ML20072F366
| ML20072F366 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/31/1993 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19304C524 | List: |
| References | |
| WCAP-13787, WCAP-13787-R01, WCAP-13787-R1, NUDOCS 9408230311 | |
| Download: ML20072F366 (78) | |
Text
HESTINGHCUSE CLASS 3 (N n-Proprietary)
WCAP-13787 Revision 1 ELIMINATION OF PRESSURE SENSOR RESPONSE TIME TESTING REQUIREMENTS WOG Program MUHP-3040 Revision 1 December, 1993 Westinghouse Technical Specification Program Services l
WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division l
P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 9408230311 940917 PDR ADOCK 05000348 P
/
- 1993 Westinghouse Electric Corporation All Rights Reserved
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l LEGAL NOTICE "This report was prepared by Westinghouse as an account of work sponsored by the Westinghouse Owners Group (WOG). Neither the WOG, any member of the WOG, Westinghouse, nor any person acting on behalf of any of them:
(A)
Makes any warranty or representation whatsoever, express or_ implied, (I) with respect to the use of any information, apparatus, method, process, or similar item disclosed in this report, including merchantability and fitness for a particular-purpose, (II) that such use does not infringe on or interfere with privately owned rights, including 'any party's intellectual property, or (III) that this report is suitable to any particular user's circumstance; or (B)
Assumes responsibility for any damages or other liability whatsoever (including any consequential damages, even if the WOG or any WOG representative has been advised of the possibility of such damages) resulting from any selection or use of this report or any information apparatus, method, process, or similar item disclosed in this report."
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l ACKNOWLEDGEMENTS The following personnel are recognized for their extended efforts in the preparation of this report and for their team participation and support in the validation of material for this report.
C. E. Morgan Westinghouse Electric Corporation R. L. Bencini Westinghouse Electric Corporation R. B. Miller Westinghouse Electric Corporation J. D. Campbell Westinghouse Electric Corporation T. A. Lordi Westinghouse Electric Corporation J. L. Zielinski Westinghouse Electric Corporation L. Bush Commonwealth Edison Company - WOG H. Pontious Commonwealth Edison Company - WOG M. Eidson Southern Nuclear Operating Company - WOG T. A. Green General Electric Company D. Spencer Commonwealth Edison Company - BWROG i
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TABLE OF CONTENTS Section Title Pace 1.0 Executive Sumary 1-1 2.0 Introduction 2-1
3.0 Background
3-1 4.0 Sumary of Electric Power Research Institute (EPRI) 4-1 Report NP-7243 Revision 1, " Investigation of Response Time Testing Requi rements".
4.1 Program Organization 4-1 4.2 Program Objectives 4-2 4.3 Plant Data collection and Assessment 4-3 4.4 Failure Modes and Effects Analysis (FMEA) 4-3 4.5 Summary of EPRI Recomendations 4-4 5.0 Westinghouse Analyses 5-1 5.1 Barton Model 752 Similarity Analysis Report 5-1 5.2 Barton Model 332 Similarity Analysis Report 5-7 5.3 Barton Model 351 Sealed Sensor System Analysis Report 5-13 5.4 Barton Model 763A Similarity Analysis Report 5-17 5.5 Foxboro Model E11GM Similarity Analysis Report 5-20 5.6 Foxboro Model N-E11AH Similarity Analysis Report 5-23 5.7 Tobar Model 32DP2 Similarity Analysis Report 5-25 5.8 Tobar Model 32PA2 Similarity Analysis Report 5-30 5.9 Veritrak Model 76DP1 Similarity Analysis Report 5-35 5.10 Veritrak Model 76PG1 Similarity Analysis Report 5-40 5.11 Veritrak Model 76PH2 Similarity Analysis Report 5-45 6.0 Safety Benefits 6-1 7.0 Cost Benefits 7-1 iii
8.0 Safety Assessment For Increased Response Times Beyond 8-1 Current Technical Specification Limits 9.0 Program Methodology 9-1 10.0 Technical Specifications 10-1 11.0 Conclusions 10-1 12.0 References 10-1 List of Appendices AJoendix Title fagg A
Technical Specification Markups A-1 B
Sample No Significant Hazards Evaluation (50.92)
B-1 List of Figures Fiaure Title fagg a
Figure 5.1-1 Barton Model 752 DPU 5-5 Figure 5.1-2 Barton Model 752 Block Diagram 5-6 Figure 5.2-1 Barton Model 224 Differential Pressure Unit 5-10 l
Figure 5.2-2 Barton Model 224 Mechanical Operation 5-11 Figure 5.2-3 Barton Model 332 Block Diagram 5-12 Figure 5.3-l' Typical Barton Model 351 Sensor Assembly 5-16 l
Figure 5.4-1 Barton Model 763A Block Diagram 5-19 Figure 5'5-1 Foxboro Model E-11GM Block Diagram 5-22 Figure 5.6-1 Foxboro Model N-E11AH Block Diagram 5-24 Figure 5.7-1 Tobar Model 320P2 Amplifier Block Diagram 5-28 Figure 5.7-2 Tobar Model 320P2 Capsule Assembly 5-29 Figure 5.8-1 Tobar Model 32PA2 Amplifier Block Diagram 5-33' Figure-5.8-2 Tobar Model 32PA2 Capsule Assembly 5-34 Figure 5.9-1 Veritrak Model 76DP1 Amplifier Block Diagram 5-38 Figure 5.9-2 Veritrak Model 76DP1 Capsule Assembly 5-39 iv
List of Figures Fiaure Title Paae Figure 5.10-1 Veritrak Model 76PG1 Amplifier Block Diagram 5-43 Figure 5.10-2 Veritrak Model 76PG1 Capsule Assembly 5-44 Figure 5.11-1 Block Diagram of the AC Amplifier for Veritrak 5-47 Model 76PH2 Figure 5.11-2 Veritrak Model 76PH2 Capsule Assembly 5-48 List of Tables Table Title Pace 9-1 Sensor Response Times 9-2 J
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1.0 EXECUTIVE
SUMMARY
WOG Program MUHP-3040, Revision 1, deals with an effort by the industry to eliminate the specific Technical Specifications requirement to perform periodic response time testing on pressure and differential pressure sensors in Reactor Trip System and Engineered Safety Features Actuation System instrumentation loops. These pressure type sensors are utilized for flow, level, and pressure process measurement functions.
Specifically, MUHP-3040 Revision 1:
o Utilizes the reconnendations contained in EPRI Report NP-7243 Revision 1, " Investigation' of Response Time Testing Requirements" for justifying elimination of response time testing surveillance requirements on certain pressure and differential pressure sensors identified in the report.
o Contains similarity analysis to sensors in the EPRI report of sensors not identified in the report to establish justification for elimination of response time testing requirements for those sensors.
o Qualifies as a Cost Beneficial Licensing Action (CBLA) as described in section 7.0.
o Provides reasonable assurance that elimination of the specified response time testing requirements does not create a significant impact on the safety of the plant as described in section 8.0 o
Provides a methodology for substituting response times in lieu of values obtained from response time testing for each sensor covered by this WCAP.
WOG program MUHP-3041, scheduled for completion in-April of 1994, is complimentary to MUHP-3040. MUHP-3041 is being conducted to eliminate the specific requirement to perform periodic response time testing on the process instrumentation and logic portions _ of the Reactor Trip System and Engineered Safety Features Actuation System' instrumentation loops.
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2.0 INTRODUCTION
Response Time Testing (RTT) of Reactor Trip System (RTS) instrumentation and Engineered Safety Features Actuation System (ESFAS) instrumentation has been required by Technical Specifications since the mid 1970's.
The purpose of the RTT was to demonstrate that the instrumentation met the response time performance requirements assumed in the plant safety analyses.
Since its inception, RTT has proven to be resource intensive.
RTT is generally performed in discrete steps, with the sensor response time being one of the steps.
RTT of sensors is especially expensive, since many of the tests require special equipment and technical skills in addition to extensive test times. Furthermore, many of the sensors are located in radiation areas and substantial man rem exposure is incurred by plant maintenance staffs. A data review conducted by the EPRI has shown that RTT has not detected response time failures. This can be attributed in a large part to the fact that other surveillance (calibration) is typically performed first and has discovered failures that would affect response time. As a result, EPRI initiated a program to determine if RTT requirements could be eliminated for specific pressure and differential pressure transmitters and switches.
The results of the EPRI program are delineated in EPRI Report NP-7243 Revision 1 (Ref. 1).
This WCAP provides the technical justification for deletion of periodic RTT of selected pressure sensing instruments using the EPRI report reconnendations and additional analysis by Westinghouse for selected sensors not addressed by the EPRI report.
An example "No Significant Hazards Consideration" evaluation is included as Appendix B to aid the utilities in preparation of a plant specific license amendment request.
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3.0 BACKGROUND
In 1975 RTT requirements were added to the Westinghouse Standard Technical Specifications. As a result, all plants licensed after 1975 had to include RTT in routine surveillance testing. The first plant required by Technical Specifications to perform RTT was D. C. Cook Unit 2.
The Standard Technical Specifications contain definitions for both Reactor Trip System and Engineered Safety Features Actuation System Response Times.
The response time definitions are:
"The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage."
"The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach 1
their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable."
The Bases section states that the response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured. This approach is also consistent with ISA Standard 67.06. Given this guidance and the complexity of testing an entire instrument channel from the sensor to the final device, plant surveillance procedures typically test a channel in two or more steps.
One individual step in most plant test methodologies is the instrument sensor. Separate procedure (s) using specialized test equipment and/or outside vendors are typically used solely for sensor testing.
The first industry RTT guidelines were established by the Institute of Electrical and Electronic Engineers in ANSI /IEEE Standard 338-1975,
" Criteria for the Periodic Testing of Class 1E Power and Protection Systems."
In 1977 this Standard was revised and accepted by the NRC as documented by NRC Regulatory Guide 1.118, " Periodic Testing of Electric Power and Protection Systems," Revision 1.
Following Revision 2 of Regulatory Guide 1.118, the Instrument Society of America approved Standard ISA S67.06, " Response Time Testing of Nuclear Safety-Related Instrument Channels in Nuclear Power Plants," August 29, 1986. These guidelines have also been used by industry when developing plant specific test procedures.
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4.0 Summary of EPRI Report NP-7243 Revision 1," Investigation of Response Time Testing Requirements".
In May of 1991, EPRI issued a report " Investigation of Response Time Testing Requirements", Report No. NP-7243.
EPRI is publishing revision 1 to this report to address additional comments by the industry (Ref.
1).
The EPRI report provides recomendations for reducing or eliminating RTT requirements which apply to typical pressure and differential pressure sensors used in safety-related protection system instrumentation in the nuclear power industry.
The primary purpose of the EPRI investigation was to determine if deleting RTT could be justified for specific pressure, level, and flow sensors.
IEEE Standard 338-1977 defines a basis for eliminating RTT.
Section 6.3.4 states in part:
" Response time testing of all safety-related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety system equipment is verified by functional testing, calibration check, or other tests, or both."
In addition, the standard states:
"This is acceptable if it can be demonstrated that changes in response time beyond acceptable limits are accompanied by changes in perfonnance characteristics which are detectable during routine periodic tests."
4.1 Program Organization The EPRI report was developed through the efforts of EPRI, the RTT Utility Advisory Committee (UAC), 39 nuclear power plants, 6 pressure sensor manufacturers, Science Applications International Corporation (SAIC) and Performance Associates, Inc.
EPRI coordinated the program and acted as technical director and reviewer. The RTT UAC acted as advisors on the formulation of the project and as reviewers during the development of the report.
The 39 plants provided detailed response time data used in the investigation to determine failure trends, test methods and data repeatability. The 6 manufacturers, which supplied technical infonnation regarding the design and performance of their product (s), represented the majority of manufacturers currently providing qualified pressure sensors to the nuclear industry. SAIC and Performance Associates, Inc. performed the RTT investigation and developed the report.
Rosemount FMEA data included in the report was provided by the EPRI Project Manager.
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4.2 Program Objectives In support of industry efforts to improve plant availability and reduce plant personnel exposure levels, EPRI established a program to. determine whether or not requirements for RTT of specific pressure and differential pressure transmitters and switches could be eliminated.
The investigation included only the sensor since the premise for deleting RTT of the sensor was not directly applicable to.the other components in the instrument loop.
Therefore, the conclusions and recommendations of the EPRI report are applicable only to the sensor portion of the instrument loop.
The EPRI program was devised to detemine: 1) how RTT pe'rforms as a unique indicator of sensor (pressure and differential pressure) response time degradation; 2) sensor failure modes, if _any exist, which result in response time degradation that would not be detected by other periodic (non-RTT) testing methods; 3) the level of redundancy between RTT.and other periodic tests; and, 4) the RTT methods best suited to detect, where necessary, response time degradation.
Assessment of plant response time data and performance of FMEAs on sensor hardware provided the mechanisms for obtaining the necessary infonnation.
For the plant data assessment, evaluations were performed to establish any similarities and differences in response time testing 3
at selected plants and to determine the repeatability of RTT measurements. Sensor types were identified as either having failed response time testing, or as trending toward failure. Root causes of known sensor failures were evaluated to determine if response time degradation was the key indicator of the failure.
Additionally, the use of response time measurement data to identify sensors that were trending toward response time failures was evaluated.
FMEA considerations included the evaluation of sensor failure modes which could affect response time, the ways that sensors which are experiencing response time degradation exhibit other operability changes, and the identification of routine plant tests that essentially provide the same information as RTT for each response time failure mode.
4.3 Plant Data Collection and Assessment For the EPRI investigation, 39 nuclear power plants supplied over 4200 RTT measurements including information regarding sensor type, sensor use, instrument range, trip setpoint, allowable response time, test method, and type of test system.
The-plant data covered all but one manufacturer (Statham) supplying safety-related pressure and differential pressure sensors to the nuclear industry.
EPRI used this data to determine failure trends and evaluate test methods and data repeatability.
Six manufacturers, including Statham, supplied EPRI with technical information regarding design and performance of their 4-2
products. Sensor failure modes affecting response time were evaluated to the guidelines provided in ANSI N41.4-1976/IEEE Std 352-1975 "IEEE Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection Systems" to confirm that response time failures could be detected by surveillance tests other than specific RTT.
Based on the 4200 measurements supplied, the EPRI investigation found "RTT has not identified any sensors that have failed response time requirements using hydraulic or electronic white noise analysis -
techniques.
However, calibrations and other tests have detected transmitters with excessive response times".
The EPRI report concludes that the current RTT program for pressure and differential pressure sensors adds very little to the identification of failed sensors and verification of loop response times.
With a few exceptions (noted in Section 4.5), the existing instrumentation surveillance requirements, such as channel checks and sensor calibrations, have proven effective in identifying failed sensors in a timely manner.
4.4 Failure Modes and Effects Analysis The EPRI report included FMEAs on the following sensor-types:
Barton 288/289 Differential Pressure Indicating Switches Barton 763 Gauge Pressure Transmitter Barton 764 Differential Pressure Transmitter Foxboro/ Weed N-E11DM Differential Pressure Transmitter Foxboro/ Weed N-E13DM Differential Pressure Transmitter Foxboro/ Weed N-E130H Differential Pressure Transmitter Foxboro/ Weed N-E11GH Gauge Pressure Transmitter Foxboro/ Weed N-E11GM Gauge Pressure Transmitter Tobar 32PA1 Absolute Pressure Transmitter Tobar 32PG1 Gauge Pressure Transmitter Tobar 32DP1 Differential Pressure Transmitter Rosemount Differential Pressure Transmitter Models 1151, 1152, 1153, 1154 Rosemount Pressure Transmitter Models 1151, 1152, 1153, 1154 Statham PD-3200 Differential Pressure Transmitter Statham PG-3200 Pressure Transmitter 50R Differential Pressure Switch SOR Pressure Switch.
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These sensor types were selected as being representative of the pressure and differential pressure sensor instrumentation being utilized in safety-related systems and are representative of the various sensor designs (e.g., bourdon tube, force-balance, capacitance and strain gauge).
The FMEA method of analysis provided a systematic approach for identifying failure modes.
The FMEAs permitted the identification and analysis of failure modes associated with each principal design component of the pressure and differential pressure sensors that could affect response time.
4.5 Sumary of EPRI Recomendations The EPRI report provides recomendations for modifying the current RTT program for pressure and differential pressure sensors in the nuclear utility industry. These are recomendations to consider when enhancing or upgrading existing plant RTT programs and are not intended to require changes to current plant RTT programs.
The report also provides the basis for deletion of all periodic pressure and differential pressure sensor response time testing requirements subject to following exceptions / limitations:
1.
Perform hydraulic response time test prior to installation of new transmitter / switch or following refurbishment.
2.
For transmitters and switches that use capillary tubes, RTT should be performed after initial installation and after any maintenance or modification activity that could damage the capillary tubes.
3.
Perform periodic drift monitoring on all Rosemount pressure and differential pressure transmitters in accordance with Rosemount Technical Bulletins and NRC Bulletin 90-01 (affects certain model numbersonly).
Note: NRC Bulletin 90-01 has been superseded by NRC Bulletin 90-01 Supplement 1.
4.
Assure that variable damping (if used) is at the required setting and cannot be changed or perform hydraulic or white noise response time testing of sensor, following each calibration.
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5.0 Westinghouse Owners Group Analyses The primary objective of the following similarity analysis is to provide justification for deleting Technical Specifications requirements for RTT on pressure and differential pressure sensors not included in EPRI Report NP-7243.
In 1991 a survey was made of Westinghouse Owners Group plants to provide information identifying those pressure and differential pressure sensors currently in operation with periodic Technical Specifications RTT requirements. Following is a list of pressure. and differential pressure sensors identified from that survey which are in addition to those evaluated in the EPRI report:
Barton 752 Differential Pressure Transmitter Barton 332 Differential Pressure Transmitter Barton 763A Gauge Pressure Transmitter Barton 351 Sealed Sensor Foxboro/ Weed N-E11AH Absolute Pressure Transmitter Foxboro E11GM Gauge Pressure Transmitter Tobar 32DP2 Differential Pressure Transmitter Tobar 32PA2 Absolute Pressure Transmitter Veritrak 76DP1 Differential Pressure Transmitter Veritrak 76PG1 Gauge Pressure Transmitter Veritrak 76PH2 Absolute Pressure Transmitter.
Similarity analyses were utilized to compare the design and the functionality of the principal components of each pressure and i
differential pressure unit, to those evaluated in the EPRI report. For those instruments where similarity could not be shown, other techniques (FMEA, historical approach, circuit testing) were utilized to justify elimination of the RTT requirements. The respective sensor manufacturer has reviewed / approved each Westinghouse analysis contained herein.
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5.1 BARTON MODEL 752 SIMILARITY ANALYSIS REPORT 5.1.1 Descrintion
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6.0 Safety Benefits Reduction in the amount of response time testing requirements results in improvement in plant safety by:
Increasing the availability of equipment.
The response time tests typically require that instruments be removed from service in order to perform the test.
The time period can range several hours to several days depending on the type of test.
By eliminating or reducing unnecessary testing, the plant safety is improved.
Decreasing the possibility of unwanted engineered safety features actuations.
Some of the testing requires that equipment be placed in the trip condition.
Improper return to service or other malfunctions that occur while testing may lead to plant trips or inadvertent actuation of equipment.
By eliminating or reducing unnecessary testing, the plant safety is improved.
Decreasing radiation exposure.
Sensors are typically located in radiation areas to minimize the sensing line lengths.
Eliminating unnecessary testing on these sensors reduces exposures consistent with the guidelines of ALARA.
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4 7.0 Cost Benefits The EPRI report data indicated that RTT was both resource and exposure (ALARA) intensive.
The process noise method of RTT for pressure sensors is often performed by an outside vendor (AMS or Westinghouse) and can cost $30K - $45K per unit per outage for a 4 loop plant (based on recent bids at Zion and Braidwood Stations).
The hydraulic ramp generator test method of RTT, an alternate method to process noise and the usual method for those pressure transmitters with no process noise, can take up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per transmitter to perform (based on procedures used at Braidwood Station). An individual utility cost saving would depend on the test method (s) and plant procedures used and whether or not an outside vendor is used for some or all of the work.
The overall response time of a channel is generally obtained by combining the times from the Segments shown below:
HGURE 1 OVERLAP OVERLAP OVERLAP SEGMENT 1 SEGMENT 3 SEGMENT 2 SEGMENT 4 PROCESS SSPS OR FINAL DEVICE, SENSORS 7300, 7100, RELAY RODS FREE TO FALL, EAGLE LOGIC OR ESF COMPONENTS OBTAIN SAFETY P05.
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The manpower required to do response time testing on Segments 1, 2, and 3 shown above is estimated by the WOG to be between 900 and 2000 manhours.
Both the BWR Owners Group and EPRI estimated that the manpower required to do this response time testing could range between 1500 and 2600 manhours.
Using a conservative estimate, shown below, one can see a significant utility cost savings.
Manhours 900 (per 18 month cycle)
Cost /hr
$24.00 (includes benefits)
Success 75%
Rate
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900 hrs x $24.00/hr x 75%
$16,200.00
=
- Percent of total hours for RTT surveillance requirements eliminated 7-1
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8.0 Safety Assessment For Increased Response Times Beyond Current Technical Specification Limits The WOG programs outlined in Section 1.0 demonstrate the capability of the calibration surveillance to detect failures that would affect response time on sensors, signal conditioning, and logic equipment.
Typical allowances include 400 milliseconds for differential pressure sensors, 200 milliseconds for pressure sensors, and 100 milliseconds each for the signal conditioning and logic.
For most applications significant information exists in the form of test data from manufacturers and from the industry that pressure and differential
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pressure sensors have response times that could be a factor of ten below these allowances.
In addition, FMEAs (with test verification) have been performed on the signal conditioning and logic portions of the Westinghouse instrumentation and have shown with reasonable assurance that response times of the signal conditioning will not increase above the 200 milliseconds allowance.
Summing of the above more conservative response times results in total response times well within the values allowed in the Technical Specifications.
However, as added assurance, a maximum incredible response time for each reactor trip function was calculated by increasing the sensor, signal conditioning, and logic each by a factor of five and applying the root mean square method since the degradation of response time of each portion of the system would be independent.
The results of the root mean square method show that containment pressure measurement response time matches the Technical Specification and for level and flow functions, which utilize differential pressure transmitters, the highest response time calculated was 2.4 seconds.
In all cases, even without applying the root mean square method, the response time obtained for functions with pressure transmitters was still below the Technical Specification value.
Engineered Safety Features response times, where the required response time corresponds to the diesel start time, are excluded from surveillance by NRC Generic Letter 93-05.
Since LOCA and steamline break analyses utilize pressure and containment pressure measurements as primary functions and the SGTR analyses is conservatively performed assuming no delay, the only potential impact to the safety analyses
{
results involves transients that credit level and flow measurements as primary reactor trips or ESF functions with response times less than the i
diesel start time.
Low reactor coolant flow reactor trip is credited in the partial 4
loss of flow analysis and the locked rotor event.
With the margins assumed, the response time would increase from 1.0 to 2.4 seconds.
The partial loss of flow event is bounded by the complete
)
loss of flow analysis in terms of consequences and would l
continue to be bounded even with the projected time response increase.
This more limiting event satisfies condition _II criteria of no fuel failure.
8-1
i 8.0 Safety Assessment For Increased Response Times Beyond Current Technical Specification Limits (continued)
The locked rotor event is classified as a condition IV event
=
that allows fuel failure to occur.
With the projected increased response time, better estimate assumptions (e.g.
consistent power shape in the transient reactivity feedback and DNBR analysis, actual moderator temperature coefficient) would yield similar results.
The peak reactor coolant system pressure is expected to increase less than 2.0% and the maximum clad temperature and ZR-H O reaction criteria 2
would not be exceeded.
i High steam generator water level ESF functions are credited in the feedwater malfunction event. With the margins assumed including a
.25 second turbine stop valve closure time, the response time would still be below the Technical Specification value.
Lo-lo steam generator water level reactor trip is used in the loss of normal feedwater/ station blackout event, the feedline break j
even, the loss of load / turbine trip event, and the determination of mass / energy release outside containment.
With the margins assumed, the response time would increase from 2.0 to 2.4 seconds.
This trip function is not required for protection for a DNB condition, so no fuel failure would be expected due to the projected increase in response time. The effect of the increased response time on the DNB transient, pressurizer overfill, reactor coolant system subcooling, and the results of the mass / energy release outside containment would be negligible.
I 8-2
9.0 Program Methodology Each sensor that has been identified as a candidate for elimination of periodic response time testing requirements is listed in Table 9-1.
The l
response time to be allocated in place of response times obtained l
through actual measurement during the period of verification may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications.
There is no specific recommendation, although the value will be increasingly more conservative progressing through these methods.
Available values have been incorporated into the table. An explanation for use of Table 9-1 is provided below.
1.
Determine the type of sensor being used for each reactor trip or engineered safety features function that requires periodic response time verification per the plant Technical Specifications.
2.
Verify that these sensors are included in Table 9-1.
3.
Obtain the sensor response time from Table 9-1.
If the sensor I
response time is not provided in the Table, then neither the manufacturer nor Westinghouse currently provide this information.
F111 in the Baseline column using the most conservative data obtained from either previous plant insitu response time testing or, if replacing the transmitter, the response time obtained through testing.
4.
Ensure that the plant test procedures are written such that the sensor response time is accounted for separately from the rest of the protection channel.
(If this is not the case, incorrect overall times may be obtained because Table 9-1 values only account for the sensor portion of the protection channel.)
5.
Incorporate the sensor response time acceptance criteria in the plant procedures.
6.
Obtain the response time for the remainder of the reactor trip or engineered safety features function.
7.
Add the sensor response time to the response time for the remainder of the protection channel and verify that the total is less than the value for that function given in the plant Technical Specifications or FSAR.
l 9-1
Table 9-1 Sensor Response Times Response Time Model Baseline Manufacturer Westinghouse Manufacturer Number Description (1)
Supplied E Spec BARTON 288 Differential Pressure Indicating Switch BARTON 289 Differential Pressure Indicating Switch BARTON 332 Differential Pressure Transmitter BARTON 351 Sealed Sensor 1 sec BARTON 752 Differential Pressure Transmitter 30 msec 400 msec BARTON 763 Gauge Pressure Transmitter 180 msec 200 msec BARTON 763A Gauge Pressure Transmitter 180 msec 200 msec BARTON 764 Differential Pressure Transmitter 180 msec 400 msec FOXBOR0 E11GM Gauge Pressure Transmitter FOXBOR0/ WEED N-EllAH Absolute Pressure Transmitter FOXBOR0/ WEED N-E11DM Differential Pressure Transmitter FOXBOR0/ WEED N-E11GH Gauge Pressure Transmitter FOXBOR0/ WEED N-E11GM Gauge Pressure Transmitter FOXBOR0/ WEED N-E13DH Differential Pressure Transmitter FOXBOR0/ WEED N-E13DM Differential Pressure Transmitter (1)
Utilize the most conservative value for response time obtained from either previous plant insite respcase time testing or, if replacing the transmitter, the response time obtained through testing.
i 9-2
Table 9-1 Sensor Response Times Response Time Model Baseline Manufacturer Westinghouse Manufacturer Number Description (1)
Supplied E Spec ROSEMOUNT 1151 Pressure Transmitter ROSEMOUNT 1152 Pressure Transmitter ROSEMOUNT 1153 Pressure Transmitter ROSEMOUNT 1154 Pressure Transmitter ROSEMOUNT 1151 Differential Pressure Transmitter ROSEMOUNT 1152 Differential Pressure Transmitter l
ROSEMOUNT 1153 Differential Pressure Transmitter l
ROSEMOUNT 1154 Differential Pressure Transmitter j
l TOBAR 32DP1 Differential Pressure Transmitter 400 msec TOBAR 32DP2 Differential Pressure Transmitter 400 msec l
T0BAR 32PA1 Absolute Pressure Transmitter 200 msec TOBAR 32PA2 Absolute Pressure Transmitter 200 msec TOBAR 32PG1 Gauge Pressure Transmitter 200 msec VERITRAK 76DP1 Differential Pressure Transmitter 400 msec VERITRAK 76PG1 Gauge Pressure Transmitter 200 msec 1
VERITRAK 76PH2 Absolute Pressure Transmitter 200 msec (1)
Utilize the most conservative value for response time obtained f rom either previous plant insitu response time l
testing or, if replacing the transmitter, the response time obtained through testing.
9-3
10.0 Technical Specifications Appendix A contains a markup of the recormiended Technical Specifications to be used when eliminating the requirements for actual measurement of response times on Reactor Trip and Engineered Safety Features Systems.
l This change does not represent a significant hazard to the public as evaluated in accordance with the requirements of 10 CFR 50.92.
See Appendix B.
11.0 CONCLUSION
S By utilizing the recomendations of EPRI Report NP-7243 Revision 1,
" Investigation of Response Time Testing Requirements", justification is established for eliminating response time testing surveillance requirements for the pressure and differential pressure sensors covered by that report. Justification for eliminating additional sensors has been documented by this WCAP by showing similarity to those sensors included in the EPRI report. Where similarity could not be shown, FMEA i
or testing demonstrated that the time response would not be significantly effected by degradation of components or that such changes would be detectable by normal calibration procedures.
All sensors that have been justified by the above methods appear in Table 9-1.
12.0 REFERENCES
1.
EPRI NP-7243 Revision 1, " Investigation of Response Time Testing Requirements".
2.
EPRI TR-103436, Volume 2, Instrument Calibration and Monitoring Program
- Failure Modes and Effects Analysis.
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APPENDIX A Docket No. 50-445 3/4.3 INSTRUMENTATION August 14, 1987 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMIT!NC CONDITION FOR OPERATION 3.3.1 As a sinimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERA 8LE wita RESPONSE TIMES as shown in Table 3.3-2.
APPLICA8ILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILL NCE REQUIREMENTS 1
4.3.1.1 Each Reactor Tri6 System instrumentation channel and interlock and the automatic trip logic *sna11 be demonstrated OPERA 8LE by the performance of the Reactor Trip Systaa Instrumentation Surveillanca Requirements specified in Table 4.3-1.
4.3.1.2 The REACTOR TRIP '5YSTEM Resp 0MSE TIM of each Reacter trip function shall be 1: :--'=:d to be within its limit at least once per la months.
Each y 4 shal include at least one train such that both trains a at least o ke per months and one channel per function such that al c anels are 4ess at le t once every N times IS sonths where N is the tal nuncer of redus tc Is in a specific Reacter trip function as in the
" Total o.
C is" column of Table 3.3-1.
l v ritication vuisi.4 11 4 COMAMcHE PEAK - UNIT 1 3/4 3 1 A-1
APPENDIX A Becket %. g.445 INSTRUMENTATION Ansgust 14, [g7 SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic l
actuation logic and relays shall be demonstrated CPERA8LE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Taele 4.3-2.
I 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIM of each ESFA5 function
- ;^-d to be within the limit at least once per 18 aanths.
shall be :: :n$ include at least one train such that both trains are ' :::d at Eme,__.. shal1 least once per months and one enannel per function such that al nnels are tr--f at les t once per N times la months wnere N is the 1 number of redu t chan is in a specific E5FAS function as s n the " Total No. of Chan 1s" co men of Table 3.3-3.
verification verified verified COMANCHE PEAK - UNIT 1 3/4 3 17 A-2
APPENDIX A INSTRUMENTATION Socket No. 50 445 l
BASES August 14, 1987 l
REACTOR TRIP SYSTEM and ENGINEERED SArgrY FEATURES ACTUATION SYSTEM IhsauMiiNTATION (Continueo) the sensor from its calibration point or the v'alue specified in Table 3.3-4, in percent span, from the analysis assumptions.
Use of Equation 3.3-1 allows for a sensor draf t factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.
Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertafFnty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Being that there is a small statistical chance that this will happen, an infrequent, excessive orift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
asurement of response time at the specified frequencies provi assurance Reactor trip and the Engineered Safety Featur untion associated with eac 1 is completed within the ti asstaned in the
[n$eri safety analyses.
No credit kan in the anal or those channels with b
response times indicated as not app
_a sponse time say be demonstrated by any series of sequential, over
, or channel test seasurements provided that such tests rate the total e nse time as defined.
Sensor response ti ication may be demonstrated by ait
) in place, onsi offsite test seasurements, or (2) utilizing replac sen ith certified response time.
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.
If they are, the signals are contined into logic matrices sensitive to combina-tions indicative of various accidents events, and transients.
Once the required logic combination is completed, the systes sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition.
As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to sitigate the consequences of a steen line break or loss-of-coolant accident:
(1) Safety Injection pumps start and autcaatic valves position, (2) Reactor trip, (3) feed w.ter isolation, (4) startup of the emergency diesel generators, (5) containment
. spray pumps start and automatic valves position (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and auto-antic valves position, and (12) Control Room Isolation and Ventilation Systems start.
COMANCHE PEAK - UNIT 1 8 3/4 3-2 A-3
l APPENDIX A l
INSERT A I
The verification of response time at the specified frequencies provides assurance that the reactor trip and the engineered safety features actuation associated with each channel is completed within the time I
limit assumed in the safety analyses.
No credit was taken in the i
analyses for those channels with response times indicated as not l
applicable.
Response time may be verified by actual tests in any series l
of sequential, overlapping or total channel measurements, or by l
sumation of allocated sensor response times with actual tests on the I
remainder of the channel in any series of sequential or overlapping measurements. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications.
WCAP-13632 Revision 1, " Elimination of Pressure Sensor Response Time Testing Requirements" provides the basis and methodology for using allocated sensor response times in the overall verification of the Technical Specifications channel response time.
The allocations for sensor response times must be verified prior to placing the sensor in operational service and re-verified following maintenance that may adversely affect response time.
In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value.
One example where time response could be affected is replacing the sensing assembly of a transmitter.
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l APPENDIX B Doc. #
Revision i
Page 1 of 5 WESTINGHOUSE NUCLEAR SAFETY DEPARTMENT SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS 1)
NUCLEAR PLANT (S):
2)
SUBJECT:
Response Time Testina Deletion - Pressure Sensor 3)
TECHNICAL SPECIFICATIONS CRANGED: 4.3.1.2 and 4.3.2.2
" Instrumentation-Surveillance Recuirements, B 3/4 3.1 and B 3/4.3.2 " Reactor Trio and Encineered Safety Features Actuation System Instrumentation" 4)
A written analysis of the significant hazards consideration, in accordance with the three factor test of 10CFR50.92, of a proposed l
license amendment to implement the subject change has been prepared and is attached.
On the basis of the analysis, the checklist below has been completed.
Will operation of the plant in accordance with the proposed amendment:
4.1)
Yes No X Involve a significant increase in the probability or consequences of an accident previously evaluated; 4.2)
Yes No X Create the possibility of a new or different kind of accident from any accident previously evaluated; i
4.3)
Yes___ No X Involve a significant reduction in margin of safety.
5)
Reference Documents:
1.
WCAP-13632 Revision 1,
" Elimination of Pressure Sensor Response Time Testing Requirements" 2.
(List of Applicable Sensors) 6)
Comments:
None Prepared by (Nuclear Safety):
Date:
Verified by (Nuclear Safety):
Date:
Coordinated with Engineer (s):
Date:
Nuclear Safety Group Manager:
Date:
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APPENDIX B Doc. #
Revision Page 2 of 5 PLANT NAME PRESSURE SENSOR RESPONSE TIME TESTING DELETION SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS INTRODUCTION
{
In 1975 response time testing (RTT) requirements were included in the Westinghouse Standard Technical Specifications and were required for all plants licensed after that date.
The Standard Technical Specifications contain definitions for both Reactor Trip System and Engineered Safety Features Actuation System response times.
The response time definitions are:
"The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage."
"The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,
the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable."
The Dases section states that the response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured. This approach is also consistent with ISA Standard 67.06. Given this guidance and the complexity of testing an entire instrument channel from the sensor to the final device, plant surveillance procedures typically test a channel in two or more steps.
One individual step in most plant test methodologies is the instrument sensor.
Separate procedures using specialized test equipment and/or outside vendors are typically used solely for testing the sensors.
The first RTT guidelines were established by the Institute of Electrical and Electronic Engineers in ANSI /IEEE Standard 338-1975, " Criteria for the Periodic Testing of Class lE Power and Protection Systems".
In 1977 this Standard was revised and accepted by the NRC with NRC Regulatory Guide 1.118,
" Periodic Testing of Electric Power and Protection Systems", Revision 1.
Following Revision 2 of the Regulatory Guide 1.118, the Instrument Society of America approved Standard ISA S67.06, " Response Time Testing of Nuclear Safety-Related Instrument Channels in Nuclear Power Plants
- August 29, 1986.
This significant hazards consideration analysis applies to the proposed deletion of periodic time testing requirements for certain pressure and differential pressure transmitters and switches from the Technical Specifications.
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l APPENDIX B Doc. e Revision Page 3 of 5
~
DESCRIPTION OF THE AMENDMENT REQUEST As required by 10 CFR 50.91 (a)(1), an analysis is provided to demonstrate that the proposed license amendment to delete the requirement for certain pressure and differential pressure sensor response time testing involves no significant hazards consideration.
The proposed amendment, described in Reference 1, would modify Surveillance Requirements 4.3.1.2 and 4.3.2.2 Technical Specifications 3/4.3.1, " Reactor Trip System Instrumentation" and 3/4.3.2, " Engineered Safety Feature Actuation System Instrumentation" and B 3/4.3.1 and B 3/4.3.2 " Reactor Trip and Engineered Safety Features Actuation System Instrumentation" to indicate that the total response time will be determined based on the results of Reference 1 for pressure sensors.
EVALUATION The primary purpose of this evaluation is to determine if the deletion of periodic response time testing could be justified for specific pressure, level, and flow functions that utilize pressure and differential pressure sensors. IEEE Standard 338-1977 defines a basis for eliminating RTT.
Section
6.3.4 states
" response time testing of all safety-related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety system equipment is verified by functional testing, calibration check, or other tests, or both."
WCAP-13632 Rev. 1 (Reference 1) provides the technical justification for deletion of periodic response time testing of selected pressure sensing instruments. The program described in the WCAP utilizes the recommendations contained in EPRI Report NP-7243 Rev.
1,
" Investigation of Response Time Testing Requirements" for justifying elimination of response time testing surveillance requirements on certain pressure and differential pressure sensors. To address other sensors installed in Westinghouse designed plants, the WCAP contains a similarity analysis to sensors in the EPRI report or an FMEA to provide justification for elimination of response time testing requirements.
The specific sensors installed (Reference 2) at (Plant Name) are listed below.
- Steam Generator Water Level (Manufacturer / Model)
- Pressurizer Pressure (Manufacturer / Model)
- Steamline Pressure (Manufacturer / Model) l
- Containment Pressure (Manufacturer / Model)
- Reactor Coolant Flow (Manufacturer / Model)
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APPENDIX B ooc. #
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[
Revision Page 4 of 5 The basis for eliminating periodic response time testing for each sensor is discussed in the WCAP and/or the EPRI report. These reports provide justification that any sensor failure that significantly degrades response time will be detectable during surveillance testing such as calibration and channel checks.
Based on these results, the (Plant Name) Technical Specifications are being revised to indicate that the system response time shall be verified utilizing a sensor response time justified by the methodology described in WCAP-13632 Revision 1.
Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications.
ANALYSIS Conformance of the proposed amendment to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 is shown in the following:
1)
The proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
This change to the Technical Specifications does not result in a i
condition where the design, material, and construction standards that were applicable prior to the change are altered.
The same RTS and ESFAS instrumentation is being used; the time response allocations /modeling i
I assumptions in the Chapter 15 analyses are still the same; only the method of verifying time response is changed.
The proposed change will not modify any system interface and could not increase the likelihood of an accident since these events are independent of this change.
The proposed activity will not change, degrade or prevent actions or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the SAR.
Therefore, the proposed amendment does not result in any increase in the probability or consequences of an accident previously evaluated.
2)
The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
This change does not alter the performance of the pressure and differential pressure transmitters and switches used in the plant protection systems. All sensors will still have response time verified by test before placing the sensor in operational service and after any maintenance that could affect response time. Changing the method of periodically verifying instrument response for certain sensors (assuring equipment operability) from time response testing to calibration and l
channel checks will not create any new accident initiators or scenarios.
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APPENDIX B poc. #
Revision Page 5 of 5 Periodic surveillance of these instruments will detect significant I
degradation in the sensor response characteristic.
Implementation of the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3)
The proposed license amendment does not involve a significant reduction in margin of safety.
This change does not af fect the total system response time assumed in the safety analysis.
The periodic system response time verification method for selected pressure and differential pressure sensors is modified to allow use of actual test data or engineering data.
The method of verification still provides assurance that the total system response is within that defined in the safety analysis, since calibration tests will detect any degradation which might significantly affect sensor response time.
Based on the above, it is concluded that the proposed license amendment request does not result in a reduction in margin with respect to plant safety.
CONCLUSION Based on the preceding analysis, it is concluded that elimination of periodic sensor response time testing is acceptable and the proposed license amendment does not involve a Significant Hazards Consideration Finding as defined in 10 CFR 50.92.
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