05000245/LER-1996-062, :on 961203,failure Vulnerability of SBGTS Was Identified.Performed Engineering Calculations to Quantify Consequences of Failure of 1-AC-10 During Fuel Handling Accident Coincident W/Venting of Primary Containment

:on 961203,failure Vulnerability of SBGTS Was Identified.Performed Engineering Calculations to Quantify Consequences of Failure of 1-AC-10 During Fuel Handling Accident Coincident W/Venting of Primary Containment

ML20133B960
Person / Time
Site:	Millstone
Issue date:	01/02/1997
From:	Robert Walpole NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20133B956	List: 05000245/LER-1996-062, :on 961203,failure Vulnerability of SBGTS Was Identified.Performed Engineering Calculations to Quantify Consequences of Failure of 1-AC-10 During Fuel Handling Accident Coincident W/Venting of Primary Containment
References
LER-96-062, LER-96-62, NUDOCS 9701060308
Download: ML20133B960 (3)

LER-1996-062, on 961203,failure Vulnerability of SBGTS Was Identified.Performed Engineering Calculations to Quantify Consequences of Failure of 1-AC-10 During Fuel Handling Accident Coincident W/Venting of Primary Containment

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(ii)
2451996062R00 - NRC Website

text

f NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMS NO. 3160-0104 (4-95)

EXPIRES 04/30/98 NoWATiO OL" LEE [lON R7Q S

E TED L if.^a"'?o'LMft!"^!!%3% 'c"!s'~"Ti"%7f,a'a ^=!

LICENSEE EVENT REPORT (LER) 15;ri. u'? 'Wal?a",^=,^at a'gga?J,aa^may,,gte,c"l,c

^"

!?"A%^~"fdM"T'l#128t!. "'AW'8"ot"dse?"' "

(See reverse for required number of digits / characters for each block)

)

FACILITY NAME (1)

DOCKET NUMBER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 1 05000245 1 of 3 TITLE I4)

Standby Gas Treatment System Single Failure Vulnerability EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACiUTY NAME DOCKET NUMBER NUMBER 12 03 96 96 062 00 01 02 97 OPERATING THis REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE m N

20.2201(b) 20.2203(a)(2Hv) 50.73(an2Hi) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3Hi)

~'

50.73(a)(2Hii)

Bo.73(aH2Hx)

X LEVEL (10) 000 20.2203(an2)(il 20.2203(an3Hii) 50.73(a)(2)(iii) 73.71 20.2203(aH2)(ii) 20.2203(aH4) 50.73(a)(2Hiv)

OTHER 20.2203(aH2Hisi) 50.36(cH1) 50.73(a)(2)(v) specif y ln Abstract below

~

20.2203(aH2Hiv) 50.36(cH2) 50.73(aH2)(vii)

LICENSEE CONTACT FOH THIS LER (12)

NAME TELEPHONE NUMBER tinclude A,ea Codel Robert W. Walpole, MP1 Nuclear Licensing Manager (860)440-2191 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT f13)

CAUSE

SYSTEM COMPONENT M ANUF AC T URER REPORTABLE

CAUSE

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

EXPECTED MONTH DAY YEAR SUBMISSION

'X' YEs NO 05 01 97 (If yes, complete EXPCCTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces, t.e., approximately 15 single-spaced typewritten lines) (16)

On December 3,1996, with the plant in a COLD SHUTDOWN, a single failure vulnerability of the Standby Gas Treatment System (SBGT) was identified. During drywell vent / purge operations, a failure of 1-AC-10 to close on an SBGT initiation would not guarantee SBGT could achieve the required 0.25 inches Wg negative differential pressure in th3 reactor building, if the SBGT initiation resulted from a signal from the radiation monitors. The reduction in reactor building negative pressure under the above scenario results from the fact that on an SBGT initiation from the radiation monitors, an isolation signal is not generated to close the atmospheric control system valves.1-AC-10 is the only valve to close in this condition and its isolation is caused by an electrical interlock between 1-AC-10 and 1-SG-1 A and/or 18.

Therefore, if 1-AC-10 fails to close due to a malfunction, the flow path to SBGT is aligned to both primary and stcondary containments. This alignment potentially reduces the SBGT flow from the reactor building resulting in a reduced building negative pressure. This postulated condition was reported pursuant to 10 CFR 50.72(b)(1)(ii) as a condition outside the design basis of the plant. There were no actual safety consequences from this event. The corrective actions include completion of engineering calculations to quantify the consequences of a failure of 1-AC-10 during a fuel handling accident coincident with venting of primary containment. This coupled with a root cause evaluation that will be performed will determine the required long term corrective actions. As an interim corrective action, a Red Tag has been hung on 1-AC-10 to prohibit valve opening when secondary containment is required.

Sacondary containment is required prior to movement of fuel.

9701060308 970102 PDR ADOCK 05000245 S

PDR

,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 1 05000245 NUMBER NUMBER 2 of 3 96 062 00 TEXT (11more space is required, use additional copies of NRC Form 366A) (17) 1.

Descriotion of Event On December 3,1996, with the plant in a COLD SHUTDOWN, a single failure vulnerability of the SBGT was identified. During drywell/ torus vent / purge operations when primary containment is required, in accordance with SPROC 95-1-12, Rev 1, an automatic initiation of SBGT due to high radiation, such as from a fuel handling accident, would isolate the drywell/ torus with the closure of 1-AC-10 via its interlock with 1-SG-1 A and/or 1-SG-1B It is postulated that under these conditions, a single failure of 1-AC-10 to close on an SBGT initiation would allow the drywell/ torus to vent uncontrollably through the SBGT system. Analysis of the SBGT system in conjunction with drywell/ torus venting hes not been performed, therefore, it can not be assured that the SBGT system could achieve the required 0.25 inches Wg negative differential pressure in the reactor building. This condition was reported pursuant to 10 CFR 50.72(b)(1)(ii) as a condition outside the design

]

basis of the plant.

11.

Cause of Event

The cause of the event is being investigated through a root cause analysis and will be provided in a supplement to this LER.

Ill. Analvsis of Event Plant procedures that control the vent / purge operation of the drywell/ torus uses a dedicated operator to perform the following actions: (1) Place idle SBGT train in lockout; (2) Monitor plant parameters for evidence of LOCA: (3) Isolate the SBGT train damaged due to LOCA; (4) Verify 1-AC-10 is closed; and (5) Remove the idle SBGT train from lockout. The use of the dedicated operator prevents having to declare the standby SBGT train inoperable while it is in the " LOCKOUT" condition. When performing vent / purge operation, LCO 3.7.b.3 or LCO 3.7.B.5 are entered as appropriate. The procedure does not address actions for the operator to take on receipt of a *High Radiation" alarm condition.

It is concluded that this condition is potentially reportable under the requirements of 10 CFR 50.73(a)(2)(ii) as a condition that is outside the design basis of the plant. A failure of 1-AC-10 to close during a SBGT initiation

{

would potentially compromise the ability of SBGT to maintain the 0.25 inches Wg negative differential pressure in the reactor building during accidents which would cause SBGT to initiate from the radiation monitors (e.g. fuel handling).

There were no actual safety consequences from this event. No formal quantitative analysis now exists to determine the potential safety impact. This analysis will be completed via an engineering calculation to determine if the SBGT system is capable of performing its design function.

IV. Corrective Action

Engineering calculations will be performed to quantify the consequences of a f ailure of 1-AC-10 during a fuel handling accident coincident with venting of primary containment. This will be completed prior to movement of fuel. Currently, Millstone Unit No.1 is defueled.

This event is being investigated via a root cause evaluation. The results of the investigation and the engineering calculation will be provided in a supplement to this report prior to movement of fuel. The supplemental report will provide a cause and description of any required long term corrective actions.

_ ~.

\\

.U.S. NUCLEAR REGULATORY COMMISSION (4-95) 1 LICENSEE EVENT REPORT (LER) l TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUM8ER (2)

LER NUMBER (6)

PAGE (3) i YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 3 of 3 96 062 00 TEXT (11 more space is required, use additional copies of NRC form 366A) (17)

The immediate corrective actions include precluding a fuel handling accident during primary containment venting by prohibiting fuel movement during vent / purge operations until the above actions are completed.

Also, a Red Tag has been hung on 1-AC-10 to prohibit valve opening when secondary containment is l

required. Secondary containment is required prior to movement of fuel.

V.

Additional Information

l

Similar Events

i None Manuf acturer Data l

Not Applicable i

)

r 4

t 1-