05000245/LER-1997-016-01, :on 970401,discovered That Vent/Purge Operation Using Sgts,Opening & Closing Sequence of Atmospheric Control Created Temporary Flow Path Between Reactor Bldg & Primary Containment Prior to Isolation of AC Sys

:on 970401,discovered That Vent/Purge Operation Using Sgts,Opening & Closing Sequence of Atmospheric Control Created Temporary Flow Path Between Reactor Bldg & Primary Containment Prior to Isolation of AC Sys

ML20138F983
Person / Time
Site:	Millstone
Issue date:	05/01/1997
From:	Robert Walpole NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20138F979	List: 05000245/LER-1997-016, Forwards LER 97-016-00,which Documents an Event That Occurred at Mnps,Unit 1 on 970401.Commitment Made within Ltr,Listed 05000245/LER-1997-016-01, :on 970401,discovered That Vent/Purge Operation Using Sgts,Opening & Closing Sequence of Atmospheric Control Created Temporary Flow Path Between Reactor Bldg & Primary Containment Prior to Isolation of AC Sys
References
LER-97-016-01, LER-97-16-1, NUDOCS 9705060117
Download: ML20138F983 (3)

LER-1997-016, on 970401,discovered That Vent/Purge Operation Using Sgts,Opening & Closing Sequence of Atmospheric Control Created Temporary Flow Path Between Reactor Bldg & Primary Containment Prior to Isolation of AC Sys

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(t)(2)(i) 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2) 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
2451997016R01 - NRC Website

text

l e

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OM8 NO. 3160-0104 j

EXPIRES 04/30/98 j

(4-95) fNFO TiO COLLECT ON R UFST SO RS E O TED L t'e"'?o^"40%?."P"'!amo 'M!a".=2*o ^LM LICENSEE EVENT REPORT (LER) l'It^"u'e 'MS"^,'o*"tr&a'88"faa^o"f'n'd4^ot" E7c sinMla!3e?1'A%4708"15 0. swr"oN.Mo's?"

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

F#.CluTV NAME (1)

DOCKET NUMBER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 1 05000245 1 of 3

• lTLE (4)

Loss of Secondary Containment during Purging / Venting Operations 1

EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACluTY NAME DOCKET NUMBER I

NUMBER

'^ * **"'

04 01 97 97 016 00 05 01 97 l

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)(2)(v) 50.73(t)(2)(i) 50.73(a)(2)(viii)

LEVEL (10) 000 20.2203(a)(3)(i)

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

1 POWER 20.2203(a)(1) 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 Abstr6 ' halow 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER qinclude Area Codel Robert W. Walpole, MP1 Nuclear Licensing Manager (860)440-2191 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 NPRDS i

l SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR SUBMISSION

[

YES NO (If yes, complete EXPECTED SUBMISSION DATE).

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

)

On April 1,1997, at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />, with the plant in COLD SHUTDOWN condition, it was discovered that during vant/ purge operations using the Standby Gas Treatment System (SGTS), the opening and closing sequence of the Atmospheric Control (AC) and the SGTS inlet valves (1-SG-1 A/B) creates a temporary flow path between the Reactor Building and primary containment prior to isolation of the AC system. The flow escapirm into the Reactor Building through this path during a Loss of Coolant Accident (LOCA) could be sufficient to lift t.a blow-out panels in the Reactor Building, especially if purging / venting was conducted with the 18" AC valves. Since the differential pressure (DP) in the Reactor Building would be at the 6 in Wg relief pressure of the panels,10CFR100 limits would be exceeded before the SGTS could restore the.25 in Wg DP and begin filtering the release. The potential safety consequences of this condition is a loss of safety function since without secondary containment, SGTS would be unable to mitigate the consequences of a LOCA. It should be noted that, Technical Specification 3.7.B.3.6 requires primary containment to be purged through SGTS whenever primary containment is required.

The potential to pressurize the Reactor Building during vent / purge evolutions results from the fact that the design of the SGTS system did not consider a LOCA coincident with these operations. This event is considered outside the plant's original design basis.

9705060117 970501 PDR ADOCK 05000245 S

PDR

=

JU.S. NUCLEAR REGULATORY COMMISSION (4-95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3) l YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 2 of 3 97 016 00 TEXT (11 more space is required use additional copies of NRC form 366A) (17) i 1.

Description of Event

i On April 1,1997, at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />, with the plant in COLD SHUTDOWN condition, it was discovered that the opening and closing sequence of the AC valves and the SGTS inlet valves (1-SG-1 A/B) during a LOCA coincident j

with venting and purging primary containment creates a temporary flow path between the Reactor Building and j

primary containment. The flow escaping into the Reactor Building through this temporary path could be j

sufficient to unseat the Reactor Building blow-out panels, especially if purging / venting was conducted with the 18" AC valves. Technical Specification 3.7.B.3.6 requires primary containment to be purged through SGTS whenever primary containment is required 4

An analysis was performed to evaluate the extent of building pressurization due to the above scenario. This analysis concluded that the pressure in the Reactor Building could exceed the 6 in Wg relief pressure of the blow out panels. The Reactor Building would no longer be sub-atmospheric, and after reseating of the panels, would have an internal positive differential pressure equal to the lift pressure of the blow-out panels. Review of radiological analyses concludes that 10CFR100 limits would be exceeded before SGTS would recover the.25 in Wg differential pressure in secondary containment and begin filtering the release.

i 11.

Cause <>f Event 1

The caase of this condition is that the SBGT system was not designed to withstand the effects of a LOCA, I

]

because this accident scenario was outside of the plant's original design basis, j

l I

)

i til. Analysis of Event Several changes to the Technical Specifications are currently under development for containment vent / purge i

operations. These changes are being developed in order to comply with Three Mile Island (TMI) Action item II.E.4.2. The changes will also propose restoring the Reactor Building Ventilation System (RBVS) as the normal i

flow path for containment venting / purging operations. Investigation into the use of the RBVS for containment

]

purging revealed that there is a potential to pressurize secondary containment should a LOCA occur while in this alignment. Subsequently, an evaluation was performed to determine if a similar concern existed while purging j

primary containment via SGTS. This *nitial analysis concluded that sequencing of the AC and the SGTS inlet

{

valves (1-SG-1 A/B) creates a temporary flow path between the Reactor Building and primary containment prior to isolation of the AC system. The flow escaping into the Reactor Building through this path during a LOCA would i

be sufficient to lift the blow-out panels, especially if purging / venting was conducted with the 18" AC valves.

l Technical Specification 3.7.B.3.6 requires primary containment to be purged through SGTS whenever primary containment is required.

After the blow-out panels reseat, the Reactor Building would be at an internal positive differential pressure equal to the lift pressure of the panets.

Review of radiological analyses conclude that 10CFR100 timits would be exceeded before SGTS would recover the.25 in Wg differential pressure in secondary containment and begin i

j filtering the release.

4 The initial analysis for this accident scenario used very conservative assumptions and did not credit several aspects of the system which would reduce the overall pressurization of secondary containment. No credit was taken for throttling of system flow during valve opening and closing and the analysis did not account for the,

.,~ -..- -. -

- = -.

,#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 3 of 3 97 016 00 TEXT Uf more space is required, use additional copies of NRC Form 366Al (17) steam exhausted from the Reactor Building during isolation of the ventilation system. The cause of this event is that the original design of the system did not consider a LOCA coincident with venting or purging.

IV. Corrective Action

i The control switches for AC-10 have been tagged to prohibit vent / purge operations through the SGTS.

l e

l Additional analyses and/or design modifications will be performed to resolve this issue prior to start-up.

e 3

t l

i 1

V.

Additional Information

i One of the provisions required an analysis to demonstrate the acceptability of plant capabilities to protect j

structures and safety related equipment (ductwork, fans, filters) located beyond the " purge system" isolation i

valves from the steam / air environment that wo 'd be experience 1 from an accident during vent / purge operations.

l The related licensee position was that the intemh/ of the operatag SGTS train, and later the idle train, could not l

be assured. The NRC required structural modifications to upgradt SGTS or the incorporation of appropriate plant l

Technical Specifications. Although these items specified the actions necessary to protect SGTS, the impact on i

secondary containment pressures was never identified or evaluated.

It should also be noted that the original plant design used the RBVS and not SGTS as the normal flow path for containment venting and purging. The impact of a LOCA on RBVS and on secondary containment pressures associated with this alignment was also not evaluated, thus providing further evidence that this event was not considered in the plant's original design basis.

e I

j

Similar Events

l LER 95-10 identified a similar condition related to valve sequencing which causes a loss of SGTS function: On j

June 21,1995 while processing Proposed Technical Specification Change Request 1-94-11, Rev 1 it was idsntified that the integrity of SGTS could not be assured if a LOCA occurred while venting containment. The integrity problem results from the fact the isolation dampers to SGTS open before the containment isolation

)

valves close thus making both trains susceptible to the pressure wave from the LOCA inside containment. The expected pressure exceeds the design pressure of the SGTS ductwork.

i Manufacturer Data N/A a!