ft elevation (this level includes an' allowance for instrumentation error). The corresponding actual level will satisfy the ECCS pump NPSH, core' cooling, shutdown, and pH requirements. The'RB flood level will i' be less than 99.85 ft~with the switchover from the BWST to the RB sump completed in 10 minutes even under worst case large break LOCA flow rates.- FPC has confirmed - ~ that this switchover can be 1 accomplished in this amount of time.

Prior to CR-3 restart from this refueling outage, FPC will perform the following actions:

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May.17, 1990 3F0590-06 l Page 3-

1. A modification will be installed to add an alarm'in' thel l Main Control Room.to indicate when the RB flood level has reached the point that . operator action to initiate switchover from the BWST:to the sump is to begin.

2. Operational procedures will- be revised and -issued.-

Operator training on the use of the RB flood level as:the I new switchover' parameter will~be completed.

If you have- any questions, please contact Mr. Rolf L C. WideJl, . .,

Director, Nuclear Site' Support at (904)S63-4329. '

j sincerely,.

-tAAk P. M. Beard, Jr Senior.Vice' President i Nuclear Operations-1 PMB/JWT.

Enclosure xc: kegional Administrator, Region II-Senior Resident Inspector l

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] eso e, n 1 l1 1l5 9l0 Ant. ACv <t ,e e . - A . ,.. .p e. e . ,s A On March 29, 1990 at approximately 1400, Florida Power (FPC) determined that Crystal River Unit 3 (CR-3) was operating outside the plant design basis due to a non-conservative Reactor Building -(RB) maximum accident flood level calculation. When re-calculated using conservative assumptionst the maximum RB flood level exceeds the level necessary to prevent submergence of safe shutdown instrumentation and equipment. At the time of this determination, CR-3 was in MODE 5 (C01.0 SHUTDOWN) preparing for a refueling outage. This non-conformance was caused by an oversight in 1972 by the design enginen? when a calculation of the Net Positive Suction Head for post-accident recirculation cooling was incorrectly used for the maximum flood level. Modifications and calculations performed after 1972 did not identify this problem because the design engineers assumed the original calculation was correct or were unaware of the original calculation. FPC is evaluating possible solutions to resolve this design basis istue prior to restart from the refueling outage and is continuing to evaluate corrective actions to prevent recurrence. .

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0gol5 0;0 0j2 op 0 j6 von ~ . ..e mu.,on EVENT DESCRIPTION On January 9,1990, during an engineering review of a calculation performed for NRC Bulletin 79-01B,= non-conservative assum identified in -the calculation of the' Reactor Building (RB) (NH)ptions . flood were level. This calculation review was performed as a part of- the engineering configuration management  ;

verification and upgrade. The assumptions included using nominal or minimum tank volumes instead of maximum tank- volumes permitted by -Technical Specifications and assuming water from inside,the primary shield wall does~not L

reach the RB sump (NH, ACC). Based on these concerns, the maximum RB flood level-was re-calculated. The final results'of the corrected calculations were received  :

! by Florida Power on March 28,1990. ' Evaluation of the results on March 29,1990,  ;

l at approximately 1400,- concluded tu maximum level - would exceed the' level '

necessary to prevent submergence of matial safe shutdown instrumentation and equipment. This condition is consiw.ed to be outside the plant design basis.

At the time of the verification that the plant was outside the design basis, March 29,1990, CR-3 was in MODE 5 (COLD SHUTDOWN) in preparation for a refueling outage. No immediate actions were necessary.- t At 1410 cn March 29, 1990, a four-hour verbal report of this non-conformance was provided to the NRC Operations Center per 10CFR50.72(b)(2)(i) requirements.

This written report is being provided per- the requirements- of 3

10CFR50.73(a)(2)(ii).  ;

CAUSE This non-conformance was caused by an oversight in 1972 by the architect design }

engineer performing the original RB flood level calculation. The original calculation used minimum or nominal Borated Water Storage-Tank (BWST) (BP,TK),

NaOH Tank (BE,TK), and Core Flood Tanks (CFT) (BP,TK) volumes and assumed that 5 water within the primary shield area and other areas does not reach the RB sump.

area. These assumptions are appropriate for calculating the minimum water level available for Decay Heat Removal and Building Spray systems pump (BP,P](BE,P) ->

Net Positive Section Head. The resultant minimum water level of 99.85 ft, plant ,

datum was incorrectly assumed to be the maximum level above which critical instruments and equipment must be located.

Since the original 1972 calculation, other calculations and modifications have occurred which could have identified this problem. In 1981, during evaluation <

of NRC Bulletin 79-01B, the maximum R8 flood levol was reviewed and recalculated.

These calculations were based on the faulty assumptions and incorrectly calculated a new flood level. In 1987, the primary shield wall was modified when drain holes were drilled in the wall to allow water to drain to the RB sump area.

The utility Design Engineer contacted the Nuclear Steam Supply System vendor, the Architect Engineering firm, and a sister utility to determine why drains

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were not already installed, but no reason could be provided. The utility design engineer did not determine any impact on the original flood calculation since the impact to the original, or later calculations and assumptions was not provided. In 1989, the flood level calculation was reviewed by a contract design-engineer to determine- the impact of the addition of equipment which would displace water and thus potentially raise the maximum water level.- This calculation did not re-verify the previous level; calculation, .butl simply evaluated the impact on the level. In conclusion, it appears these calculations and modifications did not identify the faulty assumptions because the design engineers assumed the original calculation was correct or were not provided the pertinent original design assumptions.

EVENT EVALUATION FSAR Section 6.2.2.1 states:

"In the event of a postulated LOCA [ Loss of Coolant Accident), water will be pumped into the reactor building via the Reactor Building Spray _ System and Decay Heat Removal System as described in Sections 6.2 and 9.4, respectively. The reactor building will fill' to an approximate elevation of 99.85 ft. prior to the initiation' of the recirculation mode of- the Emergency Core Cooling System (ECCS)."

This statement is incorrect. The essential safe shutdown instrumentation used during a LOCA is located approximately two to three inches above the 99.85 ft, elevation. The correct flood level- elevation should be 101.7 ft.(101'8"),

approximately 1.85 feet above the -incorrect flood level. As a result, the attached list of essential safe shutdown equipment may be' subjected -to an environment for which they are not qualified to perform their safety function.

The corrected flood level elevation of 101'8" is based on maximum BWST, NaOH, and CFT tank volumes and assumes the water level .in the primary shield wall:will equalize with the level in the RB sump area following a' LOCA in the. cold leg reactor coolant pump [AB.P] suction. Additionally, the entire contents of. the Reactor Coolant System (RCS) [AB), less the Reactor Vessel [AB,RPV) volume, are assumed to contribute to the final maximum water level. These new assumptions are conservative because few accidents result in totally draining tanks and major portions of the Reactor Coolant System.

Most of the instrumentation and equipment affected will perform their automatic safety function before the water level reaches- the equipment. Automatic ~re-initiation of High Pressure Injection (HPI) [BQ) may not occur. Additionally, automatic actuation of Low Pressure Injection (LPI) [BP) may not occur. However, operators would be capable of manually initiating these safety systems. If the q affected containment isolation valves [JM,ISV) have been opened, the associated g m .. . . . . . , . . . . . . . . . . . . . . - . . . . , . . , . . . .

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One impact of the new flood level is the information provided to the operator.  !

Because many_ RCS pressure instruments (AB, PIT)-are affected, the operator may not have reliable pressure and subcooling margin information.  ;

1 CORRECTIVE ACTIONS Florida Power is evaluating possible solutions to resolve this non-conformance prior to restart from the current refueling outage. These solutions _ include  ;

relocation of equipment necessary for accident mitigation, re evaluation of the '!

analytical methods. which have .-established operator procedural steps, and 1 reduction in tank- volumes.

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Florida Power is ovaluating - the modification and calculation controls for  !

improvement-to reduce the possibility of:the. types of errors which contributed I

.to this non-conformance, in future modifications and calculations.

PREVIOUS SIMILAR EVENTS i This is the first report related to design error in' the maximum RB flood. '

elevation calculation. Two prior events were' identified that also relate' to RB equipment inoperability due to submergence. These previous events were concerned with locating equipment below the established RB flood elevation; ,

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0l0 015 0' 0 16 run,, . .. - c w .,im SAFE SHUTDOWN EQUIPMENT AFFECTED TAG NO. EQUIPNENT DESCRIPTION INSTRUMENTATION AH 536-TE [NH,TI) Reactor Building temperature instrumentation, used for post accident monitoring.

RC-1-LT1 [AB,LT) Pressurizer level transmitters. This instrumentation is RC-l_-LT3 used by the o>erator and by.the Integrated Control System to control ma(eup flow and pressurizer level.

RC-3A-PT3 [AB"PT) Reactor Coolant System (RCS) pressure transmitters-used-by RC-3A PT4 o Reactor Protection System for high pressure, low RC 38-PT3 pressure and-variable low pressure Reactor trips, o Engineered Safeguards System =for High: Pressure-Inject-

' ion and Low Pressure Injection automatic actuation.-

o Automatic Closure Interlock for overpressure protection of the Decay Heat Removal System.

L o Pressurizer - spray, heaters and -Pilot Operated Relief Valve for RCS pressure control.-

L o Main Control Board- indication' for' RCS pressure and Subcooling Margin.

RC-132-PT [AB,PT)- 3 Low range RCS pressure transmitters used for pressure '

indication on the -Main Control Board and for engi_neered safeguards actuation.

RC-158-PT[AB,PT) Wide range RCS pressure' transmitters locat'ed on the Main RC-159-PT' Control Board and the Remote Shutdown Panel.

RC-163A-LTl[A8,PT) Reactor Coolant Inventory Tracking System'1evel and RC-163B-LT1 temperature transmitters used for post-accident indication.

RC-164A-LT1 RC-1648-LTl RC-163A-TEl[A8,TT]

RC-1638-TE1 RC-164A-TE1 RC-1649-TEl WD-303A-LT[NH,LT]

Reactor Building Flood level transmitters. ,

WD-303B-LT WO 304A-LT .

WO-304B-LT

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TAG Nq. EQUIPMENT DESCRIPTION SP-18 LT [JB,LT) Emergency Feedwater Initiation and Control level SP-19-LT transmitters used for initiation of Emergency Feedwater on SP-21-LT low Steam Generator level.

SP-22-LT SP-23-LT SP 24-LT SP-25-LT SP-27-LT SP-31-LT SP-32-LT VALVES CAV-1 [KN,V) RCS sampling valves and containment-isolation valves.

CAV-3 CAV-126 MUV-40 (CB,V) Letdown Cooler' isolation valves and containment isolation MUV-41 valves.

HUV-505 I

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