ML20073J706

Forwards Addl Info Re Environ Qualification Program,Per Bj Youngblood 830315 Request.Fsar Will Be Amended to Include Info

ML20073J706
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	04/14/1983
From:	Swartz E COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
6374N, NUDOCS 8304190377
Download: ML20073J706 (45)

Text

-

~y I M [C-- \ Commonwealth Edison

) One First Natsonal Plaza, Chicago, Illinois

(

g O ' J Address Reply to: Post Office Box 767

\ j Chicago. Illinois 60690 April 14, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Additional Information Concerning Environmental Qualification Program NRC Docket Nos. 50-454/455 and 50-456/457 Reference (a): B. J. Youngblood letter to D. L. Farrar dated March 15, 1983

Dear Mr. Denton:

Reference (a) requested that the Commonwealth Edison Company provide, by April 15, 1983, certain additional information concerning our Environmental Qualification Program for Byron and Braidwood. Stations. The purpose of this letter is to provide the requested information.

Our FSAR will be amended to include the information contained in the Attachment to this letter as appropriate. Please address any questions you or your Staff may have concerning this matter to this office.

One (1) signed original and fif teen (15) copies of this letter with Attachment sie p.rovided for your use.

Very truly yours, p

E. Douglas Nuclear Licensing Administrator Attachment cc: J. G. Keppler - RIII RIII Inspectors - B/B Q/

6374N DR A 0 0 L.

A

g >

B/B-FSAR QUESTION 270.1

" Based on the information in your program, we are unable to determine if all of the systems and components requiring qualification have been identified. Provide the following additional information for our review.

a. A comparison of the systems in Table 3.2-1 of the FSAR with the systems containing equipment in a harsh environment in the June 17, 1982 submittal. Justifica-tion should be provided for the exclusion of safety-related systems in Table 3.2-1 from the environmental qualification program (e.g. not required for accident mitigation, all components located in a mild environment, etc.).

b. A list of the TMI Action Plan equipment currently in your program and its equipment I.D. number. If not in your program, describe the qualification status or your plans for qualification, including the schedule for completion of qualification in accordance with Category I of NUREG-0588.

c. Confirmation that all equipment defined by items 2a, b, and c of Appendix E, NUREG-0588 has been included in your review.

d. A justification for defining ' hot standby' as equivalent to the Class lE term ' emergency shutdown.'

e. A clarification of the scope of the program as it relates to all units of the Byron /Braidwood Stations.

For example, qualification of some equipment in Byron 2 (such as Rosemount transmitters and Bunker Ramo penetrations) and Braidwood has not been addressed in the environmental qualification submittal. The staff will review and evaluate only those plants for which adequate information is received.

f. Qualification data sheets for all equipment located in a harsh environment. Some items not currently addressed in the submittal are cable, splices, conduit, and MCC's."

RESPONSE

a. Table 2.1-1 of the EEOR (Equipment Environmental Qualification Report) lists all systems which contain Class lE equipment located in both mild and harsh environmental zones. (Systems Q270.1-1

a B/B-FSAR with Class' lE equipment located in harsh environmental zones are designated with a superscript h) . Class lE equipment in mild zones is not excluded from the program.

Identification sheets for the same are provided in Volume 4 of the EEQR.

A comparison of Table 3.2-1 of the FSAR and Table 2.1-1 of the EEQR is provided in Table Q270.1-1 with accompanying notes which disposition discrepancies.

b. Appendix E to the Byron /Braidwood FSAR presents the Byron /

Braidwood response to the TMI Action Plan. A list of the specific Class lE TMI Action Plan equipment which was not included in the qualification program identifying its Byron /Braidwood equipment ID number and.its present qualification status is given in Table Q270.1-2. All equip-ment ID numbers which are not currently listed in the EEQR will be added in the next revision.

c. All equipment defined by Items 2a, b and c of Appendix E, NUREG 0588 has been included in the environmental qual-ification review. The appropriate NUREG-0588 category for each piece of equipment is identified in the EEQR on both the ID sheet and the equipment's data sheet.

d. The design and licensing of Byron /Braidwood are based on the capability to achieve hot standby conditions. This capability is described in the responses to RSB Questions 212.6, 212.47 and 212.154. The hot standby condition is a safe, stable condition which allows operator action or plant equipment repairs to be included in the procedure to proceed to a cold shutdown condition.

The Byron /Braidwood qualification program includes all Class lE equipment required to bring the plant to a hot standby condition and/or mitigate the consequences of a postulated accident as denoted in Section 2.1 of the EEQR.

It was not intended to equate emergency shutdown to hot standby. Therefore, Section 1.0 of the EEQR will be revised as follows:

"The: equipment covered by this program includes equipment associated with systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, containment and core residual heat removal or otherwise essential in preventing signi-ficant release of radioactive material to the environment."

Q270.1- 2

f B/B-FSAR

~

.e . An informal copy of qualification data sheets for Unit 2 Class lE equipment located in harsh environmental zones was submitted to the staf f for review in January 1983.

A docketed submittal will be forthcoming.

As a result of plant arrangement, eqqipment duplicated between Unit 1 and 2 is located in the same environmental zone (e.g., ICC0lPA is located in the same environmental zone as. 2CC0lPA) except the Rosemount Transmitters identified below.

Equipment, instrument and valve ID numbers for Byron are identical to those for Braidwood. Therefore, each ID number in EEQR (although listed only once) applies to both Byron and Braidwood except for items unique to one particular unit. There are no unique items in harsh envi-ronmental zones.

Rosemount transmitters with ID numbers 1PP-AF011 through 1PT-AF018 used for Byron /Braidwood Unit 1, are located in environmental zone A8. Rosemount Transmitters with ID numbers 2FT-AF0ll through 2FT-AF018, used for Byron /

Braidwood Unit 2, are located in environmental zone A-10.

Replicate equipment was used for all four units (Byron 1& 2 and Braidwood 1 & 2). The only equipment furnished with different model numbers are the electrical penetrations.

Conax penetrations are used exclusively for Byron and Braidwood Unit 1 and for two (2) Byron and Braidwood Unit 2 penetrations: 2LV09E and 2LV10E. Bunker Ramo penetrations

-are used for the balance of the Byron and Braidwood Unit 2 penetrations,

f. Environmental qualification reports for generic electrical items which do not have ID numbers have been reviewed for Byron and Braidwood under the same review process as described in the EEQR. Summary sheets for these generic items are enclosed for your review and will be incorporated into the next revision of the EEQR. ,

Q270.1-3

P .

B/B-FSAR TABLE Q270.1-1 CLASS lE SYSTEM COMPARISON OF TABLE 3.2-1 of FSAR AND TABLE 2.1-1 OF EEOR SYSTEM CODE FSAR EEQR NOTES AF X X AP X X AR X X CC X X CQ X (1)

CS X X CV X X DC X X DG X X DO X X EF X (3)

FP X (2)

FW X X HT X (5)

IC X (1)

IP X X '

LV X (1)

MS X X

'NR X X OG X (6)

PA X (4)

PC X (6)

PL X (4)

PM X (4)

PR X X PS X X RC X (6)

RD (9)

RE X (2)

RF X (2)

RH X X j RP X X (7)

RY X X SA X (2)

SD X X l SH X (2)

SI X X SX X X TG X (8)

VA X X VC X X VD X X l VE X X, VP X X VQ X (2) vX X X WO X X 0270.1-4_ _

. . . . . . .-. . . ... - _~ . . .- __.. -- .

4 o

B/B-FSAR

. TABLE Q270.1-1 (Cont'd)

Notes

1. These systems are non-lE except for electrical penetrations.

Table 3.2-1 will be revised to indicate that system does contain Class lE equipment.

2. These systems are non-lE except for containment isolation valves and are included in the generic category in Table 3.2-1, designated as PC, " Primary Containment Isolation."

3. Class lE components in this system are identified by equip-ment numbers in various other systems included in Table 2.1-1 of the EEQR.

, 4. These systems were inadvertently omitted from Table 2.1-1 of the EEQR.and will be included in a future revision.

! 5.- This system currently does not contain any Class lE components.

Table 3.2-1 of the FSAR will be revised to delete this system in the next amendment.

1 6. This system contains. Class lE components and will be included in Table 3.2-1 of the FSAR in the next amendment.

7. The RP system is listed in both the FSAR and the EEQR.

RP system equipment is not identified by an RP system code, but rather various system codes, i

l- 8. - Table 3.2-1 of ti.e FSAR lists the turbine stop valve limit

! . switches as part of the TG system. Actually, they are part of the MS system. The FSAR will be revised accordingly.

9. The RD system was inadvertently omitted from the EEQR and the FSAR, and will be added in a future revision.

f e

Q270.1-5

f B/B-FSAR )

I TABLE Q270.1-2  :

TMI ACTION PLAN EQUIPMENT STATUS EQUIPMENT MANUFACTURER QUALIFICATION ID NUMBER DESCRIPTION AND MODEL STATUS t 1/2RC014A Reactor Coolant Valcor #V526- Complete review of 1/2RC014B System Vent 6043-5 Valcor Report for  ;

1/2RC041C (Solenoid B/B applicability 1 1/2RC014D Operated U/US) 1/2RY8000A Motor Operated Limitorque Awaiting proposal 1/2RY8000B Relief Isolation- from Westinghouse Valve with Positive Posi-tion Indication 1/2RY455A Pressurizer NAMCO EA180 Qualified by Westing-1/2456 Power Operated house EQDP HE-3 Relief Valve

• Externally Mounted Limit Switch (For

. Positive Posi-tion Indication) 1/2RY8010A Pressurizer Copes Vulcan Awaiting submittal 1/2RY8010B Relief & Safety D-100-160 of Vendor's Qualifi-l 1/2RY8010C Valves with Crosby HP-BP-8E cation Report Valve Mounted i Reed Switch for Positive Posi-tion Indication 1FIS-AF022 Auxiliary Feed- Barton 288A Review Vendor 1FIS-AF024 water Flow submittal Qualifi-IFIS-AF026 Transmitter cation Report for -

IFIS-AF028 B/B applicability

. 1/2RT-AR020 - Containment General Atomic Review Vendor

! 1/2RT-AR021 High Range Co. RD-23 submittal Qualifi- ,

Area Monitors (Detector) RM-80 cation Report for (Micro Processor) B/B applicability RM-23 (Remote Display) i

(

l Q270.1-6

5 B/B-FSAR TABLE Q270.1-2 (Cont ' d) l l

EQUIPMENT MANUFACTURER QUALIFICATION ID NUMBER DESCRIPTION AND MODEL STATUS 1/ 2PT-PC 004 Containment Barton 763 Qualified by 1/2PT-PC 005 Pressure .Trans- Westinghouse

.mitters EQDP-ESE-1 1/2LT-PC 006 Containment H 9 0 Barton 764 Qualified by 1/2LT-PC007 Level Transmitter Westinghouse 1/2Ur-PC 008 Containment Sump EQDP-ESE-3 1/2Ur-PC009 Level Transmitter 1/2PS 26J Containment H Sentry Awaiting submittal 2

1/2PS 28J Monitoring of Vendor Qualifi-Equipment cation Report (Gas Analyzer

& Monitor) 1/2LE-RC 019 Core Cooling Combustion Awaiting submittal 1/2LE-RC 020 Detection Eng. & Westing- of Vendor Qualifi-Instrumentation house cation Plan s

e Q270.1-7

B/B-FSAR QUESTION 270.2

" Normal and accident environmental conditions must be defined'for areas of the plant which experience a signi-ficant change in environment as a result of a LOCA or HELB. The following additional information is required to adequately define the environmental conditions:

a. The time-temperature profiles for high energy line breaks outside containment.

b. The definition of a ' harsh' environment used in the development of the EQ program. If equipment has been exempted from qualification because of exposure to low level doses of radiation, provide the justi-fication for this practice. For equipment with solid state devices in environments with significant increases in doses and dose rates as a result of an accident, describe the methods of qualification or bases for exemption.

, c. The postulated flood level inside containment and a reference (such as a section in the FSAR) .

d. A modification to or justification for the main steam line break temperature profile inside containment used for qualification. The profile in your program is less severe than that approved by the staff in Section 6.2.1.1 of the Byron SER (NUREG-087 6) .

e. The methodology used for estimating the magnitude of the radiation environment for equipment under normal operating conditions.

f. An example of equipment specific calculations referenced in Section 4.2.1 of the submittal used to reduce the temperature required for qualification of equipment exposed to HELB's outside containment.

g. A profile for pressure vs. time inside containment for the accident duration.

h. The basis for the LOCA duration selected for equipment qualification.

i. Confirmation that random single failures of mitigation equipment have been assumed for high energy line breaks outside containment, to be consistent with 3.6.1 of the Byron SER."

Q270.2-1

B/B-FSAR

RESPONSE

a. The time / temperature profiles for the high energy line break -(HELB) areas outside containment may be found in Attachment C3.6 of the FSAR.

b. A harsh environment is defined as any area which will experience a significant change in one or more of the environmental parameters as a result of an accident. The parameters that are considered are temperature, pressure, humidity, caustic spray, radiation, and submergence.

Harsh environments also. include areas which are exposed to an abnormally'high temperature, pressure, humidity, and/or total integrated radiation dose (TID) during normal plant operation.

The equipment included in the harsh environment sections of the Byron /Braidwood EEQR is all the equipment that is locaged in an environmental zone with a TID of greater than 'O rads. None of the equipment in harsh environmental zonc) as been exempted from radiation qualification.

The qualification for all Class lE equipment with solid state components is reviewed to the same program described in the EEQR.

c. The actual flood level inside the containment is elevation 382 feet 2 inches. This flood level is documented in Attachment D3.6 of the Byron /Braidwood FSAR. (Attachment D3.6 was included as part of Amendment 40 to the FSAR.)

d. The accident temperature profile specified for the environ-mental qualification of safety-related electrical equipment l

located inside containment is that shown in Table 3.11-2 of the' Byron /Braidwood FSAR. This profile envelopes all of the pontulated pipe breaks and LOCA profiles shown in Chapter 6 of the Byron /Braidwood FSAR. The design analysis used to generate the temperature profiles shown in the FSAR was the COCO computer code, which is widely I used in nuclear power plant containment analysis. Even though the staff calculated a slightly higher maximum temperature during the performance of the confirmatory analysis, we feel that the temperatures shown in Table 3.ll-2 are conservative and therefore justifiable to use as the basis for in-containment qualification. In fact, the Byron SER states, " initial conditions and input data, including passive and active heat removal parameters, were conservatively chosen to produce the highest contain-ment pressures and temperatures."

Q270.2-2

. ~

B/B-FSAR If the NRC chooses to review the Byron in-containment environmental qualification against the peak value of 330 F calculated by the confirmatory analysis performed using Contempt-LT/28 Code, it can be shown that the Byron in-containment Class 1E equipment is qualified for the peak temperature of 330 F with some margin. This is a result of margin that was included in the original envi-ronmental qualification testing.

e. On August 24, 1982, during a conference call involving Messrs. Akstulewicz and Chesnut of the NRC, T. Tramm of Commonwealth Edison Company, W. J. Johnson and J. D.

Regan of Sargent & Lundy , information necessary to resolve this comment was communicated to the NRC. The following is a summary of information inserted in the Byron /Braidwood Qualification Report. Note that Section 3.3.3 of the current report will be changed to Section 3.3.4.

3.3.3 Normal Ooeration Excosure Equipment located in areas containing radiation sources during normal operation may accumulate a significant level of radiation exposure prior to the design basis accident.

Bounding values of' normal operation exposure are established by using the dose rate modeling techniques described above and the following:

1) Source terms are based on defective cladding on 1%

of the fuel rods as described in FSAR Section 12.2.

2) For areas containing no radiation sources, the design dose rate shown in FSAR Section 12.3 is used.

3) The period of operation is 40 years,

f. The time / temperature profiles referenced in Section 4.2.1 of the EEQR are for normal conditions only, in lieu of aging to 40 years at maximum temperature. Equipment specific calculations have not been used to reduce the temperature required for qualification of equipment exposed to HELB's outside containment.

g. The pressure profile used for in containment qualification is given in Table 3.1-1 of the BBOR. Section 3.2.i.1 presently includes only the peak pressure value. This section will be revised to include the entire containment accident pressure profile (50 psig for the first 20 minutes, saturated steam thereafter).

Q270.2-3

- ,-- - e e

B/B-FSAR

h. Specific postaccident operability requirement for each device are developed from the following guidelines:

Required Postaccident Equipment Operability -

1. Equipment necessary 5 minutes to perform trip functions.

2. Equipment that is 2 weeks located outside containment, is accessible, and can be repaired, replaced, or recalibrated.

3. Equipment located 4 months {

inside containment (this number is and is required for based on an acceptable post-accident amount of time to monitoring. allow the instrument to be repaired, replaced, or recalibrated or an equivalent indication obtained)

4. Equipment that.is located 1 year j inside containment, is J

inaccessible, or cannot 1 be repaired, replaced l or recalibrated.

i

5. Equipment located in a Continuous '

l mild environment  !

l following an accident. -

i. Section 3.2.2.2 of the submittal will be revised as follows L to clarify further the Byron /Braidwood program for qualifi-  !

cation of equipment outside containment exposed to elevated l temperature resulting from HELB's: l 1

3.2.2.2 High Energy Line Break (HELB) ]

High energy line breaks in the auxiliary building have been identified and analyzed in accordance with Standard Review Plan, Section 3.6. High energy lines are defined as pipes in which the fluid temperature exceeds 2000 F Q270.2-4

B/B-FSAR or the pressure exceeds 275 psig during normal plant operation. Breaks are postulated in these lines. The resulting temperature, pressure and humidity conditions are included in the environmental qualification program.

The potential for pipe whip and jet impingement effects has been investigated and additional protective features incorporated where required.

Section 3.6 of the FSAR and the response to Question 010.40 describe the approach used to evaluate high energy line break effects. The results of the subcompartment analyses

.are included in Attachment A3.6 of the FSAR. The high temperatures and pressures predicted in the subcompartments are not an equipment qualificat. ion concern because the object of compartmentalization of safety equipment was to ensure that a high energy line failure will not result in additional failures which would violate the plant design basis. The plant is designed such that capability to safely shut down is maintained following an initiating event and the resulting failures, plus an independent single active failure. To verify that this design approach has been successful, the high energy line break conditions are included as accident conditions for the applicable environmental zonec in Table 3.1-1 and equipment in these zones are qualified to the accident conditions if they are required to function in the accident scenario. For environmental zones in which the conditions are not.affected by high energy line breaks, the normal levels are specified as the accident conditions in Table 3.1-1.

The only area identified as experiencing elevated temper-atures and/or pressure beyond normal or abnormal conditions following a high energy line break are, with one exception, subcompartments. The subcompartments have been designed such that failure of a high energy line in the subcompart-ment will not result in failures beyond the single train of a safety system which is in the subcompartment. As an example, Zone A13g, the centrifugal charging pump rooms, are predicted to reach a temperature of 1900 F in the event of a ireak in the charging pump discharge lines.

The resulting harsh environment affects only equipment in the failed charging train. All safety equipment used to mitigate the break is unaffected by the harsh environment.

Back flow f rom system is. prevented by redundant check valves. The loss of charging function can be mitigated by use of the redundant centrifugal charging pump, the positive displacement charging pump, isolation of the letdown system, or either of the safety injection pumps, any of which could be assumed to f ail without af fecting plant safety. None of these options for mitigation are Q270.2-5

o ,

B/B-FSAR 1

i 4

.af fected ' bye the . harsh environment in the ' charging pump room. As a result, the plant design basis is valid in spitecofl harsh environmentsi caused by high' energy line breaks and no equipment must be qualified for the harsh environments which result from high energy line breaks in auxiliary building subcompartments.

I

-The only area other than the subcompartments which could 3

experience an elevated temperature is the upper area of

-Zone A13, the containment piping penetration area. A break in a three-inch letdown line in-the-chemical and ,

volume-control system could release steam into this area i which has no natural ventilation. The temperature would

then increase above the environment specified for this area. The only safety-related items in this area are j isolation valves'on the safety injection and essential  ;

, service water systems which are not required to function in1this~ accident and are redundant, and an isolation valve on the failed line which will fail.in the closed position.

The-break flow is-limited to 120 gallons per minute by

. orifices. Immediate indications of the break will be

supplied by two main control board alarms (high flow and high letdown heat exchanger outlet temperature). The plant can then be safely shut down without tha equipment L

4 which would be affected by the increased temperature.

• =

Q270.2-6  ;

l

B/B-FSAR QUESTION 270.3 "In your response to Question 040.2, various items of equipment were described as being submerged during an accident. Provide equipment identification numbers so that this equipment can be correlated with your most recent submittal."

RESPONSE

The . response to FSAR Question 040.2 lists the following. equipment as being located in the containment building and possibly subject to a flood condition in the event of a LOCA accident:

1/2PL50J 1/2PL77JA 1/2PL52J 1/2PL77JB 1/2PL57J 1/2PL77JC 1/2PL66J 1/2PL79JA i 1/2PL67J 1/2PL79JB 1/2PL79JC The Class lE instruments mounted on these local control panels and required to perform a C. lass lE function during normal operation and/or during or m_ter an accident are as follows:

1/2Ur459 1/2Pr403 1/2LT 460 1/2PT 405 1/2LT 461 1/2Pr455 1/2LT 501 1/2PT 456 1/2Ur502 1/2Pr457 1/2LT 503 1/2Pr458 1/2Ur504 1/2Ur527 1/2Ur537 These . instruments have been relocated to an elevation not subject j

to submergence. Therefore, submergence qualification for these instruments is not required.

(

The response to Question 040.2 will be revised to be consistent with the above in the next amendment.

d Q270.3-1

B/B-FSAR QUESTION 270.4 "For equipment qualified with less than a one hour margin, discuss the approach utilized to justify this deviation and provide one example for a specific item of equipment."

RESPONSE

r Section 3.8.2 of the EEQR will be revised'as'follows to resolve the question of_ demonstrating a margin of one hour in operating time. _The , data necessary to support a margin of at least one" hour will-be'part of the equipment qualification file.

F 3.8.2 Margin Included in Ooerating Time The operability requirements for each piece of Class 1E equipment is the lengthlof time the equipment is required to remain functional during accident mitigation. A margin of at least one hour of the equipment operating time has been included in the qualification program for each piece of applicable Class lE equipment.

Some equipment, e.g., transmitters, was not specified to maintain trip-function accuracy requirements for longer than five minutes post-accident. However, peak HELB temperatures will be reached

~

within the specified operability time. The operability time was conservatively established based on the reactor trip engi-

.neered safeguards function performed'by'each equipment item considering what consequences failure of the device would have on the operator and the mitigation of the event. Margins for trip function requirements are contained in the HELB enve-1 lopes which encompass a full spectrum of break sizes and are-also justified by the fact that the signal generated by the sensor is " locked-in" by the protection system and will not reset should the sensor fail after the designated trip time requirement. Most of the equipment was also specified and qualified for much longer post-accident monitoring function times to slightly reduce accuracy requirements.

4

• i Q270.4-1

B/B-FSAR QUESTION 270.5

'" Describe your conformance with the test sequence defined in IEEE 323-1974, Section 6.3.2 for harsh environment equipment."

RESPONSE

, For equipment required to operate in a harsh environment the preferred test sequence is that recommended by IEEE-323-1974.

' When a more severe sequence is identified, the more severe

- sequence is used in 1ieu of the sequence in"IEEE-323-1974.

- Test sequences are described and justified in Qualification Data Packages, which will be available for review at the audit.

Subsection 3.11.2 of the FSAR will be revised in the next amendment to delete the reference to IEEE 323-1971 and replace it with IEEE 323-1974.

e Og E Q270.5-1

B/B-FSAR QUESTION 270.6

" Describe in general terms the program to be utilized for detecting age-related degradation in equipment, including that caused by synergistic and. low dose rate effects.

The methods for determining the items to be inspected or parameters to be measured-and the frequency of exam-ination should be discussed. Provide specific information on your approach for cables located inside containment."

(

RESPONSE Equipment located in harsh environments is qualified to address potential age related degradation. Accelerated aging techniques (Arrhenius principle) are used to simulate age. Based on this data, components with limited life are then maintained or replaced through an Equipment Qualification Maintenance and Surveillance Program to be implemented at Byron /Braidwood stations. Data for this program is derived from Qualification Data Packages and mnaufacturer's recommendations. Additional existing maintenance programs will supplement this program.

I

1. Technical Specification requirements will verify through performance tests that equipment is functional (will be included in Chapter 16 of the FSAR) .

2. Vibration monitoring will be used to do comparative testing against established baselines on rotating equipment.

3. Lubrication Program I

4. Instrument Calibration / Surveillance Program

5. Inservice Inspection Programs on pumps, valves and welds per Section XI of ASME Boiler and Fressure Vessel l

Code.

l A history / trending program will be applied to the maintenance l programs to detect changes in operability.

Known low dose rate effects and synergisms are included in the environmental qualification program.

Q270.6-1

I-

• B/B-FSAR t

QUESTION 270.7 "No explanation of accelerated aging methodology is presented in your-program. Provide a description of the procedures utilized."

RESPONSE

' The ef fects of aging .on the Class lE equipment was addressed

= for all pieces of equipment identified in the'E3QR. . Ag:.ng

, was-addressed'by either performing accelerated aging on the equipment or by developing an aging analysis program to evaluate the : stresses impose'd on the equipment which degrade pr.rformance.

The : objective of an aging analysis is to determine a qualified lifeLof the equipment. -An examination is performed-to determine i which of the. materials is susceptible to aging by either heat.

(thermal) , radiation,- or both heat and radiation, and then determine the: qualified life for the most susceptible material.-

~i- Arrhenius techniques were utilized in the determination of qualified life. The qualified life for ' the most . susceptible

- material / component would be'used to establish a periodic replace-

. ment schedule if .the qualified life is lens than 40 years.

Details of specific equipment are available in the Qualification

[ Data Packages.which are available for review at the audit.

j .

2

- For.NSSS Class-lE equipment the aging evaluation program is described in Appendix B to WCAP-8587. Accelerated Thermal -

i aging parameters are described in Appendix D to WCAP-8587.

Both documents are available for review.

! ~

1 1

t I

i 1

~

Q270.7-1 S

- . . . ._-.__..--m., . - - _ - _ _ , _ _ _ _ , _ - - _ , _ . . - . _ . . - . _ _ - _ , - _ - - . . , - _ ,_, .-- - _... ,_- -

B/B-FSAR QUESTION 270.8 "The following discrepancies or omissions were noted in the equipment qualification data sheets:

a. Operating times are insufficient as listed. The specific post accident operating time requirement and qualified operating time should be listed,

b. Some Limitorque valve operators and other equipment do not have model numbers listed.

c. Accuracy requirements are omitted for some equipment.

d. Five of the NSSS items considered to be fully qualified are still undergoing testing or analycis."

RESPONSE

a. Operating time requirements are incorporated in notes explaining the qualification operating time for applicable pieces of equipment. Data sheets can be revised, if necessary, to reference the required operating time under the Plant Design column as well as the qualification column.

Data sheets showing specified and/or qualified operating time as " continuous" means that the equipment must remain functional at all times during normal, accident, and post-accident conditions. The actual length of time that the equipment is demonstrated to be qualified for during post-accident conditions is clarified by specific notes on the data sheets.

b. Data sheets for Limitorque motor operators have been revised to show model numbers instead of serial numbers.

Model numbers are omitted on certain data sheets because some equipment manufacturers do not assign model numbers to equipment such as panels, junction boxes, large motors, electrical penetrations, etc. In such cases, the information given under the model number entry is whatever identifiable information is available.

c. Accuracy requirements, where applicable, have been provided in the remarks under the Qualification Accuracy column.

Q270.8-1

B/B-FSAR

d. The following portions of EEQR Subsection 4.2.2.1 will be revised as follows to properly summarize the qualification of the equipment:

4.2.2.1 Qualification . Status A. ' Qualified

1. EQRE HE-1 ---Safety-Related Valve Electric Motor Operators

Qualification Group A

...The Westinghouse generic qualification program

'for valve electric. motor operators is still in progress. Operators that will be qualified to the'"new" report have been ordered and will be available for use on Byron as required.

2. EQRE HE Safety-Related Solenoid Valves:

Qualification Grouo A Evaluation of the test reports is based on data for solenoid valves to be installed in the Byron /

Braidwood Stations. The solenoid valves currently installed will be replaced with models identical to the ones qualified under the Westinghouse generic program prior to= fuel. loading.

3. EQRE HE Safety-Related Externally Mounted Limit Switches: Qualification Group A It is anticipated that seals as required by the test for the~ electrical connections will not be required for the DBE environmental conditions at Byron Station. A test to verify the capability-of the limit switches to survive the DBE environment without seals will be performed or seals as required by the test report will be installed with the limit switches. The limit switches located outside the containment that will not be exposed to HELB conditions will not require seals.

4. EQRE HE Safety-Related Valve Electric Motor Operators: Qualification Grouc B The motor operators located outside the containment in harsh radiation environment are to be qualified by~this'EQRE. Radiation levels specific to the actuator location in Zone A13C have been calculated 0270.8-2

B/B-FSAR to r9 duce the specified radiation 6 ev s from 1X10 rads to a maximum of 3.7X10 rads (including 10% margin) which is enveloped by the test level of 4X10 rads.

5. EQRE ESE Pressure Transmitters: Qualification Group A

-Evaluation of the test reports is based on data for the transmitters to be installed. The trans-

'mitters currently installed will be replaced.with models identical-to the ones qualified-under the Westinghouse generic program which have already been shipped to the Byron Station.

6. EQRE ESE Dif f erential Pressure Transmitters:

Qualification Grouc A Evaluation of the test reports is based on data for the transmitters to be installed. The trans-mitters currently installed will be replaced with models identical to the ones qualified under the Westinghouse generic program which have already been shipped to the Byron Station.

9 i

Q270.8-3 r,,- ns-,,e - -r-c-- ,, ,w+ - - r --e- , , - - - e ,.. ,- - - ---

B/B-FSAR .

s QUESTION 270.9 "The description of the qualification program for safety-related mechanical equipment in Section 3.~11 of the FSAR is not sufficient for demonstrating compliance with General Design Criteria 1 and 4 of Appendix A and Sections III, KI, and XVII of Appendix B to 10 CFR Part 50. Provide the following additional information for our review:

a. Confirmation that harsh environment mechanical components

.in the-safety-re. lated. systems identified in your J une 17, 1982 submittal for electrical equipment have been included in a design verification program,

b. The criteria utilized for demonstrating qualification of the equipment and their bases,

c. The curr ant status of this equipment with respect to the toove requirements. If design verification to the eyplicable criteria is complete, so indicate.

If additional review and evaluation are required, described the tasks to be performed and the schedule for their completion."

. RESPONSE a,b. Response to parts a and b have previously been sent to the NRC on January 14, 1983 (See Ceco letter to the NRC, T. R. T r amm to H . R. Denton, dated January 14, 1983).

c. The -program for BOP equipment is about 30% complete and is scheduled to be completed by March, 1985.

j Q270.9-1

^^

l B/B-FSAR l l

QUESTION 270.10

, " Equipment whose design adequacy cannot be confirmed by fuel load must be justified for interim operation until qualification is complete. The information needed to justify interim operation is defined in the proposed rule 50.49 (irsued in January) and is required a minimum of sixty days prior to fuel load."

RESPONSE

Equipment which-is not. fully qualified to Category I of NUREG-0588 by fuel load will be justified for interim oper-ation until qualification is complete.

f I

4 e

Q270.10-1

BYRON /BRAYDWOOD STATIONS - UNITS 1 & 2 ,

SUMMARY

SilEET FOR INSTRUMENT CABLES TYPES OF EQUIPT. SPECIFICATION NO. QUALIFICATION TEST REPORT NO. .

Multi-conductor, paired, F/L-2852 Qualification test of clectric cables under stranded / solid, shicided simulated LOCA/DBE by sequential exposure to

& jacketed instrument environments of radiation, thermal aging, steam cables 600V & below & chemical spray (' Samuel Moore-Dekoron Division)

MANUFACTURER MODEL NOS.

Eaton Corp. - Samuel Moore. Operations' Trade Name - Dekorad Cables Dekoron Division ,

3 ENVIRONMENT SPECIFIED Relative Max. Integ.

Temp.("F) Humidity (%)

• Press (psig)_ , E,xposure (Rads)

Normal 65-122 70 -0.1 to +0.3 Abnormal NA NA -

.NA NA 8

Accident. 320 /10-180 sec. 100 50/0-20 min. 2.0X10 4

270 /5-20 min. Saturated-Up to 1 yr.

1700-155 /1-20 days 155 /120-365 days Ichemical spray: 0.15 GPM/ft 2 of sprayed surface area Alkaline NaOH/ Boric Acid, pH-8.5 to 10.5 .

SUMMARY

S!!CET FOR TNSTRUMENT CADLTLS (CONT' D)_ ,

ENVIRONMENT OUALIFIED Max. Integ.

Relative Press (psig) Exposure (Rads)

Temp ( a P)_ liumidity (%)_

25 MRads Normal 127 105-0 to 3 hrs. 1 7 5 M . '.. d .1 Abnormal 340-0 to 3 hrs. . 100 ,

NOTE:

240-3 to 5 hrs.

atm - 3 to 5 hrs.

- # 105-5 to 8 hrs. Total radiation 340-5 to 8 hrs. dose at LOCA-320-0 to 11 hrs. 75- 8 to llhrs.

55-11 to 15 hrs. - 200 MRads.

300-11 to 15 hrs. 15-15 hrs. to 4 days-250-15 hrs, to 4 days 200-4 to 100 days

• 10-4 to 100 days Chemical spray Rate - 0.15GPM/FT of solution: *

• Start 20 secs, after r 3000 ppm H3B03 .

steam exposure and 0.064 molar Ha2 2S 0e buffered with, continue up to 30 days .

NaOH to a ph of 9-11 OPERABILITY DEMONSTRATED OPERABILITY REQUIREMENT '

Cables mus t be capable of carrying its . The tested cables demonstrated.their capability o) electrical loads / signals with neither loss performing its required function after a simulate of continuity nor loss of shielding from 40 yr. life at the specified normal and abnormal outside interference capability at specified service condition.

normal and abnormal environment for a 40 yr. During the simulated LOCA environment tested life.

cables withstood the simulated environment Additionally, must be capable of withstanding without degradation to its . electrical capability a DDE and 1 yr. Post /DBE without degrading its below specified requirement. .

electrical capability below specified require- ,

ment. '

D g

. 2-

(CONT'D)_ ,

S'UMMAliY S!!EET FOR INSTRUMENT CABLES -

t

' OUALIFICATION MET!!OD Tho qualification of cables was by type test' summarized as follows: the cables (2) represent Test Samples - A total of 12 samples, of which two These two were; 16 gauge; supplied at nyron/Braidwood Stations.

7 strand tinned copper conductors with 20 mil EPDM cxtruded primary insulation; 10 mil llypalon extruded conductor jacket;0.85 mil aluminum conductors twisted together; 7 strand tinned copper drain wiro; (25% overlap) and 16 gauge, Total icngth of each is 45 mil llypalon extruded outer jacket. h 35 feet.The cables supplied to Dyron and Braidwood facilities have muc higher capabilities than the test specimen as they are ilmade NOTII: Alum with / 3'.0 mil Alum / mylar tape shield as compared to the samples 0.85 m '

mylar tape shicid. e Thermal Aging - Six for 7 dayslifewhich were calculated using Arrh'enius prin of 40 years at an, ambient temperature of 127'F.

an equivalent ,

total of 25.5 MRads, Radiation Aging - which Initial was radiation exposure to the specimen was at a dose rats accomplished Nextby.cxposing irradiation specimen was done for 34.0 hrs. at by exposing a samples to

• of 0.75 MRads/ hour.

a rate of 0.75 MRads/hr. for 234.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which has a total of 175.8 MRads.

for Total integrated dose of radiation exposure from a Co-60 gamma .

source .

the samples is 200 MRads.

Sei_smic - Not Applicable. . e e

. 4

• g g 7 .. .

SUMMARY

SllP.F.T FOR INSTRUMENT CABLES (CONT'D)

LOCA SIMULATION (a) carry connected load at rated Performance / Acceptance Criteria - Test Samples mustvoltage during the ,LOCA exposure duration, (b) withstand the Post LOCA withstand test (40XOD bend), and dielectric withstand test as outlined in IEEE 383 paragraph 2.4.4.

Variables Monitored - Temperature, pressure, relative humidity, voltage, current and ,

After LOCA simulation, voltage -

insulation resistance were monitored.

withstand test at 2X rated voltage and 80 VAC/ mil per IEEE 383 was conducted on the cables.

Tes t Equipment. - Appendix B to Isomedix Test Report lists the instruments used at Isomedix and their corresponding calibration.

Test Results - Values from each test conducted demonstrates the specimen capability to withstand the environment simulated without loss of electrical continuity and insul,ation capability.

FLAME TEST ' Test cables passed the fire testing'of.IEEE 383 Sects. 2.5.4 & 2.5.6.

Therefore, cables will not propagate fire'.

e

• e  %

• e

1. e

• e e

O 4

• I .

1

• e

• es e

4 .

SUMhnny S!!EET FOR INSTRUMENT CABLES (CONT'D) , ,

QUALIFIED LIFE - Cables have an equivalent life of 40 years of continuous operation ' -

at 127 0F Ambient.

TEMPERATURE - The subject cables are qualified for ambient temperature of 127'F.

~

RELATIVE HUMIDITY - The cables are qualified for 100% RH. . . ,

the total radiation dose the cables RADIATION -withstoodPer Radiation was Certification 201.3 MRads. from Isomedix,Thus, it is qualified for 200 MRads at a dose rate of .

0.75 Rads /hr.

PRESSURE - Specimens were subjected to a pressure environment that envelopes specified.

Therefore, subject cables are qualified to.specified pressures.

Temperature, pressure, relative humidity and chemical spray specified were all LOCA - Total LOCA duration was.100 days as enveloped by the simulated environment.

against specified 365 days. The above shorter duration was justified by a more severe condition (2-340 F, 100 psig peak). Calculations using the Arrhenius ,

principle applied to the consumed life ratio presented in the report justified -

the 300 days LOCA duration.

MARGIN - The LOCA simulation had; +15"F margin over specified, 110% on. pressure, 278.6% .

Post LOCA exposure. -

that the testing CONCLUSION - Based on the above simulated test, we therefore conclude. -

demonstrates a 40 year qualified life for inside containment at the Byrogr Braidwood Nuclear Stations.

e O

, e

-5 ,

• BYRON /BRAIDWOOD STATIONS - UNITS 1 & 2 ,.

h:

~ ,

SUMMAltY SilEET FOR OKONITE CABLES PROCUREMENT

• SPECIFICATION SAFETY CATECORY LOCATION TYPE OF EQUIPMENT 600V Cable F/1.-2823 Will experience dcaign basis accident and must function to mitigata dux111ary is1Jg 4

said accident Containment ttANUFACTURER MODEL NO. QUALIFICATION REPORT NO.

f Okonite Okonite Okonice Report No. NQRN-1A EPR Insulated /

Cable s

~ ENVIRON!!ENT ENVIRONMENT QUALIFIED

- Maximum Maximum Relative Integrated Relative Integrated Tep(*F) llumidity(%) Press (psig) Exposure (rac)

Temp (*F) llumidity(%) Press (psig) Exposure (rad)

Numal 122 20-50 , -0.1 to 0.3 1.1 x 107 . Normal 122 100% Atm -

Abnormal NA NA NA NA 8

Accident Accident 345-3hra Saturated 112-3hre 2 s: 10 320 /10-180 secs. 100 50/0-20 min. 345-3 hrs Steam 112-3 hrs ,

270 /3-20 min. . saturated / '

335-3 hrs 95-3 hrs -

170 -155 /1-20 days 20 min-lyr. 315-4 hrs 69-4 hrs 155 /120,365 days 265-3 days, 24-3 days, 9 hrs 9hre 212-126 0-126 days '

days l Cl c.mical : Alkaline NA0ll/ Boric Acid Chemical: .28 molar H 3B03 Spray Spray Rate of .15 GPM/Ft.E pil-8.5 to 10.5 of surface Spray .064 molar Na2S0 23 ..

area sprayed. Na0li approx. .59% to make pH of 10.5 6 77'F Spray Rate - 0.5GPM/Ft.2 of sprayed surface. , l OPI itAh!!.ITY REQUIRI. MENT OPERABILITY DEMONSTRATED l

C..bleu are required to carry their rated elce.rical The qualification report demonstrates . hat che cables loads under normal service conditions without loss will perform their intended functions under r.armal and of electrical continuity over a 40 year life plus accident conditions.

accidcat and 1 year post LOCA environment. -

1

SUMMARY

SilEET FOR OKONITE CABLES (CONT'D) -

s QUALIFICATION METl!OD: ,

I' Qualification vau by type testing oummarized as follows:

l Test sample - 1/c unjacketed 600V, #12, 7x, tinned copper, .030" extruded Okonite (dPR) insulation.

Two 25 ft sampics: one unaged, one thermally aged.

Thermal Aging - The cable sample was thermally aged for 3 weeks at 150* c;in order to simulate 40 yrs. . ,

j at 90oc (50 C nmbient with 40 C rise). -

1 /

1 l

j Radiation Aginf, - Cable samples were exposed to a minianim of 201 megarade of cobalt 60 radiation at a rate of i .67 to .75 negarada per hour for 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />. ..

Seismic - Not applicable i l Performance / Acceptance Criteria - Sample must a) maintain electrical load through entire LOCA profile, j b) withstand the 30 day and 130 day post-LOCA voltage withstand test ((40x0D) bend, 80V.m11 se) in accordance with ,

4 IEEE 283-1974.

i yariabics Monitored - During testing: temperature, pressure, voltage and current are monitored. In addition, i capacitance, % PF and IR tests were performed prior to and after testing. A dielectric test was also performed 1 af ter 130 day post LOCA simulation. -

i

• l Test Equipment - The instruments used during testing are listed in Appendix Five of the report. Calibration records are available at manufacturer for- audit upon request.

Test Results - Values for the variables monitored during testing demonstrate that the cable samples maintained -

t i

their cicctrical continuity throughout the required testing.

Anomalies -

1. The accident environment to which the cables were subjected, envelope the required temperature ramp 4 for the first 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />. The specified environment requires an additional 1 year at 150*F. The sampics were subjected to 3 days 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> at 265' and 126 days at 212'F following the initial ,

15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. Based on the Arrhenius technique, the 126 days at 212*F is equivalent to 10 years at -

150*F. .

- i Maintenance & Surveillance - None required , , ,

Flame Test - Samples were subjected to and passed the flame test in accordance with 1EEE 383-1974. , ,

i i  !

i

SUMMARY

SilEET FOR OKONITE CAllLES (CONT'D) ,

CONCI.USloN:

Time - The ouhject cabics are qualified for 40 years of continuous operation. ,

Tenferature - The subject cabics are qualified for a service temperature of 90*c (59'c ' ambient; 40*c riae)

Radiation - The subject cables were exposed to 201-225 megaradeappliegin300hoursatarateof.67to'.75 megarads/ hour. Cabics are considered qualified for the specified 2x10 rada. -

Pressure - The cabics were subjected to a pressure environment which envelopes the specified pressures and a::e considered qualified for same.

IInmidity, - Through exposure to saturated steam during LOCA testing, the cables are considered qualified for the rcquired aaximum 100% relative humidity.

Ltl CA - The test profile envelopes the requirements of time, temperature, humidity and pressure. The cables are qualified for the peak temperatures of 325'F with 6.1% margin and peak pressure of 50 psig with 100%

margin. The cables are qualified for the post LOCA exposure of 150*F for one year with 1000% margin on time or 28% margin on tempe,rature.

We conclude, based on the above, 'that the testing performed demonstrates a 40 year qualif.ied lafe for inside containment at the Hyron/Braidwood Nuclear Stations.

SU? DIARY S!!EET FOR 600V POW.R & CONTROL CABLE .

.o f (ROCKilESTOS FIREWALL SIS) .

PROCURE!!ENT TYPE OF EQUIPMENT SPECIFICATION SAJETY CATECORY LOCATION 600V Switchboard wire F/L-2823 Will experience design basis accident and Containment / Auxiliary must function to mitigate said accident Building

}!A!;UFACTURER MODET. NO.

~ LUALIFICATION REPORT OtocWelitos)

Rockbestos Firewall SIS Qualification of Firewall III Class 1E Electric Cables ENVIRONMENT SPECIFIED ENVIRONMENT QUALIFIED ,

Maximum Maximum i Integrates

~

Relative Integrated . Relative Temp ( F) Humidity (%) Press (psig)_ Eicpos. (rad) Temp ( F) Humidity (%) Press (psia) Expos.(r0 Normal 122 20-50 -0.1 to 0.3 Normal 122 100 AIM -

Abnormal NA NA , NA NA Accident: .

Accident: . .

g 346(2)3hr. 113 2x10 .

320*/10-180 secs. 1LJ 270 /3-20 mins. 50/0-20 min. 335+315+265 170 -155 /1-20 days saturated /20imin-lyr. 3hr 4hr 3 days 93-69-28 155 /120-365 days 9 hrs 212-26 days 0 200-100 days 100 (Sec Anomaly #1)

Chemical AlkalineNA0!!/DoricAcipph8.5-10.5 Chemien1 Boric Acid Spray, 9-11ph at rate of .15 Spray

• At a rate of 0.15 GPM/f t Spray gal / min /ft 2 OPERABILITY REQUIRE!!ENT OPERABILITY DEMONSTRATED Cables are required to carry their rated cicctrical The qualification report demonstrates thct the' cables.will loads under normal service conditions without loss perform their intended functions under narms1 and accident of c1ccerical continuity over a 40 year life plus conditions.

accident and one year post LOCA environment. .

,i

SUMMARY

SHEET FOR 600V POWER & CONTROL CABLE (CONT'D)

) ACCURACY / RESP 0NSE TIME REQUIRFgi yE _

ACCURAC'f/RESPCSSE TIME LE40NSTRATED ,

N/A N/A QUALIFICATION METHOD:

Qualification vas'by type-testing summarized as fol' lows:.

Test Sample - 1/c #12, 600V, 30 mils of Flame Retardant XTEE Insulatio'n. . Three samples of 10 f t, Thermal Aging - The cables were thermally aged for 1300 houra at 150*C to simulata 40 years life at 90' -

4 (50 C Ambient, 40 C rise). .

Radiation Aging - Cable sampics were exposed to a minimum of 201 megarads of cobalt 60 radiation at a rate of .65/.78 megarads per hour for 78/188 hours. .

Seismic - Not Applicable Performance /Accepynce CriterM - Sample must a) maintain electrica? load through eritire LOCA profile, b)withstadd the 30 day and 100 day post-LOCA voltage withstand test ((40XOD) bend, 80V/ mil ac) in accordance with IEEE 383-1974.

Variables Monitored - During testing: temperature, pressure, voltage and current are monitored.

Test Equipment - The instruments used during testing are listed in Appendix VII attached to Rockbestos letter dated April 29, 1982. Calibration records.are available for audit on request at Rockbestos Company at Newhaven, Connecticut.

Test Results - Values for the variables monitored during testing demonstrate that the cable samples maintained their electrical continuity throughout.the required testing.

. Anomalies - ,

1. Ihe accident environment to which the cables were subjected, envelbpe the required temperature ramps ,

i for the first 30 days. The specified environment requires an additiccul 6ne year at 150 F. The sampics were subjected to 100 days at 200 F following the initial 30 efays. Based on the Arrenius -

technique, the 100 days at 200 F is equivalent to eight years at 150*F. Therefore, it is qualified at the specified temperature and time. ,,

l ,

Maintenance & Surveillance - None Required _

SUMMARY

SilE'dT FOR 600V POWER & CONTROL CABLE (CONT'D) ,

CONCLUSION Time - The subject cables are qualified for 40 years of continuous operation.

Temperature - The subject cables are qualified for a service temperature of 90 C (50*C ambient, 400 C rise)

Radiation - The subject cables were exposed to 201 megarads applied at a rate of .65 to .78 megarads/ hour.,

Cables are considered qualified for the specified 2x10 8 rads Pressure - The cables were subjected to a pressure environment which envelopes the specified pressures and are considered qualified for same.

Munidity - Through Post-LOCA exposure of 100% Relative Humidity at 200 F for 100 day.., the cables are considered qualified for the required maximum 100% Relative Humidity.

LOCA - The test profile envelopes the requirements of time, temperature, humidity and pressure. The cables are qualified for the peak temperatures of 325 F with 6.1% margin and peak pressure of 50 psig with 100% margin. The cables are qualified for the post LOCA exposure of 150 F for one year with 800% margin on time and 19% margin on temperature. ' '

k'e conclude, based on the above, that the testing performed demop,etrates a 40 year qualified life for inside containment at the Byron /Braidwood Nuclear Stations. ,

4

> b BYRON /BRAIDWOOD - UNITS.1 & 2 .

SUIDIARY SilEET FOR TERMINAL BLOCYS e

MANUFACTURER MODEL NO. -

-QUALIFICATION REPORT

\

Marathon Series 6000 & 1600 Wyle Report No. 45011-1 ENVIRONMENT SPECIFIED Relative Maximum' Temp. ( F) Humidity (%) Press (psig)_ Expos.(rads) -

^

2x10 0 Normal 122 ,

20-50 -0.1 to 0.3 Abnormal: NA NA . NA NA Accident: 3200/10-180 sces. 100 50/0-20 min. '

270 /5-20 min. saturated /

• 170 -155 /1-20 days 20 min /1 yr.

155 /120-365 days ,

Chemical Spray: Spray rate of .15 GPM/ft2 of sprayed surface Alkaline Na0ll/ Boric Acid pH between 8.5 & 10.5 e . *

• o .

~ ~

a sum!ARY SilEET FOR TERMINAL BLOCKS (CONT'D)

ENVIRONMENT QUALIFIED Relative Max. Integ.

Temp.(F ) Humidity (%) Press (pnig) . . Exposure (RADS)

Normal 122 0-90 -0.1-0.3 Included in Accident i

Abnormal 345-1st 3hr. 50-1st 3hr 345-2nd 3hr. . 50-2nd 3hr.

325-next 3hr. 45-next 6hr 250-fo110 wing 3hr. ~

25-fo110 wing 3.Shr -

325-last 27hr. 20-last 27hr. , ,

Chemical Spray: -

Alkaline Na0il/ Boric Acid pH between 8.5 and 10.5.

Due to facility limitation, which has -

• been concurred, the initial apray rate for the 1st and 2nd -

approximately0.04gpm/ft}hrwas  ; et the 6hr point was increased to 0.5 gpm/ft2, ,

OPERABILITY REQUIREMEE OPERABILITY DEMONSTRATED Terminal blocks should be capable of performing its The test report, demonstrates . the terminal blocks (Series required function which is to conduct, insulate, 1600 and 6000 type 6012 DJ, 600V, 75A, fixcd barrier, isolate and terminate cicctricity under normal 12-point) capacity to perform its required function up to nuclear plant service conditions over a 40 year 40 years at normal nuclear plant environment and withstand life and in addition, be abic to withstand a a DBA and DBE/OBE without degrading its integrity to perform DBA and DBE/OBE without degrading its required requirement.

function.

e

SUMMARY

SHEET FOR TERMINAL BLOCKS (CONT'D) .

! QUALIFICATION METHODS

! Performance / Acceptance Criteria: Specimen must: (a) Maintain a safe insulation resistance value throughout -

the entire LOCA profile, (b) pass the post LOCA continuity and insulation resistance test, (c) pass the post radiation and thermal aging insulation resistance and continuity tests.

Variables Monitored: Temperature, pressure, continuity and leakage current were monitored during the tent.

Insulatica resistances were also measured af ter thermal aging, radiation exposure, scismic and LOCA simulation.

Test Results: Visual inspections and values obtained during and af ter each test demonstrates the specimens capability and endurance to the simulated environments.

Maintenance & Surveillance: Non Required Thermal Aging: Specimens have been thermally aged at an accelerated aging temperature of 120 C for 932 hours0.0108 days <br />0.259 hours <br />0.00154 weeks <br />3.54626e-4 months <br />.

Pressure: Test specimens were subjected to pressure environment that enveloped specification.

Humidity: On the qualification plan, it vcs justified that Relative Humidity has negligible affect on

. the specimen aging. In LOCA simulation however, the introduction of chemical spray lato the .

chamber demonstrated that the terminal blocks are qualified for 100% RH. .

SUMMARY

SilEET FOR TERMINAL BLOCKS (CONT'D) ,

aa CONCLUSION Qualified Life: Specimen having been subjected to an accelerated aging temperature at 120 C for 932 hours0.0108 days <br />0.259 hours <br />0.00154 weeks <br />3.54626e-4 months <br /> simulated 88 years of life at 122 F.

Temperature: The subject terminal blocks are qualified to an ambient temperature of 122 F.

Relative Humidity: Introduction of spray during LOCA, demonstrated that the terminal blocks are qualified for 100% Ril. ,

Radiation: Specimen having been irradiated to 200 H Rads is therefore qualified at that radiation dose.

LOCA: Specified LOCA temperature, pressure, Ril and chemical spray were all enveloped by the simulated.

Total duration was however 'less (27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />). The shorter duration was justified by a more severe

! environmental condition ,(2 peaks at 345 F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> each and 50 psig). Calculation using Arrhenius

' principle,ucing lowest activation energy of all of the components of the specimen has resulted in an equivalent duration of three years at the specified 325 F.,

Based on the above, we *therefore conclude that the te' sting performdd demonstrated a 40-year qualified life for inside containment at the Byron /Braidwood Nuclear Station.

I e

o BYRON /DRAIDWOOD - UNITS 1 & 2 <

SUMMARY

S!!EET FOR INSTRUMENT CABLE SPLICES ,

i TYPE OF EOUIPMMENT, ,

SPECIFICATION NO. QUALIFICATION REPORT NO. }

Raychem Final Report F-C4033-3 Nuclear penetration splice F/L-2852 kit made of Raychem's WCSF l materials

• MANUFACTURER MODEL NOS.

Raychem Corporation WCSF-ll5-6-N and WCSF-200-6N ~(Test specimen)

WCSF-070-N, WCSF-650-N) Byron /Braidwood supplied

' WCSF-300-N, WCSF-200-N) splice kits (rated lkv)

ENVIRONMENT SPECIFIED Relative - . Max. Integ.

Temp.(F )' U Humitidy(%) Press (psig) Exposure (Rads )-

Normal 65-122 . '70 0. -0:1 to +0.3 NA NA NA ,

Abnormal NA 0

0 50/0-20 min. 2 x 10 w/ max 6 Accident 320 /10-80 sec. 270 /5-2n min. saturated / rate 1.6 x 10 /hr.

170 -155 /1-20 days 20 min-lyr.

155 /120-365 days Chemical Spray:

Spray rate - 0.15 GPM ft. of sprayed surface area e

e e e 9

~

o a. o ,-

SUMMARY

SHEET FOR INSTRUMENT CABLE SPLICES (CONT ' D) 9 ENVIRONMENT QUALIPIED_

Relative Max. Integ.

Temp. ('F) Humidity % Press (psig) Exposure (Rads)

ormal 194 5 x 107 1.5 x 10 8 ibnormal 351-1st 10 hr. ' Sat. Steam 70-1st 10 hr.

275-next 4.5 days 31-next 4.5 days 212-next 26 days 10-next 26 days Note: 203.75 MRads. TID per radiation certification Chemical Spray:

Spray rate - 0.15 GPM/Ft.2 of spray area Duration - simultaneous with steam and irradiation up to 30 days.

Composition - solution of: 3,000 ppm of H BO3 0.064 molar Na Sf03 adjusted with NaOH to pH of .5 to 11.0.

OPERABILITY REQUIREMENT OPERABILITY DEMONSTRATED Splice kit material must be compatible with cable The results of the test conducted by the cable man-insulation within the cables specified normal and ufacturer (Samuel Moore-Eaton), which include abnormal environment up to the normal plant life thermal aging, irradiation and LOCA simulation of 40 years. Additionally, splice kit materials using the cable's qualification test parameters, ,

must be capable of withstanding a DBE and up to a demonstrates compatibility between splice kit year of post DBE environment without degradation material and cable. Qualification test conducted of its integrity to perform its required function. by splice kits manufacturer (Raychem), using specimens of different types of splices, sizes and cable products, but with splice kit material (WCSF) the same as that supplied to Byron /Braidwood demon-strates the splice kit material capability to with-stand harsher simulated accident environment.

The splice kits manufacturer's test also simulated the equivalent to a 40 year normal nuclear power plant life plus a LOCA and a post-LOCA of one year duration.

J .0 - O O '

O '

bOMMARYSIICETFORINSTRUMUNTCADLESPLICES (COMT' D) ,

OUALIFICATION METilOD Both compatibility demonstration and qualification test of splice kits were by type test summarized as follows:

Test Samples: Compatibility test samples consist of 2/C #16 7 strand tinned copper conductors with a splice on one (1) conductor (black) overlapped by a splice on the jacket; another sample was 2/C #16 tinned, 7 strand copper conductors with both conductor spliced. Raychem's WCSF-070-N sleeves were slipped over the reconnected conductors and WCSF-200-N sleeving was slipped onto the cable over splice area.

Qualification tent by Raychem had six (6) specimen of different size cables using different type connectors. On all of the specimen Raychem's Thermofit WCSF splice kits were used.

Thermal Aging: Qualification test commenced with combined thermal (150t) and radiation aging for seven (7) days. With 50% retention of original clongation as end of the life, an Arrhenius plot was constructed on which the accelerated aging temperature of 150*C has an equivalent life of 40 years @ 90'C. The heat of activation for the thermal oxidation of WCSF compound was calculated to be 29K cal / mole.

Cable ganufacturer compatibility test also thermally aged their policed specimen at 121 C for 7 days. Using Arrhenius equation tl.is is calculated.to be equivalent

~

to 40 years at 127 F.

Radiation Aging: Qualifjcationtesthadradiationagingsimultaneouswiththermal. Total exposure 5 x 10 Rads - Compatibility test had their samples irradiated prior to thermal aging. Total accumulated done was 25 MRads gamma.

Seismic Not Applicable

~

LOCA SIMULATIONS Performance / Acceptance Criteria: Compatibility test.- Test samples must show that no detrimental _

affect would occur between the cables and the splice materials during and after the simulated environments.

Qualification Test: Test samples must (a) Maintain electrical load during'the simulation, (b) pass insulation resistance test, (c) pass high-potential test after performed bending

+

requirement and (d) pass both (b)E (c) tests

. '* e r',

(*} {~)

BUMF1ARY S!!EET FOR INSTRUMENT CABLE SPLICES (CONT'D) ,

Variables Monitored: Both test (compatibility and qualification) monitored similarly temperature pressure, relative humidity, current and insulation resistance. Both tests also had their test samples energized and loaded during test. At the con-clusion of their respective test, both also had their samples subjected to the IEEE 383 standard post LOCA test.

Environment Simulated: Compatibility test - Test sampics underwent an accident radiation dose of 175.8 MRads. prior to exposure to LOCA simulation.

LOCA simulated profile all enveloped specified parametersc Post LOCA duration of 26 days as compared to 1 year specified is, justified by a higher temperature maintained during post LOCA (200*F, 10 psig tested vs.

150*F, 5 psig specified). The tested environment had-an equivalent of 17.71% of qualified life consumed, whereas specified environment has 5.5% of qualified life consumed.

Qualification test - Raychem test samples were exposed to a simultaneous environment of steam, chemical spray and radiation. dose o f 1. 5 x 108 LOCA simulated profile enveloped specified parameters. Post-LOCA duration of 26 days as compared to 1 year specified is, however, justified by a higher temperature maintained during post LOCA simulated -(212'F, 10 psig tested vs. 150*F, 5 psig specified).

Both tests had their respective samples thermally and radiation aged prior to.this test.

Tast Results:' Values from both tests demonstrates that no detrimental effects occur between splice materials and cable insulating materials. The bend (40 x OD) tests results further demonstrate splice capability to withstand simulated. environment without detrimental effect. .' '

9 4

e A

y

< n n vJ.

SUMMARY

SIIEET FOR INSTRUMENT CABLE SPLICES (CONT'D):

=

Time: Compatibility test subjected .the test samples to an accelerated aging temperature of 121*C for 7 days. Using Arrhenius equation, this is equivalent to 40 years at an ambient temperature of 127 F.

Qualification test by Raychem aged their specimen, WCSF compound, at 136*C, 150'C, 162*C and 175 C for 4,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, A570 hours, 521 hours0.00603 days <br />0.145 hours <br />8.614418e-4 weeks <br />1.982405e-4 months <br /> and 194 hours0.00225 days <br />0.0539 hours <br />3.207672e-4 weeks <br />7.3817e-5 months <br /> respectively where retention of original clongation was 50% (considered end of life). Based from Raychem's Report EDR 2001 (8/10/78) calculation, the specimen (WCSF compound) has an equivalent life of 40 years at a service temperature of 90*C.

Temperature: The. subject splice kits are qualified for a service temperature of 90*C. Compatibility test had a qualified. compatibility temperature of l27'F ambient.

f Relative ilumidity: Splicekitswereexposedtoasaturatedsteamenvgronment, thus it is qualified at specified 100% Ril .

Radiation: Certification o5 radiation from Isomedix indicated that Raychem's Qualificatio: test samplec had a T.I.D. of between 197.7 to 209.8 MRads. Dose rates were approximately 0.224 MRads/hr. .

Compatibility test samples which were also irradiated at Isomedix had.a T.I.D. i of 200 MRads. Dose rates were 0.75 MRads/llr. for aging and'DBE. Therefore, splice' kits are qualified at specified radiation dose of 200 MRads. ,

Pressure: The splice kits are considered qualified to 70 psig.

Conclusion:

It is concluded that both the qualification and c.ompatibility test demonstrate -

a 40 year qualified life for inside containment at the. Byron /Braidwood Nuclear, ,- ,'

Stations. .

4 l

I t

l j -

.