ML20080K951

Criteria for Determining Duration of Integrated Leakage Rate Tests of Reactor Containments

ML20080K951
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Issue date:	12/31/1983
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EPRI NP 3400 Containment systems Project 1393-5 Containment build lngs Final Report Electric Power Leaks December 1983 l

Research Institute Flow rate Nuclear power plants Reactor safety l

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~7"9 Criteria for Determining the M}

Duration of Integrated Leakage Rate Tests of Reactor Containments l

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Prepared by Quadrex Corporation Tulsa, Oklahoma i-l it J,..

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1 ORDERING INFORMATION Requests for copies of this report should be directe $ to Research Reports Center l

(RRC), Box 50490, Palo Alto, CA 94303, (415) 965-4081. There is no charge for reports requested by EPRI member utilities and affiliates, U.S. utility associations, U.S. government

_l agencies (federal, state, and local), meJia, and foreign 0:gaizations with which EPRI has an information exchange agreement. On request, RRC will send a catalog of EPRI reports.

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Copynght 01983 Electnc Power Research Institute, Inc. All rights reserved NOTICE Th;s report was prepared by me organizaten(s) named below as an account of wort sponsored by the Electnc i

Power Research institute. Inc. (EPRI). Neither EPRt. members of EPRI. the organization (s) named below, nor any person acting on behalf of any of them (a) makes any warranty express or amphed, with respect to the use of any information, apparatus, me TM or process dtSClosed M this r6 port or that such use may not intnnge pnvate-ly owned rghts or (b) assumes any liabilities with respect to the use of. or for damages resultir,; from the use of, any informaton. apparatut, method, or process disclosed n tNs report.

Prepared by Quadrew Corporaton Tu?sa, oklahrs.a e

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EPRI PERSPECTIVE PROJECT DESCRIPTION The singular purpose of a power reactor containment system is to mitigate the consequences of a loss-of-coolant accident. This design objective is attained by completely enveloping the nuclear steam supply system with the containment system so that a barrier stands between the fission products in the reactor core and the plant environs. The effectivenec*, of the contain-3 ment system as a barrier to the radioactive material transport must be test-ed under conditions approximating those expected to exist immediately fol-lowing the design basis accidant. The testing is accomplished by filling the air space within the confines of the containment system with pressurized air, usually to several atmosp eres, and then measuring the rate at which the air escapes.

Since the air space within containment systems is usually quite large--in most instances in excess of two million cubic feet--and since the rate of leakage is generally only on the order of 100 cubic feet per hour, a consi-derab?e amount of time for testing can be expended even under optimum condi-tions. The three majcr phases of the test are pressurizatior., measurement of leak rate, ud depressurization. As a practical matter, as well as being a regulatory requirement, the plant must be in a shutdown conditicn during a containment test, and plant owners-operators are therefore keenly interested in keeping its duration to a minimum. The effect of containment leak-rate tests (CLRTs) on plant availability is exacerbated by the fact that rhile the test is in progress, equipment inspection and maintenance are essential-ly limited to balance of plant.

This project (RP1393-5) was directed toward the development and validation of technical criteria to be used by analysts in making the critically in-portant judgment of when to end the leak-rate measurement phase of a CLRT.

Currently, there is a wide variation in the criteria used by analysts in making that judgment and the relevance of some to the matter at hand is tenuous, at best. A case in point is that a test must run for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, iii

r frrespective of the fact that the leak-rate may be clearly established i

sooner.

Because containment system integrity is vitally important to the maintenance of nuclear power plant safety, the data collected and the analytic methods used during the course of a CLRT must be able to stand up to the critical review of independent parties at a later date, It is just as important, therefore, that the decision to terminate testing not be made prematurely on the basis of criteria lacking a sound technical foundation. Even though rurining a CLRT longer than necessary is bad in terms of not being able to do more productive things with the facility, the need to repeat a test as a re-sult of refutable findings from an earlier one generally presents a worse situation and should be scrupulously avoided.

PROJECT OBJECTIVES The overall objective of this project was to develop technical acceptance criteria for the nuclear power industry to perform containment integrated leak-rate tests. A more specific objective was to es'.ablish criteria for determining when the containment leak-rate has been successfully measured with an acceptable degree of confidence.

PROJECT RESULTS The project objectives were met. The results of the study show conclusively that the duration of an acceptably accurate CLRT is primarily_a function of test method and instruaentation. Attempts to foster test accuracy by prede-fining a minimum duration are based on specious reasoning.

Anyone concerned about the subject of containment system integrity and the corollary matter of containment testing will find the rept ; informative.

This report is recommended for nuclear power plant technical personnel who are responsible for containment leak-rate testing. A.chitect-engineering firms and consultants engaged by utilities to assist in conducting such tests will also find the report useful. For additional information per-taining to the procedural and hardware aspects of containment testing, the iv l

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I reader is referred to EPRI Final Report NP-2726, Containment Inteardted Leak-Rate Testina Improvements,. November 1982.

Thomas M. Law, Project Manager Nuclear Power Division 2

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AE TRACT An integrated leakage rate test (ILRT) is typically conducted on the outage critical path for a nuclear power plant. It is prudent for an electric utility operating a nuclear power plant to reduce the test duration to the minimum that can be techni-cally justified and accepted by the regulatory authorities. The objective of limiting the duration of ILRTs is to improve plant economics by reducing the outage critical path time, while maintaining an acceptable and technically accurate ap-proach. It was the purpose of this project to clearly develop and validate a set of technical acceptance criteria inditating when an ILRT may be terminated. Review of 53 ILRT reports, which represented approximately 21% of industry experience, showed that when the technical acceptance criteria were met, successful completion of the test could be predicted. In those cases where the test results only marginally met the technical acceptance criteria, the criteria would not allow test termination.

The use of these technical test duration criteria resulted in an average savings of about 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> (53 percent) per ILRT investigated. This time savings is economical-ly attractive, while the criteria formed a technically sound basis for compressing nuclear power plant critical path outage time.

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CONTENTS Section Page 1

INTRODUCTION 1-1

Background

1-1 Project Objective 1-4 Comments on Current Industry References 1-5 2

METHODOLOGY 2-1 Data Collection 2-1 Definition of Criterion 2-2 Validation of Criterion 2-2 3

TECHNICAL ACCEPTANCE CRITERIA 3-1 Analysis Techniaue 3-1 Instrument Selection Guide 3-9 Temperature Weighting Factors 3-10 Environmental Factors 3-13 Net Leakage Rate 3-13 Statistical Considerations 3-15 Confidence Limits 3-18 Extrapolation Considerations 3-21 Data Averaging Considerations 3-24 4

VALIDATION OF CRITERION 4-1 Final Criterion 4-1 Application of Criterion 4-1 Criteria Validation 4-2

.Results of Criteria Utilization 4-10 5

CONCLUSIONS AND RECOMMENDATIONS Conclusions 5-1 Recommendations 5-1 6

REFERENCES 6-1 APPENDIX A CRITERIA COMPLETION

SUMMARY

BY EVA'.UATED A-1 PLANT REPORTS APPENDIX B

SUMMARY

OF ACQUISITION AND PROCESS METHODS B-1 APPENDIX C PLANT

SUMMARY

OF SENSORS C-1 APPENDIX D PLANT

SUMMARY

OF TESTING TIME DURATIONS D-1 APPENDIX E TEMPERATURE WEIGHTING FACTOR SENSITIVITY E-1 ix

ILLUSTRATIONS Figure Page 3-1 Point-to-Point Technique 3-2 3-2 Total Time Method 2-5 3-3 Mass Poir.t Technique 3-8 3-4 Typical LSF Leakage Rate Estimate versus Test Time 3-16 3-5 Illustration of Equation 6 3-19 3-6 Typical Convergence of 90%, 95%, and 99% UCL with LSF 3-20 Leakage Rate Estimate versus Test Time 3-7 Illustration of Equation 7 3-22 3-8 Illustration of Data Averaging 3-25 4-1 Typical LSF Leakage Rate Trends - Trend 1 4-3 4-2 Typical LSF Leakage Rate Trends - Trend 2 4-5 4-3 Typical LSF Leakage Rate Trends - Trend 3 4-6 4 Typical LSF Leakage Rate Trends - Trend 4 4-9 Xi

TABLES Table Page 4-1 Categorized Reports Under Trend 1 4-4 4-2 Categorized Reports Under Trend 2 4-7 4-3 Categorized Reports Under Trend 3 4-7 4-4 Categorized Reports Under Trend 4 4-8 4-5 Affect of Technical Acceptance Criterion on Test 4-11 Duration and Leakage Rate 5-1 All ILRTs Distributed by Test Number and Duration 5-3 5-2 Evaluated Containments Distributed by Containment Type 5-4 5-3 ILRT Study Reports Distributed by Termination lime 5-5 t,

xiii

SUMMARY

An integer.ted leakage rate test (ILRT) is typically conducted on the outage critical path for a nuclear power plant. It is prudent for an electric utility operating a nuclear power plant to reduce test duration to the minimum that can be technically

. justified and accepted by the regulatory authorities. The objective of limiting the duration of ILRTs is to improve plant economics by reducing the outage critical path time,=while maintaining an acceptable and technically accurate approach. It was the purpose of this project to develop and validate a set of technical acceptance criteria indicating when an ILRT may be terminated.

The project was performed in three phases:

e Data collection e

Definition of technical criteria for test termination e

Validation of the derived technical acceptance criteria to ensure acceptable test results Approximately 144 ILRT reports were reviewed, representing about 58 percent of the total ILRT reports submitted to the NRC. Fifty-three reports (21 percent of all submitted) contained information adequate'for use in this study. Because this study required specific details, this does not imply the other reports were inadequate ILRT reports for their intended purpose.

In the definition phase, preliminary lists of potential test termination criteria were developed by considering all known factors which would both influence the successful completion and the duration.of the test. As the study progressed, the

-lists were revised and condensed into the final criteria list below:

)

1.

Use the absolute method, mass point technique.

2.

The containment must be adequately modeled.

' 3.

The final 95 percent confidence level leakage rate must not be a negative value.

4.

The calculated least squares fit (LSF) leakage rate must be less than 75 percent of the plant's allowabie leakage rate criteria at test pressure.

S-1

5.

The calculated 95 percent upper confidence level leakage rate must be less than 75 percent of the plant's allowable leakage rate at test pressure.

6.

The calculated LSF leakage rate as a function of time shall have stabilized with a neglible positive or negative slope.

7.

The calculated 95 percent upper confidence level leakage rate shall be converging with the LSF leakage rate.

In general when all technical acceptance criterion are concurrently met, sufficient data has been obtained so that the ILRT can be terminated with the containment leakage rate accurately predicted. Test continuance through additional data points should not critically influence or reverse the test results unless significant physical change occurs. A review of 53 ILRT reports indicated that when the techni-cal criterion were met, successful completion of the test could be predicted. How-ever, for those tests in which there was little margin between calculated and allow-able leakage rate, applicttion of the technical acceptance criterion did not permit early test termination. The use of these technical acceptance criterion would result in an average savings of about eleven hours (53 percent) per ILRT investi-gated. This time saving is economically attractive, while the criterion formed technically sound basis for compressing nuclear power plant critical path outage time.

S-2

Section 1 INTRODUCTION BACKGROUND in early 1959 work was initiated by the American Nuclear Society (ANS) to develop a reactor containment leakage rate testing standard. An initial draft of ANS draft Standard 7.6 was prepared by E. F. Wilson of Allis-Chalmers Manufacturing Company. This draft subsequently underwent 15 or more

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reviews 'and revisions over a 13-year period before formal issue.

In 1966 the Atomic Encrgy Commission ( AEC) funded a study to develep a technical safety guideline titled Reactor Containment Leakage Testing and Surveillance Requirements (TID-24085). This study recommended a framework for the Type A, B, and C testing prograns later outlined in 10 CFR 50

.%ppendix J, but provided no guidance for test duration other than to reference the ANS 7.6 draft dated October 13, 1966.

In 19G9, a report by F. C. Zapp, Testing of Containment Systems used with Light Water Cooled Power Reactors, ORNL-NSIC-26, was issued. This report provided broad detailed containment leakage rate testing guidelines and was the basis for much of the analytical work done in this area. However, it did not provide guidance for test duration.

In 1971 the AEC issued 10 CFR 50 Appendix J, reference (1), Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Although this regulation did not specify the duration of the Type A test, it did reference ANS draf t stando'd 7.6 and was later updated to reference ANSI N45.4-1972.

In 1972 the ANS 7.6 committee through the American National Standards Institute ( ANSI), issued ANSI N45.4-1972, reference (l2), Leakage-Rate Testing of Containment Structures for Nuclear Reactors. This standard 1-1

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l specifically mentioned a 24-hour test, but also provided for shorter test durations under certain conditions as follows:

" Period of Test. The leakage-rate test period, for any method, shall L

extend to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of retained internal pressure. If it can be demon-strated to the satisfaction of those responsible for the acceptance of the containment structure that the leakage rate can be accurately determined during a shorter test period, the agreed-upon shorter period may be used."

Also in 1972, the AEC approved a Bechtel Corporation Topical Report, BN-TOP-1, reference (3), titled Testing Critaria for Integrated Leakage

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Rate Testing of Primary Containment Structures for Nuclear Power Plants.

This report defined criterion for a test duration as follows:

"1. The Trend Report based on Total Time calculations shall indicate that the magnitude of the calculated leak rate is tending to stabilize at a value less than the maximum allowable leak rate

.(L,).

[ Note: The magnitude of the calculated leak rate may be increasing slightly as it tends to stabilize. In this case the average rate shall be determined from the accumulated data over the last five hours or last twenty data points, whichever provides the most points. Using this everage rate the calculated leak rate can then be linearly extrapolated to the 24th hour data point. If this extrapolated value of the calculated leak rate exceeds 75% of the maximum allowable leak rate (L,) then the leak rate test is continued.]

and

2. The end of test upper 95% confidence limit for the calculated leak rate based on Total Time calculations shall be less than the maximuia allowable leak rate, and

3. The mean of the measured leak rates based on Total Time calcula tions over the last five hours of test or last twenty data points, whichever provides the mo'.t data, shall be less than the maximum allowable leak rate.

and 1-2

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4. Data shall be recorded at approx'aately equal intervals and in no case at intervals greater than one hour.

and

5. At least twenty (20) data points shall be provided far proper statistical analysis.

and

6. In no case shall the minimum test duration be less than six (6) hours."

The above guidelines are the basis for a number of opera ing nuclear plant technical specifications that permit tests of less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Typically, the words utilized are:

"...shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972."

or "tne test duration shall not be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless test experiences of at least 2 prior tests provide adequacy of shorter test duration."

However by 1977, this technical specification provision had significantly dimin13hed in use for new operating licenses.

During the mid and late 1970's the ANS sponsored another standards group to update ANSI N45.4-1972. This group was known as the ANS 56.8 or N274 group.

The result of their effo-ts became. ANSI /ANS 56.8-1981, reference (4),

Containment System Leakage Testing Requirements. This standard specifies a 24-hour test with provisions for reducing the test duration to eight hours as follows:

"A Type A test shall last a minimum of eight hours after stabilization and shall have a total of not less than 20 sets of data points at approximately equal time intervals."

In January 1982, the American Nuclear Society Reactor Operations Division sponsored a CLRT Workshop in San Diego, reference (5). Of the 15 technical papers presented, a number of papers mentioned the feasibility of minimizing 1-3

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!'the test ' duration. During and after the Short uuration Testing Session, in-depth discussions were held among the ' utilities, vendors, consultants, and "NRC.

These discussions were important because all the NRC. regions were

represented, as well as the Office of Research,' Nuclear Reactor Regulation,

- and Inspection /Fnforcement Headquarters.

At-this workshop, a number of current test programs were discussed and the -

philosophy of reduced duration. testing was aired. In ~ addition, a paper was presented titled " Suggested Criteria for a Short Duration ILRT" by T. Brown and L. Estenssoro (see Reference 5_).

The workshop focused attention on the desirability of plant owners shortenira the duration of testing to minimize the impact on the outage critical path time. A number of alternative methods to develop technical acceptance criteria to minimize test duration were discussed:

9 e - Develop an ANSI standard.

e Prepare an EPRI' study report.

e Issue a NRC regulation or regulatory guide, e Submit a utility, industry group, or consultant topical report.

e ' Submit a number of individual ~ utility docket reports.

From the above alternatives, the EPRI. study report approach was chosen as the fastest and most efficient vehicle to develop criteria to allow plants to minimize the duration of Type A'(ILRT) tests. It war, tentatively planned that an ANSI standard and/or. Regulatory Guide would then be developed to

_ give the findings either an industry consensus and/or a regulatory endorsement.

~ PROJECT OBJECTIVE The project objective is to minimize the duration of ILRTs, which would improve plant economics by reducing the critical path outage time, while maintaining an acceptable and technically accurate approach. Because an

1LRT necessitates plant shutdown, it is imprudent for any electric utility 1-4

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. operating a nuclear power plant to perform ILRTs of longer duration than is absolutely necessary for reactor safety. Therefore, this project has

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intended to clearly define the technical acceptance criterion indicating wh:n an ILRT may be terminated, elapsed time notwithstanding, and to initiate the use of such criteria by the nuclear power indust ry at large.

COMMENTS ON CURRENT INUUSTRY REFERENCES Code of Federal Regulation 10 CFR 50, A.ppendix J, Reference (1) refers totally to ANSI A45.4-1972 on this topic. Reference (1) is not a problem, in that it is not specifically restrictive on the subject of test auration.

Standard ANSI N45.4-1972, Reference (2_) specifies 24-hour ILRTs unless shorter periods can be accurately determined. Reference (2) is not a problem, except it does= not define the meaning of "accurs.tely determined."

-This vagueness has allowed the nuclear industry to fall into two broad categories: the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test duration group and the less than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test duration group.

Bechtel Report BN-TOP-1, Reference (3_) provides technical acceptance criteria and was a step in the right direction. However, it does not endorse the mass point analysis method, which is the industry accepted and preferred technique. Ironically, this topical report was approved by the NRC and since its approval has not been withdrawn, it is a valid reference for test durations less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Standard ANSI /ANS 56.8-1981, Reference (4_) provides some technical acceptance criteria and is also a step in the right direction. However, it does not go far enough and has some criteria that differ with ANSI N45.4-1972 and BN-TOP-1.

This EPRI research project attempts to solve the above described problems, so that the nuclear industry will have a commonly accepted and utilized set of technical acceptance criti.ria for minimizing the duration of ILRTs.

1-5

J Section 2 METHODOLOGY

. Since the technical acceptance criterion developed in this study have to be applicable to both the current 80 operating nuclear plants as well as the 50 plants under construction, a broad range of plant applications is required.

Therefore, the study was performed in three phases:

e Data collection, e

Development of the technical acceptance criterion, e

Validation of.the technical acceptance criterion developed herein

-to ensure acceptable test results.

DATA COLLECTION The ILRT-reports reviewed in this study were copies of the reports obtained through the NRC public document room in Bethesda, Maryland. Approximately 144 reports were reviewed, representing about 58 percent of all reports submitted. The time constraints of this project limited the number of reports reviewed. Each report was reviewed for specific information pertaining to the technical acceptance criterion. More than one third of the reports contained information adequate for validation of the proposed cri terion. This study required more detailed test data than is required to demonstrate the adequacy of an ILRT. Therefore it should not be implied that the other reports are inadequate ILRT reports.

2-1

o i-v DEFINITION OF CRITERION Technical acceptance criterion were developed by considering those factors which would influence the successful completion and the duration of an ILRT.

In addition to the criterion previously developed by others (as referenced in'Section 1) other criterion were developed considering analysis technique, instrument selection guide, temperature weighting factors, environmental factors, net-leakage rate, statistics, confidence limits, extrapolation considerations, and data collection. Based on this, a preliminary set of criterion was-formulated.

s VALIDATION OF CRITERION After the technical. acceptance criterion were considered finalized, those

- ILRT reports which contained sufficient information to be evaluated against the criterion were selectea. Secticn 4 discusses the validation of the study criterion utilizing information from the reports.

2-2

k Section 3 TECHNICAL ACCEPTANCE CRITERIA' ANALYSIS TECHNIQUE Two methods, the absolute method and the reference vessel method, have been used for Type A leakage rate tests. The absolute method is the most commonly ' applied test method. This method requires the measurement of containment pressure, temperature, and vapor pressure at discrete points in time.- From these measurements, the containment air mass is calculated at each point in time. ~re test data, therefore, consists of air mass values at a number of time points. The leakage rate (percent per day) is defined as the reduction in containment air mass extrapolated to a 24-hour period divided by the initial containment air mass.

While the principle discussed above is simple, practical problems arise because instrumentation sensitivity may be comparable in magnitude to the leakage rate to be measured. Three different analysis techniques have been

. applied to overcome this difficulty. These techniques are discussed briefly below.

Point-to-Point Technique-The point-to-point technique is illustrated in Figure 3-1.

3-1

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W%l l

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t t

j 2

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5 (a)

TIME (hours) l l

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Mean L LEAKAGE 9

LSF L RATE

(%/ day)o o

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I Figure 3-1.

Point-to-Point Technique 3-2

The leakage rate, L, is calculated at each point in time, t, by g

g j and t,1:

calculating the slope in the air mass line between t j

i i-"

Equation (1)

L

=

i t{j) - tgj_g) 1bm th th between i -i and i time point Where L4 = calculated leakage rate, p,

W = measured cuntainment air mass in ibms t = time, hours i = value of parameter at ith time point 1-1 = value of parameter at ith-1 time point The point-to-point method has several problems. First, if the point-to-point calculated leakage rate values, L, were to be fitted with a least 3

squares fit (LSF) regression line, as shown in figure 3-lb, a leakage rate as a function of time would result. This is in conflict with the fundamental assumption that the containment leakage rate is constant. The assumption of a constant leakage rate is physically realistic and is basic to achieving a successful test and, even more importantly, it is basic to the underlying reasons for running a test. One would not expect changes in the leakage rate as a function of time over the duration of the test unless the leakage paths experienced physical changes or unless the drivirg pressure changed with time. Leakage path changes are not expected once the containment has stabilized at the test pressure, and the pressure decay over the duration of the test is miniscule. Furthemore, the leakage rates to be measured are small compared with the instrumentation system sensitivity and the effects of minor leak path changes. Therefore, significant instrumentation or leak path changes during the course of the test would most likely prevent convergence of the test data and delay, if not prohibit, a successful test conclusion. Finally, the underlying reason for running a test is not to demonstrate an acceptable leakage rate on the day of the test; the purpose is to give reasonable assurance that leakage rates to be expected over the period between tests is such that, should the containment 3-3

l l

oesign function be required, the contairiment performance will be adequate.

This assumption certainly implies a leakage rate that is constant for all practical intents and purposes during the comparatively short period of the test (typically 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less).

The constant leakage rate assumption can be accommodated by calculating the mean leakage rate rather than the LSF leakage rate. However, as shown in Reference Q), this is effectively the same as calculating the leakage rate as the difference between the air mass at the last point and the first point divided by the total test time, neglecting all intervening points. In addition, the calculated values of leakage rate, L, do not tend to converge 4

with additional test points. Thus, if one holds t constant but collects additional test points by extending the test duration, the LSF and mean leakage rate variability will begin to decrease but the individual values of calculated L will continue to exhibit the same level of variability. On 4

the other hand, if one reduces the At, by either reducing total test time or increasing the number of test points, the scatter in the values of L will 4

increase. This can be seen from equation (1) if one assumes a constant value for the magnitude of random error in the Wi values but reduces the at values. In the limit, as At approaches zero, the calculated L becomes j

dominated by the random errors in the W 's and thus becomes meaningless.

j The above weaknesses in the point-to-point method have resulted in a reluctance to apply this method. The last characteristic discussed above renders the point-to-point method particularly unsatisfactory for reduced time tests.

Total Time Technique The total time method is illustrated in Figure 3-2.

This method calculates the leakage rate at time t by dividing the total loss in containment air j

3-4 l

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W (1b m) l l

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5 TIME (hours) l I

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LSF Regression Line LEAKAGE O

(%/ day) 1 I

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I Figure 3-2.

Total Time Method 3-5

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L' mass between t and t by the total test time t. Thus 1

g g

L W -W i 1 Equation (2)

.i

=

t -t g g This method is preferable to the point-to-point method because the calcu-lated leakage rates, L, begin to converge to the true value at larger g

values of t, due to the fact that the magnitude of the random errors in the 9

numerator of equation (2) become smaller relative to the magnitude of the denominator. In addition, a spurious error in a measured W affects only a g

and L _1) as in the point-to-single value of L, rather than two values (L g

g g

point method.

The total time method also has drawbacks. FO st, calculating leakage rate (lbm/hr) requires subtraction of all measured masses fro.n the initial air mass, W. Since W is subject to the same magnitude of random error as all g

g other measured W 's, tnis conversion imposes a systematic error on the final g

test result. In fact, judicious selection of the test initial point (i.e.

t and the corresponding W ) can significantly influence the final result.

g g

Also, the application of stanoard statistical techniques (i.e. LSF regression analyses and development of confidence limits) is inappropriate unless some sort of time-related weighting factor is applied to each successive test point. This is due to the fact that the accuracy of the calculated L 's increases with increasing time, while standard statistical g

techniques assume that all data points have the same error structure and are independent measurements rather than sharing an initial data point.

3-6

Mass Point Technique The mass point technique is illustrated in Figure 3-3.

This technique is recommended by Reference (4_) and has been widely adopted as the preferred method of analyzing leakage rate test data. It is also judged to be the most sensitive technique for short duration tests (i.e. less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

In the mass point technique the measured containment air masses are fitted with a LSF regression line. At each point in time, the slope of this line divided by its intercept represents the leakage rate (%/ day). Thus IA I x 2400 Equetion (3)

L

=

g g

i il Where Lj = calculated leakage rate, percent / day over and including the th i

time point.

.Aj = slope of the LSF regression line based on all data points up th to and including the i data point.

By = intercept of the LSF regression line based on all data th points up to and including the i data point.

The detailed equations for A, and B and their derivations, are given in g

g Reference (4_).

L begins to converge for large values of i (i.e. > approximately 20),

g because each additional data point nas increasingly less influence on the LSF line. Equation (3) yields a leakage rate independent of time at each 1.

That is consistent with physical expectations.

t 3-7 iu'

i i

i l

l I

I I

I I

I I

BY l

5 l

A l

l l

l B

T 4

4 CONT NMENT A5 I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

t t

t t

t t

t j

2 3

4 5

6 7

TIME (hours) l I

I I

I I

l l

l l

I t

1 I

I 1

1 4

t 1

+

I I

I I

I i

RATE

(%/ day)O l

I l

l l

l I

I I

I 1=(A/B)X2400 4 $

l l

I l

l l

l l

l l

l l

l l

Figure 3-3.

Mass Point Technique 3-8

INSTRUrtENT SELECTION GUIDE

- The instrument selection guide (ISG) is a parameter derived from the repeatability of the instrument system used to measure containment temperature, pressure, and dew point. The ISG is, therefore, a function of the individual instrument design characteristics and the total number of instruments used. The ISG is not influenced by actual test conditions such as temperature stability and leakage rate, nor does it reflect the data scatter actually Observed during the test.

Based on the equation used to derive the ISG, the ISG may be physically interpreted as one standard deviation in the 24-hour leakage rate if that leakage rate were calculated by using the given instrument system to take two containment mass measurements, one at time o and one at time t, and the leakage rate calculated from the two measurements then extrapolated to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

From Reference (4_), the fcnnula for the ISG is:

ISG =

2400 f

1 f

T

• 2 *pv Equation (4) 2 h

+2 e I t

( P; (T;

Pj where all parameters are as defined in Reference (4).

Reference (4_) gives the criterion that the ISG wust be less than 25 percent

.of the leakage criteria. This should not, however, be interpreted to mean that a total test time, t, which yields an ISG withir. the 25 percent criterion,-is an acceptable test duration. Rather, it should be interpreted to mean that if one were to calculate a 24-hour leakage rate based on containment air mass measured at time 0 and time t, with the given instrument system, then one would expect, from a statistical perspective, that the uncertainty of this calculated leakage is such that one standard deviation on the calculated leakage is less than 25 percent of the leakage cri teria. The ISG, then, is a rule of thumb for quantifying instrument 3-9

I

(

~ system performance such that the data scatter to be expected, using the

' selected instrw.ent system, will yield meaningful test results. The ISG is

[.

' not an acceptabie parameter for setting either minimum or adequate test duration.- The.. traditional. testing practice of continuing the test until the

. time given by the ISG is ratisfied has most likely been'due to inadequate understanding of reasons. for the ISG.

jTEMPERATUREWEIGHTINGFACTORS c

Multiple temperature-sensors typically are required to measure average containment ~ temperature of. subcompartments which may not be closely coupled, due' to convection air flow,' therrial insulation between subcompartments, and radiaton heat. sources. Each sensor must be weighted appropriately to represent the fraction of containment subvolume associated with that

. particular sensor.. Improper weightng factors, coupled with containment

temperature transients, can lead to spurious indications of results. -

Reference (4) gives criteria for selecting temperature weighting factors.

Basically, these criteria state.that each temperature sensor shall be

~

.. associated with at most 10 percent of the containment volume. This may be a

~ reasonable rule of thum0; however, in practice it is sometimes not met. For

= example, many leakage rate tests rely on a single sensor for the containment dome" area, which may represent 20 percent toc 30 percent of the total volume.

However, for some plants the dome is subject to,asynnetric. diurnal effect and temperature: stratification if air flow is not maintained. For these plants' the use of multiple dome sensors is more important. In other-areas the presence of heat sinks, sources,.or subcompartments may dictate multiple sensors for volumes' considerably -less than 10. percent of' the total volume.

Reference (4) also. gives criteria for temperature ' stability. Similar to the

- weighting factor criteria, these stability criteria are based on empirical obse'rvation-and experience rather than scientific principles. Note that the weighting factors and temperature stability are interrelated. For example, if-the containment temperature is perfectly stable, the accuracy of the

. weighting Lfactors is not'significant. Conversely, if the weighting factors 3-10

are sufficiently accurate, temperature variations will be adequately accounted for in the calculation of containment mass.

The containe:$nt must be accurately modeled to calculate any changes in the mass of etr due to temperature. At least one of the following conditions must be present for the containment to be considered accurately modeled:

e The containment temperature must be modeled with a minimum of 18 RTDs.

e Accurate and validated temperature weighting factor, must be utilized.

e The average temperature change must be less than 0.5'F during the duration of the ILRT.

Note that the current average containment has 22 RTus and some have as many as 52 RTDs. The summary of sensors by plant in Appendix C, makes obvious that many plants have close to or alreaay exceed the 18 RTO minimum.

However, each subcompartment having heat sources or sinks must have a representative RTD.

If accurate temperature weighting factors were utilized, then the containment model'would very closely represent actual conditions inside containment.. Accurate and validated temperature weighting factors can be developed by an iterative analytical process using scquential temperature increases and decreases.

To determine the sensitivity of the leakage rate results to temperature weighting factor perturbations, this study performed sensitivity analyses on 25 different ILRTs. Seven hypothetical cases were postulated. Analysis of the effects of the seven hypothetical cases on the 25 ILRTs were performed.

The seven cases are:

o Case 1 -- The effect of using only one sensor was evaluated by using only the temperature sensor that varied the most.

3-11

~

e Case 2 -- The effect of using'only one sensor was also evaluated by using only the temperature sensor that varied the least.

e Case 3 -- The effect of losing a temperature sensor was evaluated by deleting the temperature sensor having the largest weighting factor. The weighting fac. tor for the deleted sensor was added equally to the second and third largest sensor weighting factors.

e Case 4 -- The effect of losing a temperature sensor was evaluated by deleting the temperature sensor with the largest product of temperature weighting factor times the greatest temperature r

transient. The weighting factor for this sensor was then added equally to the second and third largest sensor weighting factors, e Case 5 -- The effect of weighting factor errors was evaluated by arbitrarily increasing the weighting factor for half of the

_ temperature sensors by 25 percent. The weighting factors for the remaining sensors were proportionally reduced so that the sum of all weighting factors equaled 1.0.

o Case 6 -- The effect of weighting factor errors was evaluated by arbitrarily increasing the weighting factor for half of the temperature sensors by 50 percent. The weighting factors for

.the remaining sensors were proportionally reduced so that the sum of all weighting factors equaled 1.0.

e Case 7 -- All temperature weighting factors were set equal.

Each of the seven cases were applied to all 25 ILRTs. The results of the 175 case studies (25 ILRTs times 7 cases) are displayed in Appendix E.

The following are observations from these sensitivity case studies:

3-12 I

e For those plants with overall average temperature variance of less than 0.5'F, the weightir.g factors selected made virtually no difference in cases 3 through 7'and relatively little difference in cases 1 and 2.

9

-e For those plents with a low number of RTDs (less than 18), the containment mass calculation was relatively sensitive to the assigned seighting factor.

ENVIRONMENTAL FACTORS-7

~

- The most significant environmental factor that can affect test results is

': the diurnal effect. - This effect is generally not a major consideration for later model concrete containments. Many are shielded from direct

- environmental temperature and solar effects by a secondary containment

~ structure. Most of the remainder are sufficiently massive concrete fstructures with sufficient heat capacity' which prevents significant diurnal effects. Some plants may see. diurnal effect caused by variations in inlet cooling water temperatures.

If a reduced-duration test is contemplated, past' test-data should be reviewed to verify lack of significant diurnal effects. If such a review

-indicates significant diurnal effects can be expected, and if accurate temperature weighting factors are unavailable, at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of test data may_well be required to produce adequate data convergence and con-fidence level. Increasing the number of temperature sensors.in the dome

.will permit more accurate modeling of diurnal temperature effects. The significance of the diurnal effects may be determined by comparing the day-to-night tryerature swings.against the temperature stability criteria of

. Reference, (4_).

h NET LEAKAGE RATE Many tests show a negative net LSF leakage rate early in the test.

' Typically, the leakage rate eventually increases to a positive value, as would be expected. This temporary calculation of a negative leakage rate, 3-13 t.

.n

,,,._,.n.,

.,n

. followed by an. eventual increase to a positive leakage rate, can be caused by a. low measurement of the initial containment air mass.

Any poteritial source' of 'in-leakage to the containment from pressurized containers such as fire extinguishers or gas bottles, must be identified and sealed or removed prior to the test. Occasionally, however, a leakage rate test will still yield data which consistently _ indicates a negative leak

. rate. This indicates a source of air in-leakage to the containment or improper temperature weighting factors coupled with temperature variations.

Leakage rate test data which indicates a consistent and continuing negative

. leakage rate. is of little value in determining the actual: level of containment leakage, since the measured leakage is actually the net leakage and not an absolute value. ;To illustrate:

Equation (5)

. L, = Lout.- Lin

' Where L,.

= measured leakage rate t

= leakage rate out of the containment out L

=. leakage rate into the containment in One would physically expect the L term to be zero due to the containment in g 4 pressure,and due to the concerted pretest effort to. identify and remove or -

3

. seal all' sources of possible in-leakage. ;However, if L, remains a

, consistent negative value, this is indicative that L is not zero or that in temperature weighting. factors are incorrect.

The above discussion, indicates that one criterion for test conclusten should be that the 95 percent upper confidence level -(UCL) of the measured

~

_ leakage be zero or a positive value.

I r.

. If the 95 percent UCL of a test leakage rate remains consistently negative, several corrective measures are possible. These include stopping the test to determine the source and value of in-leakage and adding the in Teakage to the measured. test leakage to estimate a realistic value for ou~-leakage.

c i

I 4

3-14 g

E Also, if the in-leakage was a temporary phenomenon (i.e., a short burst of in-leakage), its effect can be eliminated by restarting the test af ter a The testing organization must perforr a case-by-restabilization period.

case assese-ment of each occurrence to determine the most efficient approach.

STATISTICAL CONSIDERATIONS s

The discussions below are based en the mass point analysis technique.

As discussed above, the LSF leakage rate calculation requires developing the LSF regression line, using the measured values of containment air mass for The slope of this line, divided by its intercept, is the best each point.

estimate of the leakage rate represented by the various data points.

If a new LSF leakage rate is calculated each time a single new data point is added to the test and the results plotted against time, a plot similar to Figure 3-4 will typically result.

Note that Figure 3-4 does not represent a leakage rate that varies as a function of time, but represents a time-varying estiinate of a constant leakage rate (i.e., each time a test data point is added, the accuracy of the estimate improves). As illustrated in Figure 3-4, the plot of LSF estimated leakage rate as a function of time will eventually level out and The time required for the curve to level out is become relatively constant.

a function of the number of mass data points, scatter in the mass data, and

=

In other words, after a consistency in the trend of mass data points.

g number cf points have been taken (typically arcund 20 to 30), the LSF curve E

as a fu1ction of time will become relatively constant unless, subsequent

[

points tegin to exhibit a greater scatter or begin to deviate significantly f

from the previous trend. Oscillations in the LSF curve indicate A consistent positive

[

insuf ficient data points to dampen out data scatter.

h slope on the LSF curve indicates that the more recent mass pointThis could measurements are yielding values below pre-established trends.

be caused by an actual increase in the leakage rate (due to the introduction y

of a new leak) or the introduction of temperature instabilities coupled with K

1 K

r r_

5 3-15

i l.

,iL I

0

., 1 E

L B

E

. 9 A

G W

A OTKE LSAT LEEA ATLR

. 8 em i

T ts e

T

. 7 susrev

)s e

.. 6 r t

u a

o m

~

i h

t

(

s e

E m

e i

t

.. 5 T a

R e

~

g a

k ae

.. 4 L

F S

L lac

.. 3 ipy T

4

.. 2 3

erug i

F

.1 FS L

~

~

0 0

0 5

0 1

D" bn T 8 ta a

c OU$a r

i iE

' inaccurate weighting factors. A consistent downward trend in the LSF indicates the more recent mass points are above pre-established trends.

This could be caused by a reduction in actual leakage, by temperature instabilities coupled with inaccurate weighting factors, or by introduction of a gas source into the containment (e.g. leakage of high pressure nitrogen bottles).

From the above discussion, it is clear that variations in the plot of LSF leak &ge rate versus time indicate either insufficient data to achieve convergence or a change in the physical system or instrumentation. A relatively flat curve, on the other hand, indicates data convergence to a stable leakage rate that is likely to remain unchanged barring changes in either containment integrity or instrumentation system performance. Further extension of the test to collect additional data is unnecessary. Based on thee conclusions, the fo? lowing criterion, in addition to that given in Reference (4), is proposed:

The calculated LSF leakage rate as a function of time shall have stabilized with a negligible positive or negative slope.

A proposed quantification of this criterion is as follows:

x 100 1 10 Equation (6)

(L - 'n-1 )

n GREATEST (t

-t d'c ~ 'n) n n-i

'where L

= final test point LSF leakage rate n

Ln-i = leakage rate a for data point taken within previous hour t

= time in hours of the last data point n

n-i = time in hours for the data point used for Ln-1 t

L

= test leakage criteria (75 percent of allowable leakage at g

test pressure).

I 3-17

l This comparison is made using all leakage rates (Ln-1) calculated in the previous hour. The equation calculates the slopes of the LSF curve for each data point taken during the last hour The greatest slope is (t

-t n-if n

then compared to the margin betweer. the last calculated leakage rate and the leakage rate criteria (L ~ 'n).

The absolute value of this fraction is c

expressed as a percent. If equation (6) is satisfied, using the highest LSF slope calculated during the previous hour for the entire hour would not change the tast margin by more than 10 percent. Note that the greater the margin (L ~ 'n), the more variability allowed in the LSF curva and vice c

versa. Figure 3-5 further illustrates the use of equation (6).

CONFIDENCE LIMITS Reference (4_) requires the develop 1ent of 95 percent confidence limits for the calculated leakage rate. Reference (4) also requires that the 95 percent confidence limit leakage rate value be below 75 percent of the allowable leakage rate at the test pressure.

Figure 3-6 illustrates, for a typical test, the convergence of the calcu-lated L with the L Figure 3-6 also shows the convergence of the 90 95 LSF.

percent upper confidence limit and the 99 percent upper confidence limit.

It should be stressed that the actual measured leakage rate is represented by LLSF, not by the upper confidence level leakage rate of L95 (or L90 or Lgg); These upper confidence level values measure the confidence to be placed in the accuracy of the actual leakage rate, LLSF, and are a function of the number of._ data points and data scatter. Selection of L as the 95 basis of the acceptance criteria is a reasonable and traditional choice.

Selection of a lower confidence limit (L90) or a higher confidence limit 3-18

( TEST ACCEPTANCE CRITERIA J L l

L-E c

n a

'n-2 _

LSF Ln-4 ~

~ ~ ~ ~

l LEAKAGE L L_T-1 L" 3 -

n-RATE

.- - - - -I _

1-I I

f

(%/ day)

L a

w n

- - - - - -l - - T -

-l--

l l

l 'n-En-2l, bn- 'n-1l 3 'n' ln-4l l

t - t,_g I

I l

l l t-t t-t n

n-2 n

n-1 n

i I

l l

I l

1 1

I I

l I

l l

lll-En-3l n

l l

l

' t-t n

n-3 l<

l I HR ',

l

, THEREFORE, IF t t

g g

I O

X S 10 n

n-2 c

n g

i l

l l

l 1 THEN EQUATION (6) IS SATISFIED.

I I

I I

l l

I e

I I

I I

I l

T T

T T

T n-4 n-3 n-2 n-1 n

TIME +

Figure 3-5.

Illustration of Equation 6 I

l

UCL DE UCL95% 995

\\

90%

.),00-TU

~

~

tt aa w"

i o

g 0.50-ad 1

ALLOWABLE TEST

\\

LEAKg g RATE T

2 i

s 6

1 g

9 zgg (hours)

Figure 3-6.

Typ' cal Convergence of 90%, 95%, and 99% UCL with LSF Leakage Rate Estimate versus Test Tirne

r (L99) would effect the time required to complete the test, but would have no effect on the actual measured value of containment leakage rate.

indicates that scatter A time-dependent decrease of the quantity L95 - LLSF in the data is constant or decreasing, and/or that confidence in the L LfF value is increasing due to additional data points (see Figure 3-7).

The increase in L confidence is due to the decrease in the t factor (from g3p 95 the Student's t distribution) with increasing number of data points. The t 95 factor becomes an almost negligible effect after about 20 data points.

Additionally, with a relatively constant measurement error magnitude, a longer time duration yields less variability in the slope of the mass versus time line. On the other hand, an increase over time in the quantity L95 ~

LLSF. indicates increasing data scatter and a corresponding decline in confidence in the L val ue. For this reason, even though the test data LSF may have met the criterion proposed in Equation (6) above, for sufficient data stability to have been reached-to justify test termination, the

_ quantity (L95 - 'LSF) should be constant or decreasing. The proposed criterion for evaluating (l'95 ~ 'LSF) is stated quantitatively below:

g 95,1 - 'LSF,i Equation (7)

L Define: O

=

D _g g D for all i during the last hour of the test Require:

j 9

EXTRAPOLATION CONSIDERATIONS Results of a test of less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration are vulnerable to speculation about their sensitivity to perturbations in the fictitious "next data point." Their vulnerability is, of course, a function of the data scatter, the quantity of data, and the margin between test results and the allowable ~ leakage rate. Consequently, one method of resolving this issue is 3-21 L

l l

I Lc n

05 i

g l

L l

u 8

?

I i

..I g,y i

e i

e l

I LSF I

I i

I f

g s

i I

e i

g l

l t

g 8

I e

e 1HR r,

i a

i s

l s

i e

l t

E I

t l

I s

8 g

i f

j j

t -3 t

t -1 t

t -4 n

n-2 n

n n

l TIME l

2 2D, THEN EQUATION (7) IS SATISFIED.

IF D5 ED4 2D3 2D j

Figure 3-7.

Illustration of Equation 7

r-to postulate a reasonable '%orst case" next data point and assess its impact on the test ret.11ts. Following on this line of thought, the following procedure was examined:

a)

Let Standard deviation i'i observed values of containment

=

air mass, Wi, about the LSF line.

b)

Let W24 =

A postulated data point taken at t = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> c)

Let N =

The LSF estimated value of W at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

24 9

d)

Let M(t) =

The allowable leakage rate minus L at the time of 95 test termination e)

Postulate W24 " 24 - t95 (n 2) 8

+

+l Et 2 - (n-2)T2 g

f)

Calculate L95 (at' t = 20) incorporating the postulated W24

. g)

Calculate 'M(t = test termination) - M(t=24)' x 100 M(t = test teniination) t i

.The calculation in step (g) above represents the percent change in the margin between test results and test allowable at the time of test termi-nation if the worst case mass ~value at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is postulated. This

. postulated erosion in margin was calculatcd for several tests with the time of test termination assumed to be the time at which the completion criteria were met. The results showed a minimal effect. This minimal effect from a postulated worst case data point lends a degree of assurance that terminating the test is acceptable. While this procedure is not proposed as a criteria, it does illustrate the validity of terminating a test when the proposed criterion are satisfied.

3-23

DATA AVERAGING CONSIDERATIONS Several testing organizations use data. loggers for Type A tests. In soma cases, these data loggers scan the dsta at frequent intervals (e.g. seconds)

I but only report the data as a mean value for a longer time interval (e.g.10 minutes). It appears that the standard statistical techniques recommended

. in Reference (4) are not adequate for such timeesveraged data. Eor example, time averaging tends to smooth the observed data. The statistical technique of Reference (4) does not account for this kind of preconditioning of the data and may therefore lead to an underestimate (or an overestimate) of the 95 percent confidence level leakage rate. This possibility is illustrated in Figure 3-8.

Based on the above discussion, the statistical criterion proposed herein are not recommended for application to tests using time-averaged data. Further investigation is required to develop the appropriate statistical techniques necessary to address time avercged data.

l 3-24

I l

l-l l

I.

r N

\\

N N

G G

o O

\\

X s

G.

N s

g g

O

,O O

o o

W (1b m)-

g G

e X

O O

9 Actual Data Readings X Smoothed Data Readings UCL of Smoothed Data

--- UCL of Actual Data t (hours)

Figure 3-8.

Illustration of Data Averaging 3-25

-n n:

1 Section 4 VALIDATIOh 0F CRITERION FIMAL CRITERION

Seven criterion for cetermining the completion of an ILRT were finally

~ proposed:

1. Use the absolute method, mass point technique.

2. The containment must be adequately modeled.

3. The 95 percent upper confidence level leakage rate must be zero or a positive value.

4. The calculated LSF leakage rate must be less than 75 percent of the l plant's allowable leakage rate criterion at test pressure.

5. The calculated 95 percent upper confidence level must be less than 75

. percent of'the plant's allowable leakage rate criterion at test pressure.

6. The calculated' 95 percent upper confidence level leakage rate as a function of time shall have stabilized with a negligible positive or

- negative slope.

7.- The calculated 95 percent upper confidence level leakage rate shall be converging with the LSF leakage rate.

W APPL'ICATION OF CRITERION-

-It was' postulated that when all seven proposed technical acceptance

-criterion are concurrently ~ met, the test can be terminated and be considered

' uccessfully complete.. - Test continuance should not significantly change s

- the test results.unless a significant physical change occurs.

r t

4-1 t

=

1 w

e+

+

r r,,

4 T-1r---r ev-y

-r rw

-^e

-^t w eit

-Few

-t-

-w

+

r

-+-----m-'

-, - '-ee.

-itt---+----pw

-n-** e v c t--t

-v

CRITERIA VALIDATION l

As stated previously,144 ILRTs were reviewed to determine usable informa-tion for this study. (See Appendix A for a detailed list.) Of those reviewed, fifty-three (53) had sufficient information avariable. The balance of reports lacked key elements rendering them insuificient for use in this study. Therefore, the validation is based on those fifty-three usable ILRTs. In forty-seven of the tests the technical acceptance criterion would have allowed a reduction in test duration.

The LSF leakag'e rate tests were categorized in four trends. Naturally, hybrids of these trends exist but case-by-case examination for validation purposes would be far too cumbersome and yield a similar result. These four distinct trends of LSF leakage rates may be described as follows:

1. The indicated leakage rate is always positive and always under the plant's technical specifications.

2.- The indicated leakage rate begins above plant's technical specificatica.

and gradually decreases to a value below the plant's technical specifi cation.

3. The indicated leakage rate oscillates above and below the plant's t;c.inical specification (perhaps several times), then levels off to a value under the plant's technical specification.

4. The indicated leakage cate initially shows a negative leakage rate then stabilizes but does not exceed the plant's technical specification.

Termination based on the study criterion of a test exhibiting the first trend'(see Figure 4-1) is restrained by the 95 percent upper confidence

' level-leakage rate. The LSF leakage rate is relatively stabie and consistently under the plant's accept <.nca criterion. Therefore, when a sufficient quantity of points are taken, the 95 percent upper confidence level leakage rate will typically both drop under the plant's technical specification and converge with the LSF leakage rate to allow test termination. Table 4-1 lists 15 reports which fall into this category.

4-2

4 i

1 0.5 -

4 0.4 --

ALLOWABLE N

0.3 -

TEST g,

LEAKAGE j

g g0.2 RATE Etg0.1-a l

0 LSF e

I I

I I

I I

I I

T 0.0 i

i i

u

$ -0.1 -

Time (hours) 8 9

10

-0.2 -

-0. 3 -

-0.4 -

-0. 5 -

4 i

l Figure 4-1.

Typical LSF Leakage Rate Trends - Trend I i

1 I

1 1

l Table 4-1 CATEGORIZED REPORTS UNDER TREND 1 PLANT NAME DATE ARKANSAS NUCLEAR UNIT 1 1981 BROWN'S FERRY UNIT 3 1979 CRYSTAL RIVER UNIT 3 1980 FARLEY UNIT 2 1980 FITZPATRICK 1982 GRAND GilLF 1982 LASALLE 1982 MILLSTONE 1 1976 0CONEE UNIT 2 1980 PALISADES 1982 PILGRIM 1982 QUAD-CITIES UNIT 2 1980 SAN ONOFRE UNIT 3 1962 SEQUOYAH UNIT 2 1981 TURKEY POINT 3 1981 TOTAL TREND 1 REPORTS:

15 32% OF 47 REPORTS ARE TREND 1 Termination of tests based on the study criterion exhibiting the second and third trends (see figures 4-2 and 4-3 respectively) is restrained by the amount that the LSF leakage rate and the 95 percent upper confidence level leakage rate are below the plant's technical criterion. For example, if the LSF leakage rate and 95 percent upper confidence level leakage rate appear to stabilize only slightly below the plant's technical specification (see the dotted line identified as LSF, in Figure 4-2), the very small variance in the values of the points will require continuance of the test. However, if the margin is sizable (e.g. over 30 percent), the restraint is effec-tively eliminated and the study criterion will allow test termination.

Three reports fit the second trend and 17 reports fit the third trend, as illustrated in tables 4-2 and 4-3.

4-4

E L

B E

A G

W A

OTKE LSAT LEEA 0

ATLR I

i1 F'

FS S

LI i9 L

I i8 2

d ner i7 T

I DN s

E y

F d

G g

S n

E g

L e

L r

i6 T

I e

ta

)

R s

ru e

o g

a 5 h k

(

a e

e L

m i

F T

SL 1

i4 lac ipy T

li3 2

4 erug l;2 i

F l

i1 6

5 4

3 2

0 2

3 4

5 1

1 0

0 0

0 0

0 0

0 0

0 0

0

$ Ea oE GEE Ea

=in

0.5 -

0.4 -

/

ALLOWABLE en

[oS 0.3 -

TEST h

uo 0.2

/

N' =N a

0*1 -

LSF

?

e cn en I

I I

I I

(

l i

I 0

0.0 i

i i

i i

I g

1 2

3 4

5 6

7 8

9 10 a

-0.1 -

i Time (hours) a

-0. 2 -

-0.3 -

d

-0.4 -

-0.5 -

1 i

l Figure 4-3.

Typ cal LSF Leakage Rate Trends - Trend 3 l

4

g_

Table 4-2 CATEGORIZED REPORTS UNDER TREND 2 PLANT NAME DATE P0lNT BEACH UNIT 1 1981 SURRY UNIT 1 1981 THREE MILE ISLAND 1981 TOTAL TREND 2 REPORTS:

3 6% OF 47 REPORTS ARE TREND 2 Table 4-3 CATEGORIZED REPORTS UNDER TREND 3 PLANT NAME DATE COOPER 1980 FITZPATRICK 1978 MILLSTONE UNIT 1 1981 NINE MILE POINT 1979 NORTH ANNA UNIT 1 1981 NORTH ANNA UNIT 2 1979 OCONEE UNIT 1 1980 Pt.LISADES 1978 PILGRIM 1980 POINT BEACH UNIT 2 1982 H. B. ROBINSON UNIT 2 1982 SAN ONOFRE UNIT 2 1980 SURRY UNIT 2 1982 THREE MILE ISLAND UNIT 1 1978 YdRKEY PolNT UNIT 3 1982 YANKEE ROWE 1980 TOTAL TKEND 3 REPORTS:

16 34% OF 47 REPORTS ARE TREND 3 4-7 L

Termina. tion based on the study criterion of lests exhibiting the last trend (see Figure 4-4) has the Same restraints as the first trend but may also be subject to the restraint trends 3 and 4.

However, the latter restraint was not observed in the test reports analyzeo. As discussed in Section 3, trend

' 4 is caused by a relatively low initial mass point value, which, when com-pared to the values of-the subsequent mass points, produces a negative leakage rate. The final value must be positive at the termination point and not negative, as the dotted line in Figure 4-4 indicates; otherwise, the study criterion would not-be met. Thirteen ILRT reports are categorized in

~ this trend, as shown in Table 4-4.

Table 4-4 CATEGORIZED REPORTS UNDER TREND 4 PLANT NAME DATE ARKANSAS NUCLEAR UNIT 2 1981 BEAVER VALLEY UNIT-1 1978 BRUNSWICK UNIT 2 1977 CALVERT CLIFFS UNIT 2 1979 DAVIS-BESSE UNIT 1 1960 FARLEY UNIT 1 1981 MILLSTONE UNIT 2 1979

~OCONEE' UNIT 3 1981 QUAD-CITIES UNIT 1 1979 SEQUOYAH 1 1979 SURRY UNIT 2 1980 TROJAN HALF PRESSURE 1975

-TURKEY-PolNT UNIT 4 1980 TOTAL TREND 4 REPORTS 13 28% OF 47 REPORTS ARE TREND 4 In each trend category and in' every test report evaluated, when the techni-cal acceptance criterion were met, successful completion of the test could

'be predicted. Furthermore, the study criterion remained satisfied through-out the balance of the test. However for those tests in which there was little margin between calculated and allowable leakage rate, application of the technical-acceptance criterion did not permit early test tennination.

4-8

=+w--

y---*i-*

g y

y y-

.m

+r=-

-av 9yp-

--+

y er&v'

1 0.6 -

0.5 -

• >,0. 4 54 ALLOWABLE TEST L' ' 0*3 -

LEAKAGE EE RATE eg 0.2 3

ce 8,o, 0.1 -

- L7F

=

i 1

1 I

[

I I

I 2

c; 0.0 i

i i

i i

p

-== -- ama aus"" ""~3

-0.1 -

1 2

3

1. 4 5

6 7

8 9

10 Time (hours!

-0.2

-0.3 -

-0.4 -

LEGEND:

-0.5 -

LSF - example of positive final LSF LSF' - example of neystive final LSF' Figure 4-4.

Typical LSF Leakage Rate Trends - Trend 4

s

[

RESULTS OF~ CRITERIA UTILIZATION

. This study has found that using the validated technical acceptance criterion results in a shorter test without a significant change in test results. In theory, one might think by using the technical acceptance criterior.- the -

plant's"LSF leakage rate per day may be eroded through reporting a higher percentage. rate, but this study shows that this is not normally the case.

Table 4-5 compares the actual final reported duration and LSF. leakage rate

.to the technical acceptance criterion duration anc reported LSF leakage rate

at that timepoint. This comparison indicates that substantial time can be saved 'while-producing LSF leakage rate results well within the plant's technical' speci fication. In those instances where a test was only marginally acceptable,~ the technical acceptance criterion justifiably did not allow any time reduction.

T I

4 1

k Y

4 f

1 f

-t

---k-4-10

r.

Table 4-5 AFFECT OF TECHNICAL ACCEPTANCE CRITERION ON TEST DURATION AND LEAKAGE RATE DURATION OF TESTS IN HOURS LEAKAGE RATE IN %/ DAY Allowed Allowed by Differ-by Differ-Actual Criterion ence Actual Criterion ence

-1. ANO 1 1981 9.75 6.75 3.00

.0382

.0161

.0021

2. ANO 2 1981 10.00

'9.25

.75

.0271

.0292

.0021

'3; BEAVER VALLEY 1 1978 24.00 13.00 11.00

.0305

.0414

.0109

4. BROWN'S FERRY 3 1979 24.23 2.67 21.56-

.1576

.2410

.0834

5. BRUNSWICK 2 1977 25.00 7.67 17.33

.3054

.2845

.0209

~

6. CALVERT CLIFFS 2 1979 8.25 4.75 3.50

.0523

.0380

.0143

7. COOPER 1980 13.67 10.67 3.00

.4209

.4115

.0094

'8.

CRYSTAL RIVER 3 1980 24.00 21.00 3.00

.1333

.1420

.0093

9. DAVIS-BESSE 1980 8.00 3.25 4.75

.0642

.0883

.0241

.10.'FARLEY 1 1981 25.25 17.25 8.00

.0441

.0768

.0327

11. FARLEY 2 1980 24.00 6.00 18.00

.0331

.1053

.0722

.12. FITZPATRICK 1978 26'.00

~13.00 13.00

.2914

.2938

.0024

13. FITZPATRICK 1982 24.00 9.00 15.00

.2004

.2210

.0206 14' GRAND GULF 1982 8.00 2.75 5.25

.0736

.1038

.0302

15. LASALLE 1982 26.18 8.84 17.34

.3780

.3805

.0025

16. MILLSTONE 1.1976 24.00 8.00 16.00

.6128

.7767

.1639

17. MILLSTONE 1 1981-24.50 11.25 13.25

.2979

.5774

.2984 18.-MILLSTONE 2 1979 15.25 3.25 12.00

.0622

.0020

.0602

19. NINE MILE POINT 1975 24.75 14.00 10.75

.7439

.8046

.0607

20. NORTH ANNA UNIT 1 1981 24.10 7.23 16.87

.0111

.0306

.0195

21. NORTH ANNA UNIT 2 1979 24.00 18.00 6.00

.0331

.0451

.0120

22. OCONEE UNIT 1 1980 24.00 8.75 15.25

.0220

.0377

.0157

23. OCONEE. UNIT 2'1980 15.00' 2.75 12.2i,

.0537

.0413

.0124

24. OCONEE UNIT 3 1981 14.92 9.92 5.00

.0307

.0228

.0079

25. PALISADES 1978-24.00 17.00 7.00

.0084

.0028

.0056

~ 26. PALISADES 1982 23.00 8.00 15.00

.2001

.1103

.0898

'27. PILGRIM 1980 28.00 8.67 19.33

.4170

.2992

.1178

.28. PILGRIM 1982 23.00 11.00 12.00

.2002

.1855

.0147 4-11 l

~

Table 4-5 I

(Continued)

DURATION OF TESTS IN HOURS LEAKAGE RATE IN %/ DAY Allowed Allowed by Differ-by Differ-Actual Criterion ence Actual Criterion ence 29.' POINT BEACH UNIT 1 1981 12.00 8.75 3.25

.0804

.0967

.0163

30. POINT BEACH UNIT'2 1982 12.00 10.00 2.00

.0480

.0390

.0090

31. QUAD-CITIES UNIT 1 1979 22.75 15.00 7.75

.0054

.0002

.0052

32. QUAD-CITIES UNIT 21980 23,75 10.00 13.75 4369

.5136

.0767 33.'H. B. ROBINSON 2 1982 24.00 13.00 11.00

.0186

.0001

.0185

34. SAN ONOFRE UNIT 2 1980~

24.00 9.25 14.75

.0576

.0E14

.0038

35. SAN ON0FRE UNIT 3 1982 8.00 5.50 2.50

.0157

.0254

.0097

36. SEQUOYAH UNIT 1 1979 27.38 3,50 23.88

.0005

.0019

.0014

37. SEQUOYAH UNIT 2 1981 25.67 7.33 18.34

.1380

.1488

.0108

38. SURRY UNIT 1 1981 13.00 12.00 1.00

.0328

.0339

.0011 39.-SURRY UNIT 2 1980 15.67 5.00 10.67

.0353

.0424

.0071

40. SURRY UNIT 2 1981 12.00 9.00 3.00

.0187

.0155

.0032

41. THREE MILE ISLAND 1 1978 44.50 17.00 27.50

.0607

.0343

.0254

42. THREE MILE ISLAND 1 1981 24.00 9.30 14.70

.0230

.0402

.0172

-43. TROJAN 1975 HALF PRESSURE 9.00 9.00.0007

.0007

.0000

44. TURKEY POINT UNIT 3 1979 12.00 6.00 5>00

.0975

.1207

.0232

45. TURKEY POINT UNIT 31981 12.00 5.25 6.75

.0314

.0358

.0044

46. TURKEY POINT UNIT 4 1980 15.67 6.33 9.34

.0318

.0414

.0096

47. YANKEE R0WE 1980 30.50 13.75 16.75

.0477

.0734

.0257 AVERAGES 19.67 9.22 10.45

.1278

.1354

.0333 DATA

SUMMARY

. Distribution by Duration Time using criterion:

Between 4 and 8 hrs: 19....... 40%

=Between 8 and 12 hrs: 18....... 35%

Between 12 and 16 hrs: 4.......

9%

- Between 16 -and 20 hrs: 6....... 13%

TOTAL.

47...... 100%

4-12

y-:

3 i

Section 5 CONCLUSIONS AND RECOMMENDATIONS COSCLU310NS r

~ For every case where the test net tnese technical acceptance criteria, the t

measured leakage rate remained below the allowable leakage rate for the duration of the test. This is evidence that these criteria are sufficiently conservative to provide for duration testing substantially less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The average required test duration using the technical acceptance criteria on the forty-eight tests investigated was approximately nine hours. The average actual test duration of these same 48 tests was 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, so a net average savings of 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> was demonstrated. Using a replacement cost value of 3 cents per kilowatt hour generated, an 800 megawatt electric plant will save $24,000 tor every hour that the test is shortened. Thus, an 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> savings could produce a $264,000 savings.

l

. The duration of containment integrated leakage rate testing can be reduced while yielding acceptable and technically accurate results.

' RECOMMENDATIONS Hore validation is recommended because this study only applied the criterion for fifty-three tests. Of the 252 tests estimated to be available, the fifty-three tests represent a ?l% sample. See Table 5-1 for distribution by 5-1

duration and test number. For distribution by containment type, see TaDie 5-2.

See Table 5-3 for distribution by termination time.

Temperature weighting factors can have a large impact on the calculations for containment mass. Since operating plant containments have substantial j

heat sources and sinks, a more accurate method of selecting temperature weighting factors should be developed. This would create a better containment mass model, which would not only make the tests more accurate, but would permit even earlier test termination through the use of technical acceptance criteria developed in this project.

The portion of the project which studied temperature weighting factor sen-sitiv'ity only analyzed data frcm 25 ILRTs out of 252 ILRTs (approximately a 10 percent sample). Since 11 out of 25 were from pre-operational ILRTs without operational containment heat loaos and a different 11 out of 25 had temperature changes of less than 0.5'F over the test, some expansion of the data base may' be warranted to develop a more definitive evaluation of the effects of temperature weighting factors.

Data loggers were used in at least 63 of the 144 ILRT reports reviewed. All of these plants used the data logger for collection and peccessing of data.

The number of plants which also use the data logger to report the mean value of the. data is not known. The use of data loggers can affect the statistical calculation of the LSF of leakage rate and the 95 percent upper confidence level leakage rate. The potential exists for data loggers to be used in a manner that is inconsistent with the statistical basis for LSF and UCL calculation. The use of data loggers needs to be further studied to ascertain the impact on containment mass calculations.

5-2

r,-

Table 5-1 ALL ILRTs DISTRIBUTED BY TEST NUMBER AND DURATION Total No.

<24 Hour

=24 Hour

>24 hour Reports Test Number of Tests Juration Duration Duration Unavailable

• Preoperational 59 13 22 6

18 First 56 13 19 12 12 Second 56 12 13 9

22 Third 38 2

4 6

26 Fourth 17 0

4 0

13 Fifth 10 0

0 0

10 Sixth 6

0 1

0 5

Seventh 5

0 2

1 2

Eighth 4

0 3

1 0

Ninth 1

0 1

0 0

TOTALS 757 W

ET 3T IUF

• Some preoperational reports show a full pressure and a half pressure test. Ten (10) preoperational reports show both full and half pressure tests with a duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> each. These two test types have been counted as one report, to make a total of 144 tests.

5-3 L

Table 5-2 1

EVALUATED CONTAINMENTS DISTRIBUTED BY CONTAINMENT TYPE Description TYPE 1 TYPE 2 TYPE 3 TYPE 4 TYPE 5 TYPE 6 TOTALS 1

l Raw numbers 1

3 28 11 2 45

' Percentages

. per type that meet criteria 2.0%

7.0%

62.0%

24.6%

4.4%

0.0%

100%

E33533535553335353533355 3 35 35535 35 3555 33533 3 33 335 5553 33333E3R335ESSEEESSEE Totals for operating plants TYPE 1 TYPE 2 TYPE 3 TYPE 4 TYPE 5 TYPE 6 TOTALS Raw numbers 5'

8 39 22 3

1 78 Percentages per type per operating plants 6.6%

10.0%

50.0%

28.0%

4.4%

1.0%

100%

E53533 33553333333333333333335F33333335233355333335535333233333 333333 333555

-Difference in percentages-

-4.0%

-3.0%

12.0%

-3.4%

0.0%

-1.0%

DEFINITIONS:

Type 1 = Steel sphere Type 2 = Steel cylinder Type 3 = Reinforced concrete cylinder with steel liner Type 4 = Steel drywell and wetwell

. Type 5 = Reinforced concrete drywell and wetwell with steel liner Type 6 = Reinforced concrete pressure ve. sal 5-4

Table 5-3 ILRT STUDY REPORTS DISTRIBUTED BY TERMINATION TIME TEST DURATION DATA

SUMMARY

< 8 HRS.

13 8.3%

> 8 AND < OR = 12 HRS.

18 11.5%

> 12 AND < OR = 16 HRS.

5 3.2%

> 16 AND < OR = 20 HRS.

2 1.2%

> 20 AND < 24 HRS.

0 %

'= 24 HRS.

77 49.0%

>.24 HRS.

23 14.7%

DURATION NOT SHOWN/NO REP:

15 12.1%

TOTAL NUMBER OF TESTS:

157 100.0%

1 i

I 5-5

(<-

F Section 6 REFERENCES STANDARDS AND ARTICLES REFERENCES 1.10 CFR 50, Appendix J, Primary Reactor Containment Leakaoa Testing for Water-Cooled Power Reactors, Atomic Energy Commission,1971.

2. ANSI N45.4-1972, Leakage Rate Testing of Containment Structures for Nuclear Reactors, American Nuclear Society through American Nuclear Standards Institute,1972.

3. BN-TOP-1, Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants, Bechtel Corporation, approved by AEC,1972.

4. ANSI /ANS 56.8-1981, Containment System Leakage Testing Requirements, American Nuclear Society through American Nuclear Standards Institute, 1981.

5. EPRI Report NP-267, Sensor Response Time Verification,_ Project #503-1, October 1976.

6. EPRI Report NP-834, Insitu Response Time Testing of Platinum Resistance Thermometers, Project #503-3, July 1978, Volumes 1 and 2.

7. ANS Rod Workshop Proceedings, Integrated Leak Rate Tests, State of the Art Technology, San Diego, California, Jan 18-19, 1982.

8. EPRI. Report NP2726, Containment Integrated Leak-Rate Testing Improvements, Project a1390-1, November 1982.

9. David L. Fleshood, Containment Leak Rate Testing, Why the Mass-Plot Method is Preferred, Power Engineering, February 1976.

10. Walter Sommer, Burns & Roe Inc, Automated Containment Tests Cut Downtime, Save $2.5 M, Electric Light and Power, January 1979.

REPORT REFERENCES

11. Arkansas Power & Light Co., Arkansas Nuclear One, Unit 1, Reactor Containment Building Integrated Leak Rate Test Report, Docket #50-313, 19/3.

6-1

12. Arkansas Power & Light Co., Arkansas Nuclear One, Unit 1, Reactor Containment Building Integrated Leak Rate Test Report, Docket #50-313-147, 19/8.

13. Arkansas Power & Light Co. with Bechtel Corp., Reactor Containment Building Integrated Leakage Test for Arkansas Nuclear Unit One, Docket

1. 50-313, 1981.

14. Arkansas Power & Light Co. with Bechtel Corp., Arkansas Nuclear One, Unit 2, Reactor Containment Integrated Leakage Rate Test, Docket #50-

'368-551, 1977.

15. Arkansas Power & Light Co., Primary Reactor Containment Integrated Leakage Rate Test for Arkansas Nuclear Unit Two, Docket #50-368,1981.

16. Duquesne Light Co.', Report on Reactor Containment Building Integrated Leak' Rate Test Type A, B, and C on the Beaver Valley Power Station, Unit 1,. Docket #50-334-253,19/b.

17. Duquesne Light Co., Reactor Containment Integrated Leakage Rate Test Report for Beaver Valley Power Station, Unit l, Docket 50-331, 1978.

18. Consumer's Power Co., Special Report #17, Reactor Containment Integrated Leak Rate Test for Big Rock Point, Docket #50-155-300, 1974.

' 19. Consumer's Power Co., Reactor Containment Building Integrated Leak Rate Test for Big Rock Point, Docket #50-155, 1978.

20.~ Tennessee Valley Authority, Brown's Ferry Nuclear Power Station, Unit 1, Reactor Containment Building Integrated Leak Rate Test Report, Docket #50-259-102,1973.

21. Tennessee Valley Authority, Brown's Ferry Nuclear Power Station, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-259-928, 1976.

22. Tennessee Valley Authority, Browa's Ferry Nuclear Unit 1, Containment Integrated Leak Rate Test, Docket #50-259,1980.

23. Tennessee Valley Authority, Brown's Ferry Nuclear Power Station, Unit 2, Reactor Containment Building Integrated Leak Rate Test, Docket #50-260-219, 19/4.

- 24. Tennessee Valley Authority, Brown's Ferry Nuclear Station, Unit 3, Containment Building Integrated Leak Rate Test, Docket #50-259,1979.

25. Carolina Power & Light Co., Brunswick Steam Electric Plant, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-325-405, 1976.

' 26. Carolina Power & Light Co., Brunswick Steam Electric Plant, Unit 1, Containment Building Integrated Leakage Rate Test, Docket #50-325, 1981.

6-2

27. Carolina Power & Light Co., Brunswick Steam Electric Plant, Unit 2, Integrated Primary Containment Leak Rate Testing, Docket #b0-324-1268, 1978.

28. Baltimore Gas & Electric Co., C&1 vert Cliffs Nuclear Power Plant, Unit 1, Final Report on Reactor Conij' ment Building Integrated Leak Rate Test, Docket #50-317-169, 1973

29. Baltimore Gas & Electric Co., Calvert Cliffs Nuclear Power Plant, Unit 1, Reactor Containment Building Integrated Leakage Rate Test Report, Docket #50-317,1982.

30. Baltimora Gas & Electric Co., Calvert Cliffs Nuclear Power Plant, Unit 2, Primary Reactor Containment Integrated Leak Rate Test, Docket

1. 50-318, 1979.

31. Quincy Billingsly III, Connecticut Yankee Atomic Power Co., Containment Operation & Testing,1976.

32. Connecticut Yankee Power Co., Containment Leakage Rate Testing, Docket

1. 50-213, 1981.

33. Indiana & Michigan Power Co., Donald C. Cook Nuclear Plant, Unit 1, Peactor Containment Building Integrated Leak Rate Test Report, Docket

1. 50-315,-19/8.

34. Indiana & Michigan Electric Co., Donald C. Cook, Unit 2, Reactor Containment Building Integrated L eak Rate Test, Docket #50-315,1981.

35. Nebraska Public Power District, Cooper Nuclear Station, Reactor Containment Building Integrated Leak Rate Test, Docket #50-298-134, 1974

36. Nebraska Public Power District, Cooper Nuclear Station, Reactor Containment Integrated Leak Rate Test, Docket #50-298, 1976.

37. Nebraska Public Power District, Cooper Nuclear Station, Summary Technical Report on Primary Integrated Leak R:te Test, Docket #50-298, 1980.

38. Gilbert Associates and Florida Power Corp., Crystal River Unit 3 Nuclear Generating Plant, Reactor Containment Building Integrated Leak Rate Test, Docket #b0-302-320, 19/6.

39. Gilbert Associates and Florida Power Corp, Crystal River Unit 3 Nuclear Generating Plant, Reactor Containment Building Integrated Leak Rate Test, Docket #50-302,1980.

40. ioledo-Edison Co., Davis-Besse Nuclear Power Station Unit 1, Containment Vessel Integrated Leak Rate Test, Docket #50-346,1976.

41. Toledo-Edison Co., Davis-Besse Nuclear Power Station Unit 1, Containment Vescel Integrated Leak Rate Test, Docket #50-346,1980 6-3

N

42. Pacific Gas and Electric Co., Diablo Canyon Nuclear Power Plant, Unit 1, Reactor Containment Building Integrated Leak Rate fest and Structural 1;.tegrity Test, Docket #50-275-203,1976.

43. Pacific Gas and Electric Co., Diablo Canyon Site, Unit #1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-2/5-OL,

-1978.

44. Pacific Gas and Electric Co., Diablo Canyon Site, Unit #2, Reactor Containment Integrated Leak Rate Test, Docket #50-275, 1980.

45. Commonwealth Edison Co., Dresden Nuclear Power Station, Unit 1. Primary Containment Integrated Leak Rate Test, Docket #50-010-878,1977.

46. Commonwealth Edison Co., Dresden Nuclear Power Station, Unit 2. Reactor Containment Building Integrated Leak Rate Test, Docket #50-237-1092, 19/6.

47. Commonwealth Edison Co., Reactor Containment Building Integrated Leak Rate Test for Dresden Nuclear Power Station, Unit 3, Docket #050-249,DPR-25, 198".

48. Iowa Electric Lighc and Power Co., Reactor Containment Integrated Leakage Rate Tcat for the Duane Arnold Energy Center,1973.

49. Iowa Electric Light and Power Co., Reactor Containment Integrated Leakage Rate Test for the Duane Arnold Energy Center,1978.

50. Iowa Electric Light and Power Co., Reactor Containment Integrated Leakage Rate Test for the Duane Arnold Energy Center,1980.

51. Iowa Electric Light and Power Co., Reactor Cbntainment Integrated Leakage Rate Test for the Duane Arnold Energy Center,1981.

52. Iowa Electric Light and Power Co., Reactor Containment Integrated Leakage Rate Test for the Duane Arnold Energy Center,1983.

53. Alabama Power Co., Final Report on the Primary Reactor Containment Integrated Leak Rate Test for Joseph M. Farley Nuclear Plant - Unit 1, Docket #50-348, IV/7.

54. Alabama Power Co., Final Report on the Primary Reactor Containment Integrated Leak Rate Test for Joseph M. Farley Nuclear Plant - Unit 1, Docket #50-348,1981.

55. Alabama Power Co., Joseph M. Farley Nuclear Plant - Unit 2, Containment Integrated Leak Rate Test, Docket #50-364,1980.

56. Power Authority of the State of New York, James A. Fitzpatrick Nuclear Power Plant Reactor Containment Building Integrated Leak Rate Test Type A, B, & C., Docket #50-333,1978.

6-4

r x

.57. Power Authority of the State of-New York, James A.~ Fitzpatrick Nuclear

Power Plant Reactor Containment ' Building Integrated Leak Rate Test Type A, B, C, Docket #50-333,195Z.

O

58. Omaha Public Power District and Nuclear Services Corp., Fort Calhoun

? Station 1 Reactor Containment Building Leak Rate Test, Docket #50-285-105, 1973.:

2 59. Omaha Public Power District and Nuclear Services Corp., Fort Calhoun l Station, Unit 1. : Reactor Containment Building Integrated Leak Rate Test, First Periodic Type A Test, Docket #50-285-734,1976.

. 60. Omaha Public Power District and Nuclear Services Corp., Reactor

. Containment Building Integrated Leakage Rate Test at Fort Calhoun Station, Docket 750-Zub, 1950.

61. Gilbert ' Associates.and Rochester Gas and Electric Corp., The Robert

Emmit Ginna Nuclear Power Station Type A, B, & C Reactor Containment Building Periodic Retest Results, GAI-Report #1783,197Z.

62. Rochester Gas and Electric Corp., Robert Emmit Ginna Nuclear Power Plant,' Unit I, Reactor Building Integrated Leak Rate Test, Docket #50-244-539, 1976..

63. Mississippi Power. and Light Co., Grand Gulf Nuclear Station Unit 1 Primary' Reactor Containment Integrated Leakage Rate Test Report, Docket j

tbO-416, 195Z.

~ 64. Bechtel Power Corp. and Georgia Power Co., Reactor Containment Building 2

Integrated Leakage Rate Test for the Edwin I. Hatch Nuclear Plant Unit

+

Z,, Bechtel-Job #6511, Docket #50-366,19/8.

- 65. Gilbert Associates and Consolidated Edison Corp., Preoperational _

Integrated Leak Rate Test of the Reactor Containment Building, Indian 1

Point Unit 2, Docket #50-247, 1971,

^

66. Ebasco Services Inc. and Consolidated Edison Co., Reactor Containment

Building Integrated Leak Rate Test for Indian Point Unit 2, Docket #50-

Z47-811, 1976.

t

[

67; Ebasco Services Inc, Plant Operations and Betterment Dept and Consolidated Edison Co., Indian Point Unit 2, Reactor Containment Building Integrated Leak Rate Test, Summary Technical Report, Docket

~

_#50-528, 1979.-

68. ' Gilbert Associates and Consolidated Edison Corp., Indian Point Unit 3 Preoperational Integrated Leak Rate Test of the Reactor Containment p

Building, gal report #1861,19/5.

i:

- 69. Ebasco Services Inc. Plant' Operations. and Betterment Dept and Power Authority of the State of New York, Indian Point Nuclear Power Plant Unit 3, Reactor Containment Integrated Leak Rate Test, Docket #50-Z86, L

1978 I'

r.

6 !

~

. ~.

70. Wisconsin Public Service Corp., Kewaunee Nuclear Power Plant Shield Building Integrated Leak Rate Test, Docket #50-305-180,1974.

71. Wisconsin Public Service Corp., Kewaunee Nuclear Power Plant Reactor Containment Building Integrated Leak Rate Test, Docket #50-305P,1980.

72. Dairyland Power Cooperative, Lacrosse Boiling Water Reactor Type A Containment Building Integrated Leak Rate Test, Docket # 50-409-298, j

-19/5 1

73. Dairyland, Power Cooperative, Lacrosse Boiling Water Reactor, Reactor Containment Building Integiated Leak Rate Test (ILRT), Docket #50-409, 19/8

74. Dairyland Power Cooperative, Lacrosse Boiling Water Reactor, Reactor Containment Building Integrated Leak Rate Test (ILRT), Docket #50-409, 1979

75. Dairyland Power Cooperative, Lacrosse Boiling Water Reactor, Reactor Containment Building Integrated Leak Rate Test (ILRT), Docket #50-409, 1980.

76. Commonwealth Edison Co., Lasalle County Nuclear Power Station Unit 1, Reactor Containment Building Integrated Leak Rate Test,1982.

77. Maine Yankee Atomic Power Co., Maine Yankee Atomic Power Plant Reactor Containment Integrated Leak Rate Test, Docket #50-309-379,1975.

78. Maine Yankee Atomic Power Co., Maine Yankee Atomic Power Plant Reactor Containment Integrated Leak Rate Test, Docket #50-309,1982.

79. Duke Power Company, McGuire Nuclear Station, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-369,1979.

80. Duke Power Company, McGuire Nuclear Station, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-369,1981.

81. Northeast Nuclear Energy Co., Millstone Nuclear Power Station Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-245, 19/U.

82. Northeast Nuclear Energy Co., Millstone Nuclear Power Station Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-245-23, 1973.

83. Nortbeast Nuclear Energy Co., Millstone Nuclear Power Station Unit 1 Primary Containment Leak Rate Test Report, Docket #50-245-934,1976.

84. Northeast Utilities, Millstone Nuclear Power Station, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-245,1981.

85. Northeast Nuclear Energy Co., Summary Technical Report of the Primary, Containment Leak Rate Test Performed at Millstone Nuclear Power Station, Unit 2, Docket #50-336-237,1975.

6-6

(L 86.' Northeait Utilities, Millstone Nuclear Power Station, Unit 2, Reactor Containment Building Integrated Leak Rate Test, Docket #b0-336,1919.

87. Northern States Power Co., Monticello Nuclear Generating Plant, Unit 1, Containment Building Integrated Leak Rate Test, Docket #50-263,1980.

88. Niagra Mohawk Power Corp., Nine Mile Point Nuclear Station, Unit 1, Reactor Containment Buildino Integrated Leak kate Test, Docket #50-220-

~

6//, 1975.

89. Niagra Mohawk Power Corp., Nine Mile Point, Unit No.1, Reactor Containment Building Integrated Leakage Rate Test, Docket #50-220, 1979.

90. Virginia Electric and Power Co., North Anna Power Station, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Type A, B, and C, Docket #50-338,1981.

91. Virginia Electric and Power Co., North Anna Nuclear Power Stetion, Unit 2, Reactor Containment Building Integrated Leak Rate Test, Type A, B, and C Preoperational Test, Docket #50-339,1979.

92. Duke Power Co. and Bechtal Corp., Oconee Nuclear Station, Unit 1, Integrated Leak Rate Test of the Reactor Containment Building, Docket

1. 50-269, 1971.

93. Duke Power Co., G;onee Nuclear Station, Unit 1, Report on Integrated Leak Rate Test of the Reactor Containment Building, Docket #50-269-5/8, 1976.

94. Duke Power Co., Oconee Nuclear Station, Unit 1, Reactor Containment Integrated Leak Kate Test, Docket #b0-269,1960.

95. Duke Power Co., Oconee Nuclear Station, Unit 2, Integrated Leak Rate Test of the Reactor Containment Building, Docket #50-2/0-111,19/3.

96. Duke Power Co., Oconee Nuclear Station, Unit 2, Integrated Leak Rate Test of the Reactor Containment Building, Docket #50-2/0,1971.

97. Duke Power Co., Oconee Nuclear Station, Unit 2. Reactor Containment Building Integrated Leak Rate Test, Dccket #50-270,1980.

98. Duke Power Co., Oconee Nuclear Station, Unit 3, Report on Integrated Leak Rate Test of the Containment Building, Docket #50-287-160,1974.

99. Duke Power Co., Oconee Nuclear Station, Unit 3, Integrated Leak Rate Test of the Reactor Containment Building, Docket #50-287,1978, 100. Jersey Central Power and Light Co., Summary Technical Report of the Reactor Containment Building Integrhted Leak Rate Test for the Oyster Creek Nuclear Generating Station,1969.

F 6-7

T 101. Jersey Central Power and Light Co., Reactor Containment Building Integrated Leak Rate Test for the Oyster Creek Nuclear Generating Station, Docket #bO-Zl9, 19/B.

' 102. Jersey Central Power and Light Co., Summary Technical Report of the Reactor Containment Building Integrated Leak Rate Test for the Oyster Creek Nuclear Generating Station, Docket #50-219 1980.

l 103. Consumers Power Co., Palisades Plant Integrated Leak Rate Test of the Reactor Containment Building, Bechtel Job 5935, 1970.

104. Consumers Power Co., Special Report 8, Palisades Plant Reactnr Containment Building Integrated Leak Rate Test, Docket #50-255-327, 1974.

105. Consumers Power Co., Special Report 10 Palisades Plant Reactor Containment Building Integrated Leak Rate Test, Docket #50-255,1978.

106. Consumers Power Co., Special Report 11, Palisades Nuclear Plant Reactor C_o_ntainment Building Integrated Leak Rate Test, Docket #50-255,1982.

o 107. Philadelphia Electric Co., Peach Bottom Atomic Power Station, Unit 2, Reactor Containment Building Integrated Leakage Rate Test Report, Docket #b0-Z//-863,19/6.

108. Philadelph'a Electric Co., Peach Bottom Atomic Power Station, Unit 2, Primary Reactor Containment Integrated Leakage Rate Test Report, Docket

1. 50-277, 1980.

109. Philadelphia Electric Co., Peach Bottom Atomic Power Station, Unit 3, Reactor Containment Building Integrated Leak Rate Test Report, Docket

1. 50-278, 1977.

110. Philadelphia Electric Co., Peach Bottom Atomic Power Station, Unit 3, Primary Reactor Containment Integrated Leak Rate Test, Docket #50-278, 1981.

111. Boston Edison Co., Pilgrim Nuclear Power Station, Unit 1, Report on i-Primary Containment Integrated Leak Rate Test, Second Periodic Test and Local Leak Rate Test Results Conducted January Through May,1980, Docket #50-293,1980.

112. Boston Edison Co., Reactor Centainment Building Integrated Leak Rate Test Types A, B, and C for the Pilgrim Nuclear Power Station, Unit 1,

[

Docket f50-Z93, 1982.

-113. Bechtel Corp. for Wisconsin Electric Power Co. and Wisconsin Michigan Power Co., Point Beach Nuclear Plant, Unit 1, Initial Integrated Leak Rate Test of the Reactor Containment Building, Bechtel Job #6118,1970.

114. Nuclear Services Corp. for Wisconsin Electric Power Co., Point Beach Nuclear Plant, Unit No.1, Reactor Containment Building Integrated Leak Rate Test,1974.

6-8

f 116. Quadrex Corp. for Wisconsin Electric Power Co., Reactcq Containment Building Integrated Leak Rate Test for Point Beach Nuclear Plant Unit N o. 1, 1981.

117.' Bechtel Corp. for Wisconsin Electric Power Co., Initial Integrated Leak Rate Test for the Point Beach Nuclear Plant Unit 2,1972.

118. Nuclear Services Corp. for Wisconsin Electric Power Co., Point Beach Nuclear Piant, Unit No. 2, Reactor Containment Building Integrated Leak Rate Test, 1974.

119. Nuclear Services Corp. for Wisconsin Electric Power Co., Reactor Containment Building Integrated Leakage Rate Tect for Point Beach Nuclear Plant Unit No. Z,19/8.

120. Quadrex Corp. for Wisconsin Electric Power Co., Reactor Containment Building Integrated Leakage Rate T(st for Point Eeach Nuclear Plant Unit No. 2,1982, 121. Northern States Power Co., Prairie Island Nuclear Generating Plant, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Dnck'et

1. 50-262-166, 1973.

122. Northern States Power Co., Prairie Island Nuclear Generating Plant, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket

1. b0-262, 19//.

123. Northern-States Power Co... Prairie Island Nuclear Generating Plant, Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket

1. 50-282, 1980.

-124. North' ern States Power Co., Prairie Island Nuclear Generating Plant, Unit No. 2, Reactor Containment Building Integrated Leak Rate Test, Dockct #50-306,1977, 125. Northern States Power Co., Prairie Island Nuclear Generating Plant,

~

Unit No. 2, Reactor Containment Building Integrated Leak Rate Test, Docket #50-306,1981.

126. Commonwealth Edison Co., Report on Reactor Containment Building Integrated Leak Rate Test for Quad-Cities Nuclear Power Station, Unit One, Docket #50-Zb4-803,1976, 127. Connonwealth Edison Co., Reactor Containmant Building Integrated Leak Rate Test for Quad-Cities Nuclear Power Station, Unit 1, Docket #50-

-254, 1979.

128. Commonwealth Edison Co., Reactor Containment Building Integrated Leak Rate Test for Quad-Cities Nuclear Power Station, Unit 2, Docket #50-265, 1980.

129. Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Reactor Containment building Integrated Leak Rate Test, Final Report, Dccket #50-312,19/4.

6-9

I.

T, 130. Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Reactor Containment Building Integrated Leak Rate Test, First i

. Type A Test, Docket #50-312, 19/8.

131. Gilbert Associates and Carolina Power and Light Co., H. B. Robinson Steam Electric Plant, Unit No. 2, Reactor Containment Building Integrated Leak Rate Test, Docket #50-261, 1982.

1 132. Public Service Electric and Gas Co., Salem Nuclear Generating Station, Unit 1, Reactor Containment Integrated Leak rate Test, First Retest, Docket #50-272,1979.

133. Southern California Edison Co., San Onofre Nuclear Generating Station, Unit 2, Reactor Containment Building Integrated Leak Rate Test Report, Docket #50-361,1981.

'134. Southern California Edison C3., San Onofre Nuclear Generating Station, Unit 3, Primary Reactor Containment Integrated Leak Rate Test, Docket

1. 50-362, 1982.

135. Tennessee Valley Authority, Division of Power Production, Reactor Building Containment Integrated Leak Rate Test for Sequoyah Nuclear Plant, Unit 1, Docket #50-327,1979.

136. Tennessee Valley Authority, Division of Nuclear Power, Reactor Building Containment Integrated Leak Rate Test for Sequoyah Nuclear Plant, Unit 2, Docket #50-328,1981.

137. Florida Power and Light Co., St. Lucie Plant, Unit 1, Summary Technical Report on Reactor Containment Building Integrated Leak Rate Test, Docket #50-335-216,19/5.

138. Ebasco Services Inc. for Florida Power and Light Co., St. Lucie Nuclear Power Station, Unit No.1, Reactor Containment Building Integrated Leak Rate Test, Summary Technical Report, Docket #50-335, 1979.

139. South Carolina Electric and Gas Co., Virgil C. Summer Nuclear Station, Unit 1, Summary Technical Report of the Integrated Leak Rate Test, Docket #50-395,1982.

140. Virginia Electric and Power Co., Surry Nuclear Power Station, Unit 1, Reactor Containment Building Integrated Leak Rat? Test Types A, B, and C, Docket #50-280,1981.

141. Virginia Electric and Power Co., Surry Power Station, Unit 2, Reactor Containment Building Integrated Leak Rate Test, Docket #50-281-126, 1972.

142. Virginia Electric and Power Co., Surry Power Station, Unit 2, Summary Report of Containment Integrated Leak Rate Testing, Docket #50-281-567, 19/6.

6-10

f 143. Virginia Electric and Power' Co., Surry Nuclear Power Statfor., Unit No.

2, Reactor Containment Building Integrated Leak Rate Test Types A, B, and C, Docket #50-281,1980.

144. Virginia Electric and Power Co., Surry Nuclear Power Station, Unit No.

' 2, Reactor Containment Building Integrated Leak Rate Test Types A, B, and C, Docket #50-281,1981.

145. Gilbert Associates for Metropolitan Edison Co., Three Mile Island Nuclear Station Unit 1, Reactor Containment Building Preoperational Integrated Leak Rate Test, GAI Report #183/, Docket #50-289-264,19/4.

146. Gilbert Associates for Metropolitan Edison Co., Three Mile Island Nuclear Station Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-289, 1978.

147. Gilbert Associates for Metropolitan Edison Co., Three Mile Island Nuclear Station Unit 1, Reactor Containment Building Integrated Leak Rate Test, Docket #50-789,1981.

148. Bechtel Corp. Preoperational Test for Portland General Electric, Trojan

' Nuclear Power Plant Ur.it No.1, Reactor Containment Building Integrated

- Leakage Rate Test, Bechte) Job # 6478, Docket #50-344-208,1975.

149. Portland General Electric Co., Trojan Nuclear Power Plant, Final Report on the Reactor Containment Builoing Integrated Leak Rate Test, Docket

1. 50-344, 1979.

-150. Bechtel Corp. for Florida Power and Light Co.,, Turkey Point, Unit 3, Initial Integrated Leak Rate Test of the Reactor Containment Building, Docket #50-250,19/2.

151. Florida Power and Light Co., Turkey Point Plant, Unit 3, Summary Technical Report on Reactor Containment Building Integrated Leak Rate Test, Docket #50-250-453,19/5.

152. Florida Power and Light Co., Turkey Point Plant, Unit 3, Reactor Containment Building Integrated Leak Rate Test, Docket #50-250,1979.

153. Florida Power and Light Co., Turkey Point Plant, Unit 3, Reactor Containment Building Integrated Leak Rate Test, Docket #50-25U~~T982.

154. Bechtel Corp. fcr Florida Power and Light Co., Turkey Point Plant, Unit 4, Initial Integrated Leak Rate Test of the Reactor Containment Building, Docket #50-251,1972.

155. Florida Power and Light Co., Turkey Point Plant, Unit 4, Reactor Containment Building Integrated Leak Rate Test Summary Technical Report, Docket #50-251-55/,1976.

156.' Florida Power and Light Co., Turkey Point Plant, Unit 4, Reactor

~

-Containment Building Integrated Leak Rate Test Types A, B, and C, Docket #50-251,1981.

6-11 e

,,e r

-r w e m u

\\

157.. Vermont Yarlee Nuclear Power Corp., Vertront Yankee Nuclear Power Station, Integrated Primary Containment Leakage Rate Test, Docket #50-271, 19/2.

158. Vermont Yankee Nuclear Power Corp., Vermont Yankr.e Nuclear Power Station, Renort on Primary Containment Leakage (ate Testing, Types A, r

. B, and C, Docket #50-271-469,1975.

159. Vermont Yankee Nuclear Power Corp., Vertolt Yankee Nuclear Power Station, Primary Containment Leakage Rate Testing, Types A, B, and C, Docket #50-Z/1,1978.

160. Vertriont Yarite Nucleat Power Corp. Vermont Yankee Primary Containment Leakage Rate Testing, Types A, B, and C, Docket #50-271,19/9.

161. Vermont Yankee Nuclear Power Corp., Vermont Yankee Primary Containment Leakage Rate Testing, Types A, B, and C, Docket #50-271,1980.

162.. Yankee Atomic Electric Co., Yankee Nuclear Power Station, Reactor Containment Building Integrated Leak Rate Test, Docket #50-029-363, Report #10/4, 19/4.

163. Yankee Atomic Electric Co., Reactor Containment Building Integrated Leak Rate Test, Docket # 50-029-887, 1977.

164. Yankee Atomic Electric Co., Reactor Containment Building Integrated Leak Rate Test, Docket #50-029,1980.

165. Com:aanwealth Edison Co., Zion, Unit 1, Reactor Containment Building Integrated leak Rate Test, Docket #50-295,1981.

166, Commonwealth Edison Co., Zion, Unit 2, Reactor Containment Building Integrated Leak Rate Test, Docket #50-304,1980.

6-12

p Appendix A CRITERIA COMPLETION

SUMMARY

BY EVALUATED PLANT RESORTS PLANT -ILRT

---

CRITERIA NUMBER----------------

NAME

YEAR 1.METH 2.MODEL 3.+95% 4.LSF 5.95% 6. STAB 7. CONY RESULT ANO 1-73 0

X X

X X

INS ANO 1-78 X

X X

X X

INS AMO 1-81 X

X X

X X

X X

YES**

ANO 2-77 X

X X

X X

INS ANO 2-81 X

X X

X X

X X

YES**

BEAVER VALLEY 1-75 X X

X X

INS

-BEAVER VALLEY 1-78 X X

X X

X X

X YES**

INS BIG ROCK POINT-74 0 0

X X

X INS BIG ROCK P0lNT-78 X 0

X X

INS BROWN'S FERRY 1-73 X X

INS BROWN'S. FERRY 1-76 X X

p BROWN'S FEDRY 1-80 X X

X X

X INS

' BROWN'S FERRY 2-74 0 X

INS BROWN'S FERRY 3-79 X X

X X

X X

X YES**

BRUNSWICK 1-76 0

X X

X X

INS BRUNSWICK 1-81 X

X X

X X

INS BRUNSWICK 2-77 X/0 X

X X

X X

X YES**

INS X

CAL. CLIFFS 1-73 0

X X

CAL. CLIFFS 2-79 X

X X

X X

X X

YES**

CONN. YANKEE-76 X

0 0

X X

YES COOK 1-78 0

X X

X X

INS INS COOK 2 0 X

X X

)

INS C00PER-73 0

X X

X INS C00PER-76 X/0 Y

X X

X C00PER-80 X

X X

X X

X X

YES CRYSTAL RIVER 3-76 X X

X X

X INS CRYSTAL RIVER 3-80 X X

X X

X X

X YES**

A-1

)

PLANT -ILRT

---

CRITERIA NUMBER----------------

NAME YEAR 1.METH 2.MoDEL 3.+95% 4.LSF 5.95% 6. STAB 7. CONY RESULT DAVIS'BESSE-76 0

X INS DAVIS BESSE-80 X

X X

X X

X X

YES**

X INS j

DIABLO CANYON 1-75 X 0

X DIABLO CANYON 1-78 0 X

X X

X INS DIABLO CANYON 2-77 X X

X X

X INS DRESDEN 1-75 X

X X

X INS DRESDEN 1-77 0

X X

X X

INS DRESDEN 2-76 X

X X

X X

INS DRESDEN 3-82 0

X X

X X

INS FARLEY 1-81 X

X X

X X

X X

YES**

FARLEY 2-80 X

X X

X X

X X

YES**

FITZPATRICK-78 X

X X

X X

X X

YES**

FITZPATRICK-82 X

X

-X X

X X

X YES**

FORT CALHOUN-72 0

X X

X X

INS FORT CALH00N-76 X

X X

X X

INS FORT Call 10VN-80 X

X X

X X

0 X

YES

'GINNA-72 X

X X

X X

INS GINNA-76 X/0 X

X X

INS GRAND GULF 1-82 X

X X

X X

X X

YES**

HATCH 2-78 X/0 0

X X

X INS INDIAN FOINT 2-71 X X

INS INDIAN POINT 2-76 X/0 X

0 X

X X

X YES INDIAN POINT 2-79 X X

X X

INS INDIAN POINT 3-75 0 X

X X

INS INDIAN POINT 3-78 X/0 X

X X

X INS KEWAUNEE 74 0

X X

X X

INS KEWAUNEE-80 0

X X

X INS LACROSSE-75 0

0 X

X X

INS LACROSSE-78 X

0 X

X X

INS LACROSSE-79 X

0 X

.X X

INS LACROSSE-80 X

0 X

X X

INS LaSALLE 1-82 X

X X

X X

X X

YES**

A-2

[

PLANT -ILRT

---

CRITERIA NUMBER----------------

NAME YEAR 1.METH 2.MODEL 3.+95% 4.LSF 5.951 6. STAB 7.CONV RESULT INS X

X MAINE YANKEE-75 X.

McGUIRE 1-79 X

X X

X X

INS X

INS X

MILLSTONE 1-70 MILLSTONE 1-73 0

X X

X X

INS MILLSTONE 1-76 X

X X

X X

X X

YES**

MILLSTONE 1-81 X

X X

X X

X X

YES**

MILLSTONE 2-75 0

X X

X X

INS MILLSTONE 2-79 X

X X

X X

X X

YES**

MONTICELLO-80 X

X X

X X

INS INS X

NINE MI. PT.1-75 X

X X

NINE MI. PT.1-79 X

X X

X X

X X

YES**

NORTH ANNA 1-81 X

X X

X X

X X

YES**

NORTH ANNA 2-79 X

X X

X X

X X

YES**

OCONEE 1-70 0

X X

X X

INS INS OCONEE 1-76 0

X X

X X

OCONEE 1-80 X

X X

X X

X X

YES**

OCONEE 2-73 0

X X

X INS INS OCONEE 2-77 0

X X

X X

'0CONEE 2-80 X

X X

X X

X X

YES**

INS OCONEE 3-74 0

X X

X X

INS OCONEE 3-78 0

X X

X OCONEE 3-81 X

X X

X X

X X

YES**

INS X

OYSTER CREEK-78 X/0 X

X INS 0YSTER CREEK-80 X

X X

X X

INS PALISADES-74 0

X X

X X

PALISADES-78 X

X X

X X

X X

YES**

PALISADES-82 X

X X

X X

X X

YES**

INS X

PEACH BOTTOM 2-76 X 0

X PEACH BOTTOM 2-80 X 0

X X

X X

X YES INS PEACH BOTTOM 3-77 X/0 X

X X

X PEACH BOTTOM 3-81 X X

X X

X X

X YES A-3

)

PLANT -ILRT

---

CRITERIA NUMBER----------------

NAME YEAR 1.METH 2.MODEL 3.+95% 4.LSF 5.95% 6. STAB 7. CONY RESULT PILGRIM-80 X

X X

X X

X X

YES**

PILGRIM-82 X

X~

X X

X X

X YES**

POINT BEACH 1-70 0

X X

X X

INS POINT BEACH 1-77 X

X X

X X

INS POINT BEACH _1-81 X

X X

X X

X X

YES**

POINT BEACH 2 0 X

X X

INS POINT BEACH 2-78 X

X X

X X

INS POINT BEACH 2-82 X

X X

X X

X X

YES**

PRAIRIE IS. 1 X X

X X

INS PRAIRIE IS. 1-77 X

X X

X X

INS PRAIRIE IS. 1-80 X

X X

X X

INS PRAIRIE IS. 2-77 X

X X

X X

INS PRAIRIE IS 2-81 X

X X

X X

INS QUAD-CITIES 1-76 0

X X

X INS QUAD-CITIES 1-79 X

X X

X X

X X

YES**

QUAD-CITIES 2-80 X

X X

X X

X X

YES**

RANCHO SECO-74 0

X X

X X

INS

-RANCHO SECO-77 X

X X

X INS ROBINSON 2-82 X

X X

X X

X X

YES**

SALEM 1-79 X

X INS SALEM 1-80 INS SAN ONOFRE 2-80 X

X X

X X

X X

YES**

SAN ON0FRE 3-82 X

X X

X X

X X

YES**

SEQUOYAH 1-79 X

X X

X X

X X

YES**

SEQUOYAH 2-81 X

X y

X X

X X

YES**

ST LUCIE 1-75 0

X X

INS X

ST LUCIE 1-79 0

X X

X INS SUMMER-82 X

X X

INS SURRY I-81 X

X X

X X

X X

YES**

SURRY 2-72 X

X X

X X

INS S'URRY 2-76 X

X X

X X

INS A-4 i_

T

' PLANT -ILRT

---

CRITERIA NUMBER----------------

NAME YEAR 1.METH 2.MODEL 3.+95% 4.LSF 5.95% 6. STAB 7. CONY RESULT SURRY 2-80 X

X X

X X

X X

YES**

SURRY 2-81 X

X X

X X

X X

YES**

INS THREE MI. IS. 1-74 X X

X X

X THREE MI. IS. 1-78 X X

X X

X X

X YES**

THREE MI. IS. 1-81 X X

X X

X X

X YES**

TROJAN 1 (HALF)-75 X/0 X

X X

X X

X YES**

TROJAN 1 (FULL)-75 X/0 X

X X

X YES TROJAN 1-79 0

X X

X X-INS TURKEY POINT 3-72 0 X

X X

X INS TURKEY POINT 3-75 0 X

X X

X INS TURKEY POINT 3-79 X/0 X

X X

X X

X YES**

TURKEY POINT 3-82 X/0 X

X

.X X

X X

YES**

TURKEY POINT 4-73 0 X

X X

X INS INS TURKEY POINT 4-76 0 X

X X

X TURKEY POINT 4-80 X X

X X

X X

X YES**

VERMONT YANKEE-74 X X

X X

INS VERMONT YANKEE-78 X X

X X

X INS VERMONT YANKEE-79 X X

X X

X INS INS VERMONT YANKEE-80 X X

X X

X YANKEE-74 0

X X

X INS YANKEE-77 0

X X

X X

INS YANKEE-80 X

X X

X X

X X

YES**

X INS Z:CA 1-81 X

X X

INS ZION'2-80 X

X X

X X

LEGEND:

"X" = MEETS CRITERION "0" = DOES NOT MEET CRITERION

" " = NOT SHOWN OR PROVIDED IN REPORT

" INS" = INSUFFICIENT AVAILABLE DATA IN REPORT TO INDICATE CRITERIA RESULTS "YES" = DATA AVAILABLE IN REPORT

"**" = CRITERIA ALLOWS TERMINATION A-5

)

DATA SU M RY TOTAL NUMBER OF.. REPORTS INCLUDED:

144

-TOTAL NUMBER OF REPORTS WITH SUFFICIENT DATA: 53 TOTAL N'JtSER OF REPORTS WITH INSUFF. DATA:

91 REPURT CONTENT

SUMMARY

RAW NUMBERS PERCENTAGES 1.

MASS POINT TECHNIQUE:

100 70 ANOTHER TECHNIQUE:

38 26 NOT SHOWN:

6 4

2. MODEL:

130 90

3. POSITIVE 95% CONFIDENCE LEVEL:

132 92

4. LEAST SQUARES FIT LEAKAGE RATE:

111 77

5. REPORTED 95% CONFIDENCE LEVEL:

134 93

6. REPORTED LSF LEAKAGE RATE STABLIZATION:

48 33

'7. CONVERGENCE OF LSF LEAKAGE RATE AND 43 33 95% CONFIDENCE LEVEL:

)

A-6

-Appendix B SUP94ARY OF ACQUISITION AND PROCESS _ METHODS ACQUISITION MANUAL COLLECTION FROM ILRT SENSOR CONSOLE:

61 DATA LOGGER COLLECTION:

36 DATA LOGGER COLLECTION INTO PLANT COMPUTER:

34 NOT SHOWN IN REVIEWED REPORTS:

13

~

TOTAL:

144

- PROCESSING.

DATA LOGGER INTO PLANT COMPUTER:

34 REMOTE TERMINAL INTO PLANT COMPUTER:

22

. STAND ALONE COMPUTER:

35

' USE A SERVICE BUREAU COMPUTER:

21

NOT SHOWN IN REVIEWED REPORTS:

32 TOTAL:

144 i

o B-1 i

.w-

-w c-

,,, - - -, ~

,..,,,,,,,,r,n,,,

,---ew.,

,,,,-a,-c...,,,,

[

[-

Appendix C PLANT SIA4 MARY OF SENSORS

~

- _ PLANT NAME RTD'S DEW CELLS PRESSURE ANO 1 20 6

2 ANO 2 18 6

2 BEAVER VALLEY 1 18 5

4 BIG ROCK POINT 4

4 1

BROWN'S FERRY 1 24 4

-1 BROWN'S FERRY 2' 19 4

'1 BROWN'S FERRY 3 18 4

1 g-BRUNSWICK 1 24 10 1

" ~ ' ~

BRUNSWICK 2 24 10 2

CALVERT CLIFFS 2 20 8

2 CALVERT CLIFFS 1 20 8

2 CONNECTICUT YANKEE 10 2

2 COOK 1 46 6

2

. COOK 2 46 6

2 COOPER-20 4

1 CRYSTAL RIVER 3' 24 10 2

DAVIS-BESSE 20 6

2 UIABLO CANYON 1 24 12 1

- UIABLO CANYON 2 24 12 1

ORESDEN 1 29 10 2

- DRESDEN 2 27 6

2 URESDEN 3 30 10 2

DUANE ARNOLD 12 6

2 GINNA 24 6

2 FARLEY UNIT 1 18 6

2

. FARLEY UNIT 2.

17 5

1 FITZPATRICK 18 1

4 s

C-1

A

^

l i

1 PLANT NAME RTD'S DEW CELLS PRESSURE FORT CALHOUN 20 9

1 GRAND GULF 1 22 6

1 HATCH 1 12 8

2 HATCH 2 15 3

2 INDIAN POINT 2 24 6

1 INDIAN POINT.3 24-6 2

KEWAUNEE 18 6

2 LACROSSE 12 6

1 LaSALLE 1-30 10 2

MAINE YANKEE 13 3

1 McGUIRE 1 52 3

3 MILLSTONE 1 20 6

1 MILLSTONE 2 18 4

2 MONTICELLO 20 6

1 NINE MILE POINT 1 17 4

1

~ NORTH ANNA i 18 5

1 NCRTH ANNA 2 18 5

1 OCONEE 1 28 2

2

-OCONEE 2 28 2

1 OCONEE 3 28 2

1 OYSTER CREEK 30 10 1

PALISADES 19 6

2 PEACH BOTTOM 2 12 6

1 PEACH BOTTOM 3 12 6

2 PILGRIM 18 4

2 POINT BEACH 1 21 6

2 POINT BEACH 2 21 6

2 PRAIRIE ISLAND 1 24 6

2 PRAIRIE ISLAND 2 22 6

2 QUAD-CITIES 1 20 6

2 QUAD-CITIES 2 30 8

2 C-2

/.

4

' PLANT'hAME RTD'S DEW LELLS PRESSURE te ~

RANCHO SECO 19 7

2 ROBINSON 24 5

2 SALEM i 2?

5 2

SAN ON0FRE 1~

18 -.

4 1

SAN ONOFRE 2 16 4

1

. SAN ONOFRE 3 18 -

4 1

SEQUOYAH 1 46 10 1

.SEQUOYAH 2 48 13 1

ST. LUCIE 1 40 10 1

-SUMMER 24-5 1

SURRY 1 18 2

1 SURRY.2 18 4

1 HREE MILE ISLAND 1 ' 24 10 2

TROJAN 18 6

3

-TURKEY POINT 3--

20-8 2

TURKEY POINT 4 20 7

2 VERMONT YANKEE

.18 6

1 YANKEE 20 4

1 ZION 26 6-2 TOTALS-1,705 474 126 AVERAGE SENSORS BASED ON 77 PLANTS:

ROUNDED:

'22 6

2

[

[.

i L

i C-3 l-.

E

gr7 Appendix D PLANT

SUMMARY

OF TESTING TIME DURATIGNS PLANT NAME YEAR DURATION IN HOURS ANO 1 1973 8.5 ANO 1 1978 24 AN0 '1 1981 10 ANO 2 1977 8

ANO 2 1981 8

BEAVER VALLEY 1 1975 24 BEAVER VALLEY 1 1978 24 BIG ROCK POINT 1974 BIG ROCK POINT 1978 - 24 BROWN'S FERRY 1 1973 24 FULL /24 HALF

-BROWN'S FERRY 1 1976 26 BROWN'S FERRY 1 1980 27.05 BROWN'S FERRY 2 1974 25 FULL /24 HALF BROWN'S FERRY 3 1979 24 BRUNSWICK 1 1976 24 FULL /24 HALF BRUNSWICK 1 1981 24 BRUNSWICK 2 1977 24 CALVERT CLIFFS 1 1973 9 FULL /8.75 HALF CALVERT CLIFFS 2 1979 8.15 CONNECTICUT YANKEE 1976 24 CONNECTICUT YANKEE 1980 24

' COOK 1 1978 14 COOK 2 1981 24 COOPER 1973 24 FULL /24 HALF COOPER 1976 NO REPORT COOPER 1980 14 CRYSTAL RIVER 3 1976 24 CRYSTAL RIVER 3 1980 24

.DAEC 1978 10.25 DAEC 1980 8

D-1 i

N PLANT NAME YEAR DURATION IN HOURS DAVIS-BESSE

-1976 24 DAVIS-BESSE 1980 8

DIABLO CANYON 1 1975 35 FULL /24 HALF DIABLO CANYON 1 1978 20 DIABLO CANYON 2 1977 24 DRESDEN 1 1975 24 DRESDEN 1 1977 24 DRESDEN 2 1976 25.5 DRESDEN 3 1982 24 GINNA 1972 NO REPORT GINNA 1976 NO REPORT FARLEY 1 1981 25.25 FARLEY 2 1980 12.5 FITZPATRICK 1978 26 FITZPATRICK 1982 24 FORT CALHOUN 1976 13 FORT CALHOUN 1980 24 GRAND GULF 1 1982 8

HATC;I 2 1978 8

INDIAN POINT 2

-1971 24 INDIAN POINT 2 1976 24 INDIAN POINT 2 1979 24 INDIAN POINT 3 1975 24 FULL /24 HALF INDIAN POINT 3 1978 24 KEWAUNEF 1974 KEU?UNEE 1980 24 LACROSSE 1975 24 LACROSSE 1978 24 LACROSSE 1979 24

' LACROSSE 1980 24 LaSALLE 1 1982 26.18 MAINE YANKEE 1975 24 McGUIRE 1 1979 24 D-2

PLANT NAME YEAR DURATION IN HOURS MILLSTONE 1 1970 MILLSTONE 1 1973 MILLSTONE 1 1976 8

MILLSTONE 1 1981 41 MILLSTONE 2 1975 MILLSTONE 2 1979 57 MONTICELLO 1 1980 8.20 NINE MILE POINT 1979

'71.5 NORTH ANNA 1 1981-26 NORTH ANNA 2 1979 24 OCONEE 1 1971 10 OCONEE 1 1976 10 OCONEE 1 1980 65 OCONEE 2 1973 11 FULL /10 HALF OCONEE 2 1977 10 OCONEE 2 1980 14.5 OCONEE 3 1974 10 OCONEE 3 1978 24 OCONEE 3 1981 20 OYSTER CREEK 1978 24 0YSTER CREEK 1980 24 PALISADES 1974 24 PALISADES 1978 PALISADES 1982 PEACH BOTTOM 2 1976 8

PEACH BOTTOM 2 1980 8

PEACH BOTTOM 3 1977 PEACH BOTTOM 3 1981 8

PILGRIM 1980 34

. PILGRIM 1982 24 POINT BEACH 1 1970 24 FULL /24 HALF POINT BEACH 1 1977 12 POINT BEACH 1 1981 8

D-3

\\

s PLANT NAME YEAR DURATION IN HOURS POINT BEACH 2 1971 24 FULL /24 HALF POINT BEACH 2 1978 12 POINT BEACH 2 1982 12 PRAIRIE ISLAND 1 1973 24 FULL /24 HALF PRAIRIE ISLAND 1 1977 PRAIRIE ISLAND 1 1980 10.2 PRAIFIE ISLAND 2 1977 24 PRAIRIE ISLAND 2 1981 NO REPORT QUAD-CITIES 1 1976 24 QUAD-CITIES 1 1979 24 QUAD-CITIES 2 1980 24 RANCHO SECO 1974 24 FULL /24 HALF RANCHO SECO 1977 24 ROBINSON 2 1982 24 SALEM i 1979 24 SALEM i 1980 SAN ON0FRE 2-1980 24 SAN ON0FRE 3 1982 8

SEQUOYAH 1 1979 NO REPORT SEQUOYAH 2 1981 24 ST. LUCIE 1 1975 24 FULL /24 HALF ST. LUCIE 1 1979 24 SURRY 1 1981 33.51 SURRY 2 1972 24 SURRY 2 1976 24 SURRY 2 1980 12 SURRY 2 1981 63.99 THREE MILE ISLAND 1 1974 24 FULL /24 HALF i

THREE MILE ISLAND 1 1978 44.5 THREE MILE ISLAND 1 1981 24 TROJAN 1 1975 9 FULL /8.75 HALF TROJAN 1 1979 24.5 4

D-4

[

PLANT NAME YEAR DL' RATION IN HOURS TURKEY POINT 3 1972 NO REPORT TURKEY POINT 3 1975 24 TURKEY POINT 3 1979 8

TURKEY POINT 3 1982 24 TURKEY POINT 4 1973 NO REPORT TURKEY POINT 4 1976 24 TURKEY POINT 4 1980 41 VERMONT YANKEE 1974 24 VERMONT YANKEE 1978 37 VERMONT YANKEE 1979 24 VERMONT YANKEE 1980 YANKEE R0WE 1974 24 FULL /26 HALF YANKEE R0WE 1977 NO REPORT YANKEE R0WE 1980 31 ZION 1 1981 215 ZION 2 1980 26 DATA

SUMMARY

< 8 HRS.

13 8.3%

> 8 AND < OR = 12 HRS.

18 11.5%

> 12 AND < OR = 16 HRS.

5 3.2%

> 16 AND < OR = 20 HRS.

2 1.2%

> 20 AND < 24 HRd.

0 %

= 24 HRS.

77 49.0%

> 24 HRS.

23 14.7%

DURATION NOT SHOWN/NO REP:

19 12.1%

TOTAL NUMBER OF TESTS:

-157 100.0%

D-5

I l-t

}..

Appendix E TEMPERATURE WEIGHTING FACTOR SENSITIVITY r

f.

1 t

b Table E-1 TEMPERATURE VARIANCE COMPARISON DATA RTO TEMP VAR BASE CASE 1 CASE 2 CASE 3 CASE 4 CASE 5-255 CASE 6-505 CASE 7 PLANT - TEST POINTS

1. Ist&LAST/ MAX.

1st4LAST/ MAX.

1st&L AST/ MAX.

1st&LAST/ MAX.

Ist&LAST/ MAX.

1staLAST/ MAX.

Ist4LAST/ MAX.

Ist&LAST/ MAX.

1.

Am01 39 20.243/5AME 1.531/SAME

.1699/SAME

.2522/5AME

.2345/5AME

.2542/5AME

.2653/SAME

.3004/5AME 2.

Amol-78 97 20 1.8369/1.9583 2.0239/2.1439 1.811/SAME 1.8505/1.9591 1.8223/l.9523 1.8251/1. % 21 1.8133/1.9655 1.4782/1.9698 -

3.

Aw02-81 41 18. 1026/.2534

.13/.38

.1699/.2699

.1103/.2539

.0885/.2504

. 1214/.2586

.1401/.2639

.1027/.2563 4.

Au02-FULL-76 36 18

.3946/5AME

..48/.5

.11/.12

.4113/SAME

.379/5AME

.41/SAME

.07/5AME

.39/5AME 5.

AN02-HALF-76 33 18

.118/.2237

. 19/2.18

.06/.15

.1233/.1374

. 1076/.2217 '

.1242/.2003

..1303/.1767

.1178/.2322 6.

00E4 TANK-80 95 10.007/5AME

.3778/5AME

.-7409/SAME

.0891/5AME

.0891/5AME

.1268/SAME

.1659/5AME

.1527/SAME 7.

DAEC-78 42 12

.7928/.9486

-1.9501/-2.25.. 63/. 83ul

.7928/.8976

. 7928/.8976

.71/.9206

.6272/.895

.7816/.8926*

8.

DAEC-83 33 14

.6717/.765 3.1899/SAME

.05/. 1411

.6446/.7154

.6446/.7154

.5919/.7218

.5279/.6789

.8722/.9142 9.

DAV BE55E-80 33 20

.98/5AME 1.854/1.9751

.7271/.818

.9799i5AME 1.ou6/5AME

.9272/5AME

.8744/5AME

.995/5AME ret

10. FARLET 2-FULL-80 51 17

-08u5/.252

.04/1.1801

.15/.21

.0834/.2355

.u?!/.26u9

.0942/.2665

.1u80/.2809

.0847/.2565 b

al. FARLET 2-MALF-80 9J 17 4032/.4u89

.62/SAME

.0799/.1801

.3916/.4172

.4179/SAME

.3948/.4u44

.387/.4

.4057/.4047

12. FARLET l-FULL-77 44 le 1.1076/5AME 1.6501/1.84

.87/.9u99 1.1176/bAME 1.u792/hAML 1.1149/5AME 1.1227/hAME 1.0967/SAME

13. HATCH-78 37 11

.6394/5AME 1.69/SAML

.11/SAML

.806/SAME

.5Su/5AML

.7622/5AME

.866/SAME 1.0164/SAME

14. MILL 2-FULL-75 18

.6239/.6781

.485/3.U05

.614/.721 53U3/.7Uu1

.6269/.6984

.6t5/.7023

.626/.7042

.6338/.6959

15. MILL 1-HALF-75 39 18

.0911/.2246 1.098/1.231

.u24/.121

.0964/.19u3

.1103/.2049

. 0 5 3/.2459

.0615/.2665

.1179/.2101 16 PEACH 2-80 33 12

.6154/.635 4.021/4.031

.3721/.422

.8626/.8671

.3944/.4636

.817!.f.825 1.0189/1.0228 '.8962/.8996

17. PEACH 3-77 33 12

.1457/1.7784

.41/6.06

.14/.28

. 2187/1.9485

.07B1/1.4946

.1686/1. % 67

.1914/2.155

.!!84/2.442

16. PT BEACH 1-81 47 21

.7892/.9149

- 9/-1.1 0/.3

. 87U4/-1.0832

.6668/.8815

.8034/-1.u394

.8176/-1.095

.6619/.7762

19. PT BEACH 2-82 47 21

.8372/5AML 1.4/SAME

.5/SAME

.8256/.8447

.8261/.8319

.8169/.8185

.7965/.8037

.8714/.a048

20. TROJAN-FULL-75 36 18

.1580/.1725

.0501/.0601

.0501/.0701

.1685/.1826

.1787/.1901

.1601/.1663

.1620/.1701

.1460/.1460

21. TR(MAN-HALF-75 37 18. 4914/.4914

. 8/.98

.38/.38

.4773/SAME

.462/.4715

.5104/SAME

.52%/5AME

. 4862/.4862

22. TRCUAM-79 99 18 2.2388/SAME 2.6/SAML 1.87/2.3 2.5632/SAME 1.3204/SAME 2.595/5AME 2.6184/5AME 2.53/5AME

23. TRY PT 3-FULL-72 49 20 2.5759/5AME

-1.77/SAME 1.93/5AK 2.725/5AE 2.5994/5AME 2.4171/5AME 2.2586/5AME 2.535/5AME

24. TKT PT 4-FULL-72 34 20 1.1451/1.1451 1.25/SAME

.94/SAME 1.423/5AME 1.1198/SAME 1.1697/SAME 1.1943/SAME 1.122/SAME

25. TET PR 4-HALF-72 25 20

.5478/.5478 1.55/1.01

.40/.42

.5514/5AME

.555/5AME

.5408/$AME

.5338/S ut

.5615/SAME

Table E-2 l

l MARGIN COMPA iSON l

l OA7A RTD BASE j

PLANT - TEST POINTS

1. L5F MARGIN CASE 1 CASE 2 C ASE 3 CASE 4 CA5t 5-251 CASE 6-50%

CASE 7 l

l 1.

AHO 1-81 39 20

.1028

.0422

.13%

.0994

.1041

.09%

.0952

.0668 2.

Amo 1-78 97 20

.1491

.1133

.1695

.1483

.1504

.1493

.1454

.1463 3.

ANO 2-81 41 18

.0482

.0556

.0776

.0$u8

.0419

.0545

.0621

.0471 4

ANO 2-FULL-76 36 18

.0547

.1202

.0982

.0642

.0452

.0641

.0744

.0526 5.

Ano 2-HALF-76 33 18

.0649

.2261

.0129

.0663

.0582

.0685

.U717

.0646 6.

00ept TANA-80 95 10

.235t

.1574

.2847

.2344

.2344

.2312

.2270

.2188 7.

uAEC-78 42 12 1.1612 1.6064 1.0218 1.1552 1.1552 1.1441 1.1273 1.1362 8.

0ALC-83 33 14 1.1744

.1938 1.!574 1.1873 1.1873 1.2115 1.2479 1.07J2 9.

DAV stSSE-80

$3 20

.3057

.2w4 438r

.3114

.299

.3406

.37U9

.Ju25

10. FAALEY 2-FULL-80 51 17

.1441

.1444

.1456

.1444

.1440

.1449

.1454

.1445 r9k II. F ARLEY 2-MALF-80 93 17

.0741

.u6e5

.0973

.u746

.072J

.0743

.0752

.0741

12. FARLET 1. FULL-77 44 18

.0809

.1819

.2705

.0769

.0954

.0792

.0769

.0878

13. HA'CH-78 17 11

.7748

.2759 1.147

.6953

.8120

.7199

.6660

.5%6

14. MILL 2-FULL-75 18

.1107

.1469

.1163

.1081

.1068

.1120

.1133

.1085

15. MIL'. 2-MALF-75 39 18

.2896

.3829

.2431

.2379

.2889

.2916

.2944

.2899

16. PEACM 2-80 33 12

.401

-1.5565

.5591

.2616

.5273

.2849

.1687

.2388

17. PE ACM 1-77 33 12

.0663

-1.1613

.4305

.0712

.0629

.0358

.0059

.097

18. PT BEACM 1-81 47 21

.1229

.3115

.0111

.1366

.0996

.1243

.1258

.1028

19. PT SCACM 2-82 47 21

.1535

.0433

.3077

.1521

.1533

.1627

.1716

.13 bu

20. TROJ-FULL-75 36 18

.0274

.0136

.0284

.0278

.0281

.0267

.0256

.0228

21. TROJ-HALF-75 37 18

.047

.4652

.104

.0492

.0527

.0428

.04

.048

22. TRM-79 99 to

.0162

. 0074

.1173

.0168

.0168

.0135

.0106

.0227

23. TRY PT 3-FULL-72 49 20

.2687

-1.4413

.0016

.1040

.2771

. 2329 J237

.3049

24. TKY PT 4-FULL-72 34 20

.0745

.1322

.0315

.0739

.0643

.0864

.0984

.0622

25. TRY PT 4-HALF-72 25 20

.1889

.2067

.1347

.1884

.1918

.1824

.1764

.1884

Table E-2 l

PERCENTAGE COMPARISON DATA RTO PLAuf - TEST POINT 5 CASE 1 CASE 2 CASE 3 CASE 4 CAM 5-25%

CASE 6-50%

CASE 7 1.

Ano 1-81 39 20 41 135 96 101 9b 93 64 2.

ANU 1-78 97 20 76 114 99 101 100 IUU 95 3.

AND 2-81 41 16

-115 161 105 87 114 129 98 4.

Ano 2-FULL-76 36 18 220

-180 117 83 117 136 5.

Amo 2-HALF-76 33 i8 34 8 20 102 90 106 110 10's 6.

Coum YAu-8U 95 10 67 121 100 100 95 91 93 7.

DAEC-78 42 12 138 88 99 99 99 97 9H 8.

DAEC-83 33 14

-165 133 101 101 103 106 91 9.

08E55E-80 33 20

-82 144 102 98 112 122 99

{

10. FARLEY 2-FULL-80 51 17 100 101 100 100 100 tot 100 W

11. FARLEY 2-HALF-80 93 17 92 131 101 98 100 101 100

12. FARLEY 1-FULL-77 44 18

-225 334 95 118 98 95 109

13. MATCM-78 37 11 35 148 90 105 93 86 77

14. MILL 2-FULL-75 18 133 105 98 101 102 98

15. MILL 2-HALF-75 39 18 132 84 99 100 101 102 100

16. PEACM 2-80 33 12

-388 139 65 131 71 42 60

17. PE ACM 3-77 33 12

-1751 649 107 95 54 9

186

18. PT BEACM 1-81 47 21 253

-9 Ili 81 101 102 84

19. PT BEACH 2-82 47 21

-296 200 99 100 106 112 88 20 Tral-FULL-75 36 18 50

-104 101 103 97 93 83

21. Tral-HALF-15 37 18

-990 221 105 112 91 85 102

22. 'Ra!-79 99 18

-457 724 104 104 83 65 140

23. TKY PT 3-FULL-72 49 20 536 1

113 103 87 83 88

24. TKY PT 4-FutL-72 34 20 177

-42 99 86 116 132 83

25. TKY PT 4-HALF-72 25 20 109 71 100 102 97 93 100 1

1 1

i l

EPRI NP 3400

[ j % !2 Below are five index cards that allow for filing according to the four f

e 4. O 2 cross-references in addition to the title of the report. A brief abstract Qo { Z describing the major subject area covered in the report is included on o

N cach card.

s a g a

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o-m EPRI NP 3400 Criteria for Determining the Duration of Integrated 3 i 2 i;;

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@O RP1393-5 Leakage Rate Tests of Reactor Containments ij o

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Final Report

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3 December 1983 1

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Technical acceptance critena have been developed for integrated leakage rate 3"

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tests of reactor containments Using these criteria, utehty engineers can shorten

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the duration of the tests by 50%, thereby reducmg plant downtime. 104 pp.

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Do EPRI Project Manager: T M. Law h $'

M Cross References.

3

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1. EPHI NP 3400 2 RP1393-5 3 System Performance Pmgram E. e

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3 EPRI NP-3400 Criteria for Determinina the Duration of Integrated Leakage Rate Tests of heactor Containments i

RP1393-5 Final Report Contractor: Quadrex Corporation c 712r m g.' g 3 Technical acceptance criteria have been developed for integrated leakage rate

• 5N5 3ygg tests of reactor containments. Using these criteria. utihty engineers can sturten g e g, i the duration of the tests by 50%, thereby reducing plant downtime. 104 pp.

,E N

EPRI Project Manager: T. M. Law 2

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Cross-Referrnces:

1 E PRl hP 3400 2 RP1393 5 3 System Performance Program

• "Q s g lis' T"" Q 4 Containment Systems fj! q$E! p

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tests of reactor containments. Using these criteria, utility engineers can shorten Ee$

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EPRI Project Manager: T. M Law

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1. EPAI NP 34n0 2 PP1393 5 3 System Nrformance Program 3

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