ML20151U419

Discusses Review of Two Addl Concerns Identified in 970407 Memo to Recipient from Special Insp Branch Re Addl Open Items Associated W/Dec 1996 Design Insp Rept 50-289/96-201

ML20151U419
Person / Time
Site:	Three Mile Island
Issue date:	05/23/1997
From:	Marsh L NRC (Affiliation Not Assigned)
To:	Milano P NRC (Affiliation Not Assigned)
Shared Package
ML20151U412	List: IA-98-334, First Final Response to FOIA Request for Documents.Records Subj to Request Identified in App a & Encl ML20151U410 ML20151U414 ML20151U419
References
FOIA-98-334 50-289-96-201, NUDOCS 9809100389
Download: ML20151U419 (4)

See also: IR 05000289/1996201

Text

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NUCLEAR REGULATORY COMMISSION

WASHINoTON, O C. 2 A6 Hoot

• • May 23, 1997

MEMORANDUM T0: Patrick M. Milano, Acting Director [hiff

Project Directorate 1-3

Division of Reactor Projects 1/11 .

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FROM: Ledyard B. Marsh, Chief /' //,r.'is/, , . - l

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Plant Systems Branch '

Division of Systams Safety and Analysi's l

SUBJECT: THREE MILE ISLAND, UNIT 1 - ADDITIONAL OPEN ITEMS I

ASSOCIATED WITH DECEMBER 1996 DESIGN INSPECTION ,

(SSFI INSPECTION REPORT 50-289/96-201) l

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The Plant Systems Branch (SPLB), in coordination with the Reactor Systems l

Eranch (SRXB), has reviewed the two " additional" concerns identified in the

April 7, 1997 memorandum to you from the Special Inspection Branch (PSIB). In

that memorandum, PSIB requested NRR to review 1..c two concerns and provide NRR

staff positions to be used as guidance during the followup of those concerns.

The two concerns were identified by one of the inspection team members and

,<ere delineated in the PSIB memorandum as follows:

1. The inspector's assessment of the system impact of a letdown line failure  !

identified a sequence of events which merits further assessment. The l

basic concern is that failure of this line coupled (a complete line

severance is not necessary) with the postulated failure of the single

check valve that interconnects this line with the gaseous space of the ,

makeup tank could lead to depressurization of the hydrogen gas from the  !

makeup tank into the El. 281' area; a condition that aside from concerns

on the local explosive effects could also result in insufficient NPSH for

the high pressure injection (HPI) pumps which take suction from the j

makeup tank prior to turnover to the borated water storage tank (BWST). j

2. The auxiliary steam system is also ir ated on the same elevation of the l

auxiliary building as the HPl. It i routed in an alleyway that leads to

the HPI cubicles. However, since its complete routing could not be

observed the licensee was requested to provide design information on the

full routing of this line within the building. The EQ impact of failure

of this system was discounted in the FSAR and in a licensee evaluation of

NRC I&E Information Notice 90-53, " Failure of Auxiliary Steam Piping"

since its operating conditions did not meet the HELB criteria of both

>200 psig and >200*F. (The licensee concluded that crack breaks in the

piping in the auxiliary building would not result in a significant change

in the environmental conditions. Also, the HELB study concluded that the

postulated HELBs are acceptable because one string of each engineered

safeguards system would remain functional.) A review of the design

indicated that it (auxiliary steam system piping) was nonseismic; J

typically safety analysis would include consequences of such a failure i

during a DBA since a safe shutdown earthquake is also postulated to g

occur. '

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Patrick H. Milano 2-

The first concern postulates the random failure of a check valve coincident

with a postulated pipe break. The staff considers this beyond the licensing

basis for THI-l and most other, if not all, operating plants. The failure of

a che d valve is consid ed a passive failure and would not be postulated

coincident with a pipe break unless the pipe break itself could cause such a

failure of the check valve due to jet impingement, pipe whip, or other induced

mechanical force resulting from the postulated break. Therefore, Concern

No. I should only be followed up to ensure that no postulated pipe breaks can

result in the direct malfunctioning of the check valve.

With respect to the second concern, the licensee appears to have done an

inadequate analysis of the failure of the auxiliary steam (AS) piping.

Although the AS piping may not be high energy as defined by the licensee's

criteria, it is nonseismic Category I. Unless this piping has been

seismically analyzed to remain intact following a design basis earthquake, the

complete failure of this line must be taken into consideration, not just a

moderate energy crack. Additionally, postulated worst-case single active

failures must be assumed coincident with the break. As stated in NRC

Information Notice 90-53, a seismic event could result in the failure of

..rseismic Category I auxiliary steam lines. Staff requirements to fully

break nonse'smic Category I moderate energy lines go back to generic letters

dated SeptemL=r 26, 1972, sent to each licensee, in those letters we

requested licensees to determine (and correct if necessary) whether the

failure of any nonseismic Category I equipment (particularly in the

circulating water system or fire protection system) could result in a

condition that might potentially adversely effect the performance of safety-

related equipment required for safe shutdown or to limit the consequences of

an accident.

However, we disagree with the last sentence ef.the Concern No. 2 where it is

stated that " typically safety analysis would include consequences of such a

failure during a DBA since a safe shutdown earthquake is also postulated to

occur." In general, the staff does not postulate a safe shutdown earthquake

(SSE) coincident with design basis loss-of-coolant accidents (LOCAs) or any

other DBA and it is not part of most plants' (including THI-1) design and

licensing basis. Since the reactor coolant system is designed to seismic

Category I requirements, it is not expected to fail as a result of an SSE.

However, all equipment necessary to mitigate a LOCA must be designed to

seismic Category I requirements in order to ensure it is designed to the

highest standards for quality and reliability.

For the seismic induced failure of the auxiliary steam lines, the licensee

should assume a loss of offsite power (which can be caused by the seismic

event) coupled with the worst-case single active failure of any seismic

Category I equipment. All nonseismic Category I equipment should be

considered unavailable for mitigation of this event. However, the

consequences of additional nonseismic Category I piping system failures do not

have to be considered concurrent with the consequences from the failed

auxiliary steam piping. The consequences (flooding, spray, temperature and

pressure effects) from failures of nonseismic Category I piping systems need

only be evaluated on an individual basis because the staff does not expect

each and every nonseismic pipe to catastrophically f ail during a seismic

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Patrick M. Milano -3- ,

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event. If all pipes catastrophically failed there would be no means of

transporting the fluid to the area of concern (compartment being analyzed for

pipe break) and the flooding event would be terminated after the piping within

the compartment emptied it's limited volume to the compartment's atmosphere.

If you have any questions regarding this memorandum please contact Bill LeFave

at 415-3285.

DISTRIBUTION (w/o encl.):

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DATE 5/ 1 3/97 5//Y/97 5/ J'l /97 5/ b9[

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event.

, If all pipes catastrophically failed there would be no means of

transporting the fluid to the area of concern (compartment being analyzed fo

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pipe break) and the flooding event would be terminated af ter the piping with

the compartment emptied it limited volume to the atmosphere.

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If 415-3285..

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DISTRIBUTION (w/o encl.):

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