Information Notice 1979-06, Stress Analysis of Safety-Related Piping

Stress Analysis of Safety-Related Piping

ML031180121
Person / Time
Site:	Indian Point, 05000000
Issue date:	03/23/1979
From:	NRC/IE
To:
References
IN-79-006, NUDOCS 7904270155
Download: ML031180121 (4)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555 March 23, 1979 IE Information Notice No. 79-06 STRESS ANALYSIS OF SAFETY-RELATED PIPING

Summary:

Commission issued Orders

On March 13, 1979, the Nuclear Regulatoryshut down within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

to

to licensees for five power reactors remain shutdown pending

and to show cause why they should not

piping systems and pending

reanalysis of certain safety-related by the reanalysis. This

completion of any modifications indicated

a potentially unconservative

action was based on the discovery of program that has been

calculational technique within a computer of certain piping.

used by an architect-engineer for analysis

Description of Circumstances

Beaver Valley Power Station

During construction of Unit 1 at the in the above orders, a

and the four other reactor units involved was used by Stone &

version of a computer program, PIPESTRESS, analyses of piping requiring

Webster in performing "as built" stress was responsible.

a seismic analysis for which the architect-engineer

system at BVPS-1 to

Later, modification of the safety injection

correction of an error in the

improve net positive suction head and led to re-evaluation of the

weight of some components in that systempipe attachments. These

stresses in this piping and associated entered were six-inch

components whose weights had been incorrectly

company. The error in weight

check valves manufactured by Velan Valve on Velan drawings. As a

was caused by an incorrect weight shown reported the existence of

result of this reanalysis, the licensee FSAR to the NRC on October 26, stress levels above those stated in the Report and inspection followup

1978. NRC review of the Licensee Event discrepancies between stresses

identified the existence of significantby NUPIPE, a code currently

calculated by the PIPESTRESS code and

used by Stone & Webster.

790427 0I5 5

IE Information Notice No. 79-06 March 23, 1979 In some instances, NUPIPE yielded stress results which were

significantly higher than those obtained with PIPESTRESS. The

differences in the results are attributable to the force

summation method utilized by SHOCK 2, a subroutine of the

PIPESTRESS computer code. Based on an examination of a sample

run, the NRC staff learned on March 8, 1979 that this computer

program subroutine algebraically summed horizontal response

components from the seismic input components. It also

algebraically summed vertical response components from the

seismic input components. Such response loads should not be

algebraically added (with predicted loads in the negative direction

offsetting predicted loads in the positive direction) unless far

more complex time-history analyses are performed. Rather, to

properly account for the effects of earthquakes, as required by

General Design Criterion 2 for systems important to safety, such

response loads should be combined absolutely or, as is the case

in the newer codes, using techniques such as the square root of

the sum of the squares (SRSS). This conforms to current industry

practice and Reg Guide 1.92.

Stone & Webster also used the SHOCK 2 subroutine of PIPESTRESS in

the analysis of piping in safety systems for FitzPatrick, Maine

Yankee, and Surry 1 and 2. NRC review on March 10-13, 1979, of

preliminary results from the reanalysis of portions of the Beaver

Valley piping at Stone & Webster's offices in Boston, Massachusetts, indicated several instances of pipe stress beyond allowable limits.

In the face of this deficiency information, the NRC concluded that

until full reanalysis of all potentially affected piping systems

important to safety has been completed with a piping analysis

computer code which does not contain the algebraic summation method, it would be prudent to assume that the potential for reducing

intended design margins at each of the facilities in question exists

in the event of an earthquake and could be sufficiently widespread

such that the basic defense-in-depth provided by redundant safety

systems may be compromised.

On March 13, 1979, NRC issued to licensees for these facilities, Orders to Show Cause why: (1) potentially affected safety system

piping should not be reanalyzed using an appropriate computer program;

(2) modifications indicated by reanalysis should not be done; and

March 23, 1979 IE Information Notice No. 79-06 suspended pending completion

(3) facility operation should not be significance of this problem, of this work. Because of the safety to require that these

these Orders were made effective immediately

condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of

facilities be in the cold shutdown that condition until further

receipt of the Order and remain in

Orders are issued.

of a significant

This Information Notice provides detailsby the NRC staff. After

occurrence that is still under review Information Notice will be

completion of the staff review, this

be taken by licensees.

followed with specific actions to

you desire additional information

No written reponse is required. If the Director of the appropriate

regarding this matter, please contact

NRC Regional Office.

Enclosure:

List of IE Information

Notices Issued in 1979

IE Information Notice No. 79-06 March 23, 1979 LISTING OF IE INFORMATION NOTICES

ISSUED IN 1979 Information *Subject

Notice No. Date Issued To

Issued

79-01 Bergen-Paterson Hydraulic

Shock and Sway Arrestor 2/2/79 All power reactor

facilities with an

OL or a CP

79-02 Attempted Extortion - 2/2/79 All Fuel Facilities

Low Enriched Uranium

79-03 Limitorque valve Geared

2/9/79 All power reactor

Limit Switch Lubricant

facilities with an

OL or a CP

79-04 Degradation of

Engineered 2/16/79 All power reactor

Safety Features facilities with an

OL or a CP

79-05 Use of Improper Materials

In Safety-Related Components 3/21/79 All power reactor

facilities with an

OL or a CP

79-06 Stress Analysis of Safety- 3/23/79 All power reactor

Related Piping

facilities with an

OL or a CP

Enclosure

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