ML20133G882
| ML20133G882 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 10/14/1985 |
| From: | Bells S, Byam T, Hable A ILLINOIS POWER CO. |
| To: | |
| Shared Package | |
| ML20133G872 | List: |
| References | |
| NUDOCS 8510160171 | |
| Download: ML20133G882 (142) | |
Text
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CLINTON POWER STATION, UNIT 1 ILLINOIS POWER COMPANY EQUIPMENT SEISMIC ASSESSMENT PROGRAM FOR SAFETY-RELATED MECHANICAL & ELECTRICAL EQUIPMENT PREPARED BY:
/
s K. J. HaWIe Staff Engineer -
Technical Assessment REVIEWED BY:
d.
T. A. ffam Staff Engineer -
Technical Assessment CONCURRED BY:
lh Jh
'S. R. Bell Supervisor -
Mechanical Engineering CONCURRED BY:
// q,m j R. A. Parson Sargent & Lundy Engineers CONCURRED BY:
L)88 Y M
/} W. Blattner
%rgent & Lundy Engineers APPROVED BY:
- 7. /
w D. L. Holtzschr. f Director - Nur. lear Sahty and Engineering Analysis K
3 n-SUhMARY i
This report describes the Equipment Seismic Assessment Program which provides additional assurance of the seismic capability of the Emergency Power Supply System and the Decay Heat Removal System at Illinois Power Company's Clinton Power Station.
The program was developed in response to a request made by the Advisory Committee on Reactor Safeguards and the Nuclear Regulatory Commission. The program was implemented in three phases; piping design, which examined the adequacy of the design methods used for small bore piping (2" diameter and under); interaction analysis.
which examined as-built equipment configurations fot seismic concerns; and equipment evaluation, which evaluated the ability of equipment i
to survive earthquake acceleration loadings. No problems were identified except in the area of piping design. These problems were identified during the piping design review and were reported to the NRC as a potential reportable deficiency under 10CFR50.55(e). All problems hate been corrected and IP is monitoring ongoing work in this area. The report concludes that there is sufficient assurance that the emergency power and decay heat removal systems can withstand the safe shutdown earthquake.
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F Table of Contents Section Page 1
Introduction 1-1 1.1 Program Summary 1-1 1.1.1 Phase I - Piping Design 1-1 1.1.2 Phase II Interaction Analysis 1-2 1.1.3 Phase III - Equipment Evaluation 1-3 1.
2 Achievement of the ESAP Objectives 1-3 2
Phase I - Piping Design 2-1 2.1 Small Bore Piping Design Methods 2-1 2.1.1 Small Piping Procedure 2-2 2.1.2 PIPSYS Analysis 2-3 2.1.3 Small Tap Line Details 2-3 2.1.4 Prequency of Design Method Use 2-4 2.2 Assurance that Small Bore Piping Was 2-4 Designed by Engineering Methods 2.2.1 Review of Piping Line List 2-5 Against P&ID's 2.2.2 deview of Design Calculation 2-5 Comp 1.eteness 2.3 Small Piping Procedure Review 2-6 2.3.1 Review of the Application of 2-6 the Small Piping Procedure 2.3.2 Review of the Small Piping 2-7 Procedure 2.4 PIPSYS and Small Tap Line Review 2-7 2.5 Phase I Summary 2-8 3
Phase II - Seismic Interaction Analysis 3-1 3.1 Description cf the Interaction 3-1 Analysis Program 3.2 NSED Review of the Interaction 3-3 Analysis Program 11
Table of Contents (continued)
Section Page 3.3 Supplemental Walkdowns for the Residual 3-5 Heat Removal, Shutdown Service Water, Diesel Generator and Diesel Fuel Oil Systems 3.3.1 NSED Walkdown Description 3-5 3.3.2 Burns & Roe Special Walkdowns, 3-6 Description _and Schedule 3.4 Summary for Phase II 3-6 4
Phase III - Equipment Evaluation to the Revised 4-1 Response Spectra 4.1 Background for the ESAP Equipment 4-1 i
Evaluation 4.2 Equipment Qualification to the Design 4-1 Response Spectra.
4.2.1 Qualification by Test 4-1 4.2.2 Qualification by Analysis 4-2 1
4.3 Equipment Evaluation to the Revised Response 4-2 Spectra 4.3.1 Evaluation of Qualification Test 4-2 Results 4.3.2 Evaluation of Qualification Analysis 4-2 Results j
4.3.3 Equipment Chosen for Evaluation to 4-3 the Revised Response Spectra 4.3.4 Evaluation Results 4-3 4.4 Summary for Phase III 4-4 5
Conclusions 5-1 6
References 6-1 Appendix A:
NSED Walkdowns and Problem A-1 Resolution Report Appendix B:
Equipment Evaluation Results B-1 l
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SECTION 1 INTRODUCTX0M The Equipment Seismic Assessment Program (ESAP) was developed in response to a request made as a result of the Advisory Committee on Reactor Safeguards (ACRS) meeting on the Clinton Power Station (CPS). The ACRS stated that " specific attention should be given to the seismic capabilities of the emergency AC Power supplies, the DC Power supplies, and small components such as actuators and instru-ment lines that are part of the decay heat removal system." A proposed program to investigate the seismic capabilities of this equipment was prepared and submitted to the NRC in the Illinois Power Company letter U-0484 dated May 19, 1982 (Reference 1).
The NRC had minor comments concerning the program but generally felt it responded to the ACRS concerns (Reference 2).
This report discusses how the program was implemented at CPS, how the IPC program objectives were met, the types of problems identified by the program and how the problems were corrected.
1.1 Program Summary The ESAP was implemented in 3 phases; Phase I - Piping Design, Phase II - Interaction Analysis and Phase III - Equipment Evaluation.
1.1.1 Phase I - Piping Design Phase I was a review of small bore piping (2" diameter and under) designed by the CPS Architect Engineer, Sargent & Lundy (S&L) to ensure that this piping was designed to withstand Safe Shutdown Earthquake (SSE) acceleration loadings. This review was accomplished by ensuring that all safety related small bore lines in the Residual Heat Removal (RH), Shutdown Service Water (SX), Diesel Generator (DG) and Diesel Fuel Oil (DO) systems have a design calculation performed for them, and by investigating each of these design methods to determine whether they were adequate.
Phrse I was initiated with a review of the Piping Line Lists for the DG, DO, RH and SX systems against the Piping and Instrumentation Diagrams to ensure that the Piping Line Lists were complete.
After these lists were confirmed to be complete, a review of S&L's small bore piping support and location calculations was performed to verify that each safety related small bore line in these systems had a design calculation performed for it.
After this, a representative sample of the calculations prepared by S&L using their Small Piping Procedure was reviewed to determine whether the procedure had been properly followed. The review found that there were discrepancies in the calculations. This condition was reported to the NRC on September 2, 1982 as a potential reportable deficiency (Reference 3) under 10CFR50.55(e).
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The discrepancies in the small piping procedure calculations have been corrected, and the design program has been revised to prevent recurrence.
IP has monitored S&L's ongoing work in this area, and found it acceptable.
Computer calculations are performed to demonstrate the adequacy of the following:
- 1) small bore piping details such as high point vents and low point drains; 2) small bore piping systems that are too complex to be adequately handled by the small bore piping procedure; and 3) small bore piping that has a significant effect on large bore piping systems. The computer code used for these analyses is PIPSYS which is a NRC and industry accepted code.
PIPSYS is also used for large bore piping design.
IP has monitored the S&L work associated with coding the input data for PIPSYS during reviews of the large bore piping design (the method is the same for small bore piping).
S&L has performed this work acceptably.
Therefore, small bore piping verified by PIPSYS is considered to be adequate to withstand design basis seismic events.
IP will continue to monitor this effort in the future.
Phase I is described in detail in Section 2.
1.1.2 Phase II - Interaction Analysis Phase II consisted of walkdowns of selected safety related equipment in the decay heat removal and emergency power supply systems to identify equipment configurations that may be susceptible to damage from seismic events. Two different types of walkdowns are being utilized. One is conducted by Burns & Roe (B&R) personnel who have been trained in performing walkdowns for the CPS Interaction Analysis (IA) Program. The IA Program is used to identify, evaluate, and correct (if necessary) equipment configurations where the potential exists for equipment to interact during a seismic The B&R personnel will use the same criteria for identifying event.
potential interactions during the two special ESAP walkdowns as they did for the IA Program. The second type of walkdown was conducted by personnel from Illinois Power's Nuclear Station Engineering Department (NSED).
These walkdowns provided an independent review of the as-installed equipment. No specific set of investigation criteria were used during these walkdowns except engineering judgement. The personnel conducting these walkdowns looked for poorly restrained equipment, potential interactions, or any other anomalies.
The NSED walkdowns have been performed.
Problems noted during these walkdowns were referred to S&L for their evaluation and to initiate corrective action, if necessary.
The S&L evaluations and proposed corrective actions have been provided in this report.
S&L was able to determine in most cases that the existing configuration was adequate as is.
For the remainder, they initiated design changes that would correct the problem.
The two special ESAP walkdowns utilizing B&R Interaction Analysis personnel for identifying problems have been scheduled. The first 1-2
i valkdown will occur when the systems involved are turned over to IP for startup testing. The second walkdown will occur when construction is completed in thd areas surrounding the equipment.
Phase II is described in detail in Section 3.
l.1.3 Phase III - Equipment Evaluation i
Phase III consisted of evaluations of equipment stress levels for loadings based upon the Revised Response Spectra.
The revised spectra were developed using the elastic half space approach for soil structure interaction analysis which is different than the finite-element approach that was used in developing the design basis qualification spectra. The purpose of the evaluations was to demonstrate the ability of selected equipment important to decay heat removal and emergency power supply to survive an earthquake of the type described by the revised response spectra. Two types of l
stress evaluation were utilized for Phase III. For equipment that t
was seismically qualified by test, the Test Response Spectra were shown to envelope the Revised Response Spectra in the frequency range of interest for the equipment.
For equipment that was being i
seismically qualified by analysis, the acceleration values used in the analysis were shown to exceed the corresponding acceleration values from the Revised Response Spectra, or the component's critical stresses were calculated and shown to be within applicable code allowable limits.
The Phase III evaluations are nearly complete and are being performed concurrently with the ongoing CPS Equipment Seismic i
Qualification Program. The stress evaluations are scheduled to be complete in late 1985.
For all the evaluations completed to date, the equipment has been determined to be able to survive an earthquake as defined by the Revised' Response Spectra.
Phase III is discussed in Section 4.
1.2 Achievement of the ESAP Objectives 1
The three phases of the ESAP examine the seismic aspects of the systems important to decay heat removal and emergency power. Phases I and III verify that the designs of piping and equipment are adequate for seismic acceleration loadings. The Phase II walkdowns provide confidence in the ability of the equipment to survive seismic interactions and earthquake acceleration loadings from examining the as-built equipment configurations. As part of the ESAP, any concerns identified are evaluated and corrective action is initiated as necessary.
The following sections describe on an item by item basis how the objectives of the program proposed in letter U-0484 (Reference 1) are met by the existing program:
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ESAP Analytical Aspect Proposed Program -
"The seismic assessment program will consist of a combination analytical calculations and field inspection.
The analytical aspect of the program will compare the component stress level resulting from the Revised Seismic Response spectra (See Appendix A) to the maximum allowable stress level for which the components are qualified. The specific components which will be included in this analysis are listed in Parts IV and V.
An example of how the results of this analysis will be tabulated is provided in Appendix B."
Existing Program -
The analytical aspect of the program is accomplished in Phase III. The list of equipment being evaluated generally encompasses the equipment list proposed in letter U-0484. A comparison of these lists is provided in Section 4.
ESAP Field Inspection Aspect Proposed Program -
"The field inspection aspect of the program will consist of field inspections of limited safety-related components and their surroundings for the following systems:
A.
Typical components for one equipment train in the shutdown service water system.
B.
Typical components for one equipment train of the residual 1
heat removal system-shutdown cooling mode.
C.
Typical components for one division of onsite AC and DC Power Systems."
Existing Program -
The walkdowns conducted for Phase II examined all trains of these systems. The diesel generators, however, were not examined during the walkdowns.
Instead, credit is being taken for the equipment qualification program for the diesels.
ESAP Field Inspection Purpose Proposed Program -
"The purpose of the field inspection will be as follows:
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t A.
Review major component configuration to assure that small lines supplied with the equipment are adequately supported.
i B.
Review major component surroundings to determine if nearby components are fabricated, located or restrained in such a manner so as not to pose a hazard during a seismic ever.t.
C.
General review of small' piping (2" and under) for the previously listed mechanical systems to assure the following potential problem areas are adequately supported:
i)
Lumped masses tttached to lines with center of gravity locatel at distances greater than one foot from the lines, ii)
Inadequately supported lines, iii) Disproportionately large lumped masses in lines, f
iv) Any other anomalies."
Existing Program -
Items A,B and C are accomplished in a qualitative manner by Phase II.
In addition the piping design review performed for Phase I ensures that items C-1, C-11 and C-iii are not problems with the CPS piping design.
4 ESAP Field Inspection Documentation Proposed Program -
" Inspection results will be documented on the form provided in j
Appendix C and evaluated as appropriate.
Problems found with an equipment configuration will then be investigated on similar configurations of redundant divisions or equipment trains."
i Existing Program -
The forms provided in Appendix C were not used to record the inspection results, however in most cases the configurations of concern were photographed. Sketches were made to record equip-ment configurations with dimensions where deemed appropriate.
t' Because all equipment trains were examined, similar configura-tions on redundant divisions require no special investigation.
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ESAP Field Inspection of Surroundings l
Proposed Program -
"That portion of the Field Inspection Program which will address the review of component surroundings will be carried out as a part of the systems interaction program currently underway.
Existing Program -
The walkdowns that Burns & Roe Incorporated perform for Phase II will be conducted under NSED Instruction ME-2, " Interaction
{
Analysis".
In general, the existing program accomplishes all the proposed program objectives, i
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SECTION 2 PHASE ? - PIPING DESIGN Phase I of the ESAP was a piping design review to determine the adequacy of the seismic design of safety-related small bore piping designed by the Clinton Power Station Architect Engineer (AE). Small bore piping is piping I
or tubing of nominal diameter two inches or less. This category includes instrument lines.
The Phase I review was limited to those safety related j
systems routed and supported by the AE.
The AE is responsible for designing plant piping systems not supplied as ancillary parts of plant components. The factory assembled' plant component piping is seismically qualified by the vendor and verified during part of the equipment qualification program discussed in Phase III.
The Phase I piping design review consisted of two parts. First, i
all safety-related small bore piping of the plant decay heat removal and emergency power systems was reviewed to determine if its seismic capabilities had received an engineering verification. Second, the methods 1
of engineering verification and their application were examined for 4
adequacy.
2.1 Small Bore Piping Design Methods The Phase I review was concerned with the seismic aspects of piping designed by the AE, Sargent & Lundy (S&L). Thus it covered the structural j
- mechanical aspects of piping design; i.e., the methods of locating piping supports.
The purpose of the small bore piping load and location calculations made by the Clinton Power Station AE is to verify that piping complies with the stress limitations delineated in the ASME Boiler & Pressure Vessel Code I
(1974 and 1974 Summer Addenda). Three approaches were used by S&L for small bore safety related pipe seismic design:,
j The Small Piping Procedure, a simplified load and location l
calculation procedure.
PIPSYS Analysis, a computer analysis for code compliance used primarily for large bora piping design.
Small Tap Line Design Guidelines, used to design vents, drains, and instrument taps connecting to larger lines for flow induced vibration loads.
To make the design process more manageable, piping systems (both large and i
small bore) were broken down into smaller portions called subsystems.
Subsystem boundaries occur at equipment nozzles, pipe anchors (an anchor is j
defined as a point where the piping is restrained in all orthogonal directions and against torsion and bending) or at connections with large bore piping if the polar moment of inertia of the smaller line is less than 1/7 of that of the larger line.
In cases where the smaller line has a polar moment of inertia greater than this, the smaller line is considered j
part of the larger line subsystem.
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1 2.1.1 Small Piping Procedure The Small Piping Procedure for Restraint of Seismically Qualified Small Piping (S&L procedure EMD-015666) provides a methodology for restraining and thermally designing Seismic Category I small bore piping and instrumentation systems. The procedure is applicable for ASME Boiler &
j Pressure Vessel Code,Section III, Class 2 and 3 small bore piping, ANSI B31.1 small bore piping and cubing sizes as noted in the applicable sections of the procedure, The procedure is not applicable for transient loadings and break exclusion areas; these are evaluated separately by the AE.
The fundamental concepts of the procedure require the piping be adequately restrained in three mutually perpendicular directions for the effect of dynamic loads. The procedure specifies the vertical, lateral and axial restraint locations for each applicable leg of piping. Also, specific span tables based on building response spectra, process fluid medium, insulation and pipe j
schedule are given to determine the spacing between adjacent supports.
4 Piping systems with a maximum operating temperature of 200*F or greater and/or significant anchor movements receive additional design rules to 1
ensure the system has adequate flexibility. Conformance with the flexibility check rules ensures that piping thermal stresses will be within code allowables and thermal loads will be included in the restraint design.
The procedure was written to account for the most limiting piping configurations expected and was verified by means of test cases which provide the greatest possible challenge to the procedure (the test cases were analyzed using a finite element computer program and compared to the i
ASME Code allowables). Therefore, application of the procedure to most piping configurations yields highly conservative design.
The AE design personnel located at the construction site utilize the small piping procedure for the initial design and subsequent acceptance of construction requested field changes. The use of the Small Piping Procedure is documented with a calculation covering each unique subsystem.
The AE's Quality Assurance Procedures and specific project instructions require each calculation be prepared and reviewed in detail. Any deviations from the small bore procedure requirements are to be reconciled by the AE's Pipe Stress Analysis Engineers. Each calculation has:
dimensioned drawings showing the pipe routing and support locations a list of the line numbers in the subsystem, their size, thickness, maximum temp 6rature, weight, insulation weight, valve weights, standard recommended spans, and standard estimated loads i
a flexibility check for the thermal expansion of each leg of pipe and the thermal movements of the pipe anchors (e.g. a large-bore line) i a span check for each span a load estimation for each support (used as a basis for support design) 2-2
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2.1.2 PIPSYS Analysis The Clinton Power Station AE has developed a piping and structural analysis computer program, PIPSYS, to verify system compliance with the AS:1E Boiler
& Pressure Vessel Code. This program has been verified and benchmarked in accordance with nuclear safety-related design standards and is currently in use for all of the AE's projects. The computer code is the primary means f verifying the design of all safety-related piping systems; either directly for large bore systems and specific small bore lines, or indirectly for small bore systeme via verification of the Small Piping Procedure.
Small bore piping is analyzed with PIPSYS as part of large bore subsystems when the small bore lines have a significant loading effect on the connected large bore lines. Small cap line details are also verified by PIPSYS and are discussed in Section 2.1.3.
Information concerning the piping system configuration, location, orientation, structural characteristics, pressure, temperature, building response spectra loading and loading combinations is input to the PIPSYS computer code. PIPSYS uses finite element analysis methods to determine the behavior of the piping system. From this PIPSYS generates a stress report for the piping system. From the stress report the adequacy of the piping can be determined.
2.1.3 Small Tap Line Details Safety related small tap lines are designed primarily for prevention of fatigue failure caused by flow induced vibration in the header to which they are connected. To do this, the tap lines are subjected to a generic PIPSYS Analysis to qualify them for bounding vibration loads which also qualifies them for the safe shutdown. earthquake loadings. Examples of small tap lines are high point vents, low point drains, and instrumentation taps. Typically a small tap consists of a 3/4 inch line connected to a large bore pipe terminating with two hand operated globe valves, in series.
Large bore valve bypass and other similar small lines are not small tap lines but are designed and analyzed in the same manner. For the purposes of the ESAP, they are classified and discussed with the small tap lines.
Small tap lines are subject to fatigue failure caused by flow induced vibration in the large bore line to which they are attached. An effective way to protect tap lines is to design them to have a high fundamental frequency.
For this reason tap lines are designed to be short and stiff and all valves and other large masses are supported. The supports for small tap lines are tightly secured to the large bore line to allow no movement relative to the large bore line, because for the frequencies and geometries involved even low amplitude vibrations could be damaging. To obtain high stiffness, differential thermal expansion is minimized by routing the tap line in parallel with the header pipe, making it short and supporting it from the header. The design requirements for resisting fatigue failure caused by flow induced vibration are generally more restrictive than the seismic requirements.
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Small tap lines are installed primarily from standard design deta,ils.
In this way one design can be used for the installation of =any similar tap lines. For unusual configurations where the standard details cannot be applied, individual designs have been developed.
Af ter each design was completed in accordance with the methods for increasing the fundamental frequency, it was analysed for compliance with the ASME Boiler & Pressure Vessel Code. The AE uses its finite element ecmputer program, N0 HEAT, to determine the axial temperature distribution in the tap lines (tap lines are under no flow conditions, except when being flushed or when system operation requires manual venting or draining). This data is input to the PIPSYS analysis. The PIPSYS analysis establishes ASME code compliance, as described earlier, including system response to cycled thermal, seismic, safety relief valve and Loss of Coolant Accident loadings.
2.1.4 Frequency of Design Method Use The schematic Piping & Instrument Diagrams and Control and Instrument Drawings identify unique line numbers. A line is defined as a length of piping or tubing of one size, code class and AE Piping Design Table.
There are nearly 2000 safety related small bore lines in the station. Approxi-mately 50% of them were verified to be seismically qualified by the use of the Small Piping Procedure. Approximately 45% were verified by the analysis of small tap line details. The remaining 5% were verified concurrently with large bore subsystems by the PIPSYS program.
2.2 Assurance that Small Bore Piping Was Designed by Engineering Methods In nuclear stations licensed prior to the accident at the Three Mile Island Station it was an industry practice to field route safety related small bore piping. For this reason it was decided that the ESAP shoul'd include, as a minimum, a review of all safety related small bore piping in the emergency power and decay heat removal systems to determine if it was designed by engineering methods and to determine what method has been applied to each line.
The following Clinton Power Station systems contain all of the safety related small bore lines designed by the AE that are considered to be in the emergency power and decay heat removal systems:
The Diesel-Generator System (DG):
The standby emergency power source for the three divisions of safety-related AC power; consisting of the three diesel-generator sets and their air start equipment.
The Diesel Fuel Oil System (DO):
The fuel oil storage and supply system for the standby emergency diesel-generators; consisting of fuel oil storage tanks, day tanks, piping, and pumps except for that on the diesel-generator skids.
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The Residual Heat Removal System (RH):
A system which removes decay heat from the reactor, cools the suppression pool, provides containment spray, and provides low pressure coolant injection into the vessel; including the three low pressure coolant injection pumps and the two heat exchangers which allow cooling of the reactor coolant and suppression pool water.
The Shutdown Service Water System (SX):
A system which provides lake water for cooling safety-related plant equipment; most notably the diesel-generators, the switchgear heat removal chillers, and the reactor coolant in the RH heat exchangers.
2.2.1 Review of Piping Line List Against P& ids A list of all small bore lines designed by the AE in the DG, D0, RH, and SX systems was developed. This list was based upon the AE's piping line list which provides the size, Piping Design Table, ASME code classification and design / operating conditions for each line in the plant. The piping line list was verified by IPC to be complete and accurate for all safety and non-safety related small bore lines in the subject systems by comparing it to the system Piping & Instrument Diagrams (P& ids) and Control & Instrument Diagramt. These two types of drawings are schematic process and instrument line drawings which identify all lines, valves and equipment by number and indicate their mechanical function.
2.2.2 Review of Design Calculation Completeness Once all lines were accounted for, IP determined the method of seismic design verification used for each line.
In this way IP assured that each line had been seismically design verified through accepted engineering methods. Most of the safety related lines could be easily classified as having been designed by either the Small Piping Procedure, the Small Tap Line Detail method, or the Large Bore method of PIPSYS analysis. When a line was found to be in a Small Piping Procedure calculation it was verified that every span of the line was accounted for in the calculation (in a few cases one part of a line was verified in one calculation and the other part was verified in a second calculation). When a line was found to be in a large bore PIPSYS calculation, it was verified to be completely covered by examining the analytical drawing from which the PIPSYS input was made.
In a few cases one part of a line fell in a large bore PIPSYS calculation and the other part fell in a Small Piping Procedure calculation. The remaining lines were checked to determine if they had configurations for which the small tap line standard details could apply or if they were shown on an individual small tap line detail.
In some cases lines were verified to have adequate seismic designs by 1
methods that were slightly different than the three major methods such as a j
number of lines which were not of the small tap configuration but required the same type of design considerations.
These were usually small warming lines that bypassed large valves.
The design methods used for these lines were sufficiently similar to the methods of small tap line design that they j
were classified as small tap lines.
There were also several small bore i
lines of the DG, DO, RH, and SX systems which were either not scfety related or were temporary lines for startup testing instrumentation and 1
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thus required no verification of ASME Code compliance (the status of lines as temporary das indicated on the system P&ID).
The review found that all lines that required a calculation had either been, or were in the process of being verified with a calculation.
In all, 753 lines were accounted for. These included all of the small bore lines of the DG, DO, RH, and SX systems. Of these, roughly 43% were verified by means of the Small Piping Procedure, 39% by the Small Tap Line methods, and 5% by Large Bore PIPSYS analysis.
In addition, roughly 10% of the lines were non-safety related, 2% were safety related with verification in progress and 1% were temporary.
2.3 Small Piping Procedure Review A review of the Small Piping Procedure design program was performed.
The scope of this review included:
i a review of Small Piping Procedure calculations to determine the quality of the procedure's application.
a review of the adequacy of the Small Piping Procedure itself.
The purpose of the review was to determine if the safety related piping
-I qualified by the Small Piping Procedure was being adequately verified to be in compliance with the reauirements of the ASME Boiler & Pressure Vessel Code (which includes seismic design).
2.3.1ReviewoftheApplicationoftheSmalkPipingProcedure A representative sample of the Small Piping Procedure calculations was reviewed by IP for ESAP. This technical review identified discrepancies in the sample. As a result, IP reported this condition to the NRC as a potentially reportable quality problem in accordance with 10CFR50.55(e)
(potential deficiency 82-09).
The AE responded to this condition by performing a. detailed technical review of all calculations utilizing the small bore procedure. The AE's review identified numerous discrepancies in the calculations which exceeded the tolerance allowed by the small bore procedure. However, from a safety l
perspective, all the small bore piping calculations and support designs were verified as not jeopardizing the pressure integrity of the reacter vessel, the safe shutdown of the plant, or the release of offsite dose. As j
a result of the review, the small bore procedure and program were modified and the calet!1stions vera revised to be within the intent of the small bore procedure and to restore the inherent conservatism associated with the simplified hand calculations.
IP concludes that the AE's review and change adequately resolved the potential concerns associated with the small bore procedure. IP also agreed with the AE's position that the condition did not constitute a reportable item. To ensure the calculations continue to be of acceptable quality, IP has established a program to review a minimum of 10% of the small bore procedure calculations. Based upon these reviews IP is confident in the quality of the small bore procedure design.
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2.3.2 Review of the Small Piping Procedure 1
i The Small Piping Procedure was reviewed for clarity and adequacy by both the AE and IP as a part of the potential deficiency 82-09 investigation.
As a result of these reviews the following revisions were made to the procedure:
Discrepancies in the Small Bore Piping Procedure were eliminated.
The validation and references for the Small Bore Piping Procedure were expanded and summarized to ensure proper historical documentation.
The scope of the procedure was clarified to make the applicability and limitations of the procedure more clear.
With the improvements made, IP and S&L are satisfied that the Small Piping Procedure is a valid, consistent and properly verified design methodology.
Furthermore, because S&L reviewed and revised the applicable subsystem calculations affected by the modifications to the Small Piping Procedure, IP is confident that all Small Piping Procedure calculations have been performed using sound engineering methods.
2.4 PIPSYS and Small Tap Line Review Safety related small bore lines analysed with large bore piping subsystems and small tap line details are both verified to be in compliance with the ASME Boiler & Pressure Vessel Code by means of PIPSYS. Since PIPSYS has -
been verified and benchmarked in accordance with nuclear safety related design standards and is in extensive use, the ESAP program did not include a review of the PIPSYS computer software.
IP-NSED has established a program to review and monitor the AE's large bore safety related piping design program.
The review program concerned itself not only with piping stress analysis, but with the specification of piping material and schedule, pipe support calculations, penetration design, and the effort to account for as-built piping configurations. One large bore subsystem calculation was reviewed each month. Twenty safety related subsystems have been reviewed by IP-NSED (as of August 31, 1985) and no evidence of a possible quality problem exists.
The program has now been superceded by the IP ASME N certificate holder program, in which all stress analyses are reviewed.
Although small cap lines are designed to meet the more challenging demands of flow induced header vibrations, they too are analysed for their capability to withstand the safe shutdown earthquake by means of PIPSYS.
Because the coding techniques are similar to those for large bore subsystems and the details are less numerous and far simpler than the large bore subsystems, the large bore review program of IP-NSED would detect problems in the qualification of small tap lines if they existed. The design guidelines used for small-fap lines require the small tap line designer to make the tap line far stronger than is necessary for withstanding the safe shutdown earthquake accelerations.
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2.5 Phase I Summary Safety related small bore piping designed by the AE is designed by one of three methods:
Small Piping Procedure PIPSYS Analysis Small Tap Line Details IP verified that all of the saall bore lines of the DG, DO, RH, and SX systems (emergency power and decay heat removal systems) with the exceptions identified in section 2.2.2 above, are designed by one of the above three engineering methods.
Each of the design methods and their application were reviewed.
Discrepancies were found with the application and quality of the Small Piping Procedure and a potential deficiency investigation was initiated by IP.
As a result, all of the Small Piping Procedure subsystem calculations and the Small Piping Procedure itself have been reviewed and corrected.
As a result of the ESAP, IP is confident that the safety related small bore piping at the Clinton Power Station designed by the AE is capable of withstanding the loadings of the safe shutdown earthquake. Not only have the designs been subject to extensive review, but IP has established a program to monitor and control the quality of the AE's work.
2-8 e
-.~
SECTION 3 PHASE II - SEISMIC INTERACTION ANALYSIS Phase II of the Equipment Seismic Assessment Program (ESAP) is 1
primarily concerned with potential seismic interactions of safety-related components during an earthquake. Using the existing interaction investigation program supplemented by ESAP walkdowns.
Phase II improves confidence in the seismic design in this area.
The existing interaction program is the Interaction Analysis (IA)
Program (all areas of the plant where safety-related equipment exists are walked down to discover potential seismic interactions) and IP NSED's formal surveillence program for monitoring the quality of the Interaction Analysis Program. To provide added assurance that the area of seismic interaction is adequately addressed, IP has conducted supplemental walkdowns and has requested the IA valkdown team to conduct special walkdowns to search for seismic interaction problems for key components of the decay heat removal and emergency power supply systems.
Detailed descriptions of these programs are provided in the following subsections.
3.1 Description of the Interaction Analysis Program j
A program has been established in which all areas of the plant, where safety related equipment exists, will be examined in the as-built condition for potential seismic interactions of safety-related compo,nents. The purpose of the program is to eliminate the possibility that safety related equipment could be damaged by mechanical interaction during a design basis earthquake. The program is defined by NSED Instruction ME-2 " Interaction Analysis Program". This document delineates the personnel and methods used for identifying Potential Interactions and for writing and dispositioning Potential Interaction Reports (PIRs).
The IA program is the final step in the seismic design verifica-tion of safety-related equipment. Safety-related piping, conduit, cable trays and HVAC ductwork have all been designed to withstand self imposed seismic loadings (i.e. acceleration loadings).
For safety-related equipment such as pumps, heat exchangers and motor control centers, the equipment qualification program (as described in Chapter 3 of the FSAR) provides assurance that the equipment can survive seismic acceleration loadings. The IA program is intended to prevent damage to safety-related components caused by seismic interactions with any other component.
The program is carried out via plant walkdowns rather than drawing reviews because of the difficulty of finding all possible j
interactions from plan and section drawings, and because the
~
as-built dimensions and clearances may be different than those shown on design drawings because of construction tolerances. The walkdowns are conducted on an area basis with the areas being defined by the AE's composite drawing boundaries. A given area is walked down a number of times during the construction of the f
3-1 l
---L-,
g-.i g
,.--,n--
4 g.,g..
,s... - - - -.
plant, once every 3 months for congested areas, every 6 months for noncongested areas. Two or more surveillance teams are used to conduct the walkdowns. Each surveillance team normally consists of two Burns & Roe Incorporated field walkdown personnel.
Potential interactions are identified on the basis of clearance criteria.
If a safety-related component is located nearer to another component than the clearance criteria allow, a potential interaction exists. The clearance criteria used are the sums of two partial clearance criteria, one associated with the safety related component, the other associated with the interacting component (either safety or nonsafety-related). When a potential interaction is identified, a member of the surveillance team completes a PIR and signs as the preparer. The PIR includes a brief description of the potential interaction and its location.
A sketch is attached to the PIR showing the equipment involved in the potential interaction along with important dimensions. An independent reviewer reviews the PIR and the attached sketch and signs as the reviewer if he finds them acceptable.
Each PIR is tracked in two separate log books. One of the logs is used to record inspection activities and PIRs by area. This ensures that all areas have adequate inspection coverage. The second log is used to record the disposition status of the PIRs.
This ensures that each PIR is being addressed.
PIRs can be dispositioned "Use-As-Is" at the construction site by a Burns & Roe Interaction Engineer, using one of the following methods:
1.
Use of refined clearance data, such as actual dynamic and thermal displacements at the exact location of the potential interaction. In the case of piping, such dpta can be obtained from the stress analysis reports at the site.
2.
Use of a calculation sheet to reduce clearance criteria for nonsafety related piping.
3.
Engineering rationale or judgement based on installed conditions can be used. This would justify using reduced criteria or eliminate the potential for interaction.
Examples of this approach include:
a)
Reduced partial clearance for components such as ductwork, cable trays and safety related piping, when the interaction occurs closer to lateral restraints rather than at midspan where the maximum values occur.
b)
Presently all partial clearance criteria are taken as positive or negative. However, for some components, based on installed conditions the actual direction can be logically concluded to occur in only one direction.
3-2
When such calculations or rationale are utilized to disposition PIRs, then all supporting calculations and justifications must be formally completed, signed, checked and filed with the PIR.
PIRs dispositioned by an Interaction Engineer using one of the above three methods are reviewed by another Interaction Engineer and approved at the construction site. When the PIR cannot be field dispositioned "Use-As-Is", the PIR is sent to S&L's home office for dispositioning.
There the Mechanical Engineer assigns the PIR to the design division responsible for the safety-related component (s) involved in the PIR.
For PIRs involving safety-related components for which two divisions are responsible, both divisions will work together to disposition the PIR.
The responsible divisions may perform more detailed displacement calculations to show that the equipment involved in the PIR will not collide.
Or alternately, an impact analysis can be performed to show that safety-related equipment will not sustain enough damage from collisions with other equipment to render it unable to perform its function. PIRs resolved by either of these methods are documented by calculations prepared, reviewed and approved in accordance with the AE's quality assurance procedures.
PIRs not resolved by the previous methods will be dispositioned by redesign.
In this case, the divisions responsible for the compo-nents involved in the interaction agree upon which components should be redesigned to disposition the PIR. The division responsible for the redesign summarizes the recommended redesign and writes the Engineering Change Notice or the Field Engineering Change Notice identification number on the PIR. The surveillance walkdown team verifies cae adequacy of the redesign by walkdowns after the redesign has been installed.
If the redesign results in a new violation of the clearance criteria, a new PIR is initiated.
Originally, IP-NSED was directly responsible for performing IA walkdowns and resolving PIRs. On January 4, 1982 Sargent & Lundy assumed responsibility for both of these tasks. On September 10, 1984 Burns & Roe Incorporated personnel reporting to NSED assumed responsibility for the walkdowns and field resolutions.
Sargent &
Lundy retained responsibility for resolving those PIRs that cannot be dispositioned in the field.
As of August 31, 1985, 8331 PIRs were identified that require dispositioning. The status of these is as follows:
Disposition Complete:
Total 6615 Use-As-Is 6611 Redesign or Rework 4
PIRs Requiring Dispositioning:
Total 1716 (By Field or S&L Personnel) 3.2 NSED Review of the Interaction Analysis Program Illinois Power Company's Nuclear Station Engineering Department (NSED) is responsible for performing surveillances of the 3-3
l Interaction Analysis Program. Direction for performing these surveillances is given in NSED Procedure D.5, Interaction Analysis Program. The purpose of the surveillances is to monitor the progress and quality of the IA Progran.
To accomplish the first objective, monitoring progress in identifying and dispositioning PIRs, NSED reviews Burns & Roe and Sargent & Lundy system interaction analysis program reports which include summaries of the following:
The walkdowne which have taken place since the last report.
The number of potential interac'tions identified during the walkdowns.
The status of PIRs which have previously been filed.
l To accomplish the second objective, verifying the quality of the work, NSED reviews the field work of the Interaction Analysis Group at least once each 60 days. The NSED engineer responsible for the field review performs the following tasks during the review:
Participates in an area walkdown with Burns & Roe field personnel to ensure that potential interactions are being identified.
Conducts a mini-walkdown (independent of the walkdown performed with Burns & Roe field personnel) to find potential interactions, and verifies that Burns & Roe has written PIRs for any potential interactions found.
Reviews the status of unresolved PIRs with the Burns & Roe Interaction Group Supervisor to ensure that adequate action is being taken.
A report to IP Management is prepared by the NSED engineer documenting'the results of the field review. Illinois Power Company management may increase the frequency of the field reviews or may request that changes be made to the IA program based on the findings in this report.
Few problems have been noted during the surveillances, and because of the multiple walkdowns that are conducted in each area, it.is likely that all Potential Interactions will be detected. Overall, Illinois Power Company's surveillance plan has found that the Interaction Analysis Program has been successful in accomplishing its objectives.
3-4
3.3 Supplemental Walkdowns for the Residual Heat Removal, Shutdown Service Water, Diesel Generator and Diesel Fuel Oil Systems.
In addition to the Interaction Analysis field walkdowns conducted by Burns & Roe, special NSED and Burns & Roe walkdowns have been scheduled for the Equipment Seismic Assessment Program. These additional walkdowns provide added assurance that Decay Heat Removal and Emergency Power Systems will be able to function af ter a seismic event.
3.3.1 NSED Walkdown Description The NSED Technical Assessment Section, assisted by an S&L equipment qualification engineer, conducted special walkdowns for the Equipment Seismic Assessment Program. These walkdowns were intended to provide an overview of the seismic design for the decay heat removal and emergency power systems. The walkdowns addressed a list of equipment determined to be important to Decay Heat Removal or Emergency Power Supply (see table 3-1).
When the walkdowns were conducted, attention was not only given to the equipment on the list but also to the area surrounding the equipment.
The walkdown team looked for the following types of problems during the walkdowns:
Possible seismic interactions of piping, valves, conduit, cable trays, hangers and other equipment.
Poorly supported piping, valves and conduit that have the potential to be damaged because of seismic acceleration loadings.
Equipment that has the potential to be damaged because of seismic acceleration loadings.
Obvious non-seismic problems such as equipment interferences and damaged or improperly assembled equipment.
Since these walkdowns were conducted early in the program (1982-1983), problems were identified on the basis of engineering judgement, rather than on the basis of a strict set of criteria.
Each problem noted was evaluated after the walkdowns to determine its significance and to determine what corrective action, if any, would need to be taken.
The results of these walkdowns and the corrective actions taken for problems noted are reported in Appendix A.
The equipment deceribed in this appendix are categorized by area and include all the equipment listed in Table 3-1.
3-5
The following observations can be made concerning the NSED walkdowns:
Typical concerns found were potential interactions, inadequately restrained piping and valvec, and equipment interferences.
The construction in many of the areas covered by the walkdown was incomplete (some so much so that no review was performed). However, a large enough sample of the equipment was walked down to allow a good overview of the seismic design for these systems.
Many of the problems identified in the early walkdowns were noted to have been corrected without any referral by NSED being necessary (i.e. the existing programs for finding and correcting problems of this nature were working effectively).
IP believes that all problems found by the walkdown team would eventually have been corrected through existing programs.
For those concerns referred to Sargent & Lundy, S&L deter-mined for most that the existing configuration was adequate as is.
For the remainder, they proposed and.then initiated design changes that would correct the problem.
Because of the NSED walkdowns, confidence in the seismic design of the equipment listed in table 3-1 has been increased.
3.3.2 Burns & Roe Special Walkdowns, Description and Schedule Burns & Roe Incorporated, under the guidance of NSED Instruction ME-2, will perform Special Interaction Analysis Walkdowns for the ESAP. The criteria they will use are the same as those used for the IA walkdowns (see section 3.1).
The walkdowns will address interactions concerning the equipment listed in Table 3-2.
The Burns & Roe walkdowns have been divided into two groups. The first group has been scheduled for when the system is turned over to IPC Startup and the second group has been scheduled for when construction is completed in the areas.
These walkdowns will serve as the final check that potential seismic interactions have been identified and addressed for those systems and components which provide long term decay heat removal capability or emergency electrical power. Any PIRs identified will be dispositioned and necessary corrections will be made prior to commercial operation.
3.4 Summary for Phase II A program is in place for identifying and correcting potential seismic interactions. This program is defined by NSED Instruction ME-2.
NSED surveillances of the program, which are required by NSED procedure D.5, show that the program has been successful in accomplishing its objectives.
3-6 e
y--
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.-r-
I In addition to the Interaction Analysis Walkdowns normally conducted, NSED Technical Assessment Section conducted walkdowns of the equipment listed in Table 3-1.
Illinois Power has also requested Burns & Roe Incorporated to perform special walkdowns for the equipment listed in Table 3-2 using the criteria in NSED Instruction ME-2.
These walkdowns were requested to increase confidence in the capability of the equipment for decay heat removal and emergency power supply to function after a seismic event.
i r
e 4
/
e i
1 l
I i
3-7
Table 3-1 Significant Equipment Observed During the NSED Equipment Seismic Assessment Program Walkdowns.
A)
Equipment for Decay Heat Removal 1)
Residual Heat Removal.(RHR) System Equipment Number Description IE12C002A RHR Pump 1A lE12C002B RHR Pump LB IE12C002C RHR Pump IC lE21C002 Low Pr' essure Core Spray Water Leg Pump (for LPCS & RHR Pump 1A)
LE12C003 RHR Water Leg Pump (for RHR Pumps 1B and IC) 1E12B001A RHR Heat Exchanger 1A LE12B001B RHR Heat Exchanger IB IE12F014A Shutdown Service Water Inlet Valve to RHR Heat Exchanger lA 1E12F014B Shutdown Service Water Inlet Valve to RHR Heat Exchanger IB IE12F068A Shutdown Service Water Discharge Valve from RHR Eeat Exchanger lA LE12F068B Shutdown Service Water Discharge Valve from RHR Heat Exchanger IB IE12F006A RHR Shutdown Cooling Mode Valve, Loop A LE12F006B RHR Shutdown Cooling Mode Valve Loop B 1E12F008 RHR Shutdown Cooling Mode Outboard Icolation Valve IE12F009 RHR Shutdown Cooling Mode Inboard Isolation Valve LE12F003A RHR Shutdown Cooling Mode Valve, Loop A 1E12F003B RER Shutdown Cooling Mode Valve, Loop B 3-8 l
1
L l
l Table 3-1 (continued) i Equipment Number Description IE12F047A RHR Shutdown Cooling Mode Valve, Loop A lE12F047B RHR Shutdown Cooling Mode Valve, Loop B IE12F048A RHR Low Pressure Core Injection Mode Valve, Loop A lE127048B RHR Low Pressure Core Injection Mode Valve, Loop B IE12F026A RHR Steam Condensing Mode Valve, Loop A lE12F026B RHR Steam Condensing Mode Valve, Loop B LE12F065A RHR Steam Condensing Mode Valve, j
Loop A LE12F065B RHR Steam Condensing Mode Valve, Loop B 2)
Shutdown Service Water (SX) System Equipment Number Description ISXOlPA SX Pump 1A ISX0lPB SX Pump.lB ISX0lPC SX Pump IC 1SX01FA SX Strainer LA ISX01FB SK Strainer IB ISX01FC SK Strainer LC B)
Equipment for Emergency Power I)
DC Power Systems Equipment Number Description IDC01E 125 Volt Battery, Division I IDC02E 125 Volt Battery, Division II lE22-S001D 125 Volt Battery, Division III IDC03E 125 Volt Battery, Division IV 3-9
Table 3-1 (continued)
Equipment Number Description IDC06E Battery Charger, Division I IDC07E Battery Charger, Division 11 lE22-S001E Battery Charger, Division III 1DC08E Battery Charger, Division IV 1DCl3E 125 Volt DC Motor Control Center IA 1DC14E 125 Volt DC Motor Control Center 1B 1DC15E 125 Volt DC Motor Control Center ID 2)
AC Power Systems Equipment Number Description LAP 07E 4160 Volt Switchgear lAl 1AP09E 4160 Volt Switchgear IB1 1APilE 480 Volt Substation lA 1AP12E 480 Volt Substation IB 1AP29E 480 Volt Screenhouse Motor Control Center lA LAP 30E 480 Volt Screenhouse Motor Control Center 1B 1AP31E 480 Volt Screenhouse Motor Control Center IC 1AP60E 480 Volt Diesel Generator Building Motor Control Center IA LAP 61E 480 Volt Diesel Generator Building Motor Control Center IB 1AP72E 480 Volt Auxiliary Building Motor Control Center lAl LAP 73E 480 Volt Auxiliary Building Motor Control Center lA2 1AP74E 480 Volt Auxiliary Building Motor Control Center IA3 3-10
Icble 3-1 (continued)
Equipment Number Description LAP 75E 480 Volt Auxiliary Building Motor Control Center IBl LAP 76E 480 Volt Auxiliary Building Motor Control Center IB2 1AP77E
' 480 Volt Auxiliary Building Motor Control Center IB3 1AP78E 480 Volt Auxiliary Building Motor Control Center ICI lE22-S002 480 Volt Motor Control Center, Division III C)
Diesel Generator System Equipment Number Description 1DG065A Diesel Generator Control Panel, i
Division I 1DG065B Diesel Generator Control Panel, Division II 1DG065C Diesel Generator Control Panel, Division III IDG01TA Diesel Fuel Day Tank, Division I 1DG01TB Diesel Fuel Day Tank, Division II IDG01TC Diesel Fuel Day Tank, Division III f
a I
3-11 9
8
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TABLE 3-2 Equipment List for the Burns & Roe Equipment Seismic Assessment Program Walkdowns A.
Mechanical Equipment 1.
Diesel Cenerator System a.
Local Control Panels b.
Air Start Piping System c.
Fuel Oil Supply to Day Tank 2.
Decay Heat System (RHR) a.
Pump (lE12-C002A,B&C: Aux, Bldg, 707 '-6")
b.
Heat Exchanger (1E12-B001A&B: Aux. Bldg.
707'-6")
c.
Valves (lE12-F003A&B, IE12-F006A&B, IE12-F008, IE12-F009, IE12-F047A&B, IE12-F048A&B; Aux. Bldg. 707'-6")
{
3.
Shutdown Service Water System
~
a.
Strainer (ISX01FA,B&C: Cire. Water Screen House) b.
Pump (ISX0lPA,B&C: Cire. Water Screen House) c.
Instrument Panel (ISX11J,12J & 13J: Cire.
Water Screen House) i d.
Valves (lE12-F014A&B and lE12-F068A&B: Aux.
i Bldg.-707'-6"; ISX001A,B&C, ISX003A B&C and 1SX004A,B&C: Cire. Water Screen House)
B.
Electrical Equipment 1.
AC Power Systems 4160 V Switchgear Assemblies (IAP07E & 09E:
a Aux. Bldg. 781')
b.
480 V Unit Substations (IAPilE & 12E: Aux.
Bldg. 781')
c.
480 V MCC's (LAP 29E, 30E & 31E: Cire. Water Screen House 699'; 1AP72E, 73E, 74E, 75E, 76E
& 77E: Aux. Bldg. 781'; 1AP60E & 61E:
Control Bldg. 737'; 1AP78E & lE22S002:
~
Control Bldg. 781')
)
2.
DC Power Systems I
125 V DC MCC and Distribution Panel (IDCl3E &
a.
14E: Aux. Bldg. 781'; IDC15E: Control Bldg.
781')
b.
Batteries and Racks (IDC01E, 02E, 03E and 1E22-S001D: Aux. Bldg. 781')
c.
Battery Chargers (IDC06E & 07E: Aux. Bldg.
781'; IDC08E and 1E22-S001E. Control Bldg.
781')
3-12
2 t
4 1
SECTION 4 PHASE III - EQUIPMENT EVALUATION TO THE REVISED RESPONSE SPECTRA 3
]
Phase III is an evaluation of the seismic capability of equipment critical to the plant functions of decay heat removal and emergency power supply. The evaluation is based on plant seismic response spectra determined by difierent soil structure interaction analysis methods than t
were used to generate the design basis response spectra. Equipment qualification data were used to evaluate the equipment based on these different spectra. The purpose of evaluating key equipment with these different spectra, which are referred to as the Revised Reponse Spectra, j
is to improve confidence in the capability of the equipment to withstand safe shutdown earthquake acceleration loadings.
I 4.1 Background for the ESAP Equipment Evaluation 1
j As a result of CPS FSAR reviews the NRC staff identified a concern j
regarding the CPS seismic analysis. The staff questioned the use 9:
of finite element methods for soil-structure interaction analysis ll sud required that CPS seismic design bases be reevaluated using I;
}
the elastic half space approach for soil-structure interaction.
ll analysis including a variation of soil properties. Such an 4
{
analysis for CPS was performed using the ground motion parameters that describe the site-specific spectra equivalent to a design basis earthquake of Mb equal to 5.8.
The comparison of reanalysis i
results with the original design basis results showed that the CPS seismic design is adequately conservative. The NRC staff and ACRS accepted this conclusion. However, the ACRS recommended that l
" specific attention be given to the seismic capability of the emergency AC power supplies, the DC power supplies, and small j
components such as actuators and instrument lines that are part of j
the decay heat removal system" (Reference 4).
A program to address this ACRS recommendation was submitted by Illinois Power in Reference 1 and was accepted by the NRC in Reference 2.
4.2 Equipment Qualification to the Design Response Spectra There are two methods used for seismic qualification of plant
{
equipment; qualification by test and qualification by analysis.
Sometimes a combination of these methods is used.
4.2.1 Qualificztion by Test For active components that must perform their safety function j.
during and after a postulated seismic event, a dynamic test method is preferred.
i i
t; Tests are conducted per the requirements of IEEE Standard 344-1975 l'
and S&L Standard MSS-6.2-D, " Standard Specification for Dynamic ll Qualification Criteria for Nuclear Safety-Related Equipment".
i.
Equipment is mounted on a test fixture which simulates the actual i
service mounting. For active components, operability is demonstrated during the test.
The Procurement Specification Response Spectra (or greater) at the point of attachment to the supporting structure define the input motion. The input motion is i
t 4-1 i
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,,y
,.,.,-.,.,,-m--.--.-
-,,,,.a,,
,_v~.,..
, - - ~ - -,
w~m-c, r,y.
applied in the vertical and horizontal directions simultaneously, unless it can be shown that the equipment does not have cross-coupling between the two axes. When a random vibration input is used, the test input motion envelopes the required response spectra. When a single-frequency input is used, the frequency of the test input is the natural frequency of the equipment, or the Zero Period Acceleration (ZPA) frequency if the equipment is shown to be rigid.
4 4.2.2 Qualification by Analysis Equipment can be qualified by analysis by following the guidelines of IEEE Standard 344 and ASME Code requirements or Standard MSS-6.2-D.
Structural integrity is demonstrated by showing the critical stresses in the components are lower than the allowable stresses.
For active components, deflections are calculated and checked with allowable tolerances to ensure operability.
For flexible equipment, a dynamic analysis or a simplified dynamic analysis is performed. A dynamic analysis uses the required response spectra along with component structural parameters to determine actual (or bounding) stresses and deflections throughout the piece of equipment. When using the simplified dynamic analysis, the input acceleration value used is 1.5 times the peak of the applicable floor response spectra if the fundamental natural frequency of the equipment is not known.
For rigid equipment, a static analysis using the zero period acceleration from the applicable floor response spectra can be used.
4.3 Equipment Evaluation to the Revised Response Spectra Reanalysis using the elastic half space approach for soil-struc-ture interaction resulted in the Revised Response Spectra which is different than the response spectra used for equipment qualifica-tion.
In Reference 1, which describes ESAP, Illinois Power committed to evaluating the seismic capability of equipment important to emergency power supply and decay heat removal to the Revised Response Spectra (for the specific equipment list, see Section 4.3.3).
This eauipment has been or is in the process of being seismically qualified to the design basis response spectra as part of the equipment qualification program. The equipment qualification data are also being used to demonstrate the ability of the equipment to survive earthquake loadings as defined by the Revised Response Spectra. This work is described in the following sections.
4.3.1 Evaluation of Qualification Test Results For equipment qualified by test, the Test Response Spectra were shown to envelope the Revised Response Spectra.
If the test response spectra did not completely envelope the Revised Response Spectra, a justification was provided to show that the equipment is acceptable for the Revised Response Spectra.
4-2
4.3.2 Evaluation of Qualification Analysis Results For equipment qualified by analysis, the acceleration value used in qualifying the equipment (i.e. in the seismic analysis report) was shown to be greater than the corresponding acceleration value from the Revise.' Response Spectra.
If the acceleration value from the Revised Response Spectra was not enveloped by the seismic analysis report, the component's critical stresses were calculated and shown to be within required ASME Code allowable limits.
4.3.3 Equipment Chosen for Evaluation to the Revised Response Spectra In Reference 1 Illinois Power committed to performing a stress assessment of the equipment listed in the second column of Table 4-1.
The actual equipment for which an assessment was made is listed in the third column of Table 4-1.
The reason for any deviation from the original list is identified in the fourth column.
4.3.4 Evaluation Results An item by item description of the evaluation results is provided in Appendix B.
The following observations can be made concerning the results:
For most equipment qualified by test, the test response spectra either enveloped the Revised Response Spectra over their entire frequency range or enveloped the Revised Response Spectra for all but the low frequency range (below the natural frequency of the equipment).
In either case the equipment is acceptable.
In one case (the RHR heat exchangers) the Vendor Response Spectra used in the analysis were lower than the Revised Response Spectra in a narrow frequency range above the natural frequency of the equipment.
For this case the component stresses were scaled up based on the highest ratio of acceleration values between the Revised Response Spectra and the Vendor Response Spectra, and the equipment was shown to be acceptable.
For those components qualified by analysis the accelerations due to the Revised Response Spectra were found to be less than the accelerations from the Procurement Specification Response Spectra if the accelerations were found to be higher than the accelerations from the Procurement Specification Response Spectra, the equipment stresses due to the Revised Response Spectra were calculated and found to be within acceptable limits.
4-3
4.4 Summary for Phase III The equipment evaluated in Phase III have all been shown capable a
of withstanding an earthquake of the form predicted by the Revised Response Spectra. The Division III diesel generator has not yet been evaluated because the equipment qualification report for the generator has not been completed. Once this report becomes available, the Division III diesel will be evaluated against the Revised Response Spectra. Based on the other equipment evalua-tions performed, no problems are anticipated with the capability of the Division III Diesel Generator to withstand an earthquake of the form predicted by the Revised Response Spectra, a
4 J
4 I
i a
l 4-4
TABLE 4-1 EQUIPMENT FOR EVALUATION TO THE REVISED RESPONSE SPECTPA Electrical Equipment Assessed ITEM Equipment No.
Equipment Justification for Change Listed in Reference I being assessed AC Power Systems 6900 V Switchgear IAP06E Equipment deleted because Assemblics (Typographical 6900 V switchgear is not Error in Reference 1 part of the emergency should read IAPOSE)
AC power supply system.
4160 V Switchgear IAP06E LAP 07E, IAP09E A non-safety grade switch-Assemblies gear assembly has been replaced by 2 similar a
safety grade switchgear assemblies.
480 V Unit IAPilE lAPilE, IAP12E An additional safety-
' Substations grade substation similar to the original has been added.
480 V Motor Control 1AP27E 1AP29E, IAP30E, A non-safety grade Centers IAP31E, IAP60E, Motor Control Center IAP61E, IAP72E, has been replaced by LAP 73E, IAP74E, 13 safety grade Motor 1AP75E, IAP76E, Control Centers.
IAP77E, IAP78E, 1E22-S002 i
DC Power Systems 125 V DC Motor Control IDCl3E IDCl3E, IDCl4E 2 additional safety grade Centers and Distribution
. Panels the original have been added.
4-5
-e
,.~
TABLE 4-1 (cont.)
EQUIPMENT FOR EVALUATION TO Tile REVISED RESPONSE SPECTRA Electrical Equipment Assessed ITEM Equipment No.
Equipment Justification for Change Listed in Reference 1 being assessed Batteries and Racks IDC01E IDC01E, IDC02E Scope was increased to IDC03E, IE22-S001D include batteries from all four divisions of emergency DC power.
Battery Chargers IDC06E IDC06E, IDC07E Scope was increased to IDC08E. IE22-S001E include battery chargers from all four divisions of-emergency DC power.
Inverters IC71-S001A IC71-S001A o
4-6
TABLE 4-1 (cont.)
EQUIPMEliT FOR EVALUATION TO TIIC REVISED RESPONSE SPECTRA Mechanical Equipment Assessed ITEM Equipment No.
Equipment Justification for Change Listed in Reference I being assessed Decay Heat System (RIIR)
Pumps IE12-C002A IE12-C002A Scope was increased to IE12-C002B include pumps from all IE12-C002C three loops of RHR.
Ileat Exchangers IE12-B001A lE12-B001A Scope was increased to IE12-B001B include both RIIR heat exchangers.
Shutdown Service Water System Strainers ISX01FA ISX01FA Scope was increased to ISX01FB include strainers from ISX01FC all three divisions of SX.
Pumps ISX0lPA ISX0lPA Scope was increased to ISX0lPB include pumps from all ISX0lPC three divisions of SX.
Instrument Panel ISX11J ISX11J Scope was increased to ISX12J include instrument panels ISX13J from all three divisions of SX.
4-7
TABLE 4-1 (cont.)
EQUIPMENT FOR EVALUATION TO Tile REVISED RESPONSE SPECTRA Mechanical Equipment Assessed ITEM Equipment No.
Equipment Justification for Change Listed in Reference 1 being assessed l
Diesel Generator System IDG0lKA IDG0lKA Including IDG0lKB Engine IDG0lKC Scope was increased to Generator Including for all include all three diesel-Auxiliary Lube Engine generators.
System Piping Generator An additional auxiliary Filters Auxiliary Lube was added.
Fuel Storage &
System Piping i
Supply Filters Engine Cooling Fuel Storage Water Heat Supply Piping Exchangers Engine Cooling Oil Coolers Water Heat Air Start Control Exchangers Panels Oil Coolers Local Control Air Start Control Panels Panels Local Control Panels Air Start System Piping
+
4 4
3 1
I t
4-8 1
SECTION 5 CONCLUSIONS The Equipment Seismic Assessment Program was developed in response to a request made during the ACRS meeting on the Clinton Power Station (CPS) which stated that " specific attention should be given to the seismic capabilities of the emergency AC power supplies, the DC power supplies, and small components such as actuators and instrument lines that are part l
of the decay heat removal system".
~
The three phases of the ESAP examined many seismic aspects of the systems important to decay heat removal and emergency power and, in doing so, responded to the request made during the CPS ACRS meeting. Phase I reviewed the seismic adequacy of the small bore piping designed by Sargent & Lundy. Phase II examined the as-built equipment configurations of the decay heat removal and emergency power supply systems for seismic Phase III evaluatad the ability of equipment in these systems concerns.
to withstand an earthquake o! the form predicted by the revised response spsetra.
l Although the ESAP was not an all inclusive review (it was not the intent of the program to verify all seismic design at CPS because programs already exist to do this), the ESAP does provide added confidence in the seismic capability of the emergency power supply and decay heat removal systems. Based upon the ESAP Illinois Power is confident that the seismic design of these systems is adequate.
l I
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5-1
SECTION 6 REFERENCES 1.
Wuller, G.
E., Supervisor - Licensing, IPC, Preliminary Response to ACRS Request for an Evaluation of Equipment Seismic Capability, letter to Miller, J.
R.,
- SSPB, Division of Licensing, NRC, U-0484, May 19, 1982.
2.
Bernard H., Acting Branch Chief, SSPB, Division of Licensing, NRC, Response to Request of May 19, 1982 of NRC Review and Comment on Illinoia Power's Equipment Seismic Assessment Program, letter to
- Wuller, G., Supervisor - Licensing, iPC, June 22, 1982.
3.
Campbell, R.
E., Quality Assurance-IPC, Reporting of a Potential 10CFR 50.55(e) Concerning Small Bore Piping at the Clinton Power Station, Record of Coordination with USNRC, Y-13910 August 2, 1982.
4 Miller, J.
R., Chief SSPB, Division of Licensing, NRC "ACRS Followup Actions", letter to Wuller, G.,
Supervisor Licensing, IPC, March 16, 1982.
i l
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6-1
APPENDIX A NSED WALKDOWNS AND PROBLEM RESOLUTION REPORT The Nuclear Station Engineering Department (NSED) conducted special walkdowns for the ESAP. These walkdowns were intended to provide an overview of the seismic design for the decay heat removal and emergency power systems.
These walkdowns were performed independently of the interaction analysis program and problems were identified on the basis of engineering judgement, rather than on the basis of a strict set of criteria.
The results of these walkdowns and the corrective actions taken for problems noted are included in this appendix.
The equipment described in this appendix are categorized and presented by area and include all the equipment identified in Table 3-1.
The construction in some of the areas covered by the walkdowns was incomplete. However, a large enough sample of-che equipment was walked down to allow a good overview of the seismic design for these systems.
Because of the NSED walkdowns, confidence in the seismic design of the equipment identified in Table 3-1 has been increased.
O e
NSED WALKDOWN AREAS Area 1:
LPCS Pump Room, Auxiliary Building, Elevation 707' AREA 2:
RER Loop A Pump Room, Auxiliary Building, Elevation 707' Area 3:
RHR Loop A Heat Exchanger Room, Auxiliary Building, Elevations 707', 737', 762' and 781' Area 4:
RHR Loop B Pump Room, Auxiliary Building, Elevation 707' Area 5:
RHR Loop B Heat Exchanger Room Auxiliary Building.
Elevations 707', 737', 762' and 781' Area 6:
RHR Loop C Pump Room, Auxiliary Building, Elevation 707' Area 7:
Auxiliary Building, East Side, Elevation 781' Area 8:
Auxiliary Building, West Side, Elevation 781' Area 9:
Centrol Building, Elevation 737' and 781' Area 10:
Division IV Battery Charger Room, Elevation 781' Area 11:
Division I Battery Room, Auxiliary Building, Elevation 781' Area 12:
Division II Battery Room, Auxiliary Building, Elevation 781' Area 13:
Division III Battery Room, Control Building, Elevation 781' Area 14:
Division IV Battery Room, Control Building, Elevation 781' Area 15:
Division III Diesel Generator Room, Diesel Generator Building, Elevation 737' Area 16:
Division I Diesel Generator, Diesel Generator Building, Elevation 737' Area 17:
Division II Diesel Generator, Diesel Generator Building, Elevation 737' Area 18:
Drywell, Azimuth 0*, Elevation 737' Area 19:
Auxiliary Building, Steam Tunnel Area, Elevation 767' Area 20:
Shutdown Service Water (SX) Division I Pump Room, Screenhouse Area 21:
SX Division II Pump Room, Screenhouse Area 22:
SX Division III Pump Room, Screenhouse i
^
A-1
AREA 1 LOCATION: LPCS Pump Room, Auxiliary Building, Elevation 707' IE21-C002 (Low Pressure Core Spray Water Leg Pump A)
No potential seismic interactions or other problems were noted.
Piping leading to and from the pump was adequately supported.
O A-2
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AREA 2 LOCATION: RHR Loop A Pump Room. Auxiliary Building Elevation 707' A) 1E12-F006A (16 inch Motor Operated Valve for RHR Pump 1A Suction in Shutdown Cooling Mode)
No potential seismic interactions or other problems were noted (nearest objects are structural steel and piping over 6 inches away from the valve operator).
B) 1E12C002A (RHR Pump 1A)
There is no vendor supplied piping on the motor; however the pump has a seal cooler and a suction vent with connecting S&L designed small bore lines.
The suction vent line, IRF59BA 3/4, is about 15 feet long and runs to a floor drain. At the time of the walkdown this line was equipped with 2 unattached hangers. The suction vent line will be adequately supported once constructica is completed and the hangers are attached.
The pump seal cooler, bolted to the side of the RHR pump has four lines associated with it:
1.
A firmly supported SX line.
ISX53AAlh.
2.
A poorly supported line, ISX54AA1, with a 20 foot unsupported span.
3.
The Seal Cooler vent line, 1RF60BA 3/4, routed through the motor / pump coupling access portals with 2 unattached hangers. When construction is completed these hangers will provide adequate support. 4 A firmly supported vendor
. supplied line from the seal to the seal cooler.
ACTION TAKEN:
For Line 2 the SX M07 series drawings (construction is based upon these drawings) were examined. These drawings showed that another hanger will be installed during construction which will adequately support the line.
C)
E26-1000-03A/CP-3 (Conduit Hanger Supporting Conduit Attached to the RHR Pump 1A Motor)
The hanger is unusually long and flexible. There was concern that the hanger could be damaged during a seismic event.
ACTION TAKEN: The problem was referred to the AE.
They reviewed the design calculation for this hanger. Although the center of load distance exceeded the tabular values provided in the verification procedure, the hanger was determined to be adequate by a separate analysis.
(Reference S&L Calculation SDQ45-00DG05 pages 21.520 through 21.524.)
A-3
AREA 3 LOCATION: RHR Loop A Heat Exchanger Room, Auxiliary Building, Elevations 707', 737', 762' & 781' A) lE12-F014A (18 inch Motor Operated Valve on SX Water Inlet to RHR Heat Exchanger 1A, Elevation 707')
No potential seismic interactions or other problems were noted.
B) lE12-F068A (18 inch Motor Operated Valve on SX Water Discharge from RHR Heat Exchanger 1A, Elevation 707')
No potential seismic interactions or other problems were noted.
C)
IVYO35 (RHR Heat Exchanger Room A Coil Cabinet, Elevation 707')
Valve ISX023A is a 180 lb diaphragm operated control valve located on a u-shaped dip in line ISX036AA2, the coil cabinet discharge line. This valve was found to sway easily because there were no hangers installed near it.
(See photograph A-1).
Valve ISX137 on high point vent ISX63A 3/4 of the coil cabinet inlet line, ISXFIA2 has an interference problem.
Its handle cannot be turned without hitting a neighboring 2 inch line.
(See photograph A-2)
ACTION TAKEN: The M07 series drawings for line ISX36AA2 were armnined. They showed that after construction is complete there will be an adequate number of hangers installed on this line to properly restrain it, including one within 6 inches of the valve (as required by the Small Pip'ng Procedure).
i The interference concerning valve ISX137 was subsequently observed to be corrected by rework.
D) lE51-F374 (3 inch manually operated valve on 1RI53A, Elevation 737')
There was a potential seismic interaction noted between the valve handwheel for IE51-F374 and an overhead pipe support (See photograph A-3).
ACTION TAKEN: This problem was observed to be corrected by rework since the walkdowns.
E) lE12-F026A (4 inch Motor Operated Valve on line IRH26BA4, RCIC suction from the RER Heat Exchanger, Elevation 737')
There was a potential seismic interaction noted between the valve operator for this valve and a nearby pipe support. The clearance between these two was approximately 1 inch. The line which the valve is on 1RH26BA4, was noted to be rigid.
ACTION TAKEN: Sargent & Lundy was requested to analyze this situation. They were able to determine from design calculations that the 1 inch clearance was adequate.
A-4
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F) 1E12-F065A (4 inch Air Operated Control Valve for Steam Condensing Mode of RRR, Elevation 737')
No potential seismic interactions or other problems were noted.
G) lE12-F003A (14 inch Motor Operated Valve for Shutdown Cooling Mode of RHR, Elevation 737')
No potential seismic interactions or other problems were noted.
H) lE12-F048A (14 inch Motor Operated Valve for Low Pressure Coolant Injection Mode of RHR, Elevation 737')
Conduit hanger E26-1000-02A-CC29 passes within 1 3/4 inch of the valve body for valve LE12-F048A. This presents a potential seismic interaction for the valve.
ACTION TAKEN: This problem was referred to Sargent & Lundy.
The problem has been corrected by removal of the conduit hanger.
I) 1RH100AA2 (1 inch Water Level Instrument Line for the RHR Heat Exchanger lA, Elevation 737'-781')
This line extends most of the length of the heat exchanger. There were no hangers installed on it at the ' time of the walkdowns.
ACTION TAKEN: The RHR M07 series drawings were examined. They show that 6 hangers will be installed on this line after construction is complete. When they are installed the line will be properly restrained.
J)
LE12-F047A (14 inch Motor Operated Valve on RHR Heat Exchanger Inlet, Elevation 781')
No potential seismic interactions or other problems were noted.
K)
LE12-F051A (6 inch Air Operated Control Valve for Steam Condensing Mode of RER, Elevation 781')
No potential seismic interactions were noted.
L) 1RF95AAl (1 inch vent on top of RHR Heat Exchanger lA - vents shell side of the Heat Exchanger to the Room)
The vent line on top of the heat exchanger is a cantilever section, with two 1 inch valves located on it.
This arrangement, without pipe hangers installed, would during a seismic event, place a large bending stress on the welds connecting the vent line to the heat exchanger.
(See photograph A-4)
ACTION TAKEN: The M07 series drawings for this vent line were aramined. There will be a hanger-installed af ter construction is complete downstream of the outermost valve which will adequately support this line.
A-5 _ _
AREA 4 LOCATION:
RHR Loop B Pump Room, Auxiliary Building, Elevation 707' A) lE12-F006B (16 inch Motor Operated Valve for RHR Pump 1B Suction in Shutdown Cooling' Mode)
No seismic interactions or other problems were noted.
B) lE12C002B (RHR Pump 1B)
Line ISX54AB1h leading from the pump scal cooler was found to be poorly supported. Air operated valve ISX029B on this line was subject to movements that could damage the air line supplying this valve.
Valve IE12-F308B located on a tap off of the discharge pipe for the RHR Pump IB had an interference with a nearby pipe hanger.
The hand operator for this valve could not be turned.
(See photograph A-5)
ACTION TAKEN: The SX M07 series drawings for line ISX54ABI were examined. They showed that there will be hangers ir. stalled on this line during construction, to provide adequate support. -
The interference problem concerning Valve IE12-7308 was referred.co Sargent & Lundy. They will issue a construction drawing revision to tilt the valve to eliminate the interference.
This modification sdll be completed prior to fuel load.
,)d' A-6
i AREA 5 LOCATION:
RHR Loop B Heat Exchanger Room, Auxiliary Building, Elevations 707', 737', 762' and 781'.
A) lE12-F014B (18 inch Motor Operated Valve on SX Water Inlet to RER Heat Exchanger 13, Elevation 707')
A hanger (probably for conduit or HVAC) was noted to be resting on the motor operator for this valve (See photograph A-6)
ACTION TAKEN: This problem has been corrected since the walkdowns by rework.
(The hanger has been removed).
B) 1E12-F068B (18 inch Motor Operated Valve on SX Water Discharge from RHR Heat Exchanger IB, Elevation 707')
No potential seismic interactions or other problems were noted.
C) 1E12-F026B (4 inch Motor Oparated Valve on line 1RH26BB4, RCIC suction from the RER Heat Excianger 1B, Elevation 737')
No potential seismic interactians or other problems were noted.
D) lE12-F065B (4 inch Air Operated Control Valve for Steam Condensing Mode of RHR, Elevation 737')
No potential seismic interactions or other problems were noted.
E) lE12-F003B (14 inch Motor Operated Valve for Shutdown Cooling Mode of RHR, Elevation 737')
No potential seismic interactions or other problems were noted.
F) lE12-F048B (14 inch Motor Operated Valve for Low Pressure Coolant Injection Mode of RER, Elevation 737')
No potential seismic interactions or other problems were noted.
G) 1RH100AB2 (2 inch Water Level Instrument Line for RHR Heat Exchanger IB, Elevation 737'-781')
This line extends most of the length of the RER heat exchanger.
I There were no hangers installed on it at the time of the walkdowns.
i ACTION TAKEN: The RHR M07 series drawings were examined. The drawings showed that there will be an adequate number of hangers on this line after construction is complete to properly restrain it.
H) 1E12-F047B (14 inch Motor Operated Valve on RHR Heat Exchanger IB 1
Inlet, Elevation 781')
No potential seismic interactions or other problems were noted.
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1 AREA 5 (Cont.)
I) 1RF95ABl (1 inch vent on top RHR Heat Exchanger - vents shell side of the Heat Exchanger to the Room. Elevation 781')
The vent line en top of t'he Heat Exchanger is a cantilever section, with two 1 inch valves located on it.
This arrangement, without pipe hangers installed would, during a seismic event, place a large bending stress on the welds connecting the vent line to the Heat Exchanger.
ACTION TAKEN: Tne M07 series drawings for this vent line were examined.
There will be a hanger installed during construction downstream of the outermost valve which will adequately support this line.
1 A-8
4 AREA 6 LOCATION:
RHR Loop C Pump Room, Auxiliary Building, Elevation 707' A) lE12-C003 (RHR Water Leg Pump for RHR Loops B and C)
The suction line, IRH16A2, for this pump comes from line IRH01BC20, the suction line for RHR Pump IC.
Line IRH16A2 has a pipe hanger attached to it near this juncture. This hanger may have the undesirable effect of resisting the thermal motions of the 20 inch line, in which case the welded connection between these two lines may be damaged (See Photograph A-7)
ACTION TAKEN: The thermal motion concern with line IRH01BC20 was referred to Sargent & Lundy.
From the stress report for this line they were able to determine that the thermal motions would be insignificant. Therefore no damage will occur to the welded connection because of thermal motions.
B) lE12-C002C (RHR Pump 1C)
Line ISX54ACI, leading from the pump seal cooler, was found to be poorly supported.
Line IRH01BC20, the suction line for the RHR pump, has a potential interaction with a hanger fastened to the discharge piping for the water leg pump.
ACTION TAKEN: The M07 series drawings were examined for line
'.SX54ACl. They showed that there will be hangers installed on this line during construction to provide adequate support.
The potential interaction concerning line IRH01BC20 was referred to Sargent & Lundy. They determined that the installed clearance was adequate.
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AREA 7 LOCATION: Auxiliary Building, East Side, Elevation 781' A) 1AP07E (4160 Volt Switchgear lAl)
No potential seismic interactions or other problems were noted.
B) 1APilE (480 Volt Unit Substation lA)
No potential seismic interactions or other problems were noted.
C) 1AP72E ( uxiliary Building Motor Control Center lAl)
No potential seismic interactions or other problems were noted.
D)
LAP 73E (Auxiliary Building Motor Control Center lA2)
No potential seismic interactions or other problems were noted.
E)
LAP 74E (Auxiliary Building Motor Control Center lA3)
No potential seismic interactions or other problems were noted.
F)
IDC06E (125 Volt Battery Charger lA)
No poiential seismic interactions or other problems were noted.
G)
LDC013E (125 Volt DC Motor Control Center lA)
No potential seismic interactions or other problems were noted.
i A-10 y
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AREA 8 LOCATION: Auxiliary Building, West Side, Elevation 781' A) 1AP09E (4160 Volt Switchgear 1B1)
No potential seismic interactions or other problems were noted.
B) 1AP12E (480 Volt Unit Subs'tation IB)
No potential seismic interactions or other problems were noted.
C) 1AP75E (Auxiliary Building Motor Control Center 1B1)
No potential seismic interactions or other problems were noted.
D) 1AP76E (Auxiliary Building Motor Control Center IB2)
No potential seismic interactions or other problems were noted.
E) 1AP77E (Auxiliary Building Motor Control Center 1B3.
No potential seismic interactions or other problems were noted.
F) 1DC07E (125 Volt Battery Charger IB)
No potential seismic interactions or other problems were noted.
G) 1DCl4E (125 Volt DG Motor Controi Center IB)
No potential seismic interactions or other problems were noted.
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AREA 9 LOCATION: Control Building, Elevation 737' and 781' A) 1AP60E (Diesel Generator Building Motor Control Center IA.
Elevation 737')
No potential seismic interactions or other problems were noted.
B) 1AP61E (Diesel Generator Building Motor Control Center IB.
Elevation 737')
No potential seismic interactions or other problems were noted.
C) 1AP78E (Auxiliary Building Motor Control Center ICl, Elevation 781')
No potential seismic interactions or other problems were noted.
D) lE22-S002 (Division III Motor Control Center, Elevation 781')
No potential seismic interactions or other problems were noted.
E) lE22-S001E (Division III Battery Charger, Elevation 781')
The charger is in the hall on the plant south side of the Division III Battery Room.
It is far enough from hallway traffic to have little risk of damage from passers-by. No potential seismic interactions or other problems were noted.
e A-12
AREA 10 LOCATION: Division IV Battery Charger Room, Elevation 781' A) 1DC15E (125 Volt DC Motor Control Center ID) f No potential seismic interactions or other problems were noted.
B) 1DC08E (Division IV Battery Charger)
No potential seismic interactions or other problems were noted.
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AREA 11 LOCATION: Division I Battery Room, Auxiliary Building, Elevation 781' 1DC01E (Division I 125 Volt Batteries)
Access to the battery room is controlled via a locked door.
Batteries are mounted on two racks (each rack having two trays) whfch are strong, rigid and firmly enclose the batteries. The racks are about 2.5 feet apart. The racks are securely bolted to the floor. Conduit in the room servicing the batteries is very heavily supported.
Except for the batteries the room is relatively empty. The contents in the room include:
Overhead fluorescent lights which are securely mounted directly to the ceiling with screws; An emergency light fixture placed on a wall mounted platform near one of the racks (the platform holding the lights is securely fastened to the wall);
4 A ceiling penetration for an exhaust fan exhausting into a cable tray passage area; Service air system piping to an air outlet in the room.
No potential seismic interactions or other problems were noted.
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AREA 12 LOCATION: Division II Battery Room, Auxiliary Building, Elevation 781' 1DC02E (Division II 125 Volt Batteries) i i
Access to the battery room is controlled via a locked door.
Batteries are mounted on two racks (each rack having two trays) which are strong, rigid an'd firmly enclose the batteries. The racks are about 8 feet apart.
The racks are securely bolted to the floor. Conduit in the room servicing the batteries is very heavily supported.
Except for the batteries the room is relatively empty. The contents in the room include:
Overhead fluorescent lights which are securely mounted directly to the ceiling with screws; An emergency light fixture placed on a wall mounted platform well away from the racks; A ceiling penetration for an exhaust fan exhausting into a
]
cable tray passage area; Service air system piping to an air outlet in the room.
4 No potential seismic interactions or other problems were noted.
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I A-15
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AREA 13 LOCATION: Division III Battery Room, Control Building, Elevation 781' LE22-S001D (Division III 125 Volt Batteries)
Access to the battery room is controlled via a locked door.
Batteries are mounted on a single rack which has two trays.
The rack is strong, rigid and' firmly encloses the batteries. The rack is securely bolted to the floor. Conduit in the room servicing the batteries is very heavily supported.
Except for the ba*.ceries the room is relatively empty. The contents in the room include:
Ove-head fluorescent lights which are mounted by chains to the ceiling; An Emergency light fixture placed on a wall mounted platform and located near the rack (the platform holding the lights is firmly fastened to the wall);
A wall penetration for an exhaust fan exhausting into a hallway; An Eyewash Station located about 8 feet from the rack and draining to a floor drain; Service air system piping to an air outlet in the room.
No seismic problems were noted in this room.
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i A-16
AREA 14 LOCATION: Division IV Battery room, Control Building, Elevation 781' 1DC03E (Division IV 125 Volt Batteries)
The batteries are mounted on two racks which are bolted to the floor. There are 2 sets of fluorescent lights in the room, partially hanging over the' batteries. They pose no hazard to the batteries because they are mounted to'the ceiling from sturdy hooks. No significant problems were noted in this room.
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' A-17
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AREA 15 LOCATION: Division III Diesel Generator Room, Diesel Generator Building, Elevation 737' A) 1DG06SC (Division III Diesel Generator Air Start Skid)
All the piping on the air start skid seems adequately supported, including small instrument tubing. Electrical conduit servicing the skid also appears adequately supported. Neighboring conduit and cable trays (there are no neighboring pipes) are distant enough to avoid interaction problems.
Flexible connections from both the electric and diesel engine driven compressors to their after coolers are adequately stiff.
The air dryers designed for the system are firmly mounted as is the piping from the air start skid to the dryers.
B) 1DG0lTC (Division III Diesel Generator Fuel Oil Day Tank)
The tank is in a cubicle in the Division III Diesel Generator Room. All the fuel oil piping in the cubical appears to be adequately supported. A fire protection sprinkler line appears to 4
be adequately supported.
The fuel oil piping between the diesel generator and the day tank is in a covered trench in the floor. The pipe supports in the trench are spaced 6 to 8 feet apart which appears adequate. The fuel oil lines are attached to the diesel by flexible connectors to avoid transmitting movements of the diesel to them.
The piping leading to the flexible connections is a cantilevered section about 2 f&et in length (see Photograph A-8).
A long cantilevered section such as this may be susceptible to damage from an earthquake, because of the bending stress at the base of the cantilever.
ACTION TAKEN: The piping leading to the flexible coupling has had another pipe hanger installed on it since the walkdowns.
It now appears to be well supported.
C) 1E22-S001B (Division III Diesel Generator Control Panel)
Conduit coming into the panel is adequately supported. No potential interactions or other problems were noted.
D) lH21-P028 (Division III Diesel Generator Protective Relay)
Coaduit coming into the panel is adequately supported. No rotential interactions or other problems were noted.
A-18 v
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AREA 16 LOCATION: Division I Diesel Generator, Diesel Generator Building, Elevation 737' No review was performed in this area due to lack of completeness.
9 6
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AREA 17
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j LOCATION: Division II Diesel Generator, Diesel Generator Building, Elevation 737' No review was performed in this area due to lack of completeness.
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AREA 18 LOCATION: Drywell, Azimuth 0*, Elevation 737 '
1E12-F009 (18 inch Motor Operated Valve for Opening the RHR Reactor Recirculation Crosstie for Shutdown Cooling Mode of RHR)
There is a k inch clearance between the valves motor operator and a 6 inch line, 1RT01B.
(See Photograph A-9).
1RT01B is a ASME section III class i line. This is a critical interaction problem in that failure of this RHR valve to open would prevent use of the Shutdown Cooling Mode of RHR (There is no redundant Reactor Recirculation Crosstie line).
ACTION TAKEN: This problem has been corrected by rework since the walkdowns.
It has been observed that the motor operator for valve 1E12-F009 has been reoriented to eliminate this potential interaction.
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A-21
AREA 19 LOCATION: Auxiliary Building, Steam Tunnel Area, Elevation 767' 1E12-F008 (18 inch Motor Operated Valve for Shutdown Cooling Mode i
of RHR) j No potential seismic interactions or other problems were noted.
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LOCATION: Shutdown Service Water (SX) Division I Pump Room, Screenhouse A)
ISX0lPA (SX Pump 1A)
There is a lh inch clearance between 7.ines ISX77AAl and ISX78AAl which supply cooling water, to the motor bearings. This clearance appears to be sufficient. 'Th-se lines are adequately supported.
There is an interaction problem with one of the 3 x 3 inch tube steel pipe supports for these cooling lines and the pump motor, k inch away (see Photograph A-10).
The pressure instrumentation lines ISX05AA 3/4 and ISX05AB on the pump discharge line are well supported.
ACTION TAKEN: The potential interaction concerning the 3 x 3 inch tube steel and the pump motor was referred to Sargent & Lundy.
This problem has been corrected. The hanger was shortened to eliminate the potential interaction with the pump motor.
B)
ISX0 LEA (SX Strainer 1A)
The strainer Limitorque operator has a handwheel for manual operation which cannot be turned because of an interference with the 480 Volt motor control center 1AP29E (See Photograph A-11).
The local pressure gauges and their instrument tubing are mounted securely to the strainer top.
Pressure instrument line ISXA7AB is sturdy but passes within k inch of the flange on butterfly valve ISX004A on the strainer discharge line. This is a potential interaction problem.
Valve ISX013D on line LSX29BA3 (the auto backwash line) has a large cantilevered operator.
It imposes a large bending moment on itself and the pipe. The neighboring lines, however, are very rigid.
ACTION TAKEN: The above noted problems were referred to Sargent &
Lundy. The handwheel to motor control center interference was corrected by rework. The handle on the handwheel, which is not essential for the operation of the handwheel, was removed.
Line ISIA7AB was determined to be adequate as is because the small i
seismic motions of this line and the nearby flange preclude a seismic interacton.
Valve ISX013D has been verified to be adequate for its seismic environment in S&L qualification reports CQD-000308, EMD-026594, and END-018589.
1 A-23
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AREA 20 (Cont.)
C)
ISX010A (2 inch Air Operated Control Valve on Room Cooling Unit.
Cooling Water Discharge)
This valve is located on a U-shaped dip in line ISXO8AA2.
(See Photograph A-12) The span between supports is 18'-0" with the nearest support 7'-9" away from the valve. Because this U is part of a long span of piping between pipe supports, this valve is susceptible to large sways during seismic events. This could damage the valve, the pipe it is on, or most likely the 3/8 inch copper instrument air line attached to it and the wall nearby.
ACTION TAKEN: The line the valve is on was analyzed by PIPSYS, A Sargent & Lundy Computer Code used for piping design verification.
This analysis shows that the line is adequate as is and that the valve motions will not be great enough to collide with nearby equipment, so no damage will occur to the valve.
If the instrument air line were to break or crimp the valve wculd fail open, ensuring a steady flow of cooling water through the area coolers. This is a failure in a conservative direction and is not considered a safety problem.
D) 1AP29E (Screenhouse Motor Control Center 1A)
No interaction problems are evident (except for the strainer handwheel interference mentioned earlier). Conduit in the area appears to be well supported.
A-24 4
I l
AREA 21 LOCATION: SX Division II Pump Room. Screenhouse A)
ISX01PB (SX Pump 1B) 4 ISX78AB1 and ISX77AB1, the motor bearing cooling lines, appear to i
be adequately spaced apart and supported.
The pressure instrument tap lines ISX05BA 3/4 and ISX05BBh on the pump discharge line appear to be well supported.
No potential seismic interactions or other problems were noted.
l B)
ISX01FB (SX Strainer 1B)
Valve ISX013E on the strainer auto backwash line 1SX29BB3, has a large cantilevered operator.
It imposes a large bending moment on itself and the pipe. The lines it could effect are, however, securely supported.
(See photograph A-13).
ACTION TAKEN: The concern with valve ISX013E was referred to Sargent & Lundy. It was verified to be adequate for seismic loadings in qualification reports CQD-000308. END-026594, and EMD-018589.
C)
ISX010B (Air Operated Control Valve on Room Cooling Unit Cooling
]
Water Discharge).
This valve is located on a U-shaped dip in line 1SX08AB2. The I
span between supports is 18'-0" with the nearest support 7'-9" away from the valve. Because this U is part of a long span of piping between pipe supports, this valve is susceptible to large sways during seismic events. This could damage the valve, the pipe it is on, or most likely the 3/8 inch copper instrument line attached to it and the wall nearby.
ACTION TAKEN: The line the valve is on was analyzed by PIPSYS, a Sargent & Lundy computer code used for piping design verification.
This analysis shows that the line is adequate as is and that the valve motions will not be great enough to collide with nearby equipment, so no damage will occur to the valve.
If the instrument air line were to break or crimp the valve would fail open, ensuring a steady flow of cooling water through the area coolers. This is a failure in a conservative direction and is not considered a safety problem.
l i
D) 15X04B (30 inch butterfly valve on strainer for maintenance isolation)
The Limitorque motor operator was noted to be loosely assembled at the flange between the motor gearbox and the final worm gear drive. This joint will have to transmit a large and cycling torque.
l ACIION TAKEN: The loose connection between the operator and the valve was verified to have been corrected since the walkdowns.
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AREA 21 (Cont.)
E) 1AP30E (Screenhouse Motor Control Center IB)
No potential seismic interactions or other problems were noted.
A-26
AREA 22 LOCATION: SX Division III Pump Room, Screenhouse A)
Instrument tubing on the pump discharge pipe appears to be adequately supported and so'dces the pump bearing flush line, ISX87A 3/4 which terminates in the pump. No potential seismic
(
interactions or other problems were noted.
B)
A potential interaction problem between the operator for valve ISX013F and a nearby blind flange was noted. This valve is on the auto backwash line, ISX29BC2.
Instrument tubing for local pressure gauges mounted on the strainer appears to be well supported.
ACTION TAKEN: The interaction problem concerning valve ISX013F was referred to Sargent & Lundy. They determined that the small seismic motions of the valve and the blind flange preclude the possibility of a seismic interaction.
C) 1AP31E (Screenhouse Motor Control Center IC)
/rhe strainer bypass line ISXO6AC8 passes 6 inches above the panel. This clearance seems adequate. No problems were noted.
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4 APPENDIX B i
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4 i
EQUIPMENT EVALUATION RESULTS i
i i
As a part of the ESAP, an evaluation of the seismic capability of equipment critical to the plant functions of decay heat removal and emergency power supply was performed. The i
evaluation was based on plant seismic response spectra determined by different soil structure interaction analysis methods than were used to generate the design basis response spectra. The use of these different spectra (the revised response spectra) in the evaluation improved confidence in the capability of the equipment to withstand safe 1
j shutdown earthquake acceleration loadings.
I L
4 i
An item by item description of the evaluation results is provided in this appendix. Graphs comparing the Revised Response Spectra with the spectra used for equipment qualification l
I and tables comparing revised stresses with allowable stresses are also provided in support of these descriptions.
The equipment evaluated during this phase of the program have all been shown to be capable of withstanding an earthquake of the form predicted by the Revised Response Spectra.
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3 EQUIPMENT EVALUATTON RESULTS 3
l Equipment j
AC Power Systems Number Status
-4160V Switchgear 1AP07E,9E
-The Procurement Specification Spectra and the Revised Response Spectra were not completely enveloped by the lower bound of the Test Response Spectra. However, investigation of the test report indicates i
i; that the equipment has a fundamental natural frequency of 13 Hz, and will not respond in j
the frequency range which has not been i
enveloped. Therefore, the equipment is i
acceptable.
I
-Comparison of the response spectra is j
shown in figures B-1 and B-2.
-480V Unit Substations 1AP11E 12E
-A lower bound of all the Test Response Spectra (both horizontal & vertical) was found to envelope the Revised Responsa j
Spectra. These plots are shown in Figures B-3 and B-4 i
i
-480V Motor Control 1AP60E, 61E
-Test Response Spectra envelope both the l
Centers (MCCs)
Revised Response Spectra and Procurement i
l Response Spectra, excent for frequencies below 2 Hz.
Vendor report shows that the MCC has a natural frequency of 10 Hz, and will not respond to the frequencies below 2 Hz. Therefore, this equipment is i
acceptable.
-Comparison of the response spectra is shown in Figure B-5, B-6 and B-7.
I
-Curves are plotted at 1% damping.
-480V MCCs 72E through 77E
-Test Response Spectra envelope both the l
Revised Response Spectra and Procurement l
Response Spectra, except for frequencies i
below 2 Hz.
Vendor report shows that the
{
MCC has a natural frequency of 10 Hz, and
['
will not respond to the frequencies below 2 Hz. Therefore, this equipment is accaptable.
J
-Comparison of the response spectra is
)
shown in Figures B-8 and B-9.
Curves are L
plotted at 1% damping.
4 B-1
=
.~.
i 4
I Equipment AC Power Systems (cont)
Number Status
{
480V MCC 1AP78E
-Test Response Spectra envelope both the i
Revised Response Spectra and Procurement Response Spectra, except for frequencies below 2 Hz.
Vendor report shows that the MCC has a natural frequency of 10 Hz and will not respond to the frequencies below 2 Hz.
Therefore, this equipment is l
acceptable.
Comparison of the response spectra are j
shown in Figures B-10. B-11 and B-12.
Curves are plotted at 1% damping.
}
480V MCCs 1AP29E,30E,31E
-Test Response Spectra envelope both the Revised Response Spectra and Procurement i
i Response Spectra, except for frequencies below 2 Hz.
Vendor report shows that the i
i MCC has a natural frequency of 10 Hz, and will not respond to the frequencies below 2 Hz.
Therefore, this equipment is acceptable.
-Comparison of the response spectra is shown
]
in Figures 5-13 B-14 & B-15.
-1.ower bound of the test response spectra i
(both horizontal and vertical) was shown to t
envelope the revised response spectra, j
except for the low frequency range (below 3 Hz in the horizontal direction, and between i
6 and 7 Hz in the vertical direction).
1 i
-As shown in the vendor report, the MCC does not have any resonant frequencies below 15 Hz.
Therefore, the equipment is acceptable.
-Comparison of the response spectra is shown
)
in Figures 8-16 and B-17.
f 1
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DC Power Systems l
i!>C13 E.14 E,
-Test Response Spectra envelope both the j
Distribution Panel Revised Response Spectra and Procurement Response Spectra, except for frequencies below 2 Hz. Vendor report shows that the i
MCC has a natural frequency of 10 Hz, and i
will not respond to the frequencies below 2 Hz. Therefore, this equipment is acceptable.
1
(
B-2 i
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i Equipment DC Power Systems (cont)
Number Status
-Comparison of the response spectra is shown in Figures B-18 and B-19.
Curves are plotted at 1% damping.
1DC15E
-Test Response Spectra envelopes both the Distribution Panel Revised Response Spectra and Procurement Response Spectra, except for frequencies below 2 Hz.
Vendor report shows that the MCC has a natural frequency of 10 Hz, and will not respond to'the frequencies below 2 Hz.
Therefore, this equipment is acceptable.
!j
-Comparison of the response spectra is shown in Figures B-20, B-21 & B-22.* Curves are i
plotted at 1% damping.
-Batteries & Racks IDC01E,2E.3E
-Lower bound of all of the Test Response Spectra (for both horizontal & vertical i
response) envelopes both the Revised i
Response Spectrum and the Procurement Specification Spectrum. Therefore, this 4
equipment is acceptable. For comparisons 1
see Figures B-23 & B-24 1
IE22-S001D
-Original analysis used peak accelerations l
from the Required Response Spectra at 1%
damping (1.73 horizontal & 8.0g vertical) 1
-The Revised Response Spectra (2% damping)
]
have the following peak accelerationst i
1.2g horizontal 3.4g vertical
-Since the revised seismic accelerations are j
below those used in the analysis, this equipment is acceptable. See Table B-1, item 1 for tabulation of results.
a NOTE: Use of 1% damping was a conservatism in the original analysis.
-Battery Chargers 1DC06E 7E.8E
-Lower bound of the Test Response Spectra i
was shown to envelope the Revised Response Spectra.
I
-The Procurement Specification Spectra was not completely enveloped by the lower bound of the Test Response Spectra. However.
investigation of the transmissibility plots indicates that the equipment will not i
4 B-3 i
EQUIPMENT EVALUATION RESULTS Equipment DC Power Systems (cent)
Number Status respond at the frequencies that were not enveloped (resonant frequencies are at 16, 25 and 50 Hz). Therefore, this equipment is acceptable.
-Comparison of the response spectra is shoen in Figures B-25 through B-29.
3 ttery Charger 1E2-S001E
-Lower bound of the Test Response Spectra (both horizontal and vertical) was shown to envelope the Revised Response Spectra.
-Comparison of the response spectra is shown in Figures B-30, B-31 and B-32.
-Inverter & Static 1C71-S001A
-Test Response Spectra envelope the Revised Bypass Switch Response Spectra.
-Comparison of the response spectra is shown in Figures B-33. B-34 and B-35.
D eny Heat System (RHR)
-Pumps 1E12-C002A B.C
-The Procurement Specification Spectra were compared with the Revised Response Spectra, and have been shown to envelope the Revised Response Spectra for all frequencies except for those below 7 Hz.
However, the equipment has a fundamental natural frequency of 16.4 Hz.
Therefore, the RHR pumps remain acceptable.
-Comparison of the response spectra is shown in Figures B-36 and B-37.
l
-See Table B-1. Item 2 for tabulation of maximum stresses for each component.
-Heat Exchangers 1E12-B001A,B
-Vendor report used dynamic response spectra analysis.
-Comparison of the response spectra is shown in Figures B-38 and B-39.
- w From these figures it is shown that the vendor's horizontal spectra envelope both the Revised Response Spectra, and the Procurement Specification Spectra except for the frequencies between 12 and 19 Hz, and below 5 Hz.
The vendor's vertical spectra envelope both the Revised Response Spectra and the Procurement Specification Spectra, except for the f requencies below 6.5 H3.
B-4
EQUIPMENT EVALUATION RESULTS Dscay Heat System (RHR)
Equipment (continued)
Number Status
-From the vendor report, this equipment does not have natural frequencies below 8 Hz.
Therefore, the vertical spectra are acceptable.
From Figure B-38, horizontal spectra, the highest ratio of the acceleration values between the two spectra is.48/.4=1.20, at 14 Hz.
-This ratio is used to scale up the component stress. Critical stresses of this equipment are revised and tabulated in Table B-1, Item 3.
-This tabulation shows that the revised component stresses are less than the allowable stresses.
Therefore, this equipment remains acceptable.
Shutdown Service Water System
-Motor for Pump ISXO1PC
-The original analysis used acceleration values of 2.23g horizontal (1.5 x equivalent acceleration value) and 5.20g vertical (peak acceleration value) to qualify the equipment.
-The Revised Response Spectrum has equivalentacg+eleratignygluesof]g ' 2.59g ho
[(1.5)[(1.24)
(1.20) and.97g vertical.
-The component stress is scaled up by a ratio of 2.59/2.23 = 1.16.
-See table B-3, item 1 for tabulation of results.
These results show that the component's new stress is less than the allowable stress.
Therefore, this item is acceptable.
-Pump 1SX01PC
-Original analysis shows that the fundamen-tal frequency of the equipment is 34 Hz.
-Original analysis used acceleration values as follows:
.26g in both horizontal directions and
.60g in vertical direction B-5
~. - _.
3 EOUXPMENT EUALUATRON RESULTS Shutdown Service Water Equipment 3
System (cont)
Number Status
-Revised Response Spectra has the following acceleration values:
.26 in both horizontal directions and
.25g in vertical direction
-Since the revised seismic accelerations are i
equal to or below those used in the original analysis, the equipment is acceptable, i
-See table B-3, item 2 for tabulation of results.
-Motor for Pump ISX01PA PB
-Original analysis shows that the fundamen-tal frequency of the equipment is 53 Hz.
-Original analysis used the following acceleration values:
40g in both horizontal directions and
.65g in the vertical direction i
-Revised Response Spectra have the following 7
i acceleration values at the frequency of i
50 Hz (rigid range):
.263 in both horizontal directions j
.24g in the vertical direction
-Since the revised seismic accelerations are i
lower than those used in the original analysis, this equipment is acceptable.
-See table B-3, item 3 for tabulation of j
results.
a i
-Pump ISX01PA,PB i
-Original analysis shows the following equipment fundamental frequencies:
49 Hz in the vertical direction (rigid range) and 23 Hz in the horizontal direction j
-Original analysis used vertical acceleration value of.65g and a dynamic analysis using the response spectra to j
calculate equipment stresses.
2
-Revised Response Spectra shens an
{
acceleration value of.24g in the vertical lI direction.
i I
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}
B-6
.m.
._., _,,. - _., _ - _ -, ~ _ _,. _ _... -. -.. _.,... _
.____,--_.-m.._
a Shutdown Service Water Equipment System (cont)
Number Status A comparison of the horizontal response spectra is shown below:
Combined Combined Hor. Accel.
Period Accel. In from revised (SEC) report (G) spectra (G)
.1 1.21 1.11
.076 1.101
.823
.056
.814
.711
.052
.701
.684
.047
.386
.532
.030
.355
.368
.001
.355
.368
-The highest ratio of the acceleration values between the two spectra is.532/.386
= 1.38 and is used to scale up the component stresses by this factor.
-See table B-3, item 4 for cabulation of results.
These results show that all of the component new stresses are less than the allowable stress. Therefore, this item is acceptable.
-Strainer ISX01FA,FB
-Original analysis uses a dynamic analysis.
In comparing the Procurement Specification Response Spectra with the Revised Response Spectra, a scale factor of 1.23 (shown below) is used to amplify the original stresses.
C in C in Period Revi' sed Original (SEC)
Spectra Analysis Ratio
.047
.373
/
.30g 1.23
=
The highest stress of 706 psi occurs in the lower gearmotor mount bolt. This is less than the allowable stress.
Therefore, this equipment is acceptable.
B-7
COUIPMENT EVALUATION RESULTS Shutdown Service Water Equipment System (cont)
Number Status
-The peak acceleration at the Limitorque operator mounting is also scaled up as follows:
)
(.728g) (1.23) =.895g horizontal, and
(.970g) (1.23) = 1.19g vertical
-These acceleration values are used to qualify the Limitorque operator. (see next j
item)
Limitorque operator ISX01FA.FB
-Vendor report states that the operator j
for strainer is qualified for acceleration values up to 5 i
g's.
{
-The scaled up acceleration values, according to the revised response spectra, are shown above. These values are less than the qualified acceleration of 5 g's.
Therefore, this equipment is acceptable.
Strainer ISX01FC
-Original analysis uses a dynamic analysis.
-The input accelerations are the'.ame as s
i those of the previous items (ISX01FA.FB) and therefore have the sama ratio of 1.23.
-The component stresses are scaled up by this ratio and are tabulated in item 5, i
table B-3.
These stresses are all less than the allowable stress. Therefore, this item j
is acceptable.
-The maximum accelerations at the Limitorque actuator mounting are scaled up as follows:
(.716 ) (1.23) =.881g horizontal 3
(.886g) (1.23) = 1.09g vertical These values are used to qualify the Limitorque actuator. (see next ites)
-Limitorque actuator ISX01FC
-Vendor Qualification Report states that the i
fsr strainer actuator is qualified for an acceleration of 3.0g in each orthogonal direction and has a fundamental natural frequency of greater j
than 33 Hz.
i 1
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Shutdown Service Water Equipment System (cont)
Number Status
-At a frequency of 33 Hz the Revised Responsa Spectra indicates the following acceleration values:
H
=H
= 0.26g g
2 V = 0.24g Since these values are less than those used for qualification the actuator is acceptable.
-Instrument Panels ISX11J,12J,13J
-Critical member stresses have been evaluated based on the Revised Response Spectra.
This evaluation indicates that the equipment is acceptable.
-Table B-3 items 6 and 7 contain a stress summary and a discussion of the method used to calculate the revised stresses.
Diesel Generator System
-Division I & II IDG01KA, KB
-Division I and II diesel engines supplied Emergency Diesel by General Motors Electro Motive Division Generator Sets (EMD), Division I and 11 generators supplied by Ideal Electric and Division I and II mechanical components have been assessed based on the Revised Response Spectra.
Table B-2 contains a stress summary and a discussion of the method used to calculate the revised stresses.
-Electrical components and Instruments associated with the Diesel Generator Sets have been evaluated based on a comparison of the Revised Response Spectra to the Emergency Diesel Generator System (EDGS)
Owner's Group required response spectra.
The Owner's Group required response spectra envelopes the Revised Response Spectra for all frequencies. Therefore, the original qualification is valid.
See Figures B-40, B-41 and B-42 for response spectra comparisons.
1 I
B-9
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Diesel Generator Equipment System (cont)
Number Status
-Diesel Generator IDG0lJA, JB
-Lower bound Test Response Spectra envelope and Transformer IPL12JA JB Revised Response Spectra except in the range Panels of 2.5 to SHz. However, review of the test report indicates that the equipment has a fundamental natural frequency of 15Hz, and will not respond in the range which has not been enveloped.
Therefore, the equipment is acceptable.
-Comparison of the response spectra is shown t
in Figures B-43. B-44 and B-45.
-EPCS Diesel 1H22-PO28
-Lower bound Test Response Spectra envelope Protection Panel I
the Revised Response Spectra for all I
frequencies. Therefore, the equipment is I
acceptable.
-Refer to Figures B-46, B-47 and B-48 for a comparison of the response spectra.
{
-Division III Diesel Awaiting Information.
j Generator supplied by Beloit Power Systema 4
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LAP 07E, 9E l4160V Switchgear Assemblies)
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1DG0lJA, JB j
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(Diesel Generator and Transformer Panels) i Diesel Generator Bldg. Elev. 737' Horizontal (E-W) Spectra for Service Level C 44 Damping
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B-54
IDG0lJA, JB 1PL12JA, JB (Diesel Generator and Transformer Panel)
Diesel Generator Bldg. Elev. 737' Vertical Spectra for Service Level C 4 Damping FREQUENCY IN CPS 500.0 200.0 100.0 50.0 20.0 10.0 5.0 2.0 1.0 0.5
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3-55 l
1H22-P028 (HPCS Diesel Protection Panel)
Diesel Generator Bldg. Elev. 737' i
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Figure B-46 B-$6
lH22-P028 (HPCS Diesel Protection Panel)
Diesel Generator Bldg. Elev. 737' Horizontal (E-W) Spectra For Service Level C 4% Damping FRECUENCY IN CPS 500.0 20C.0 100.0 50.0 20 0 10.0 5.0 2.0 10 0.8 f f ? f
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i Figure B-47 i
B-57 l
1H22-P028 (HPCS Diesel Protection Panel)
Diesel Generator Bldg. Elev. 737' Vertical Spectra for Service Level C 4: Damping FREQUENCY IN CPS 500.0 200.0 100.3 80.0 20.0 10 0 5.0 2.0 1.0 0.5 9 ? *
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1 i
i 1
L j
TABLE B-1
+
STRESS COMPARISIONS FOR MISCELLANEOUS EQUIPMENT 1
Ratio of Max. Stressed Revised Stress Allowable Revised /
Item Equipment Bldg /Elev.
Components Level (KSI)
Stress (KSI)
A11bwable 1
1 Batteries & Racks Control /781'
-Foundation Bolts 25.0(Tennile) 26.6 0.94 j
(IE22-S00tD) 4.5(Shear) 13.6 0.33 lI
-Battery Base 11.2 32.4 0.35 l
l
-Frame Section 14.9 32.4 0.46 I
-Brace Section 14.3 21.6 0.66 j
I
{
2 RHR Pumps Aux-Fuel /712' -Discharge Head Shell 31.1 31.5 0.99 j
j (IE12-C002A,B,C)
-Discharge Tee 17.5 18.0 0.97 r
-Discharge Column
,11.6 18.0 0.67 3
-Discharge Column A1.3 45.0 0.92 i
Flange & Bolting j
-Discharge Head Flange 35.3 45.0 0.78 l
j
-Pump Top Case 40.2 45.0 0.89 i
-Discharge Support Rib 18.2 31.5 0.58 l
-Pump Shaft 10.5 17.2 0.61 l
)
i j
3.
RNR Heat Exchangers Aux /737'
-Upper Support to Shell 23.5 (SRSS Stress) 26.25 0.97
)
(IE12-B001A,5)
-Lower Support to Shell 31.5 (ABS Stress) 31.50 0.99 q
(Pt.A)
[
]
-Lower Support to Shell 27.5 (SRSS Stress) 31.50 0.87 j
(Pt.C) f
-Tubesheet Flange 38.8 (SRSS Stress) 39.38 0.99
)
i 4
j See Attachment for Table B-1 for derivation of these stress values.
c i
I I
r I
I I
B-59 l
i ATTAC!DfENT FOR TABLE B-1
{
Derivation of Revised Stresses For the RilR Heat Exchangers (E12-B001 A.B).
l Lower Support to Shell Stress at pt. C.
Combined Emergency Level SRSS Stress is 26.5 kai. Maximum load elements from the stress report are 6, 11, 18 and 23. The maximum element stress among these elements will be used as the seismic stress, scaled up by the factor of '.20, and then added to the combined stress value, as calculated below:
Element No.
Maximum Stress (kni) l 6
3.88 11 4.89 - maximum value to be used in the calculation.
18 23 4.69 Therefore, scaled up combined stress = (26.5-4.89) + (1.2) (4.89) l
= 27.5 kai Tubesheet Flange Hub Stress.
With the same approach as above, the scaled up stress is calculated as follows:
(Emergency Level) j Scaled up combined stress = (37.8-4.89) + (1.2) (4.89) = 38.8 kai
]
1 B-60 L
TABLE B-2 1
DIESEL CENERATOR SERVICE LEVEL C STRESS COMPARISON WITil LEVEL C ALLOWABLE VALUES Max. Stressed Service Level C Revised Service Allowable Revised /
Item Equipment Bldg /Elev.
Components Stress (KSI)
Level C Stress (KSI) Stress (KSI) Allowable I
Div.I & Div.II.
Dieself
-Air Start Piping 2.9 27.0 0.11 Diesel 737'
-Aux Lube 011 12.6 27.0 0.47 Cenerator System Piping System
-Div.I & II Cenerator a) Rotor & Shaft 2.55 3.29 (1)
(2) b) Bearing Bracket 3.81 4.91 (1)
(2)
.l c) Stator 7.28 9.39 (1)
(2)
J d) Exciter Mounting 0.76 0.98 (1)
(2) e) Conduit Mounting 0.50 0.65 (1)
(2)
I f) Base 1.60 2.06 (1)
(2)
-Div.I Cenerator a) Cenerator to 26.28 (Tension) 33.90 (1) 103.90 0.33 Base Bolts 8.30 (Shear) 10.71 (1) 51.95 0.21 b) Stator Frame to 1.01 (Tension) 1.30 (1) 47.20 0.03 Bracket Bolts 6.79 (Shear) 8.76 (1) 23.60 0.37
-Div.II Cenerator a) Cenerator to 25.08 (Tension) 32.35 (1) 103.90 0.31 l
Base Bolts 7.50 (Shear) 9.68 (1) 51.95 0.19 1
b) Stator Frame to 0.83 (Tension) 1.07 (1) 47.20 0.02 I
Bracket Bolts 6.14 (Shear) 7.92 (1) 23.60 0.34 I
r
-16 Cylinder Engine a) Accessory Rack 7.75 10.0 (1)
(2)
Frame b) Lube Oil Filter 10.78 13.91 (1)
(2) c) Thermostatic Valve 5.01 6.46 (1)
(2)
Mounting B-61 i
w r-
TABLE B-2 (cont'd)
Max. Stressed Service Level C Revised Service Allowable Revised /
Item Equipment Bldg /Elev.
Components Stress (KSI)
Level C Stress (KSI) Stress (KSI) Allowable 1
Div.I & Div.II.
Diesel /
-16 Cyl. Eng.(cont)
Diesel Genera-737' d) Cooling Water 0.70 0.90 (1)
(2) tor System Piping (Cont'd) e) Water Expansion 5.04 6.50 (1)
(2)
Tank f) Base 2.19 2.83 (1)
(2) g) IIcat Exchanger 13.97 18.02 (1)
(2) h) Engine to Base 13.83 (Tension) 17.84 (1) 47.20 0.38 Bolts 6.93 (Shear) 8.94 (1) 23.60 0.38 i) Accessory Rack 1.47 (Tension) 1.88 (1) 35.95 0.05 to Base Bolts 0.67 (Shear) 0.86 (1) 17.98 0.05 J) fleat Exchanger 14.75 (Tension) 19.03 (1) 47.20 0.40 to Base Bolts 4.01 (Shear) 5.17 (1) 23.60 0.22 k) 011 Filter to 1.05 (Tension) 1.35 (1) 28.40 0.05 Rack Bolts 0.45 (Shear) 0.58 (1) 14.20 0.04
- 1) 011 Cooler to 3.39 (Tension) 4.37 (I) 12.05 0.36 Back Bolts 2.76 (Shear) 3.56 (1) 6.03 0.59 m) Water Expn. Tank O.75 (Tension) 0.97 (1) 35.95 0.03 to Rack Bolts 0.66 (Shear) 0.85 (1) 17.98 0.05
-12 Cylinder Enginer a) Accessory Rack 6.78 8.75 (1)
(2)
Frame b) 1.ube 011 Filter 9.83 12.68 (I)
(2) c) Thermostatic 1.38 1.78 (1)
(2)
Valve Mounting d) Cooling Water 2.86 3.69 (1)
(2)
Piping e) Water Expansion 4.43 5.71 (1)
(2)
Tank f) Base 1.86 2.4 (1)
(2) g) Heat Exchanger 11.31 14.59 (1)
(2) h) Engine to Base 9.08 (Tension) 11.71 (1) 47.20 0.25 Bolts 5.65 (Shear) 7.29 (1) 23.60 0.31
- 1) Accessory Rack to 1.29 (Tension) 1.66 (1) 35.95 0.05 Base Bolts 0.61 (Shear) 0.79 (1) 17.98 0.04 B-62
TABLE B-2 (cont'd)
Max. Stressed Service Level C Revised Service Allowable Revised /
Item Equipment Bldg /Elev.
Components Stress (KSI)
Level C Stress (KSI) Stress (KSI) Allowable j) llent Exchanger 11.99 (Tension) 15.47 (I) 47.20 0.33 to Base Bolts 2.91 (Shear) 3.75 (I) 23.60 0.16 k) 011 Filter to 0.96 (Tension) 1.24 (1) 28.40 0.04 Bolts 0.44 (Shear) 0.57 (1) 14.20 0.04
- 1) 011 Cooler to 2.22 (Tension) 2.86 (1) 12.05 0.24 Rack Bolts 2.38 (Shear) 3.07 (1) 6.03 0.51 m) Water Expn. Tank 0.75 (Tension) 0.97 (1) 35.95 0.03 to Rack Bolts 0.66 (Shear) 0.85 (1) 17.98 0.05 NOTES:
(1) Refer to Attachment for Table B-2 for derivation of revised Service Level C Stresses.
(2) The Seismic Qualification package states that the resultant Service Level C Stresses are below design limits for these items, but does not provide specific allowable limits. However, the resultant stresses are low by inspection, therefore, the equipment is acceptable.
4
?
e B-63
I 1
ATTACHMENT FOR TABLE B-2 1
i i
Derivation of revised service Level C Stresses for the Emergency Diesel Generator Sets.
4 i
The Emergency Diesel Generator Sets are qualified in package SQ-CLO27 for the following peak accelerations:
H 0.85g
=
g H
0.89g SSE, 3% Damping
=
y 1,
V 4.70g
=
From the revised response spectra at 3% damping, the revised seismic coefficients are obtained:
R L*18 NS H
1.lg
=
gg V
2.5g
=
Considering the " Worst Case" ratio of Revised / Qualified:
1.lg/0.85g 1.29g
=
j Therefore, the existing service Level C Stresses are multiplied by 1.29 in order to obtain the revised service Level C Stresses.
I i
l i
4
+"
i 3-64 l
m
~
4 TABLE B-3
)
SHUTDOWN SERVICE WATER SERVICE LEVEL C STRESS COMPARISON 1
WITH LEVEL C ALLOWABLE VALUES i
1 Max. Stressed Service Level C Revised Service Allowable Revised /
Item Equipment Bldg /Elev.
Components Stress (KSI)
Level C Stress (KSI), Stress (KSI) Allowable l.
Shutdown Screen Anchorage 12.156 14.10 20.0
.705 Se rvice House /
System Water 699'-0" Bolt Pump Motor ISX0lPC 4
l 2.
Shutdown Screen Column 1.32 1.32 15.0
.088 i
Service House /
i Water 699'-0" Shaft 1.324 1.324 15.0
.0883 i
Pump ISX0lPC Discharge 5.045 5.045 15.0
.3363 j
Nozzles g
3.
Shutdown Screen Anchorage 3.153 3.153 20.0
.1577 Service House /
System i
l Water 699'-0" Bolt Pump Motor ISX0lPA,B 4.
Shutdown Screen Column Flange 29.632 40.89 41.25
.9913 j_
Service House /
Bolt Water 699'-0" Pump Case Bowl 8.80 12.14 14.0
.8674 ISX0lPA,B i
Case Flange
'25.08 34.61 41.25
.839 Bolt Discharge 27.489 37.93 41.25
.9195 Flange Bolt 1
B-65 i
I i
l TABLE B-3 (Cont'd)
Max. Stressed Service Level C Revised Service Allowable Revised /
Item Equipment Bldg /Elev.
Components Stress (KSI)
Level C Stress (KSI) Stress (KSI) Allowable i
1 l
5 Strainer Screen Base Plate 14.949 18.39 22.5
.817 ISX0lFC House /
l 699'-0" Cearmount 3.558 4.376 7.00
.625 Bolting l
6 Shutdown Screen /
-Panet Support Beam 4.98(Tension) 4.98 28.8 0.17 l
j Service house 2.17 (Shear) 2.17 19.2 0.11 Water 699'-0"
-Panel Plate 0.24 0.24 28.8 0.01 i
inst rument
-Weld to Strainer 10.1 10.1 24.0 0.42
[
j Panels
-Plate Weld 1.2 1.2 24.0 0.05 l
- ISKIIJ,
-Pressure Indicator 0.06 0.06 22.5 0.003 ISX123 Mounting Bolts
-Pressure Switch 1.48 1.48 22.5 0.07 j
Mounting Bolts
)
i i
7 Shutdown Screen /
-Panel Support Beam 10.14(Tension) 13.28 28.8 0.46 Service house 7.27(Shear) 9.52 19.2 0.50 Water 699'-0"
-Panel Plate 0.14 0.18 28.8 0.01 i
Instrument
-1/8" Fillet Weld 4.29 5.62 24.0 0.23 l
l Panel
-1/4" Bolt on Instr.
16.86 22.1 25.52 0.87 l
l ISK13J Mounting Plate i
-Pressure Indicator 0.10 0.14 25.52 0.01 l
}
Mounting Bolts
[
1
-Pressure Switch 1.18 1.55 25.52 0.06 L
1 Mounting Bolts I
l t'
1 i
i i
i i
- Refer to Attachment for Table B-3 for derivation of Revised Service Level C Stresses.
I I
-1 l
+
i l
1 I
l B-66 I
e lw--.. -
ATTACHMENT FOR TABLE B-3 Derivation of revised Service Level C stresses for shutdown service water instrument panels.
I)
Panels ISX11J. ISX12J Panels ISX11J and ISX12J are mounted on strainers ISX01FA and ISX01FB respectively. Based on SQ-CL200, the strainers are identical and have natural frequencies of 26.2 Hz and 32.6 Hz.
Package CQ-CL237 determines a panel natural frequencies of 79 Hz.
Since the panels are rigid and the strainers have a flexible 1st mode (< 33 Hz), the strainer response is used in order to determine revised seismic coefficients and service Level C stresses.
Revised Coefficients:
At f 26.2 Hz At f
=
32.6 Hz
=
g 3
G 0.28g G
=
0.25g
=
y y
G 0.30g C
=
0.26g
=
NS NS G
0.34g G
=
0.26g
=
EW EW Combining Accelerations by the SRSS Method y = ((0.28)2 + (0.25)21 = 0.38g G
C = [(0.34)
+ (0.26) ]b = 0.43g (Worse case between NS & EW)
H SQ-CL237 considered the following seismic coefficientat G
0.764g
=
y CH1 0.668g
=
GH2 0.668g
=
==
Conclusion:==
The revised seismic coefficients are less than those used in the original panel analysis. Therefore, 1SX11J and ISX12J are acceptable.
II) Panel ISX13J (Mounced on Strainer ISX01FC)
SQ-CL201 provides qualification for strainer ISX01FC and deter-mines natural frequencies of 15.942 Ils and 46.728 Hz. The panel was found to have a natural frequency of 60.58 11: (Per SQ-CL238).
B-67
l ATTACHMENT FOR TABLE B-3 (cont'd)
Since this situation is the same as in Item I. the sane approach is used to determine revised seismic coefficients.
Revised Coefficients:
At f 15.94 Hx At f 46.728 Hz
=
=
g 2
C 0.43g C
0.24g
=
=
y y
C
'0 8 I
NS NS 0.45g o
C 0.25g
=
=
g g
Combining Accelerations by the SRSS Method Cy = [(0.43) + (0.24)2]
= 0.49g C = [(0.62) + (0.26) h = 0.67g (Worse case between NS & W)
H Considering the Worst-case' ratio of (Revised / Qualified) Seismic Coefficients:
0.673 Horizontal:
/0.512g = 1.31 0.493 Vertical :
/0.7943 = 0.62 Therefore, we will use a factor of 1.31 to determine service Level C Stresses for Panel 1SX13J.
8-68 N
- - -