Abstract

Nuclear power has gone through several transitional phases worldwide with early development, design, and construction in 1960s, 1970s, and 1980s, particularly in Japan and U.S. A relatively less active period followed in U.S. in 1990s followed by resurgence of interest and activities in 2000s. Currently, some countries are focusing on advanced reactor technology and smaller modular reactors instead of large light water reactors. Along with the transition in the need for power and technological advances, one other major change has occurred. Most of the workforce that participated in development, design, research, construction, and operation of currently operating plants is no longer active. There were some major research programs conducted during those years to develop methodology and prove the technology. Much of this knowledge is still very relevant and applicable and would be useful to young engineers and new professionals coming to the nuclear industry. One of such programs is seismic verification of major structures, systems, and components (SSCs) by tests in Japan during the period of 1981–2004 at a large shaking table facility. Opportunities for conducting such large tests are not likely in the current environment of the industry and, therefore, one of the objectives of this paper is to summarize some earlier tests to raise awareness for the newcomers to the industry. This paper discusses piping tests conducted in Japan at the era. In particular, this paper focuses on: insights obtained from the test results; comparisons of test results with responses during actual earthquakes; collaboration of Japan and U.S. in these tests; observations on robustness of Japanese seismic design; how the results relate to current issues; and use of these results from the future perspectives.

1 Introduction

Japan is in one of the world's highest seismicity areas and seismic safety of nuclear power plants (NPPs) has been one of the key issues related to nuclear safety.

The major construction activities of NPPs in Japan occurred during 1970s and 1980s.

Seismic design of NPPs has been based on the Standard JEAG 4601 (“Technical Guidelines for Aseismic Design of Nuclear Power Plants,” [1]), prepared by Japan Electric Association from 1970 under the leadership of Prof. H. Shibata, and “Review guideline of NPP Seismic Design” prepared by Nuclear Safety Commission of Japan (NSCJ) in 1981.

Nuclear Power Engineering Corporation (NUPEC) and subsequently the Japan Nuclear Energy Safety Organization (JNES), both established by Ministry of Economic Trade and Industry, conducted seismic verification tests of major structures, systems, and components (SSCs) of 1100 Mw class pressurized water reactor (PWR) and boiling water reactor (BWR) plants during 1981–2004, using the then world's largest shaking table located at Tadotsu in Shikoku.

Results from these tests to the design basis ground motion (DBGM), S2 (see definitions below) and more larger motions for margin evaluations have been reported in every pertinent ASME PVP conferences in a chronological order, other international nuclear engineering conferences, and delivered worldwide through individual reports.

The NUPEC and JNES collaborated with the U.S. Nuclear Regulatory Commission (NRC) and the Brookhaven National Laboratory (BNL), as a NRC's contractor, during some of these programs. The NRC collaborated by conducting analyses and developing various publicly available reports. This paper includes results from these collaborative activities.

However, even in Japan, it seems that the young engineers may not be completely familiar with these earlier tests. So, the authors' intentions are to introduce a brief outline and major result of these tests and provide a detailed reference list for convenience. It is impossible to capture all the major insights and all the details of these results in this short paper. The intention of the paper is to raise awareness of these major test programs.

This paper is divided into the following nine sections:

  1. Introduction

  2. Outline of seismic verification tests by NUPEC and JNES

  3. Outline of piping tests

  4. Japan–U.S. collaboration on Nuclear SSC's seismic verification tests

  5. Comparison of insights from tests and actual seismic response of NPPs

  6. Observations on robustness of Japanese seismic design

  7. International program for NPP piping seismic verification

  8. Post-Fukushima observations

  9. Summary

The author expresses acknowledgment to Nuclear Regulatory Authority Japan and The Institute of Applied Energy for permission of using related NUPEC reports.

The following nomenclature is used in the paper:

Seismic class As:

Highest seismic class for SSCs directly related to nuclear safety. After 2006, SSCs of this class and almost all SSCs of seismic class A are defined as “Seismic class S.” Examples of class As include: reactor pressure vessel and containment, control rod and insertion mechanism, residual heat removal system, emergency diesel generator, back-up battery system, and reactor building (RB).

Seismic class A:

Second high seismic class for SSCs related to nuclear safety. Examples include core internals and emergency core cooling system.

Seismic class B:

Structures, systems, and components with no direct function for nuclear safety but contain some radio activity. Examples include steam turbine, main steam and feed water line, and turbine building.

Seismic class C:

Structures, systems, and components with no function for nuclear safety and do not contain any radio activity. Examples include Main Transformer, Generator, and Fire protection system.

DBGM: design basis ground motion

  • S1: Maximum DBGM defined by NSCJ (1981), applied to Seismic class A SSCs

  • S2: Extreme DBGM defined by NSCJ (1981), applied to Seismic class As SSCs

  • Ss: Enhanced DBGM by NSCJ (2006); applied to Seismic class S SSCs

gal: unit of acceleration cm/sec2.

G: unit of acceleration 9.8 m/sec2.

2 Outline of Seismic Verification Tests by NUPEC and JNES

Sections 2.12.3 briefly describes the Tadotsu test facility and some general features of the various test programs. More detailed description of Tadotsu facility is included in Ref. [2].

2.1 General Features.

The availability of large-scale vibration table made it possible to conduct the tests of large-scale models closer to actual scales, which resulted in more accurate tests and contributed to refinement of seismic analysis methods for design. However, cost for construction/maintenance and tests themselves were high for the Tadotsu facility.

The vibration table of Tadotsu test laboratory was eventually closed in 2005 after completing tests of major SSCs described in Sec. 2.3 below.

This legacy of major test results needs to be shared with next generation worldwide.

2.2 Description of Test Facility.

The following is the summary of the shaking table performance and Fig. 1 shows schematics of the hydraulic actuator system, which was used to generate the shaking motions.

Fig. 1
Schematics of the shaking table
Fig. 1
Schematics of the shaking table
Close modal

Table size: 15 m ×15 m

Loading capacity: 1000 ton

Frequency range: 0–30 Hz

Hydraulic actuator:

  • Total horizontal capacity: 450 tonf × 7 (actuators)

  • Total vertical capacity: 300 tonf × 12 (actuators)

Maximum acceleration capacity:

  • With 500t load: 2.72G horizontal, 1.36G vertical

  • With 1000t load: 1.84G horizontal, 0.92G vertical

2.3 Test Objectives and Associated Schedules.

The following SSCs of 1100 Mw PWR and BWR were selected in four phases for testing at the Tadotsu vibration table.

2.3.1 Verification Tests or Proving Tests (i.e., Tests at the Design Basis Ground Motion to Confirm That an SSC Performs Its Safety Function).

Verification tests were conducted in three phases as follows:

  • Phase I: Single component tests (1981–1989) included containment vessel, reactor pressure vessel, core internal, and reactor cooling (recirculation) loop.

  • Phase II: Systems tests (1990–1993) included emergency diesel generators system, computer system with base isolation, and reactor shutdown cooling system.

  • Phase III: Evaluation of ultimate strength and new seismic technology tests (1994–2002) included main steam and feed water piping, prestressed concrete containment vessel, and reinforced concrete containment vessel, steam generator with energy absorbing support, and ultimate strength of piping.

2.3.2 Fragility Tests (i.e., Tests Conducted at Higher Levels of Ground Motion to Understand Functional Failure Levels That Can Be Used in a Fragility Calculation and Seismic Probabilistic Risk Assessments).

The fragility tests were conducted in Phase IV. The higher levels of ground motions were achieved by enhancing vibration table power up to 6 G, by adding second table on original table (2003–04). Tested components were electrical panels; horizontal/vertical shaft pumps; control rod insertion mechanism; and an overhead crane.

3 Outline of Piping Tests

This section describes the objectives, program, test models, excitation levels of vibratory input motions, few selected major results, and evaluations of each piping tests shown in Table 1.

Table 1

List of piping tests

a. PWR primary coolant loop systemd. Ultimate strength of piping system
b. BWR primary loop recirculation systeme. Eroded piping
c. Main steam (MS) and feed water (FW) piping system: MS piping of a PWR and FW piping of a BWR
a. PWR primary coolant loop systemd. Ultimate strength of piping system
b. BWR primary loop recirculation systeme. Eroded piping
c. Main steam (MS) and feed water (FW) piping system: MS piping of a PWR and FW piping of a BWR

The references [24] are for NUPEC Reports, and references [519,22] are individual documents and papers that provide additional details and they should be reviewed for further understanding. These references are highlighted in the discussions of individual tests.

3.1 Primary Loop.

The seismic performances of primary loops of a PWR and a BWR were evaluated using actual or close to actual size test models. These tests were able to directly confirm the seismic performance of PWR and BWR primary loops and functions of support structures such as snubbers and hangers. These tests were conducted in the range of excitation levels of input motions corresponding to 1.1–2.0 times S2, extreme design basis ground motion as defined earlier.

Fig. 2
PWR primary loop test model
Fig. 2
PWR primary loop test model
Close modal
Fig. 3
Reactor coolant pipe stress at increased S2
Fig. 3
Reactor coolant pipe stress at increased S2
Close modal

3.1.1 PWR Primary Coolant Loop System.

The references [2] and [57] will support reader's understanding in detail.

3.1.1.1 Model and excitation levels (input vibratory motions).

The following is the information of scaled model shown in Fig. 2. The model includes primary coolant piping, steam generator, and snubbers. This model reflects primary coolant loop system of a typical Japanese 1100 MWe class standard model of PWR plant. The scale of the model was made to be 1/2.5, by taking into account performance of the vibration table, testing conditions considering law of similitude, and other factors. Moreover, modified laws of similitude were applied as to have natural frequency ratio of 1/0.7 between the test model with mass and actual plant. Thus, the same strain (stress) is made to occur. The weight of the model is 525t.

The maximum excitation levels for S2 motions used in the tests were maximum horizontal acceleration of 1170 gal and maximum vertical acceleration of 334 gal. Duration of the motion was 28 s. The details of other tests with lesser input motions are described in Ref. [2].

3.1.1.2 Test objectives and results.

As stated in Ref. [2], the PWR primary loop tests aimed at proving the seismic reliability of the PWR primary coolant loop system, one of the major nuclear power plant facilities from a safety viewpoint, and also verifying adequacy of the then current design methods. The concrete objectives of the PWR primary coolant loop system seismic proving test were as follows:

  • To confirm the integrity of the functions of the reactor coolant pressure boundary and seismic support structures in the case of an earthquake, and

  • To confirm the adequacy of vibration characteristics in the case of an earthquake and also that of the aseismatic design methods.

In order to achieve these objectives, the model was tested under conditions similar to those of an actual plant. The tests conducted utilized vibratory motion up to 2.0 × S2 horizontal motion and 1.5 × S2 vertical motion under pressurized condition. No leak and abnormality of support structure were found for these vibratory motions.

An example of results from these tests is shown in Fig. 3. Figure 3 shows the maximum stress levels in three components of the loop with increasing intensity of the input vibratory motion. These results are important to demonstrate the adequacy of design methods of analyses used in the design. More complete test results are described in Ref. [2].

Fig. 4
Stress evaluation nodal point
Fig. 4
Stress evaluation nodal point
Close modal
3.1.1.3 Evaluation of design analysis method's validity.

The following steps were applied to confirm the validity: (i) simulation analysis of vibration test data; (ii) design analysis of vibration test data; and (iii) comparison of (i) and (ii).

Two design analysis methods were used, one using a time history approach and other using seismic response spectrum analysis approach. Results from these two analyses were compared with simulation analysis.

Following are the findings from both simulation analysis and design analysis methods paraphrased from Ref. [2].

  • The simulation analysis of the test result led to determining nonlinear rigidity characteristics of the snubbers and one-sided support, and also led to modeling the support structures of the primary coolant loop system accurately in terms of vibration analysis.

  • A result of the test revealed that the damping constant of the primary coolant loop system can be accurately estimated by a finding through the summation of the damping of the snubber dissipation energy and the other structural damping, and therefore, the adequacy of the use of design damping constant of 3% was considered confirmed.

  • By performing a comparison between the test model analysis values based on the then current aseismatic design methods and the test result values, and also performing examination by the simulation analysis, it was confirmed that vibration characteristics of the test model can be reproduced by a multi-mass system beam model for numerical analysis.

    Thus, the adequacy of the then current aseismatic design method was considered confirmed.

  • The results of the seismic response spectrum analysis and time historical modal analysis, both used for the aseismatic design methods, were compared with the simulation analysis with time history analysis of the proving test result, and the examination was carried out. As a result, analyses listed in descending order of accuracy of the seismic responses were the response spectrum analysis, the time history modal analysis, and the simulation analysis. Thus, the then current design analysis methods were confirmed to have an adequate safety margin.

3.1.1.4 Evaluation on seismic reliability of actual plant.

Aseismatic strength and adequacy of design method were confirmed for a PWR primary coolant loop system. S1 and S2 seismic response analyses were carried out for a primary coolant loop system of an actual nuclear power plant and also stress evaluation combining loads other than earthquake was performed. The primary loop represented a primary coolant loop system of a 1100-Mw four-loop PWR plant.

The maximum primary stress (combination of seismic load for S2 excitation + inner pressure + dead weight load) of the plant primary loop piping obtained by the analysis at S2 excitation was 8.91 kgf/mm2 at hot leg node 115 as shown in Fig. 4. The value is significantly smaller than the allowable value of 3Sm, which translates to 35.4 kgf/mm2.

Fig. 5
BWR primary loop recirculation system model
Fig. 5
BWR primary loop recirculation system model
Close modal

In the test, the hot leg stress was larger than other parts of primary loop piping at S2 level of excitation, observed value at hot leg node115 was 3.52 kgf/mm2 as shown in Fig. 3. The stress value estimated using the design analysis model was 7.16 kgf/mm2.

With regard to the actual plant evaluations, seismic response stresses were also calculated by a modified analysis method called, “verification analysis,” that incorporated experiences of simulation of test data. For pipe stress evaluation, this analysis was applied and hot leg stress at node 115 was 8.91 kgf/mm2 as described above. Thus, the seismic reliability of an actual loop was confirmed through several analysis methods and compared with test results.

3.1.2 BWR Primary Loop Recirculation System.

The references [3] and [8] will support reader's understanding in detail.

3.1.2.1 Model and excitation levels (input vibratory motions).

The test model shown in Fig. 5 is a full-scale mimic of one of the two primary loop recirculation (PLR) systems for the improved and standardized 1100 MWe BWR plant for high seismic zone. It consists of one test loop including two partition valves, one recirculation pump, one motor and support, support structure of a mimic of reactor pressure vessel, and gamma ray shield, which supports the piping, a pipe support frame, and a support for installation on the vibration table. Figure 5 shows the BWR test model. The weight of the model was 665t.

The maximum excitation levels for S2 motions used in the tests were maximum horizontal acceleration of 2038 gal and maximum vertical acceleration of 358 gal. The duration of the motion was 20 s. The details of other tests with lesser input motions are described in Ref. [3].

3.1.2.2 Test objectives.

The seismic proving test of BWR primary loop recirculation system was a historical first as it was the first piping test to use the vibration table of Tadotsu Engineering Laboratory.

The objectives of the test were as follows:

  • To prove the seismic reliability of the PLR, one of the most important safety components in the BWR nuclear power plants, and

  • To confirm the adequacy of seismic analysis methods used in the then current seismic design.

To achieve the objectives, the tests were conducted under conditions as near as possible to those for actual systems, using a test model of full-scale with the structure as close as possible to that of the actual plants. Several simulation methods involving nonlinear analyses were evaluated along with the adequacy of design analyses methods.

3.1.2.3 Seismic strength proving tests and results.

For the purpose of the proving tests, three input motions of S1, 1.0 S2, and 1.1 S2 were used with the system being pressurized by inner operating pressure. No leak and no abnormality of piping and support structures were found. Thus, the functional seismic reliability of the system was verified.

Figure 6 shows examples of dependence of the stresses at the major points of the test model on the excitation level as compared with S2 seismic response wave. This shows overall approximate linearity of acceleration, stress, and reaction force, with some exceptions of nonlinearity in the range of low excitation level. Thus, the test model, as a whole, showed linear behavior even for such a high seismic level.

Fig. 6
Pipe stress (reducer SN2)
Fig. 6
Pipe stress (reducer SN2)
Close modal
3.1.2.4 Evaluation of seismic design analysis methods.

Test results were utilized for verification of important factors of seismic response analysis methods, including damping factors, effect of cut of high frequencies, and modeling of support device nonlinearity. The following are some of the key findings:

  • Examination of damping factor

    In seismic design of piping, damping factor of 0.5% to 2.5% is used. Support reaction force analysis up to S2 indicated that 2.5% damping factor was conservative when compared with test data. Thus, in the case of PLR piping, the availability of 2.5% damping factor was verified.

  • Evaluation of effect of cut of frequencies in dynamic analysis

    In general, modes of 20 Hz and less are taken in consideration in dynamic analysis on response spectrum method in seismic design of piping. In this piping system, validity of analysis with cut-off frequency 20 Hz was confirmed by comparing the calculated results with the test results.

  • Modeling of mechanical snubber nonlinearity simulation analysis incorporating nonlinearity of pipe support (inner structure gap of mechanical snubber) well simulated test model seismic responses, providing confirmation of nonlinear model.

3.1.2.5 Evaluation on seismic reliability of actual plant.

The S1 and S2 seismic response analysis were carried out for the PLR system of improved and standardized 1100 MWe BWR plant for high seismic zone and also stress evaluations combining loads other than earthquake were performed.

The maximum primary stress (combination of seismic load for S2 excitation + inner pressure + dead weight load) of the plant PLR piping obtained by the analysis at 1.0S2 excitation was 29.7 kgf/mm2 at reducer in header node 85 in Fig. 7. The value is smaller than the allowable value of 3Sm, which translates to 36.2 kgf/mm2 and thus the seismic reliability was confirmed.

Fig. 7
Stress evaluation nodal point
Fig. 7
Stress evaluation nodal point
Close modal
3.1.2.6 Summary and evaluation of test results.

By conducting the vibration test of the full-scale test model of the primary loop recirculation system with the large-scale high-performance vibration table, seismic reliability of the PLR system was proved. The data obtained in the test were analyzed and evaluated to clarify the vibration characteristics of the PLR system. Furthermore, data were obtained, which could be effectively used in more rational design of future to reflect the results obtained into actual system design.

3.2 Main Steam and Feed Water Piping System.

The references [4] and [917] will support reader's understanding in detail.

3.2.1 Test Outline and Objectives.

Test models for PWR, main steam (MS) piping (M-line), and BWR feed water (FW) piping (F-line) with conventional supports such as snubbers/hangers and energy absorbing supports were tested for the following purposes in Ref. [4]:

  • To confirm reliability and margins of MS and FW piping system with conventional support; and

  • To investigate applicability of energy absorbing supports in future.

3.2.2 Model and Excitation Levels (Input Vibratory Motions).

Figure 8 shows the test configuration of both systems on the shaking table. The scale factors and weight of the test model are as follows:

Fig. 8
MS and FW piping system test configurations
Fig. 8
MS and FW piping system test configurations
Close modal
  • (a)

    Scale: 1/2.66 for MS piping

  • (b)

    Scale: 1/2.3 for FW piping

  • (c)

    Weight of model: 190t.

  • (d)

    Excitation levels of vibratory motion S2

    • Horizontal acceleration 1489 gal and Vertical acceleration 340 gal for M-line

    • Horizontal acceleration 1090 gal and Vertical acceleration 184 gal for F-line

As shown in Fig. 8, M-line was vibrated not only by shaking table but simultaneously by a computer-controlled actuator system simulating steam generator response.

3.2.3 Test Results

The following are some of the key results.

  • ln the tests on piping with conventional support, the reliability was confirmed up to l.5 S2 for M-line and 1.3 S2 for F-line. No abnormality and water leak on piping and fatigue strength of pipe penetration bellows were confirmed by component test.

  • ln the tests using 2.5 S2 motion or more and resonance wave on piping system models with EA support, M-line pipe stress reached its yield stress, while F-line EA support went into its functional loss (collision with a stopper). However, no abnormality and water leak occurred in the piping. It was confirmed that reliability of piping system with EA support is equivalent or better than piping system with conventional support.

3.2.4 Analysis Methods.

Design and simulation analyses on test model to verify conservatisms of a design analysis and validity of a simulation analysis was conducted. The simulation analysis was then used on an actual plant model to compute actual plant responses.

The results of design and simulation analyses and their comparisons with test results for F-line (BWR FW piping) with conventional supports are introduced here as a representative example.

For M-line (PWR MS piping), test and analyses results are discussed in Sec. 4 as an example of a collaborative activity between Japan and U.S.

  • Design analysis and simulation analysis of the test model

    Methods, analysis model, and results of both analyses are shown in Table 2, Figs. 9, and 10, respectively. The examination of comparisons shown in Fig. 10 confirms the conservatism of the selected design analysis method. Results of the simulation analysis also compare reasonably well considering complexities and some inherent nonlinearities in a piping system as discussed in Sec. 4.

Fig. 9
Analysis model of F-line
Fig. 9
Analysis model of F-line
Close modal
Fig. 10
Analysis result on F-line pipe stress at S2 excitation
Fig. 10
Analysis result on F-line pipe stress at S2 excitation
Close modal
Table 2

Analysis methods

ItemDeign analysisSimulation analysis
Analysis modelFigure 9 Figure 9 
Snubber modelingDesign stiffness (no gap)Actual stiffness with gap (based on performance test)
Analysis methodResponse spectrum analysisNonlinear time
history analysis
Damping constant2.0%3.2∼4.6%
ItemDeign analysisSimulation analysis
Analysis modelFigure 9 Figure 9 
Snubber modelingDesign stiffness (no gap)Actual stiffness with gap (based on performance test)
Analysis methodResponse spectrum analysisNonlinear time
history analysis
Damping constant2.0%3.2∼4.6%
  • Simulation analysis on actual BWR MARK IIR

    Figure 11 and Table 3 show the analysis model of actual BWR MKIIR plant FW piping that was used and simulation stress results, respectively.

    The simulated maximum primary stress of BWR MKIIR FW piping (due to earthquake, internal pressure, and self-weight) was 6.8 kgf/mm2 (67 MPa) at node no. 38, small enough compared with allowable stress 37.4 kgf/mm2 for Level IVAs, equivalent to ASME Level D.

Fig. 11
Analysis model of actual BWR FW piping
Fig. 11
Analysis model of actual BWR FW piping
Close modal
Table 3

BWR FW piping stress at S2 earthquake by simulation analysis verified through the test

Nodal no.Primary stress (kgf/mm2) (due to earthquake, inner pressure and self-weight)
135.5
215.9
306.1
386.8
444.2
483.5
50 N3.8
644.6
66 N5.2
Nodal no.Primary stress (kgf/mm2) (due to earthquake, inner pressure and self-weight)
135.5
215.9
306.1
386.8
444.2
483.5
50 N3.8
644.6
66 N5.2
  • Insight obtained for analysis methods

    Simulation analysis for M-line multi-input vibration test indicted conservativeness to test data and verified applicability of multi-input response spectrum analysis. Snubber model incorporating mechanical gap in inner structure of snubber well simulated piping system response.

  • Perspectives for piping system with E.A. support

    Reliability of piping system with E.A. supports was confirmed and applicability of analysis method for that was also confirmed. Application of E.A. supports will contribute to harmonization of improvement of reliability and rational design of piping system.

    Development of piping system design technology considering elasto-plastic characteristics of piping will be effective for optimizing excessive margin existing in current seismic design based on elastic deformation theory.

3.3 Ultimate Strength of Piping System.

A large-scale piping system test and elasto-plastic analyses were conducted to assure safety margin of seismic design code for piping in Japanese NPPs. References [18] and [19] describe the test program and results in detail. Insights from tests were used to inform Japanese seismic design code for piping.

3.3.1 Test Objectives

There were three objectives as follows:

  • To clarify the elasto- plastic response and ultimate strength of the nuclear piping system.

  • To ascertain the seismic safety margin of current design code and

  • To assess proposed allowable stress rules at the time.

3.3.2 Description of Test Program.

Quasi‐static loading tests of various piping components and dynamic shaking test of a piping system were carried out. The system test was conducted for two cases, for design stress levels, and for over ten times of design stress levels, by adding mass, removing the supports from normally designed piping system, and using seismic resonance wave for input to achieve the elasto-plastic behavior, on 8B sch 80 carbon steel piping system. Figure 12 shows the test configuration in latter case.

Fig. 12
System test configuration for ultimate strength level
Fig. 12
System test configuration for ultimate strength level
Close modal

3.3.3 Test Results.

One of the important results was that in ultimate stress level test, the system maintained its function up to 8.5 times stress of design level that occurred in fourth excitation. At fifth excitation, piping failure occurred by through-wall fatigue cracking at the inner surface flank of elbow 2, as shown in Fig. 13. Details of data, evaluations, and discussions on safety margins of seismic design codes are provided in Ref. [18].

Fig. 13
Failure mode of the piping system
Fig. 13
Failure mode of the piping system
Close modal

3.3.4 Discussion of Results.

Elasto-plastic and fatigue analyses of the data of quasi-static piping component tests and dynamic piping system tests were conducted. Also, frequency ratio (ratio of natural frequency of the system and peak frequency of input seismic wave Rw) dependence and influence of input wave spectrum broadening on response analysis to acceleration safety margins were evaluated.

Through these evaluations, test results were summarized as allowable acceleration safety margin MA under seismic design code at the time, as shown in Fig. 14. In the test model case where Rw was 0.7, the margins were 4.9–9.6.

Fig. 14
Frequency ratio dependence of allowable acceleration safety margin (MA)
Fig. 14
Frequency ratio dependence of allowable acceleration safety margin (MA)
Close modal

This evaluation method and insights were referred to Japanese piping design code committee for nuclear power plant. The proposed allowable stress rules based on the test data, namely, abolition of primary stress restriction and adoption of approximate coefficient for quasi-elasto-plastic analysis, were adopted in the seismic design rule JEAG4601 2008.

3.4 Eroded Piping.

Pipe thinning by aging and/or erosion is an important matter to be considered for NPP safety. For example, such failures occurred in feed water line of Surry NPP in 1986 and condensate water line of Mihama NPP in 2004, which are discussed in Refs. [20] and [21], respectively.

Tests in this program were conducted to investigate the seismic safety of eroded piping. Reference [22] described the details of this program.

3.4.1 Test Objectives

The primary test objectives were:

  • to clarify the elasto-plastic behavior and ultimate strength, and

  • to investigate seismic safety margin of thinned wall piping.

3.4.2 Test Program.

In this program, quasi-static loading tests of piping components and dynamic shaking tests of piping systems were conducted as described below.

  • Component tests

    To investigate failure mode/fatigue life, effects of location and form of eroded portion, and to confirm validity of elasto-plastic analysis for eroded components, cyclic loading tests of thinned-wall piping components such as elbows and tees were conducted. The location and the form of thinned-wall portions were to be determined from piping erosion data from actual plants and from preliminary analysis results. The wall thickness of the thinned-wall portion is the minimum thickness allowed under the seismic design code. In these tests, load for strain range of 3% at outer surface was applied, which is much larger than the allowable levels. Figure 15 shows test configuration for the component tests.

Fig. 15
Eroded piping component test configuration
Fig. 15
Eroded piping component test configuration
Close modal
  • System tests

    Three-dimensional eroded piping systems were tested under resonant conditions seismic wave in order to resolve empirically and analytically the issue of what the specific behavior of eroded piping system would be. The system response characteristics under large stress condition were investigated by adding mass and using resonance wave as an input. Figure 16 shows the test configuration.

Fig. 16
Eroded piping system test configuration
Fig. 16
Eroded piping system test configuration
Close modal

3.4.3 Examples of Test Results.

Figure 17 shows an example of component test data, indicating the hoop strain time history of the inner and outer surfaces of an elbow flank in the test of form (1) elbow.

Fig. 17
Hoop strain time history of form (1) elbow ** size: 8B Sch100, material: STS410
Fig. 17
Hoop strain time history of form (1) elbow ** size: 8B Sch100, material: STS410
Close modal

The maximum strain range of the inner surface is about four times larger than that of the outer surface and the crack penetrated from the inner surface.

Figure 18 shows an example of system test data.

Fig. 18
Hoop Strain time history of partially thinned elbow 1 by one excitation
Fig. 18
Hoop Strain time history of partially thinned elbow 1 by one excitation
Close modal

Through wall crack occurred at elbow 1, location of elbow 1 is shown in Fig. 16, during eighth run.

3.4.4 Discussion of Results.

Component tests and system tests confirmed that pipe element fatigue was the failure mode.

  • Fatigue life evaluation of eroded pipe element

By component tests, fatigue life data of eroded pipe elements were obtained. Figure 19 shows fatigue life of piping component observed at this test and Ultimate Strength of Piping System tests. During the process, appropriate material property to evaluate the behavior of eroded pipe element was also investigated.

Fig. 19
Fatigue life of piping components
Fig. 19
Fatigue life of piping components
Close modal
  • Evaluation of Eroded piping system behavior

Figure 20 shows a comparison of the relationship between the strain range of the outer surface at the elbow flank and the piping center deflection range.

Fig. 20
Comparison of relationship between elbow frank strain and piping deflection range
Fig. 20
Comparison of relationship between elbow frank strain and piping deflection range
Close modal

The usage factors and the equivalent number of cycles for the input seismic wave were evaluated for five test cases. The calculated usage factors ranged from 0.6 to 3.0 and the validity of the fatigue curve was confirmed.

4 Japan–U.S. Collaboration on Nuclear Structures, Systems, and Components Seismic Verification Tests

The references [4] and [9] will support reader's understanding in detail.

4.1 Outline.

NUPEC/JNES and U.S. NRC have been conducting collaborations on nuclear SSCs seismic verification tests from 1984 by analyzing and evaluating test data together. As an example of such a collaborative activity, the PWR main steam piping system (M-line) tests that were discussed in Sec. 3.2 are utilized.

Figure 21 shows the configuration of the test and the details of dampers and location of supports.

Fig. 21
PWR main steam system test model and details of energy absorbing supports and locations
Fig. 21
PWR main steam system test model and details of energy absorbing supports and locations
Close modal

Figure 22 shows the input S2 motion and associated response spectra. This input motion reflects DBGM S2. Figure 23 shows analysis model.

Fig. 22
Characteristics of input motion S2
Fig. 22
Characteristics of input motion S2
Close modal
Fig. 23
Analysis model of M-line
Fig. 23
Analysis model of M-line
Close modal

4.2 Comparison of Analysis Methods.

Differences between Japan and U.S. analysis methods are summarized in Tables 4 and 5.

Table 4

Comparison of design analysis methods on M-line with EA support

ItemJapanU.S.
Snubber modelingDesign stiffnessEquivalent stiffness
Floor response spectrum broadening10%15%
Cut off frequency50 Hz100 Hz
Damping constant2.5%2%
Analysis modelSingle input response spectrumMultiple input response spectrum
ItemJapanU.S.
Snubber modelingDesign stiffnessEquivalent stiffness
Floor response spectrum broadening10%15%
Cut off frequency50 Hz100 Hz
Damping constant2.5%2%
Analysis modelSingle input response spectrumMultiple input response spectrum
Table 5

Comparison of simulation analysis methods on M-line with EA support

ItemJapanU.S.
Input conditionFour-point inputTwo-point input
EA support characteristicRamberg-OsgoodBilinear
Low pass filter50∼33 Hz200 Hz
Linear damping constant2.2%2.5%
Numerical algorithmNewmark-βaNewmark-βa
Stress evaluationFour nodal pointsTwo nodal points
ItemJapanU.S.
Input conditionFour-point inputTwo-point input
EA support characteristicRamberg-OsgoodBilinear
Low pass filter50∼33 Hz200 Hz
Linear damping constant2.2%2.5%
Numerical algorithmNewmark-βaNewmark-βa
Stress evaluationFour nodal pointsTwo nodal points
a

A numerical integration method widely used in dynamic response analysis.

4.3 Comparison of Analysis Result.

Figure 24 shows design analysis and simulation analysis results of both sides, comparing with test data.

Fig. 24
Comparison of computed results by japan versus U.S. and test results (pipe stress of M-line with EA support, at S2 excitation)
Fig. 24
Comparison of computed results by japan versus U.S. and test results (pipe stress of M-line with EA support, at S2 excitation)
Close modal

4.4 Discussion.

As shown in Tables 4 and 5, there are some differences in modeling characteristics of snubbers (conventional supports) and EA supports. Differences of seismic design concepts and calculation methods between Japan and U.S. are further summarized in Ref. [23].

As an example, Fig. 24 shows results of test, design analysis, and simulation analysis with respect to pipe stress of M-line with EA support in case of S2 excitation.

In this case, design analysis of Japan seems to have large conservatisms. U.S. linear analysis followed test data well, in general, but appears to have underestimated stresses at some locations.

The following observations from Ref. [9] give some insights into complexities of piping system analyses. Reference [9] states:

“In the simulation analyses for the EA supports case, the Bouc-Wen hysteretic model properly reproduced the observed nonlinear behavior of the EA supports. However, in addition to the problems associated with the mechanical snubbers, other types of pipe support such as pin supports and sliding guides also contributed nonlinear characteristics to piping systems due to mechanical gaps in their support structures. In general, the observed analysis errors are less for higher excitation test runs as the effects of mechanical gaps in pipe supports become less significant.”

With respect to the design analysis, Ref. [9] observed:

“For both piping systems, the pipe stress results were generally more conservative than the support force results. On an average basis, the design analyses provided adequate margins for both pipe stresses and support forces.”

More recent piping benchmark project discussed in Sec. 7 also provides insights into piping analysis issues.

5 Comparison of Insights From Tests and Actual Seismic Response of Nuclear Power Plants

This section describes comparison of the above test results with the observations at NPPs during some of the recent earthquakes. The Nigata-ken Chuetsu Oki (NCO) earthquake and the Tohuku earthquake are briefly examined.

During these earthquakes, a typical experience of NPPs in Japan was that the recorded ground motions exceeded ground motion levels of DBGM S2 in several cases. The following are the detailed discussions of two events.

5.1 NCO Earthquake

Details of this earthquake are as follows:

Date: 16 July 2007

Mw (moment magnitude): 6.6

The distance of Kashiwazaki Kariwa (KK) NPP to the epicenter was 16 km and these units were most affected by the event.

Observed KK NPP responses at the RB basemat from the earthquake event for each of the seven units at the site are shown in Table 6. The table also included design acceleration values.

Table 6

Comparison of observed and (designa) Max. response acc. at RB of KK NPP units base/mat, gal

UnitTypeNorth South (NS)East West (EW)Vertical
1BWR MkII311(274)680b(273)408(235)
2MkIIR304(167)606(167)282(235)
3MkIIR308(192)384(193)311(235)
4MkIIR310(193)492(194)337(235)
5MkIIR277(249)442(254)205(235)
6ABWR271(263)322(263)488c(235)
7ABWR267(263)356(263)355(235)
UnitTypeNorth South (NS)East West (EW)Vertical
1BWR MkII311(274)680b(273)408(235)
2MkIIR304(167)606(167)282(235)
3MkIIR308(192)384(193)311(235)
4MkIIR310(193)492(194)337(235)
5MkIIR277(249)442(254)205(235)
6ABWR271(263)322(263)488c(235)
7ABWR267(263)356(263)355(235)
a

Values in parentheses are response values by S2 motion.

b

Largest value in horizontal.

c

Largest value in vertical.

The averaged observed/design ratios for NS direction, EW direction, and vertical direction are: 1.3, 2.1, and 1.5, respectively. Thus, there are significant exceedances of the design values. Therefore, it is instructive to examine the response of the plant. All units in operation shut down automatically. No damage of SSCs in seismic class As was found by inspection. Various damage and malfunction of SSCs in seismic class B, C were found. For example, anchor bolt brake, crack on turbine building wall, Turbine blade contact, buried piping brake, fire on transformer, and local liquefaction.

5.2 Tohoku Earthquake.

The second earthquake of interest is Tohuku earthquake with the following characteristics.

Date: 11 March 2011

Mw: 9.0

Two affected NPPs that were affected by the event and the measured seismic responses at RB base mats are shown in Table 7. The table also shows design values.

Table 7

Measured ground motion levels at two NPPs

Nearby NPPMax. response acc. observed (design) at RB base/mat, gal( ) : Response by backcheck Ssa
1. Onagawa (to epicenter 130 km)607 (594) gal NS at Unit 2
439 (451) gal vertical at Unit 1
2. Fukushima Daiichi* (to epicenter 180 km)550 (438) gal EW at Unit 2
302 (420) gal vertical at Unit 2
*Due to lack of redundancy of emergency power source and collapse of outer power feed line steel tower, tsunami by the earthquake caused the severe accident.
Nearby NPPMax. response acc. observed (design) at RB base/mat, gal( ) : Response by backcheck Ssa
1. Onagawa (to epicenter 130 km)607 (594) gal NS at Unit 2
439 (451) gal vertical at Unit 1
2. Fukushima Daiichi* (to epicenter 180 km)550 (438) gal EW at Unit 2
302 (420) gal vertical at Unit 2
*Due to lack of redundancy of emergency power source and collapse of outer power feed line steel tower, tsunami by the earthquake caused the severe accident.
a

Backcheck ground motion is a newly defined ground motion after the NCO earthquake and is larger than the former Ss.

Tohoku earthquake of March 2011 was larger than NCO earthquake of July 2007. As seen in Table 7, the difference between recorded and design evaluation earthquake levels were much smaller when compared to the KK responses in Table 6. Therefore, the seismically induced damage was relatively minor from the Tohoku events.

5.3 Evaluation of Kashiwazaki Kariwa Nuclear Power Plant Structures, Systems, and Components Integrity.

The largest measured response value of NPP experienced in Japan is that of KK NPP as follows, from Table 6:

680-gal horizontal (EW) at unit 1

488-gal vertical at unit 6

The NCO earthquake caused large disaster in Niigata area. However, KK NPP had safely shutdown automatically.

Behavior of KK NPP and its SCCs verified, in part, the robustness of Japanese seismic design because no damage of seismic class As and A SSCs were found by inspections.

After the earthquake, detailed integrity inspection of all seismic class As and A SSCs and major seismic class B and C SSCs were conducted for all seven KK units.

Especially, at Unit 7, stress evaluation by design analysis was conducted for condition of IIIAs, equivalent to ASME level C, on seismic class As and A SSCs, to screen out SSCs having enough margin, from detail inspection.

These evaluations are described in detail in the following documents:

Nuclear and Industrial Safety Agency in Ministry of Economic Trade and Industry (NISA) final report dated Oct. 3, 2008 and NSCJ review statement dated Oct. 31, 2008.

As for piping, residual heat removal (RHR) system piping was close to IIIAs stress limit, while FW system had significant margin, as shown in Table 8.

Table 8

Example for piping of KK Unit 7 seismic class As, A SSCs integrity evaluation after NCO earthquake, in above NISA report

Piping systema. Max. stress (MPa)b. Allowable limits (MPa, in IIIAs)a/b
FW105 primary28237%
MS186 primary28765%
RHR274 primary28596%
Piping systema. Max. stress (MPa)b. Allowable limits (MPa, in IIIAs)a/b
FW105 primary28237%
MS186 primary28765%
RHR274 primary28596%

Residual heat removal piping system was evaluated in an International Atomic Energy Agency (IAEA) international seismic program as briefly discussed in Sec. 7.

6 Observations on Robustness of Japanese Seismic Design

To understand robustness of seismic class As and A SSCs of Japanese NPPs, one of the most significant events was NCO earthquake and ground motion levels it imparted to Kashiwazaki Kariwa NPP.

These motion levels were 1.3–2.1 times in average larger than ground motion for design at RB base/mat as discussed earlier. This event did confirm robustness of seismic class As and A SSCs based on design method in use.

Main factors pointed out by NSCJ for the margins in design and robustness (NSCJ statement 20 NSCJ 25, 10.31, 2008) were as follows:

  • Use of DBGM S1 for elastic design;

  • Use of standard static seismic force 3 times larger than conventional civil structures for As class structures;

  • Use of peak broadening of floor response spectra; and

  • Verification and improvement of seismic design and evaluation methods in Ref. [1] by the tests conducted at Tadotsu laboratory.

7 International Program for Nuclear Power Plant Piping Seismic Verification

Seismic design and verification are basically a well-understood technology and practices have been established since 1970s and 1980s. To further verify these methods, the IAEA started a Seismic Extra Budgetary Program (EBP) in 2008, after NCO earthquake, with over 30 countries and organizations participating. This EBP was based on a proposal from Japan (NISA and JNES) in 2006.

In the EBP, IAEA developed a program called “KARISMA,” an acronym for Kashiwazaki-Kariwa Research Initiative for Seismic Margin Assessment. The “KARISMA” program included benchmark analyses by the participants of KK RHR piping response simulation and comparisons with the observed data.

KARISMA report [24] identified following issues for piping analysis based on experience of RHR piping benchmark:

  • Need to investigate friction versus damping in dynamic analysis;

  • Need to develop more precise procedures for multi-support input motions time history analysis;

  • Need to extend the dynamic analysis to nonlinear material behavior;

  • Need to develop criteria for allowable strain in the case of material nonlinearities; and

  • Need to investigate credible piping failure modes that should be considered for margin assessment.

The report also pointed out that typical elements for seismic response, such as damping factor and analysis methods, are somewhat defined differently in applicable codes of different countries. These differences contributed to different results from the analyses of the same model.

One of the current important issues pointed out by KK-NCO earthquake event is importance of low seismic class SSCs. IAEA Seismic EBP program is now preparing a project for gathering and evaluating seismic experience data including low seismic class SSCs.

8 Post-Fukushima Observations

It is not the intent of this paper to discuss post-Fukushima seismic activities in terms of changing requirements and additional measures that are being undertaken. However, there are several new seismic requirements established by the Nuclear Regulatory Authority of Japan, including enhancement of DBGM evaluation, requirement for a seismic probabilistic risk assessment, and criteria for fault displacement movement.

The readers are referred to Ref. [25] for more information. What is important here is to indicate that results of tests described here are equally applicable as these activities are carried out. In particular, the functional demonstration of SSCs by these tests provides insights on fragility and system modeling.

9 Summary

The Tadotsu tests that had been conducted by NUPEC and JNES with the U.S. NRC collaboration cooperating have well contributed to verify and improve seismic design and evaluation methods. These large-scale tests continue to provide important insights into the seismic behavior. The authors hope that the paper would be served as a knowledge-transfer guide to next generation and raise the awareness of available information.

Acknowledgment

Seismic proving test of NUPEC had been led by Professor Shibata H. as the chairman of executing committee. Various piping tests were led by Professor Akiyama, H., Professor Suzuki, K., and Professor Shiratori, M. The authors wish to express deep acknowledgment to these Professors.

The authors also want to express their thanks to the leaders and colleagues of NUPEC, Tadotsu Laboratory and JNES, including Mr. Kawakami, Dr. Terada, Mr. Yoshioka, Mr. Fujino, Mr. Takayama, Dr. Sasaki, Mr. Tanaka, Mr. Uchiyama, Mr. Nakamura, Mr. Ichihashi, Mr. Kuroda, Dr. Suzuki, Dr. Iijima, Mr. Inagaki and Mr. Kawauchi, Dr. Godoy, A., Mr. Samaddar, S., Mr. Morita, S., and Dr. Ebisawa, K., planned and led the IAEA Seismic EBP discussed here; and Dr. Altiniyolar, A. conducted KARISMA program.

From early stage to the completion, these tests had been conducted under collaboration of NUPEC/JNES and the U.S. NRC.

The leader of U.S. side in 1990s had been Dr. Chokshi, N. Other NRC leaders have included Dr. Costello, J., Dr. Murphy, A., Dr. Ali, S., Dr. Kammerer, A., and Dr. Pires, J. The authors also want to thank above NRC colleagues and also BNL colleagues, which included: Dr. Hofmayer, C., Dr. Park, Y., and others.

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