ML20043D234

Informs Commission of Newly Emerging Technical Concerns & Proposed Approaches for Resolution

ML20043D234
Person / Time
Issue date:	06/04/1990
From:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
TASK-PII, TASK-SE SECY-90-199, NUDOCS 9006070285
Download: ML20043D234 (4)

Text

a

........................o RELEASED TO THE PDR

-l

?p""*'%'1I (o/mleo GV I

3'

......../........ine......:.

i i

date-s,**... #

EOLICY ISSUE (Information)

Ju'ne 4, 1990 SECY-90-199 F

For:

The Commissioners

.From:

James M. Taylor Executive Director for Operations Subjeg:

QUARTERLY REPORT ON EMERGING TECHNICAL CONCERNS

Purpose:

To inform the Comission of newly emerging technical concerns and proposed approaches for their resolution.

Background:

In'the November 13, 1989, staff requirements memorandum (SRM 891025),

- the Comission asked the staff to keep it abreast of new issues on a quarterly basis, and to propose approaches for resolving these issues. The Comission also asked the staff to work closely with industry for' timely resolution of each technical issue.

Sumary:

This quarterly report identifies four issues which the NRR staff views as emerging technical concerns. The paper also charac-terizes current staff efforts to resolve these issues.

Discussion:

The staff has identified the following technical concerns which it feels should be brought to the attention of the Comission:

(1) Availability of Equipment While a Plant is in a Shutdown Condition The recent (April 1990) loss of vital ac power incident at Vogtle has caused the staff to consider whether the Technical Specifications'(or other regulatory require-ments) provide appropriate assurance that a plant can respond to an unexpected event in a safe and-controlled manner while in a shutdown. condition.

NOTE:

TO BE'MADE PUBLICLY AVAILABLE IN 10 WORKING DAYS FROM Tile

Contact:

DATE OF TIIIS PAPER Charles E. Rossi, NRR

.49-21163-f0\\o hQ QN Y f(f, Y

.l.

.7 ' >

The Commissioners 2

- When 'a' plant is at power, safety systems provided to mitigate or prevent design-basis events as well as-other postulated risk significant events (e.g.,-

anticipated transients without scram) are required to be operable. When a plant is. shut down, events that could be anticipated are not well defined.-. Also, both safety-related'and non-safety-related equipment may be out of' service for longer periods of time than r

allowed with a plant at power to facilitate performing a wide variety of maintenance and testing activities simultaneously.

Equipment availability requirements, and the need for related emergency operating procedures and-training, will

'be evaluated. This effort will_ entail examining relevant-plants-and applying.other. risk and events at' nuclear power, Generic regulatory action will be deterministic factors.

taken, as appropriate.

(2) Systematic Evaluation Program Topic Resolution In conjunction with staff efforts related-to license renewal, the status of Systematic Evaluation Program (SEP) topics is being reviewed. LThe stafficoncluded in-1984 that, in general, older plant designs compared-well with current licensing criteria, and identified-

.27. topics applicable:to operating reactors licensed before-1976. The 27 SEP-topics were recently reviewed-to determine whether or not the actions recommended M

from the'SEP have been-or are being addressed within the framework of ongoing NRC programs. The staff has m

concluded that all but four topics are being addressed

,4 by ongoing NRC programs.

In accordance with the i

May 25, 1990, staff requirements memorandum-(SRM 900516),

c4 the staff will keep the Commission informed of the

,j resolution of the SEP' issues, I

The four SEP topics that remain are:

E SEP TOPlc 1.6 - TURBINE MISSILES This topic focuses'on turbine dir integrity and

%* w overspeed protection, including stop, intercept and control valve reliability.

1 g4 f b y tl i

I 1

tx

}

I

c

~^

1 o

is Ik

- e The Commissioners?

3

.1 w1 SEP TOPlc 2.3 --CONTAINMENT DESIGN AND. INSPECTION F

This topic addresses'the adequacy of inspection-j P

programs for prestressed concrete containments.

The programs should include lift-off testing and.

q acceptance criteria, testing of.prestressingLtendons o

and surveillance for possible' deterioration of pre-stressed containments. The likelihood 'of delamination occurring ~in the shell walls or dome will also be assessed.

y SEP TOPlc 4.1 - REACTOR PROTECTION SYSTEM AND ENGINEERED 1

SAFETY l FEATURE SYSTEMS-ISOLATION'

.g This topic addresses electrical isolation of control-L and protection signals. The safety issue of concern.

is not the quality of isolation, but rather the existence of isolation devices to preclude the propagation of non-safety-related system faults to m

saf ety-related. systems.

SEP TOPIC 6.1 - PIPE BREAK EFFECTS ON SYSTEMS AND 2

COMPONENTS This topic reviews the effects of postulated pipe A,

breakston the integrity and function of systems-and components relied upon for safe. reactor shutdown and for mitigating the consequences of a postulated U

4

g.,i, pipe break.

(3) Reactor Vessel Head Cracking During-a recent refueling ) outage at-Quad-Cities 2, flaws were detected in the. reactor' vessel head.

Extensive' surface, volumetric and metallographic T'. _

examination indicates = that -there are' two types of:

s cracks in the Quad Cities 2 reactor vessel head.

Fifteen cracks (longest is.approximately 20 inches) d n

were subsurface and 34 cracks (longest is 30 inches) sL, were surface connected. The subsurface cracks were formed during original fabrication of the cladding.

1 f

The surface-connected cracks were caused.by stress o

corrosion cracking of the cladding during service.

't j -

The cracks were found in a cladding section that had been repaired-during fabrication of the reactor vessel head.. The deepest penetration of these types i

e 1

r e

s

$q s

h(w.

.y

,, +

V The Commissioners 4

i Y

of cracks was estimated to be 0;225 inch beyond the cladding-base metal interface. The fracture mechanics analysis-of these cracks indicates the reactor vessel head is acceptable for at-least one additional-fuel-cycle with the cracks present. NRR issued NRC Infor-mation Notice 90-29, " Cracking of Cladding and Its o

Heat-Affected Zone in'the Base Metal of A Reactor Vessel' Head," on this matter and the General' Electric Company issued a service information letter (SIL).

I I

f The staff is continuing to follow the licensee's-investigations to learn if generic and safety implications are involved.

7 (4) Cracks-in the Steam Generator Shell Girth Weld.

Cracks or linear indications have been~ detected on the inner circumference.of the upper shell-to-transition'

. v cone girth weld in 22 steam generators (6 plants) in i

the United States.

In addition, linear indications-i have been found at one foreign plant. The degree of cracking ranges from severe in the case off Indian Point l

-Unit'2 to dispersed at Zion Unit 1.'

In domestic plants, j

flaws have been-observed only in Westinghouse Model 44 and'Hodel 51 vertical' recirculating U-tube steam-m 1

generators, that are of the feedwater' ring design;

-l

~

Operating-experience through 1989 is summarized in NRC Information Notice 90-04, " Cracking.of the Upper' i

Shell-To-Transition Cone Girth Welds-in Steam Generators."

j

.I k'

-Westinghouse, the affected licensees and the NRC staff-G are still evaluating the available information to-..

establish the root cause of-the cracking problem and 4

determine the need for. generic regulatory action.'

j t

Our next update on emerging technical concerns will'be sent i

to the Commission in August 1990.

q

/

DISTRIBUTION:

J.es M. T or Commissioners xecutive Director OGC for Operations l

OIG j

LSS j

GPA 4

REGIONAL OFTICES

(

EDO i

ACRS.

6 ACNW ASLBP

'ASLAP SECY.'

W

-