05000423/LER-1997-015, :on 970204,review of Design Calculations Identified Min Water Level in Containment Sump at Time of RSS Pumps Below Vortex Suppression Grating During Loca. Caused by Inadequate Review/Process Control.Program Revised

:on 970204,review of Design Calculations Identified Min Water Level in Containment Sump at Time of RSS Pumps Below Vortex Suppression Grating During Loca. Caused by Inadequate Review/Process Control.Program Revised

ML20136B176
Person / Time
Site:	Millstone
Issue date:	03/05/1997
From:	Peschel J NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20136B116	List: 05000423/LER-1997-015, :on 970204,review of Design Calculations Identified Min Water Level in Containment Sump at Time of RSS Pumps Below Vortex Suppression Grating During Loca. Caused by Inadequate Review/Process Control.Program Revised
References
LER-97-015, LER-97-15, NUDOCS 9703100203
Download: ML20136B176 (5)

LER-1997-015, on 970204,review of Design Calculations Identified Min Water Level in Containment Sump at Time of RSS Pumps Below Vortex Suppression Grating During Loca. Caused by Inadequate Review/Process Control.Program Revised

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
4231997015R00 - NRC Website

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMS NO. 3150-0104 (4 95)

EXPIRES 04/30/98 N[OU/TSOI"COLTE7T M7o" ST RTE"O "L 5 S

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0 LICENSEE EVENT REPORT (LER) i$Wea 'e 'a'ag",^/oTa?,,"'88%$,sto~eo'n',%'="e o

sinh ^aosiE!5'l~^a2&"T'"'fsW2T"#d'!!a'* -

(See reverse for required number of digits /charactersfor each block) i FACluTY NAME (1)

DOCKET NUMBER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 3 05000423 1 of 5 TITLE (4)

Potential Vortexing of Recirculation Spray System Pumps EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (B)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACluTY NAME DOCKET NUMBER FACIUU NAME CKENMBER 02 04 97 97 015 00 03 05 97 OPERATINO 5

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR i: (Check one or more) (11)

MODE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii)

POWER 000 20.2203(a)(1) 20.2203(a)(3)(i)

X so 73(a)(2)(ii) 50.73(a>(2)(x)

LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v)

Specify in Abstract below

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20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER linclude Area Code)

J.M. Peschel, MP3 Nuclear Licensing Manager (860)437-5840 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE To NPRDs To NPROS SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR YES NO SUBMISSION (if yes, complete EXPECTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces,i.e., approximately 15 single spaced typewritten knes) (16) l On February 4,1997, with the unit in Mode 5, a review of design calculations identified that the calculated minimum water level in tha containment sump at the time of Recirculation Spray System (RSS) pumps start (approximately eleven minutes after a Containment Depressurization Actuation signal) would be below the level of the containment sump vortex suppression gratings during a large break Loss of Coolant Accident (LOCA). Vortex formation and the resulting air entrainment, could result in cavitation of the RSS pumps, such that they might be unable to perform their intended safety function (s) in an accident. Following evaluation, this condition was immediately reported on February 7,1997, pursuant to 10CFR50.72(b)(1)(ii)(B), and is being reported pursuant j

to 10CFR50.73(a)(2)(ii)(B) as a condition outside the design basis of the plant.

j ThTre were no adverse safety consequences from this condition, in that the unit f as not experienced a LOCA and therefore opiration of the RSS has not been necessary. However,the potentialinoperabilityof the RSS pumpsis significant because it rzpresents a condition outside the design basis of the plant. The RSS is credited in the safety analyses to cool the containment by spraying containment sump water into the containment atmosphere post-accident, provide iodine removal, and provides long-term cora cooling post-accident. Formation of vortexes due to the water level being below the level of the vortex suppression gratings, rzsulting in cavitation, could create a common cause failure of the four RSS pumps, resulting in potential loss of safety function.

Corrective actions needed to ensure operability of the RSS and design basis compliance will be implemented prior to entry into mode 4 from the current outage.

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9703100203 970305 PDR ADOCK 05000423 I

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r NRC FORM 365A U.S. NUCLEAR REGULATORY Commission (4 95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVIStoN Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 5 97 015 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (11) 1.

Description of Event

On February 4,1997, with the unit in Mode 5, a review of design calculations identified that the calculated minimum water level in the containment sump at the time of Recirculation Spray System (RSS) pumps start (approximately eleven minutes after a Containment Depressurization Actuation signal) would be below the level of the containment sump vortex suppression gratings during a large break Loss of Coolant Accident (LOCA). The RSS containment sump vortex suppression gratings are located at an elevation of minus 24.5 feet A calculation to determine head loss across the RSS containment sump screens due to insulation debris blockage during a LOCA indicated that the effective sump water Isvel for this scenario would be minus 25.2 feet. Because this effective sump levelis below the level of the vortex suppression gratings, they would not be able to prevent vortex formation. Vortex formation and the resulting air entrainment, could result in cavitation of the RSS pumps, such that they might be unable to perform their intended safety functiM(s) following a Design Basis Accident (DBA). Following evaluation, this condition was immediately reported on February 7,1997, pursuant to 10CFR50.72(b)(1)(ii)(B), and is being reported pursuant to 10CFR50.73(a)(2)(ii)(B) as a condition outside the design basis of the plant.

II,

Cause of Event

The cause of this event was inadequate review and process control within the Architect / Engineer and utility engineering organizations during construction.

]

1 In 1982 the minimum containment sump water level during a DBA was determined by calculation. Based on this value the elevation of the vortex suppression grating was established. In 1985, a calculation was performed to reanalyze the RSS pumps Net Positive Suction Head (NPSH) assuming conservative containment sump screen blockage parameters. This new calculation resulted in the projected sump water level to be below the elevation of the vortex suppression gratings. The impact of the reduced effective sump water level, relative to vortex suppression, was not recognized at the time and, therefore, was not considered.

til. Analysis of Event j

There were no adverse safety consequences from this condition, in that the unit has not experienced a LOCA and therefore operation of the RSS has not been necessary.

However,the potentiatinoperabilityof the RSS pumps is significant because it represents a condition outside the design l

basis of the plant. The RSS is credited in the safety analyses to cool the containment structure by spraying l

containment sump water into the containment atmosphere post-accident and is also credited for iodine removal. In the s:cond mode of operation two RSS pumps remain in the spray mode and the other two pumps are realigned to supply water to the Charging and Safety injection pumps for Emergency Core Cooling System cooling. The design of the l

containment sump meets the guidance of Regulatory Guide 1.82, " Sumps for Emergency Core Cooling and Containment Spray Systems," as clarified in Final Safety Analysis Report Table 1.8-1. Formation of 9ortexes due to the water level being below the level of the vortex suppression gratings, resulting in cavitation, could create a common cause failure of the four RSS pumps, resulting in potential loss of safety functions.

However, a loss of suction performance test by the manufacturer provides some assurance that pump damage to

[

internal components would not occur due to cavitation by vortexing. The "A" RSS Pump was tested in December 1980.

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NOC FORM 366A U.s. NUCLEAR REGULATORY Commission (4 95)

UCENSEE EVENT REPORT (LER)-

TEXT CONTINUATION

' FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL Revision Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 3 of 5 97 015 00 TEXT (If more space is required, use additionalcopies of NRC Form 366Al (17) by the manufacturer under severe cavitation conditions for approximately 5 minutes. As stated in the performance test trip report, "it is concluded that the loss of suction test more than adequately demonstrated satisfactory pump i

p:rformance at severe loss of suction conditions."

IV. ' Corrective Action As described in LER 97-010-00,"the calculation program is being reviewed and revised as part of the ongoing 50.54(f) effort. In addition, in order to address configuration management of the calculation program, formal work practices have been established for processing nuclear engineering calculations."

Thn following corrective action will be taken:

1.

Calculations will be performed to determine if containment sump water level requirements can be met.

Modifications will made if' necessary following the reanalysis. These actions will be completed prior to entry into Mode 4.

i V.

Additional Information

None

Similar Events

LER 96-007-02

" Containment Recirculation Spray. Quench Spray. and Safety Iniection Systems Outside Desian Basis Due to Desian Errors."

On April 3,1996, at 13:55, with the plant in Mode 5 at 0-percent power, it was determined that the plant had operated in a condition that was outside the design basis due to a deficiency in specific design conditions for a system needed to remove residual heat and mitigate the consequences of an accident. It was determined that the Containment Recirculation System (RSS) spray piping and supports were not adequately designed for thermal loads resulting from accident temperatures.

l Accident temperatures could result in stresses above the design allowable stresses for plant " Faulted" conditions.

It was initially determined that the higher RSS temperatures could result from a postulated loss cf i

Service Water (SWP) to one or more RSS heat exchangers. It was subsequently determined that: (a) l unacceptable stresses in RSS piping, Quench Spray System (OSS) piping, and portions of piping which comprise the Safety injection (SI) flowpath, and the associated supports for those systems, could also result from the design basis accident temperatures inside containment, and (b) the original design basis analyses for the RSS and OSS systems utilized support anchor movements which were nonconservative.

1 MeU.s. NUCLEAR REGULATORY Commission (4 95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVislON Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 4 of 5 97 015 00 TEXT (11more space is required, use additional copies of NRC Form 366A) (17)

At the time of discovery the plant was completing a shutdown for unrelated reasons. Plant systems responded normally to the shutdown. No Engineered Safety Features Actuation's were required or were initiated as part of the shutdown.

As conservative actions, design reviews of the RSS, QSS, and other systams are being performed, design improvements will be made, and the systems will be restored to appropriate design basis requirements prior to declaring the systems operable for other than modes 5 and 6. As action to prevent recurrence, those systems which would be exposed to a Post-LOCA or Post HELB environment and which are required to mitigate the consequences of a design basis accident are being reviewed in order to determine whether or not they are susceptible to the same types of design deficiencies.

LER 97-010-00 " Electrical Calculation Discrepancies in Minimum Voltaae Analysis for Class 1E Electrical Systems."

On January 13,1997, with the plant in Mode 5, a review of electrical calculations associated with Class 1E 480V and 120V systems identified discrepancies between related electrical calculations used to demonstrate design basis compliance. On January 29,1997, these concerns were sufficiently substantiated to question the validity of the Degraded Grid Voltage calculations. A prompt report was made, pursuant to 10CFR50.72(b)(2)(iii) as an event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to shutdown the reactor and maintain it in a safe shutdown condition.

The degraded grid voltage (DGV) relays are required to ensure that safety related equipment and devices either have adequate voltage to perform their safety functions or are not damaged due to a degraded voltage condition. These analyticallimit worst-case minimum values were not utilized as the source voltage for separate voltage drop calculations performed for the 480V and 120V bus loads. If the 4160V bus voltage were to be at its analytical limit worst-case minimum value then inadequate voltage at individual devices supplied by the distribution system could result. Therefore, these systems may not be able to perform their design safety functions under degraded grid voltage conditions.

The cause was determined to be a lack of configuration management for the comprehensive calculation program which is required to establish and maintain the design basis of the unit. The calculation program is being reviewed and revised as part of the ongoing 50.54(f) effort. In addition, in order to address configuration management of the calculation program, formal work practices have been established for processing nuclear engineering calculations.

l LER 97-011-00 " Hydrogen Recombiner Heaters Potentially Outside of Desian Basis Under Dearaded Voltaae Conditions."

l During a review of design attributes and supporting design specifications and calculations for Technical SpeciFcations, discrepancies were found between the minimum required voltages for i

safety related heaters and their capability to perform their design functions based on reduced voltage at the safety related buses. Reviews performed of heater applications found that the heaters were

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.r lU.s. NUCLEAR REGULATORY Commission (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISloN Millstone Nuclear Power Station Unit 3 05000423 NUM8ER NUMBER S of 5 97 015 00 TEXT Uf more space is required, use additionalcopies of NRC form 366A) (17) able to perform their design function with the exception of the Hydrogen Recombiner heaters. On January 29,1997, with the unit in Mode 5, a prompt event report was submitted regarding the reduction in Hydrogen Recombiner heating capability under degraded voltage conditions pursuant to 10CFR50.72(b)(1)(ii)(B) as a condition outside the design basis of the plant.

There have been no safety consequences as the result of a degraded grid voltage supply to the l

Hydrogen Recombiner heaters. A degraded voltage supply to the Hydrogen Recombiner heaters is safety significant in that it could result in reliance on operation of the Containment Hydrogen Purge system to remove hydrogen from the containment following a Loss of Coolant Accident.

The cause has been determined to be a lack of configuration management for the comprehensive calculation program which is required to establish and maintain the design basis of the unit. As j

reported on LER 97-010-00, "The calculation program is being reviewed and revised as part of the ongoing 50.54(f) effort. In addition, in order to address configuration management of the calculation program, formal work practices have been established for processing nuclear engineering i

calculations." Additional corrective action includes evaluation of Hydrogen Recombiner heater performance under degraded voltage conditions, and, if required, implementation of design changes to restore the Hydrogen Recombiners to their design basis requirements.

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Manufacturer Data Ells System Code Containment Spray System.

..BE l

Ells Component Function Identifier Pump.

..P l

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NaC FORM 366A (4 95)