Text

Forwards Justification for Continued Operation for Insulation Resistance for Amphenol Electrical Penetration Assemblies,Per Generic Ltr 88-07 Commitment
	ML20247C974
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/20/1989
From:	Cockfield D; PORTLAND GENERAL ELECTRIC CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	GL-88-07, GL-88-7, IEIN-89-029, IEIN-89-29, NUDOCS 8907240494
	Download: ML20247C974 (40)
	v • d • e

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Pbrtland General ElectricCompany r - E:

David W. Cockfield Vice President, Nuclear July 20, 1989 Trojan Nuclear Plsnt Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555

Dear Sir:

Bunker Ramo Elcetrical Penetration Assemblies Environmental Qualification Justification for Continued Operation (JCO)

Your letter of January 31, 1989, " Environmental Qualification of Bunker Ramo Electrical Penetration Assemblics at Trojan, Request for Information",

requested Portland General Electric Company (PGE) to establish environ-mental qualification of Amphenol electrical penetration ar wmblies (EPAs) installed at the Trojan Plant through testing or to replace the affected EPAs with qualified assemblies. Also requested was that a Justification for Continued Operation (JCO) be implemented in accordance with the guidance in Ceneric Letter 88-07, " Modified Enforcement Policy Relating to 10 CFR 50.49, 'Envir,onmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants'", to support interim Plant operation until testing or replacement is completed, As committed in our response of May 15, 1989, "Amphenol (Bunker Ramo)

Electrical Penetration Assemblies", a JC0 has been implemented.

It is attached for your information at the request of the Trojan Plant Project Manager. We would be pleased to answer any questions you have concerning this submittal.

Sincerely,

.e Attachment T

c:

Mr. John B. Martin okO Regional Administrator, Region V U.S. Nuclear Regulatory Commission N

00 0 Mr. David Stewart-Smith eg State of Oregon

$8 Department of Energy f

oQ Q v rY Mr. R. C. Barr ji ;

h NRC Resident Inspector Q

Trojan Nuclear Plant 121 S W S;$ron Stwt. Po@nd Ongan 97204

Portland General Electric '

% clear Division

.j

-MEMORANDUM TO:

Distribution ANR-0775-89M JCO 89-13 FROM:

A. N. Roller l

DATE:

June 26, 1989 l-l.

SUBJECT:

' Justification for Continued Operation (JCO) - Insulation l

l Resistance (IR) - Amphenol Electrical Penetration Assemblies (EPAs)

I.

EXISTING CONDITIONS l

j A.

Equipment Affected

~!

The Amphenol EPAs having safety-related and environmentally qualified post-accident-monitoring instrumentation circuits have the potential of degraded IR characteristics during an i

accident condition which may adversely affect the performance of such circuits. The affected EPAs are:

AZ-07. BZ-07. CZ-07, DZ-07, CZ-06 DZ-06,.and NZ-30.. Tatle 3 det;ignates the affected instruments.

B.

The Problem The problem pertaining to the IR issue is one of inadequate documentation of IR readings during the LOCA simulation tests.

Trojan Nuclear Plant's environmental qualification report indi-cates only one reading taken at the peak LOCA temperature pro-file.

In an effort to. prove the adequacy of the test results,

'1 PGE joined with other affected utilities.to address the IR j

issue as a group. This group known as the Nuclear Utility Group for Equipment Qualification (NUGEQ) submitted their.

j I

results to the NRC in a meeting on August 4, 1988. The NRC l

reviewed the data presented and found them unacceptable, based

.i on the premise that insufficient IR readings were taken.during-l LOCA simulation. This is documented in a letter (Reference 9).

I 1

C.

Other Documents Addressinst the problem Two documents have been issued which address the insulation resistance problem.

NRC Information Notice 89-29. " Deficiencies in Primary Contain-ment Low-Voltage Electrical Penetration Assemblies", addressed the lack of insulation resistance data obtained through the performance of environmental qualification testing of Bunker.

Ramo (Amphenc1) instrumentation penetrations.

JC0 89-13. Rev, O Page 1 of 11 Trojan Excellence-Our Way of Doing Business

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i Portland General Electnc Nuclear Division Distribution June 26, 1989 Page 2 1

NRC Ceneric Letter 88-07, " Environmental Qualification of 1

E.lectrical Equipment Tmportant to Safety for Nuclear Power

- Plants", included guidance for Trojan to follow to establish equipment qualification of the instrument circuit EPAs under Design Basis Event (DBE) conditions.

?

II.

JUSTIFICATION.FOR OPERATTON WITH EXISTING CONDITIONS

(

It is well established that the worst IR value occurs at the highest temperature. The value of one megohm was taker, at the peak LOCA temperature (303*F).

The error introduced with this value results l

in instrument error readings of less than 0.3 percent. This error is.much less than the accuracy of the instruments and will not l

result in adverse performance of their intended safety functions.

I 7n addition, the tests were performed with a configuration which rendered much worse results than the ones used on Trojan with j

l qualified Raychem splices, and EPAs located above the flood level.

l These issues are discussed in Attachment B.

III.

JCO DURATION AND SPECIAT. CONDITIONS The duration of this JC0 begins with its issuance and ends when the plant is shut down for the 1991 Refueling Outage. There are no i

special conditions associated with the JCO.

i ANR/KLJ/6202o c:

See Attachment A for Distribution CONCURRENCE AND APPROVAL:

hb 8T hhrN W

w Ction ineer General Manager, Technical Functions f

l"PlantRevIewBoard'Ch' airman Adh.

S?@

b General 'Me'r, Troj an Ach&L 6futeg Manager car Plant Engineering jfh l bAL b Man 6de D aclear Safety & Regulation JCO 89-13, Rev. O Page 2 of 11 Trojan Excellence-Our Way of Doing Business

-__.__..m.

ATTACHMENT A l

JCO COPY DISTRIBUTION JCOs:

Shift Supervisor PRB Members TNOB Members R.' E. Susee i

TNP: POW ST OP 2:JC0 TNOB:

i D. W. Cockfield, Vice President I

T. D. Walt, Chairman I

R. M. Nelson, Vice Chairman J. L. Frewing 4

A. N. Roller

)

C. P. Yundt l

D. L. Nordstrom R. W. Griebe j

c:

L. W. Erickson i

P. A. McMillan d

TNOB: ADMIN: Correspondence i

PRB:

i D. W. Swan, Chairman I

l P. A. Morton 1

R. L. Russell l

D. L. Bennett T. O. Meek R. C. Rupe c: PRB Engineer l

l

{-

62020 JCO 89-13, Rev. O Pago 3 of 11 L________._____

v.

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ATTACHMENT U to JC0 89-13 JUSTIFICATION FOR OPERATION WITH EXISTING CONDITIONS

'I.

PURPOSE This JC0 establishes the basis for continuing the operation of-Trojan Nuclear: Plant' with the present instrumentation penetration assemblies pending their replacement with fully qualified penetra-tions during the next two refueling outages or demonstration of their qualification through additional conclusive qualification testing.

II.

BACKCROUND PGE's purchase of the Amphenol penetrations was performed under Specification 6478-E20 and imposed (among others) the following requirements on Amphenol:

1.

All work was to be performed under a Quality Assurance program meeting 10 Code of Federal-Regulations (CFR)'Part 50, Appendix B " Quality Assurance Criteria for Nuclear Power Planta and Fuel Reprocessing Planta".

2.

Documentation and/or test reports were to be submitted to demonstrate compliance with the following Equipment' Qualification (EQ) standards.

1 a.

Institute of Electrical and Electronics Engineers (IEEE)-

I Standard 317-1971, "IEEE Standard for' Electrical-I Penetration Assemblies in Containment. Structures for l

Nuc1 car Fueled Power Generating Stations".

b.

Inst.ute of Electrical and Elcetronics Engineers.(IEEE) i Standard 323-1971, "tEEE Trial-Use Standard: General Guide for Qualifying Class IE Electrical Equipment for Nuclear Fueled Power Generating Stations".

Al' ins Amphenol penetrations were built and installed;in 1972 and 1973. All safety-related in-Containment terminations to field cables are currently installed using environmentally qualified

' heat-shrinkable tubing.

In accordance with NRC Bulletin 79-OlB, and later 10 CFR 50.49,

" Environmental Qualifiention of Electric Equipment Important to Safety for Nuclear Power Plants", PGE qualified the Amphenol i

penetrations to the Division of. Operating Reactors Guidelines.

JC0 89-13, Rev. O Pago 4 of 11

]

Amphenol Qualification Test Report No. 123-1275 provided the basis for the qualification (Reference 3) and.was included in Electrical Equipment Environmental Qualification (EEEQ) Record No. 8 (Amphenol Penetrations).

PGE's qualification of Amphenol penetrations was 3nitially reviewed by the Franklin Research Center (FRC) and.the results documented in their Technical Evaluation Report (TER)

(Reference 4).

The TER environmental qualification (EQ) summary form' for,Amphenol penetrations did not identify' the NRC requirement

" Criteria Regarding' Instrument Accuracy Satisfied" as a deficiency.

The NRC Safety Evaluation Report (SER)'(Reference 5)-stated ti e the NRC had " reviewed the evaluation performed by our'consultans >n the enclosed TER and concur with its bases and findings". In March 1984, PGE and NRC staff met to discuss F0E's proposed method of resolution for each of the deficiencies identified in the TER.

One of the TER deficiencies which PGE resolved related to lack of evaluation of test failures (Reference 6).

During a'May 1984 audit, the NRC inquired about Insulation Resistance (IR) measurement during high-temperature exposure and the performance of this equipment for low-voltage instrument circuits. PGE's response referenced the IR data ~(one megohm) in I

Appendix C of the test report and stated that the low IR data were i

artifacts of'the termination. methodology rather than an indication of-the true performance of the modules'(see Reference 8). 'On December 4, 1984 the NRC provided PGE with a Final Safety Evaluation addressing the environmental qualification of electric equipment important to safety for Trojan for compliance with the requirements of 10 CFR 50.49 (Reference ?).

The Safety Evaluation concluded i

that PGE's Electrical Equipment Environmental Qualification (EQ)

Program complies with the requirements of 10 CFR 50.49, and that the proposed resolutions for each of the EQ deficiencies identified in the FRC TER and SER are acceptable.

The recent NRC concerns about adequacy of IR data reflect heightened criteria regarding qualification data. Accordingly, PGE (and NUGEQ) has made a strong attempt to obtain additional information and clarifications concerning the' qualification testing of Trojan's Amphenol penetrations, and collected other EQ test data and industry experiences to support current qualification, and to provide further confidence and assurance in the performance capabilities of Trojan's Amphenol penetrations.

Additional Information Regarding the EQ Testing of Amphenol Penetrations in 197? for Troian (Appendix 1)

PGE contacted the original Amphenol test engineer, Mr. Gerald j

Sorensen, to obtain additional information regarding Amphenol Test l

Report No. 123-1275 and other EQ testing that was done in 1972.

i l

l JC0 89-13, Rev. O Page 5 of 11 I

u

Mr. Sorensen explained that low IR data in the Trojan testing were attributable to:

(1) inadequate drainage within the Loss of L'oolant Accident (LOCA) chamber. which caused flooding and prolonned immersion of cables and terminations; and (2) failure t.o properly protect loose ends of conductors from potential shorting to each other and to ground.

Mr. Sorensen also str.ted that the LOCA testing was repeated on the same test specimen.

IR readings were taken approximately five times during the first 75 minutes, and hourly thereafter for the remainder of'the 24-hour duration.

Mr. Sorensen said he recalled that the tests were' considered to have been satisfactorily completed, and further LOCA testing was not considered necessary.

File searches were performed Lsth at Wyle Laboratories in Norco, California and at ANCO Engineering, Incorporated in Culver City, California (custodian of Amphenol files). A copy of this test report was not found.

Other Industry Test Experiences (Appendix 2)

NUGEQ efforts on behalf of other utilities having Amphenol penetrations provided additional insights into other Amphenol EQ testing.

Mr. R. Perez, the test engineer for the Midland Nuclear Plant testing, provided an affidavit substantiating that all IR readings taken during the final Hidland test were above one megohm. His statements are based on the use of a one megohm criterion for determining Whether the chstber in which the test was conducted was flooded or not.

If IR readings were &bove one megohm, it was determined that the chamber was not flooded.

Mr. Perez took most of the data during the test and reviewed all the data prior to writing the final report.

For Standardized Nuclear Unit Power Plant System (SNUpFS) testing, the issue of why TR readings fell dramatically during the midstages of the test was also addressed.

Mr. Mano Aaron of ANCO has provided information reflecting that the chamber in which the SNUpPS tests were conducted flooded during this period, resulting in very low IR readings.

Based on the above information, it is clear that improper pigtail terminations or flooding will detrimentally affect instrument circuits. All of Trojan's electrical penetrations are located above the DBE ficod level and are terminated with fully qualified splices. The pigtails, manufactured by American Insulated Wire Company, are also qualified by a separate test report. This test report (Franklin Institute Research Laboratories Report No. F-C4197-2, " Qualification of Electrical Cables for a Loss of Coolant Accident") provides sufficient data to fully qualify the l

pigtails for instrument applications (see Appendix 3).

Trojan's Amphenol penetrations are not installed or terminated in a manner JCO 89-13 Rev. O Pago 6 of 11 l

l which would compromise the IR of instrument circuits. In addition, the Amphenol modules (mounted in a unitized header plate), are recessed from the in-Containment environment by approximately

{

4 feet.

This installation feature protects the penetration modules i

from direct impingement by water or chemical spray, thereby providing additional assurance of their performance capabilities.

1 Although material similarity to the tested Amphenol EPA modules has not been established, the successful testing conducted by Commonwealth Edison (Reference 9) provides additional assurance that the penetrations will perform their safety-related and post-accident-monitoring functions during and after a design basis

)

event.

Variation of Insulation Resistance With Temperature (Appendix 31 i

l Another NRC point of concern is that only one set of IR data was 5

collected during testing.

Since electrical penetrations and their l

pigtails represent a configuration characterized by an electrical l

conductor surrounded by an insulating dielectric material, pCE has

]

reviewed other test data for penetrations and cables. This review i

demonstrates that TR varies inversely with test temperature.

I 1

1.

" Environmental Test Report of Hermetic Feedthrough Connectors Used on Electrical penetrations At The Surry Power Station",

i D. G. O'Brien, Inc. Report No. C19QA061, dated February 28, i

1972, Appendix A.

Graph I portrays IR versus time.

Comparison of these two graphs clearly shows that in this EQ test the lowest IR readings were obtained at the highest test temperature.

2.

" Qualification Tests of Electrical Cables in a Simulated Loss-Of-Coolant Accident (LOCA) Environment", FRC Report F-C5115, 1

prepared for American Insulated Wire Corporation, dated April 1979. A review of Table 2 of the report, summary of IR measurements, demonstrates that the lowest IR data were obtained at the highest test temperature.

3.

" Report On Tests To Establish Insulation Resistance Versus Temperature Characteristic For Firewall III Irradiation Cross-Linked polyethylene Construction for Class IE Service In Nuclear Generating Stations", Rockbestos Repcrt QR-7804, dated January 27, 1988. A review of Figures 7 and 8 of this report (combined data) clearly establishes that IR decreases as temperature increases.

4.

" Design Qualification Test Report for Instrumentation Service Classification Electric penetration Coaxial And Triaxial Feedthrough Module Assemblies For Conax Corporation", Conax Report IpS-1045, Rev. E, dated January 4, 1985. A review of IR versus temperature measurements on pages D-4 and D-9 clearly demonstrates that IR decreases as test temperature increases.

JC0 89-13, Rev. O page 7 of 11

I III. TECHNICAL EVALUATION The NRC's concern, as stated in TEN 88-29 regarding IR values f or instrumentation circuits, has been that sensitive low-level signal 1

(milliampere) circuits will be adversely affected, due to reduction j

of IR values during a LOCA, main steam line. break, or main feedwater line break, and that qualification tests performed showed that IR readings were not taken at a frequency consistent with the guidance provided in IEEE 323-1974.

PGE has decid'ed to either replace the Amphenol instrumentation electrical penetrati'ons with fully qualified electrical penetration assemblies or establish qualification through testing that is applicable to Trojan. This would resolve the issue of inadequate l

IR readings during the LOCA simulation test (Reference 11)'.

PGE believes that continued operatiun of the Trojan plant in view of this decision is justified ~ for the following reasons:

1.

Despite the fact that the IR measurements were not taken at a frequency consistent with th'e requirements of IEEE 323-1974,'it is well established that temperature is a major factor in the degradation of IR for an insulating material. Tests conducted for similar Amphenol penetrations for Toledo Edison's. Davis-Besse nuclear plant simulating the LOCA tests with humid environ-ment (Reference 12) showed that the IR readings:taken during the LOCA simulations ranged from 1.2 x 107 to 5.4 x 1011 ohms.

There was a reading between 1.1 x 105 and 4.3 x 105 for a fan motor module. This low reading was not due to the'degrada-tion or faulty design of the module, but rather to the fact that the connectors userging.from the module body were not insulated.

A retesting of the conductors showed censurements between 3.0.

x 109 and 4.2 x 10. Although the peak temperature for the-9 LGPA test (224*F) is below the temperature for the Trojan plant (30d*F), the louest IR (one megohm) recorded for the Trojan plant correlates well with the test, even though readings were not taken frequently during the LOCA test.

2.

The repeat LOCA tests performed by Wyle Laboratories, Norco, California, on the same test specimen recorded IR data approximately five times during the first hour.(at.15-minute intervals) and each hour thereafter for the duration of the 24-hour LOCA test. Although the results of this test were not l

accepted by the NRC, due to lack of documentation, the IR problems experienced during the first test were not experienced during the Wyle test in which the modified termination configuration was utilized. Since all affected Trojan electrical penetrations are locsied above. flood level and terminated with qualified Raychem splices, the original Amphenol LOCA testing was conservative with respect to the conditions to which the conductors would be exposed at Trojan.

JC0 89-13, Rev. O Page.8 of 11

1 3.

The LOCA tests performed to show the adequacy of electrical i

penetrat' ions are done on a thermally aged and irradiated sample, simulating t he end-of-life conditions of the electrical I

penetration modules.

Since the instrumentation penetrations at

]

Trojan will be replaced during the next two refueling periods, unless conclusive qualification testing is performed, it is i

reasonable to assume that the modules will not have thermal or radiation degradation levels simulating the end-of-life l

conditions. Consequently, any TR values taken before the replacement period of these penetrations will render a more favorable'TR reading than the tested specimen which simulated the enu-of-life' conditions of the penetrations.

]

4.

The continued operation of the Trojan plant through the next i

two refueling periods will not affect the mitigating functions j

of the plant as a er suit of an accident.

In the event an l

accident should occur, the Engineered Safety Features Actuation l

Systems and the-Leactor protection System devices will virtually function instantly to mitigate the consequences of an ac cider.t.

This is long before either the rise of temperature 1

or the cirects of humidity will adversely affect the IR characteristics of the insulating material. The tests j

conducted by Conax on similar Amphenol penetrations show that i

degradation will not start taking place until approximately I

15 to 20 minutes after the start of the LOCA test i

I (Reference 12).

5.

The circuits that would be most affected as a result of a LOCA accident will be the post-accident-monitoring circuits, since the effects of high temperature and humidity will be more pronounced for circuits with continuous monitoring requirements.

1 However, since the penetrations will be replaced before the modules are completely aged and irradiated if additional conclusive qualification testing is not performed, the expected IR values will be more favorable than the worst reading of one megehm recorded during the LOCA cimulation test.

In addition, given the worst reading of one megohm for the IR of sensitive circuits, an error reading of 0.3 percent for transmitter circuits and 0.05 percent for resistance temperature detector circuits will result.

These errors are less than the accuracy of the instruments, Which are approximately 10 percent (Reference 13).

Tha expected errors, should an accident occur before the 1991 Refueling Outage, will be less than the values indicated above.

IV.

POWER AND CONTROL PENETRATIONS The concern regarding the degradation of IR values raised by the l

NRC in the Information Notice No. 88-29 pertains to low-level (milliampere) signals. Although the power and control circuits also have potential for degradation, the effects of such degradation on these circuits is insignificant.

JC0 89-13, Rev. O page 9 of 11

-l This is due to the fact that the magr.itude of currents drawn by j

these circuits is several orders of magnitude higher than any error j

introduced as a result of degraded IR values.

In' addition, power 1

and control cables generally have thicker insulation, due to the voltage requirements and larger cable size.

Since the IR of cables increases with increasing thickness of insulation [IR = K* LOG (D/d)]

(where k is a function of the material of the insulation, D/d is a function of the outer and inner diameters of the cable insulation) j ruch considerations for power and control are less important.

1 V.

CONCLUSIONS PGE believes that, based on reasons given in the technical evaluation, continued operation of the plant until the 1991 Refueling Outage is justified. This is a documentation issue which j

has been addressed by collecting additional data to respond to recent NRC concerns about instrument circuit IR values.

There is no impact on oublic health and safety issues, given the j

justifications presented.

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MLJ/62020 JCO 89-13, Rev. O I

Page 10 of 11

AMPHENOL TECHNICAL EVALUATION

, REFERENCES 1.

Letter from D. M. Crutchfield (NRC) to D. W. Cockfield (PGE), dated June 7, 1988, " Environmental Qualification of Containment Fenetra-tions at Trojan - Docket 50-344".

2.

NRC Information Notice No. 88-29, " Deficiencies in Primary Containment Low-Voltage Electrical Penetration Assemblies".

3.

" Design Verification Test Report on Glass-Filled Epoxy Modules -

Unitized Header Electrical Tor..t"ation Assemblies", Amphenol Report No. 123-1275, Revision 0, dated April 23, 1972, includin6 Attachment C (LOCA Test Data).

4.

" Technical Evaluation Report for Portland General Electric Company -

Trojan Nuclear Plant", dated Octcber 19, 1982 NRC Contract No. NRC-03-79-118/FRC Project C5257-485.

5.

" Safety Evaluation Report by the Office of Nuclear Reactor Regulation for Portland General Electric Company - Trojan, Docket 50-344",

transmitted to PGE by an NRC letter, dated December 17, 1982.

6.

Letter from B. D. Withers (PGE) to J. R. Miller (NRC), dated June 1, 1984, " Environmental Qualification of Electrical Equipment".

7.

" Safety Evaluation - Office of Nuclear Reactor Regulation to Portland General Electric Company for Trojaa Nuclear Plant Docket 50-344",

transmitted to PGE by L.

NRC letter, dated December 4, 1984.

l 8.

NRC's Comments and PGE's Resolutions for EEEQ Record 8 (Amphenol Electrical Penetrations) for the May 5-6, 1984 NRC audit, EEEQ Record 8.3.2.

9.

Letter from G. M. Holahan (NRC) to D. W. Cockfield (PGE), dated January 31, 1989, " Environmental Qualification of Bunker Ramo Electrical Penetration Assemblies at Trojan Request for Information".

10.

Wyle Laboratories Qualification Test Program on Bunker Ramo Instru-ment Penetration Assemblies for Commonwealth Edison Company for Utn in Byron /Braidwood Nuclear Power Station Unit 2, Test Report No. 17040-1, 11.

Letter from D. W. Cockfield (PGE) to US NRC, dated May 15, 1989.

12.

IPS-1077, Test Report for Evaluation Testing of Three Amphenol Module Assemblic.: for Toledo Edisen Company and Davis-Besse Nuclear Power Plant, Unit 1, Conax W/O 7-B4300.

13.

PGE Calculation TE-069, Rev. O, dated May 29, 198, "Amphenol Penetration Functional Verification", EEEQ Record 8.1.16.

62020 JC0 89-13, Rev. O Page 11 of 11

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l APPENDIX 1 Affidavit of Gerald C. Sorenson l

l l

t

4 4

AFFIDAVIT GF GERALD C. SORENSEN REGARDING ELECTRICAL PENETRATION ASSEMBLY TESTING I, Geralo C. Sorensen, being first duly sworn, hereby depose and sta' e as t

follows:

During the period of time from October 1971 to April 1973 I worked as an employee of Bunker Ramo Coporation, Amphenol SAMS Division.. Chatsworth, Californfo, in the Design and Qualification Testing of Electrical Penetration Assemblies.

I worked in various capacities during the period of my employinent, including Project Engineer and Supervisor of the Analysis Section.

I was significantly involved in the. initial planning and development of the test program to qualify Electrical Penetration Assemblies to design conditions for pr.ospective clients and tc the requirements of IEEE-317. IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generatint Plants.

This involved reviewing specifications and developing parameters to which to test the assemblies for anticipated accident conditions.

I oversaw and' witnessed or personally performed the functionality tests for_ this test program.

I collected test data, developed test data ' sheets, and at the conclusion of testing, I personally reviewed all data taken.

Test results are compiled in Amphenol Report No. 223-1275, Rev. O, " Design Verification Test Report on Glass Filled Epoxy Module Unitized Header Electrical Penetration Assemblies", dated May 5, 1972, which I authored.

I was contacted on June 9,1988, by Malcolm Philips, Jr., of Bishop. Cook, Purcell. and Reynolds, and was requested to provide information regarding the insulation resistance (IR) measurements taken during the loss of coolant accident (LOCA) exposure of the electrical penetration auemblies.

This affidavit is in repsonse to that request.

The electrical penetration test specimen was prepared by Amphenol and was subjected to preliminary tests as described in the above-referenced test report.

In order to perform the environmental test specified in the test sequence, this test specimen was transferred to the facilities of Able Boiler Works in Los Angeles, California.

Here the unit was subjected to the 24-hour saturated steam blow down conditions which would occur in the event of the maximum credible accident.

At Able Boiler Works on March 16, 1972, the test chamber was attached te a steam line from a ten (10)-hp a

Francis boiler rated at 100 psig, 330 F.

After the hookup was completed, the steam' valve was partially opened, allowing a small quantity of sgeam to escape into the test chamber.

The test chamber was heated to 125 F, and 4

y

this temperature maintained for 15 minutes.

After this 15-minute period, the steam valve was opened completely.and the full volume.of steam allowed to enter into the test chgmber.

Within 15 seconds, the chamber temperature and pressure rose to 300 F and 60 psig.

These conditions were maintained for a period of 15 minutes, during which time a random sampling of conductors were megger-tested for insulation resistance, It was found during the enurse or these tests that a majority of the conductors, as noted in the cest d' ta sheets in Appendix C of the test report, showed a

direct shorts or extremely low values of insulation resistance.

The reasorc for this ware twofold:

1.

Inadequate drainage within the LOCA chamber, which caused flooding and prolonged immersion of cables and terminations.

2.

Failure to properly protect loose endr, of conductors from potential shorting to each other and to ground.

All of the cable ends had been stripped in preparation for short-circuit testing and the ends were taped with plastic electrical tape.

All of the ends were then wrapped together with the entire cable bundle resting on the-bottom of the test chamber.

It was found at the end of the test that the tape around the bare conductor ends was missing from many conductors, allowing them to short directly to each other and to the bottom of the chamber.

In addition, a large amount of moisture was entrapped in the tape which was still in place.

Upon separation of the wire ends, all conductors were again Hi-Pot and megger-tested.

All modules successfully met the requ(red values.

Due to Amphenol's inability to obtain relevant IR data with the Able Boiler Works test facilities, a decision was made to repeat LOCA testing on the same test specimen at Wyle Laboratories in Norco, California. Modification to tne test plan provided for ' separation. of the conductors and for elevating conductor ends off the bottom of the test chamber.

Amp?anol developed a test procedure and prenred the specimen for connection to the Wyle steam source.

Wyle Laboratories provided calibrated test instrumentation and personnel to conduct the test and reccrd the data.

I personally witnessed the test setup and the initial phase of the test, and periodically monitored the test throughout its 24-hour duration.

IR readings were taken approximately five times (at 15-minute intervals) during the first 1:15 hrs of the test. Thereafter, IR readings were taken each hour.

Upon completion of this test, the test specimen and test data were shipped back to Amphenol where the test specimen was examined and more electrical tests were conducted. As noted in Amphenol Test Report 123-1275 (Page 35), these test results were acceptable.

L_

'i

-a-

.]

i I personally reviewed the test data collected, and while I do not remember the individual IR readings, I. do recall that the; tests were considered to have been satisfactorily completed and further LOCA testing was not

.j considered necessary.

The.IR problems experienced during the first test here not experienced during the Wyle Testing.

I did not personally keep a copy of the Wyle test data, and it is my understanding that others have been unable to locate the data in either Wyle or Bunker-Ramo files..

.i i

l

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i C

r/ L e.~ u.-_

GEPald Sorensen a

/

b ay of July,1988.

Subscribed and sworn to before me this d

I Q

JO e - ~

~

Notar Q blic D S D9*

l l

ATTACHMENT 1 Professional Profile of Gerald C. Sorensen Personal Data Name:

Gerald C.

(Jerry) Sorensen Residence:

2102 Austin Ct Richland, Washington 99352 Phone:

(509) 946-7800 Born:

October 25, 1938 Yuba City, California 5 ft 10 in., 150 lb, excellent health. Married; wife is college graduate and has taught elementary school. Seven children - Ages 7 through 23.

Educationa' Background Received Bachelor of Engineering Science (five-year degree) from Brigham Young University (1966). Majored'in Mechanical Engineering with minors in mathematics and language. Post-graduate courses in math, statistics, and nuclear reactor materials.

Short courses and seminars in nuclear reactor safety; quality assurance; computer programming; supervisory management skills, Kepner Tregoe, Management Oversight and Risk Tree (MORT) Analysis.

Professional Affiliations Member American Nuclear Society; American Society of Mechanical Engineers; National Management Association (Certified Manager): Atomic Industrial Forum Committee on Reactor Licensing and Safety Steering Group; Chairman of AIF Subcommittee on Anticipated Transients Without Scram (ATWS).

Employment History 1973 -

WASHINGTON PUBLIC POWER SUPPLY SYSTEM; Manager Regulatory Present Programs since 1982.

Responsible for all licensing (environmental and safety) activities for the Supply System's nuclear power plants.

At the present, the Supply System has an operating Boiling Water Reactor (BWR), and two Pressurized Water Reactors (PWR) for which license applications have been docketed, but construction has been deferred. Two other PWR projects have been canceled.

I

1 j

l Professional profile (C. C. Sorensen)

Page 2 j

Responsible for the management of Licer sing Project Managers on each of the three projects, and a small l

central staff. Responsible for all interface with State and Federal agencies'in obtaining and maintaining l

permits and licenses to construct and operate the Supply 1

i System's nuclear power plants.

I i

Performing the duties ofEthis position requires frequent contact with Senior Management at the Nuclear Regulatory Commission (NRC) and the Washington State Energy Facility Site Evaluation Council (EFSEC). The Manager, Regulatory Programs is typically the Senior Gupply System representative in meetings with these organizations.

l l

Responsible for maintaining awareness of new Federal or l

State regulations, rules, etc which could impact on-Supply System activities involving the appropriate organizations and providing comments to NRC, EFSEC, or others, as appropriate.

Responsibilities prior to present position included various supervisory and engineering'positionsLin.the licensing organization.

In addition to my position responsibilities, I have instructed in company-sponsored management skills workshops since 1981.

Presently instructing. National Management Association courses for company NMA Chapter.

I 1971 - 1973 AMPHENOL SAMS DIVISION OF BUNKER RAMO CORPORATION, Chatsworth, California, Analysis Section and Project Engineer.

Primary Responsibilities for ASME Code analysis required for electrical penetration assemblies for nuclear power plants. Also responsible for development and conduct of prototype testing for all new penetration designs.

l February 1971 - SOUTHERN CALIFORNIA E0ISON COMPANY, Los kngeles,.

October 1971 California, Licensing Engineer, San Onofre Units 2 and 3.

Responsible for PSAR amendments for the SONCS 2/3 PSAR.

1966 - 1971 DOUGLAS UNITED NUCLEAR CORPORATION, Richland, Washington, Various engineering positions in the DUN organization.

Professional Profile (G. C. Sorensen)

Page 3 Responsibilities included: Process engineer at an cperating nuclear production plant; Nuclear Safety Engineer; Fuels Development Engineer.

MLJ/38980

)

4 l

APPENDIX 2 1

l Affidavits of Other EQ Test Engineers

)

i 1

I

AFFIDAVIT OF RAYMOND PEREZ REGARDING ELECTRICAL PENETRATION ASSEMBLY TESTING IF;RODUCTION 3

1 I, Raymond Perez, being first duly swarn hereby depose and state as follows:

During calendar years 1977 and 1978 I worked as a consultant to Bunker Ramo Corporation, Amphenol SAMS Division, Chatsworth, California, in, among other things, qualification i

l testing of electrical penetrations.

(My professional qualifications are noted in Attachment 1.)

I was involved substantially in the initial planning and development of the test program to qualify to IEEE 323-1974 the electrical penetration assemblies (EPA's) for the Midland Plant Units 1 &

2.

I attended meetings with Amphenol and Bechtel Ann Arbor engineers and managers both in California and in Michigan regarding the planning and performance of these qualification tests.

I witnessed or personally performed the functionability tests for this test program.

Among other things, I developed data collection sheets, recorded data taken during the testing and reviewed data taken.

At the conclusion of the testing I personally reviewed all the data taken and wrote the following test reports:

Test Report - Qualification Testing of Electrical Penetration Assemblies For Midland Plant Units 1 & 2 of the Consumer Power Company - Project 1003-3 Aug st 1978 l

Test Report - Midland Supplementary II Qualification Testing of Electrical Penetration Assemblies For Midland Plant Units 1 & 2 of the Consumer Power Company - Project 1003-7 February 1979

l s

l On May 13, 1988, I was contacted by Malcolm Philips, Jr.,

an attorney in the Washington, D.C. law firm of Bishop, Cook, l

Purcell & Reynolds, and requested to provide information regarding the insulation resistance (IR) measurements taken during the LOCA exposure testing of the low voltage I

instrumentation electrical penetration assemblies.

This affidavit is in response to that request.

SUMMARY

AND CONCLUSIONS J

During the initial preparation for LOCA testing, it 1

became apparent that control of the steam / water environment was extremely difficult.

Because of the test set-up, 2

excessive water (and chemicals) in the chamber was a significant problem.

This problem was reflected in, among-other things, low IR readings.

While significant moisture' (and chemicals) usually resulted in lower IR values for circuits with terminal blocks, where " flooding" was present even circuits with splices (instead of terminal blocks) were significantly impacted.

By trial and error we refined out control of the test equipment to minimize " flooding" problems.

However, we had no direct method of indicating water levels during testing.

For the latter testing (Midland Supplementary II Qualification Testing), I devised informal criteria using IR readings to l

determine if " flooding" was present.

For example, if IR values of circuits with splices were. equal to or greater than 6

10 Ohms, I was fairly certain that there was no flooding.

v

-3 (It should be noted that based-on my experience with the earlier Midland tests, I was able to maintain values above 106 Ohms at peak temperature conditions as long as flooding was not present.)

In the early stages of the Supplementary II. Test, IR readings were taken on a very frequent basis (approximately once every 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).

This was more frequent than required by the test plan, and was done as a check to know, among other things, the " flooding" status of the chamber.

As the threat of flooding subsided in latter stages of the test, IR readings were taken less frequently -- even,tually only once-a day (consistent with the test plan).

Accordingly, the total' I

number of IR readings taken during the 30-day test were in excess of 40 for each circuit.

Upon completion of the Test,.I l

prepared the Test Report.

Based on, among other things, my direct observation of IR readings and my review of all IR data at the time I wrote the Midland Supplemental Test Report (Report #1003-7), I knew-that we had successfully completed the LOCA testing, there was no evidence of flooding during the test, and all IR readings for circuits with Raychem splices were at least los

Ohms, including during the early stages of the LOCA when the environmental conditions were the most severe.

These IR values increased as the severity of.the exposure environment decreased.

1

s 4

-4 DISCUSSION As background, the LOCA testing chambers for all the Midland testing consisted of two horizontal 10-foot long pipes (i.e., an 18 inch pipe standard (IPS) and a 12 IPS) stacked one upon the other which enclosed the EPA's being tested.

These chambers were in turn housed in a cinder block concrete 1

" block house" approximately 15-feet ]vng and 10-feet wide, through which the steam inlet and chemical spray piping entered overhead and the drain pipes exited below.

At the connection point of the piping to the chambers were a number of hand-controlled valves used to control temperature, pressure and chemical flow rate.

There was very little working space between the cinder block wall and these valves.

There was also very little room around the electrical connections on the EPA's which carried as much as 8800 volts and 600 amps.

The IR readings were taken from these e]ectrical connections.

The single most degrading factor in the LOCA exposure testing was the prolonged " immersion" of some of the cables and terminations in the chemical spray solution at elevated temperatures.

During the early tests (i.e.,

the December 1977 test described in Project 1003-3 Test Report), substantial flooding of the LOCA chamber occurred due to excessive amounts of chemical spray (i.e.,

25 gpm per square foot), insufficient drainage and inadequate control of the thermal hydraulic LOCA system (see NPD-1270, Attachment 2).

This was evidenced by the low IR readings and, in the extreme, by excessive amounts

s

. of fluid released to the floor by the steam traps and visual evidence of the liquid in the LOCA chambers after the testing.

(No sight. glass or direct monitoring of liquid level in the chamber was available.)

To provide perspective, simulation of a DBE LOCA event entails the control of' significant quantities of fluid and thermal energy.

Due to the non-steady state character of the initial blow-down and the requirement to maintain elevated temperature and pressure under pseudo-equilibrium conditions throughout the 30-day test period, a trade-off between sustaining elevated pressure through the introduction of steam, and minimization of flooding caused largely by the chemical spray had to be maintained.

As a result of several impromptu meetings discussing possible solutions to the flooding problems, attempts were made to improve control of the thermal hydraulic system, including (1) the use of additional steam traps on the steam inlet side, which helped only minimally, and (2) the use of an electrically powered superheater, which was found to be ineffective.

During the subsequent testing (i&gz, the October 1978 Midland Supplementary II Tests described in Project 1003-7 Test Report), the chemical spray was reduced to 0.25 gpm per square foot for the first 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of the LOCA and then reduced to 0.15 gpm per square foot thereafter; additional drainage capacity was achieved by putting two pumps in series (not shown on the drawing); and better overall control of the LOCA

. environment was maintained.

As a result, based on all evidence, no flooding occurred.

In order to monitor the possibility of flooding during I

the Midland Supplementary II Test, I used IR readings as criteria to detect flooding.

For example, if readings on 6

circuits with Raychem splices were at least.10 Chms I had confidence that no sustained flooding was present.

Again, l

based on my experience with previous Midland tests, I was able 6

to maintain values above 10 Ohms at peak temperature conditions as long as flooding was.not present.

(It should be 1

noted that even excessive moisture alone, without flooding, usually caused immediate and significantly lower IR readings l

- ]

on circuits with terminal blocks.)

Because there was no flooding during the duration of this test, the electrical and-mechanical characteristics of the low voltage EPA's-(LVEPA),

during and after the LOCA were significantly enhanced over the initial tests (i.e., Test Report 1003-3).

j Another factor which enhanced the.IR values during the Supplementary II LOCA Test for the Module D 69#16 instrumentation LVEPA was the fact that the Raychem RFR in-line splice and epoxy termination were enclosed in the'J-box (see NPD-1339, Attachment 3) oli the inboard side of the LOCA chamber.

This further reduced the effects of the direct chemical spray.

1 The viring diagram (see NPD-1337, Attachment 4) shows the IR test connections for the Module D 69#16 instrumentation EPA.

IR measurements were made using an HP4329A High l

1 Resistance Meter (see data sheet, Attachment 5) during the

____._w,_------x---

1 4

' functionability tests.

Additionally, during the Midland supplementary II tests a Simpson Model 270 VoM was used to make " quick look" IR measurements at random intervals (to check for flooding).

During the Supplementary II tests, per the test plan, IR readings were taken eight hours into the LOCA and then once daily throughout the 30 day exposure.

In addition, IR readings were taken much more frequently during the early stages of the test (on the order of every 1-2 hours) to check for " flooding" problemr The total number of IR readings taken during the LOCA test werc in excess of 40 for each circuit.

The readings were recorded on standard letter size forms which were submitted to Amphenol SAMS along with all the other supporting documentation and finished test reports.

During the testing I recorded virtually all of the IR data.

Further, I discussed with others the data taken --

particularly with respect to the l acts that (1) the data reflected that we did not have filoding in the chamber and (2) we had been able to maintain our axpected electrical performance.

At the conclusion of the testing I reviewed all data prior to writing the report.

While I canrict remember the individual IR readings, I do know that all the IR readings for the Raychem circuits were at least 106 Ohms, a criterion used l

to determine if flooding was present.

I recently visited ANCO, Inc. in Los Angeles, California, to review remaining Amphenol files to detelnine whether any of I

the original data sheets were available.

My thorough review i

q

\\

1 i

_8 indicated that the data sheets were not in the files, and I.

have no knowledge of.their location.

However, two data points contained.in a partial log-book referred to by.Mano Aaron in f

Attachments to this affidavit support my clear recollection that no flooding occurred and all IR readings for Raychcm 6

splice circuits were equal to or above.10 Ohms.

(The magnitude and time of these readings are reflected in the attached letters from ANCO Engineers Product Group, Inc.,

Attachments 6 and 7.)

Additionally, no data regarding the l

Midland Supplemental test of which,I am aware is contradictory to my clear recollection.

f r

SkuNJP+,

Raymo Perez Subscribed and sworn to before me this d day of August, 1988.

i s.L)

.J.Lb l,

Nota Q Public

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i

i i

RAYMOND PEREZ l

Education l

B.S., Mechanical Engineering, UCLA, 1971 M.S., Nuclear Engineering, UCLA, 1974

~

Professional Registration Mechanical Engineer - California No. 17525 i

' New Mexico No. 9440 Washington No. 21743 Nuclear Engineer

- California No. 784 Professional Areas Qualification Testing Digital Signal Processing Mechanical Engineering Dat'a' Acquisition Systems i

Nuclear Engineering Structural Dynamic Testing Relavent Qualifications l

Experienced in qualification. testing of systems, structures.,

and components for nuclear applications; I

' amiliar with nuclear codes, standards and regulatory guides r

and, Experienced in the design, development and direction of complex nuclear equipment qualification programs for electrical utility systems, U.S. DOE Defense Programs, and the U.S. Navy'.

Experience 1985 to present Westinghouse Electric Corporation,SunnyvAle,_Ca.

Senior Engineer, Machinery Test Group, Mari.no Division. Responsible for qualification' testing l

of marine propulsion and electrical generation machinery systems to U.S. Navy specifications.

Projects include:

Electrical Generation Systems for Trident submarine and CVN Carrier Nuclear Power Plants.

Marine Propulsion Systems for WP-85 and MP-21 Prototype Nuclear Power Plants.

Marine Propulsion Systems for CGa47, LHD and l

3 other non-nuclear power plants.

i l

I

]

l

.s.

t j

1976 to 1985 Engineering Analysis and Test Company President responsible for administrative and technical aspects of the firm's business.

Responsible for the management of nuclear and mechanical engineering projects including the performance of equipment qualification.

programs; radioactive effluent dispersion analysis; safety and environmental impact analysis, etc.. Provided engineering consulting services to architectural engineers, equipment manufacturers, utilities and government agencies.

Projects included:

IEEE-317-1972 Qualification of Electrical Penetration Assemblies for Bunker Ramo Corporation, Amphenol SAMS Division, Chatsworth, f

California.

IEEE-323-1974 Qualification of Class IE Electrical Equipment for Nuclear Power Generating Stations for:

Bunker Ramo Corporation, Chatsworth, Ca.

J3wers Regulator Co.,Skokie, Illinois Consip Inc., South E1' Monte, Ca.

I G & H Technology, Santa Monica, Ca.

IEEE-344-1975 Seismic Qualification of Class 1E Equipment For Nuclear Power Generating Stations for:

Bunker Ramo Corporation, Chatsworth,Ca.

Powers Regulator Company, Skokie, Illinois custom Control Panels, Santa Fe Springs, Ca.

Ronan Engineering Company, Los Angeles, Ca.

)

Northwest Process, Seatle, Wa.

]

Lawrence Livermore National Laboratory and others ASME Section III Pressure Vessel Code Stress Report Preparation fors Bunker Ramo Corporation, Chatsworth,.Ca.

l Amatek Division of Straza, Santa Fe Springe,Ca.

Nuclear Engineering Consulting including waste management studies, criticality and abielding analysis for Bunker Ramo Corporation, Chatsworth, Ca.

Garrett Corporation, Torrance, Ca.

Atlantic Richfield Hanford Company, Wa.

Allied Chemical Company, INEL, Idaho-

c.

e 1974,to 1976 Applied Nucleonics Company, Los Angeles Engineering consultant to architectural' engineers, equipment manufacturers, and

_ government agencies on nuclear projects;

'j Projects included:

Idaho Chemical' Processing Plant,' Idaho Fuels and Materials Examination Facility,Wa.

High Level' Waste Management Program Hanford,Wa l

High Energy Gas. Laser ; Facility LASL, N.M.

j Humbolt. Unit 3 - PG&E, Ca.

_j

. Calcine Conversion Facility, INEL, Idaho j

Retrievable Surf ace Storage - Facility, Hanford, Wa.

j 1972 to 1976-

' Instructor, UCLA Engineering Extension Responsible for the presentation of a three quarter sequence of classes in nuclear engineering covering nuclear physics, nuclear. engineering technology, safety and environmental protection.

This class sequenc9 plus.'two alectiver satisfied the requ.irements for UCLA's Enginee:ing Certificate Program and was attended by. practicing engineers new to the nuclear industry.

1972 to 1974 Lecturert Nuclear En'ergy Laboratory, UCLA ^

Subject matter dealt with nuclear safety and the environmental implications of nuclear systems. This position-was~part of an AEC

~

. Trainee. Program which was-national *in scope

' and was funded and coordinated through the Oak. Ridge National Laboratories. Participants.

conducted experimental nuclear engineerlag-studies at ORNL.

I 1970 to 1972 Research Assistant,' Energy and Kinetics,"UCLA i

Conducted computer aided engineering studies a

involving fast reactor fuel element modeling;,

light water reactor, fast breeder reactor and nuclear materials environmental and safety j

analysis performed experimental activation analysis, etc..

1967 to 1970 Research Entrineer, Microsemiconductor Corp,Ca Thermo-chem;, cal process technique development involving the use of highly toxic chemicals at temperature and pressure extremes.

l l

I I

4 l

l APPENDIX 3 Insulation Resistanee vs Temperature

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ML20247C974

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ML20247C974

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