Information Notice 1979-06, 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:
On March 13, 1979, the Nuclear Regulatory Commission issued Orders
to licensees for five power reactors to shut down within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />
and to show cause why they should not remain shutdown pending
reanalysis of certain safety-related piping systems and pending
completion of any modifications indicated by the reanalysis. This
action was based on the discovery of a potentially unconservative
calculational technique within a computer program that has been
used by an architect-engineer for analysis of certain piping.
Description of Circumstances
During construction of Unit 1 at the Beaver Valley Power Station
and the four other reactor units involved in the above orders, a
version of a computer program, PIPESTRESS, was used by Stone &
Webster in performing "as built" stress analyses of piping requiring
a seismic analysis for which the architect-engineer was responsible.
Later, modification of the safety injection system at BVPS-1 to
improve net positive suction head and correction of an error in the
weight of some components in that system led to re-evaluation of the
stresses in this piping and associated pipe attachments. These
components whose weights had been incorrectly entered were six-inch
check valves manufactured by Velan Valve company. The error in weight
was caused by an incorrect weight shown on Velan drawings. As a
result of this reanalysis, the licensee reported the existence of
stress levels above those stated in the FSAR to the NRC on October 26,
1978. NRC review of the Licensee Event Report and inspection followup
identified the existence of significant discrepancies between stresses
calculated by the PIPESTRESS code and by NUPIPE, a code currently
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
IE Information Notice No. 79-06 March 23, 1979 (3) facility operation should not be suspended pending completion
of this work. Because of the safety significance of this problem, these Orders were made effective immediately to require that these
facilities be in the cold shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of
receipt of the Order and remain in that condition until further
Orders are issued.
This Information Notice provides details of a significant
occurrence that is still under review by the NRC staff. After
completion of the staff review, this Information Notice will be
followed with specific actions to be taken by licensees.
No written reponse is required. If you desire additional
regarding this matter, please contact the Director of the
NRC Regional Office.
information
appropriate
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
Notice No.
79-01
79-02
- Subject
Bergen-Paterson Hydraulic
Shock and Sway Arrestor
Attempted Extortion -
Low Enriched Uranium
Date
Issued
2/2/79
2/2/79 Issued To
All power reactor
facilities with an
All Fuel Facilities
79-03
79-04
Limitorque valve Geared
Limit Switch Lubricant
Degradation of
Engineered
Safety Features
Use of Improper Materials
In Safety-Related Components
Stress Analysis of Safety-
Related Piping
2/9/79
All power reactor
facilities with an
2/16/79
All power reactor
facilities with an
79-05
3/21/79
All power reactor
facilities with an
79-06
3/23/79
All power reactor
facilities with an
Enclosure
Page I of I