Key Engineering Materials Vols 297-300 (2005) pp 1659-1665 Online: 2005-11-15

© (2005) Trans Tech Publications, Switzerland

doi:10.4028/www.scientific.net/KEM.297-300.1659

Alternative Fatigue Evaluation of Nuclear Piping

Designed by ANSI B31.1 Code

Y.J. Parka, Y.S. Changb, J.B. Choic and Y.J. Kimd

School of Mechanical Engineering, Sungkyunkwan University, Suwon, Kyonggi-do, Korea
a
pyj0909@skku.edu, byschang7@skku.edu, cboong33@skku.edu, dyjkim50@skku.edu

Keywords: Nuclear Piping, Implicit Fatigue Design, Explicit Fatigue Analysis, On-line Monitoring

Abstract. Class 1 piping components of a certain old vintage nuclear power plant were designed by
ANSI B31.1 code without a detailed fatigue evaluation such as the one required by recent ASME
Section III code. These components may undergo fatigue damage when considering the continued
operation beyond the design life whilst the inherent fatigue resistances of those may satisfy the
corresponding implicit limits. In this paper, the alternative fatigue evaluation has been carried out
explicitly for Class 1 piping of old nuclear power plant. At first, four representative nuclear piping
systems were selected to check the operational adequacy. After characterization of conservative
loading conditions based on design features, a series of finite element analyses have been performed
and the cumulative usage factors were calculated to guarantee if the components at each system
sustain adequate fatigue resistance. Finally, comparisons were drawn between the implicit fatigue
design specifications and alternative explicit fatigue analysis results. Even though there were some
exceptions, it was demonstrated that most components satisfied the current explicit fatigue criterion.

Introduction
The integrity of major components in nuclear power plant should be maintained during operation. As
one of important integrity issue, fatigue damage due to repetitive or fluctuating stress has been
recognized which specific effects are cracks in the material. After undergone sufficient cyclic loading,
microstructural damage can be accumulated and subsequent mechanical or thermal transients may
lead to a growth of the initiated crack. Thus, the possibility of fatigue damage of major components
should be investigated for safe operation.
The Class 1 piping of old nuclear power plant was normally designed in accordance with the
requirements of ANSI B31.1 which is an implicit fatigue design code. The code is based on maximum
stress theory and, thereby, a piping system designed by it has a thicker wall due to the lower allowable
stresses in general. Also, a stress range reduction factor is incorporated into the code for
determination of equivalent full temperature cycles. It can play a role to reduce the stress amplitude
considering thermal bending and prevent fatigue failure effectively. However, even though the
integrity of Class 1 piping designed by ANSI B31.1 code was guaranteed as a current licensing basis
(CLB), it may loose during continued operation periods beyond the design life.
Therefore, in this paper, four representative nuclear piping systems designed by ANSI B31.1 are
selected to investigate the operational adequacy. The general design information and implicit fatigue
design features of Class 1 piping are reviewed, and then, the alternative explicit fatigue analyses
according to ASME Section III as well as implicit fatigue evaluation have been carried out after under
continued operation periods beyond the design life. Finally, comparisons are drawn between the
implicit fatigue design specifications and alternative explicit fatigue analysis results.

Evaluation Scope and Design Features

Boundary Systems and Components. The Class 1 piping, basically, comprise of reactor coolant
(RC) loop pipe, branch connection pipes and associated nozzles. These components are generally
required to perform the safety functions of maintaining the RC pressure boundary and preventing the

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1660 Advances in Fracture and Strength

release of fission products. Only those parts that are passive in nature and directly support the
accomplishment of a safety function are selected for fatigue evaluation. Fig. 1 shows typical iso-
drawings for Class 1 piping and associated components of the old nuclear power plant which identify
representative locations addressed in this paper.

(a) SIS Piping Segment 3 (b) RHRS Piping Segment 2

Fig. 1 Typical iso-drawings for Class 1 piping and associated components

The scope of evaluation is limited to the pipes and nozzles up to the Class 1 boundary for reactor
coolant system (RCS), safety injection system (SIS), residual heat removal system (RHRS) and
chemical and volume control system (CVCS). It does not include components that are beyond the
class boundary (e.g., Class 2 portion of a connected system) and the major equipments such as
pressurizer, reactor vessel and steam generator etc.
Design Information. The pipes and nozzles were grouped for evaluation considering the material,
geometry, loading condition and identified in Table 1. For example, the main components of RC loop
piping such as the hot leg pipe connecting the reactor pressure vessel outlet and the steam generator
inlet, the crossover leg pipe connecting the steam generator outlet to RC pump suction, and the cold
leg pipe connecting the RC pump outlet and the reactor pressure vessel inlet were identified as RCS
Piping Segments 1 through Segments 3.
The pipes and nozzles of the old plant were made of several stainless steels, e.g., the RC loop main
pipes were centrifugally cast and the fittings were statically cast A351 CF8M except for splitter elbow
that consists of a two-piece construction. The key design information for Class 1 piping including the
material types and re-arranged transient types are also represented in Table 1. The details of the
re-arranged transient types will be discussed again in later section.

Table 1 Design information for Class 1 piping of old nuclear power plant
Identification Material Transient Type
RCS Piping Segments 1, 2, 3 A351 CF8M A (7 transients), B (6 transients), C (13 transients)
RCS Piping Segments 4, 5, 6
C (13 transients), D (16 transients), G (8 transients),
SIS Piping Segments 1, 2, 3 A376 TP316
H (7 transients), I (8 transients), J (8 transients)
RHRS Piping Segments 1, 2, 3
RCS Piping Segments 7, 8, 9
B (6 transients), E (12 transients), F (8 transients),
SIS Piping Segment 4 A376 TP304
H (7 transients), K (11 transients)
CVCS Piping Segments 1, 2
RCS Nozzles 1, 2, 3 A182 F316 C (13 transients), G (8 transients), K (11 transients)

Implicit Fatigue Design. The details of CLB for Class 1 piping of old nuclear power plant are
somewhat different with the present general design requirements (e.g., ASME Section III) since the
components were designed to the requirements of ANSI B31.1 code [1].
 Key Engineering Materials Vols. 297-300 1661

Based on the code, implicit fatigue design is possible by adopting the following definition of an
allowable stress range, SA, with adequate margin to accommodate fatigue thermal bending loads
without adverse consequences to safety [2].

SA = f (1.25Sc + 0.25Sh ) (1)

where, f is stress range reduction factor which is a function of equivalent full temperature cycles, N. In
case of N equals 7000 and less, the value of f is 1.0. On the other hand, N equals 1,000,000 and over,
the value of f is 0.5; Sc is allowable stress intensity at room temperature; Sh is allowable stress
intensity at the operating temperature. The equivalent full temperature cycle is computed using the
following equation provided in section 102.3.2 of B31.1 code.

N = N E + r15 N1 + r25 N 2 + r35 N 3 + LL + rn5 N n (2)

where, NE is the number of cycles at full temperature change, TE, for which the thermal stress has been
calculated; N1, N2, N3,……, and Nn are the number of cycles at lesser temperature change ∆T1, ∆T2,
∆T3,……, and ∆Tn; r1, r2, r3, ……, and rn are the temperature change ratios for other cyclic thermal
load sets; r1= ∆T1/∆TE, r2= ∆T2/∆TE etc. According to industry experience, it was demonstrated that N
is normally much less than 7000 [3].
In the previous research [4], two piping systems of old nuclear power plant with significant
geometric discontinuities and severe thermal transients were evaluated according to ASME Section
III. The conclusion was that most locations in the piping systems designed to the requirements of
ANSI B31.1 code were satisfied unity criterion of cumulative usage factor (CUF). It was reported that ,
also, the re-calculated thermal expansion stress of RC main loop pipes did not exceed the allowable
thermal expansion stress range reflecting design changes due to steam generator replacement [5].
However, the relevant researches were focused just on design life as well as limited piping systems,
and details of fatigue design features of specific old nuclear power plant are not unique.

Alternative Fatigue Evaluation

Re-arranged Transient. As described above, explicit fatigue parameters such a usage factor and so
on were not available for the components of certain old nuclear power plant whilst current operating
bases require implementation of inservice examination by ASME Section XI and licensing
commitments such as monitoring of operating transients, and address of environmental effects.
Therefore, it is necessary to demonstrate the adequacy of the CLBs through alternative fatigue
evaluation for the continued operation period beyond the design life.
A set of specific transients are necessary in order to perform the explicit fatigue analyses according
to ASME Section III requirements. However, since there were no sufficient data in case of old nuclear
power plant designed by ANSI B31.1 code, the most important but bothersome efforts had to be
assigned for determination of the transient set.
In this paper, at first, the transients defined in all the available documents [6-10] were assessed.
Then, considering huge amount of analyses, the specific transients were synthesized and grouped as
each conservative transient type as shown in Table 1. Fig. 2 illustrates resultant typical pressure and
temperature transients for alternative fatigue evaluation incorporated in the typical transient types A
and H. In addition, time independent external loads to simulate corresponding operating basis
earthquake loading were also considered for fatigue evaluation of each component.
1662 Advances in Fracture and Strength

(a) Transient type A b) Transient type H

Fig. 2 Typical pressure and temperature transients for alternative fatigue evaluation

(a) Element type effect (b) Ligament effect

Fig. 3 FE mesh sensitivity analyses results

Fig. 4 Typical weld FE model Fig. 5 Critical locations of typical Class 1 nozzles

Explicit Fatigue Analysis. A series of explicit fatigue analyses according to ASME Section III
were conducted to demonstrate the integrity of Class 1 straight pipes, elbows, butt welds and nozzles
after reflecting somewhat conservative assumptions [6-9].
Prior to the main stress analyses, several sensitivity analyses were carried out to check the effects
of element type, ligament and thermal sleeve. Fig. 3 depicts sensitivity analyses results on the element
type and ligament, in which the X-axis represents a thickness ratio (t/to=0.0; inner side, t/to=1.0; outer
side). From the sensitivity analyses results, optimum finite element (FE) models were selected after
considering the accuracy and analysis time.
 Key Engineering Materials Vols. 297-300 1663

Fig. 4 shows the typical weld finite element models for pipe and elbow used in the analyses. The
corresponding transient types which identify the pressure and temperature changes were applied
combined with the time independent external loads. Subsequently, Fig. 5 depicts boundary conditions
and resultant critical locations of typical Class 1 nozzles determined from finite element analyses.

Results and Discussion

Implicit Fatigue Evaluation Results. The implicit fatigue evaluation has been carried out, from
which 49-87% margins for 40-year operation and 45-86% margins for 50-year operation were
obtained using stress usage criterion. Table 2 represents the typical implicit fatigue evaluation results
for RCS Piping Segment 1 component according to ANSI B31.1 code which will be compared with
the corresponding explicit fatigue analysis results.

Table 2 Typical implicit fatigue evaluation result of RCS piping segment

Max. Allowable Thermal Expansion Stress
Stress Usage
Identification Input Data Expansion × Stress Range Reduce Factor
Stress (a) 40 years (b) 50 years (c) (a)/(b) (a)/(c)
Material: A351 CF8M
RCS Piping
Segment 1
Design Temp.: 343 [°C] 89.3 [MPa] 176.6 [MPa] 162.5 [MPa] 0.506 0.550
Design Pressure: 17.1 [MPa]

Explicit Fatigue Analysis Results. The explicit fatigue analyses were conducted according to
ASME Section III. In general, most of the analysis results by individual and superposed transient
cycles under 40- and 50-year operation were met the current acceptance criterion except a few
components of piping systems. Table 3 shows typical CUF ranges at typical locations which were
selected by highest value for each representative nuclear piping system. While the CUFs at some
components were over than 1.0, especially when using superposed cycle counting method, the main
reason is regarded as the conservatively re-arranged transients due to lack of design transients of old
nuclear power plant.

Table 3 Typical CUF ranges for Class 1 piping of old nuclear piping systems
CUF40 years CUF50 years
Identification
Individual Cycle Superposed Cycle Individual Cycle Superposed Cycle
RCS Piping Segment 1 0.000-0.001 0.001-0.825 0.000-0.001 0.002-1.031
SIS Piping Segment 3 0.461-0.700 0.700-1.030 0.576-0.875 0.875-1.287
RHRS Piping Segment 2 0.470-0.715 0.715-1.028 0.587-0.893 0.894-1.284
CVCS Piping Segment 1 0.007-0.044 0.139-0.172 0.008-0.055 0.174-0.215
RCS Nozzle 2 0.032-1.154 0.056-2.002 0.018-1.443 0.070-2.053

Discussion. The explicit fatigue analyses for Class 1 piping of old nuclear power plant led to
promising results whilst conservative assumptions were applied. Particularly, in the case of RCS
Piping Segments 1 which is one of critical location, implicit fatigue evaluation results showed stress
usage of 0.506 for 40-year operation and 0.550 for 50-year operation, respectively. The corresponding
explicit fatigue evaluation results showed CUF ranges from naught to 0.825 and naught to 1.031.
When adopting the mean value derived from the individual and superposed cycle counting methods,
the maximum CUFs were 0.413 for 40-year operation and 0.516 for 50-year operation respectively.
These values are similar to the values of stress usage as shown in Table 2.
Moreover, the critical components whose CUFs were over 1.0 in present paper were coincided
with those components such as surge line, charging nozzle, safety injection nozzle, RHR system Class
1 piping. These components were selected be addressed as a minimum for fatigue life on a sample of
1664 Advances in Fracture and Strength

critical components[11]. In addition, it is anticipated that the current exceptional critical components
may meet the acceptance criteria if more realistic operating transients are considered from on-line
monitoring. Therefore, it is anticipated that the integrity of Class 1 piping and associated components
regarding fatigue issue will be adequately maintained during the continued operation period through
explicit fatigue analyses and on-line monitoring.

Conclusions
In this paper, the Class 1 piping in four representative systems of a certain old vintage nuclear power
plant were evaluated and following conclusions have been derived.
1) Since there were no sufficient data of old nuclear power plant designed by ANSI B31.1 code, for
alternative fatigue evaluation, specific transient sets were generated conservatively after synthesizing
all the available design features.
2) Explicit fatigue analysis results as well as implicit fatigue evaluation results showed that most of
components satisfy the current acceptance criterion under continued operation periods beyond the
design life.
3) For a few components which did not satisfy the acceptance criterion, it is anticipated that the
integrity issue will be adequately maintained during the continued operation period if more realistic
operating transients are considered from on-line monitoring.

Acknowledgement
The authors are grateful for the support provided by a grant from Safety and Structural Integrity
Research Centre at Sungkyunkwan University.

References
[1] USA Standard Committee: B31.1 Code for pressure Piping (1967)
[2] Structural Integrity Associates, Inc.: Fatigue Comparison of Piping Designed to ANSI B31.1 and
ASME Section III-Class 1 Rules, EPRI/TR-102901 (1993)
[3] ASME PVP conference: Implicit Fatigue design of Piping Systems for License Renewal of B31.1
Plants, Vol. 468 (2003), pp. 139-146
[4] Westinghouse: Structural Analysis of Reactor Coolant Loop/Support System for Ko-ri Nuclear
Power Plant Unit No.1 (1974)
[5] KEPCO: Kori Unit 1 Replacement Steam Generator Safety Analysis Report (1996)
[6] KOPEC and Westinghouse: Aging Management Review for the Kori 1 Class 1 Piping (2001)
[7] Westinghouse: Structural Analysis of Reactor Coolant Loop Piping for the Kori Unit 2 Nuclear
Power Plant, WCAP-10253 Vol. 1 (1973)
[8] Westinghouse: ASME Section III Class 1 Reactor Coolant Loop Branch Nozzle Stress Analysis
for the Kori Unit 2 Nuclear Power Plant, WCAP-10190 (1983)
[9] Westinghouse: ASME Section III Class 1 Auxiliary Piping Stress Analysis for the Kori Unit 2
Nuclear Power Plant, WCAP-10189 (1983)
[10] Westinghouse: System Standard Design Criteria Nuclear Steam Supply System Design
Transients, Internal Communication
 Key Engineering Materials Vols. 297-300 1665

[11] A.G. Ware, D.K. Morton and M.E. Nitzel: Application of NUREG/CR-5999 Interim Fatigue
Curves to Selected Nuclear Power Plant Components, NUREG/CR-6260, U.S. Nuclear Regulatory
Commission, Washington D.C. (1995)
Advances in Fracture and Strength
10.4028/www.scientific.net/KEM.297-300

Alternative Fatigue Evaluation of Nuclear Piping Designed by ANSI B31.1 Code

10.4028/www.scientific.net/KEM.297-300.1659