ML20132D587
| ML20132D587 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/06/1985 |
| From: | Zimmerman S CAROLINA POWER & LIGHT CO. |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NLS-85-322, NUDOCS 8509300221 | |
| Download: ML20132D587 (35) | |
Text
.
CD&L Carolina Power & t.lght Company SEP06 $85 SERIAL: NLS-85-322 Director of Nuclear Reactor Regulation Attention:
Mr. D. B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555-BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-324/ LICENSE NO. DPR-62 JUSTIFICATIONS FOR CONTINUED OPERATION
Dear Mr. Vassallo:
On August 12 through August 16,1985, your staff held an on-site audit of the Brunswick
~
Environmental Qualification Program. At that time, a suggestion was made that Carolina Power & Light Company submit revised justifications for continued operation (JCOs) of Brunswick-2.10CFR50.49(i) provides criteria by which the JCOs are evaluated
_by the Staff. In our current JCOs, these criteria are referenced by paragraph number.
For ease of review, the JCOs have been revised to specifically state the applicable criteria of the rule rather than referring to the paragraph number. Since our June 28, 1985 submittal, two JCOs have been changed. Qualification has been established for the Limitorque motorized valve operator (E51-F019), therefore, JCO No. 2 has been
. deleted. JCO No. 3 has been revised to reflect the receipt of a final qualification report from General Electric. In addition, various editorial changes have been made to the JCOs. Enclo;ure I contains the revised JCOs.
Please refer questions regarding this matter to Mr. Stephen D. Floyd at (919) 846-6901.
Yours very truly,
",;; r w S. R. T nmerman Manager Nuclear Licensing Section MAT /rtj (1895 MAT)
- Enclosure cc:
Mr. W. H. Ruland (NRC-BNP)
Dr. J. Nelson Grace (NRC-RII)
Mr. M. Grotenhuis (NRC)
O 8509300221 850906 ADOCK 0 % g PDR 411 Fayetteville Street
- P. O. Box 1s51
- Raleigh. N. C. 27602 L
..ap o. Ag[Z3 sm.
ENCLOSURE 1 to SERIAL: NLS-85-322 Justification for Continued Operation Brunswick-2
i UNIT 2 BSEP JCO NO. 3 PAGE 1 0F 1 EFFECTIVE 9/6/85 TER NO.:
20 COMPONENT I.D. NO.:
B21-F016, Ell-F022, E51-F007 MFG / MOD. NO.:
LIMITORQUE MODEL SMB-00 VALVE OPERATOR LOCATION:
DRYWELL ELEVATION 17', 80' TECHNICAL DISCUSSION:
Component materials of the Limitoroue Motorized Valve Operators have been identified and qualification documentation located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques.
Results of this evaluation indicate that the Class H motor insulation system, melamine switches, and internal wire insulation materials are insensitive to thermal aging effects at the maximum drywell temperature of 150*F.
The valve operator and motor nonmetallic materials are exposed to the plant postulated accident profile which shows a peak temperature of 298*F.
The valve operators are qualified for 40 years at the normal and accident drywell parameters (
Reference:
Limitorque Test Report No. 600376A).
The motor, with Class H insulation, his been successfully tested to a peak temperature of 340*F (
Reference:
Franklin Report No. FC-3441) which exceeds the postulated plant accident at BSEP. Agditionally,theClassHinsulated motors were successfully tested to 2 x 10 rads gamma total integrated dose
(
Reference:
Limitorque Report No. FC-3327) which envelops the BSEP 8
requirement of 1.1 x 10 Rads gamma.
This analysis makes use of partial test data in support of qualification, thus, it is judged that the Class H insulated motors meet the criteria set
~
forth in 10CFR50.49, paragraph (i)(2).
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 NO. 5 PAGE 1 OF 2 EFFECTIVE 9/6/85 TER No.:
24, 25, 26 COMPONENT I.D. NO.:
CAC-SV-4222 CAC-SV-4223 CAC-V4-S CAC-VS-S MFG / MOD. NO.:
HT8321A5 ASCO
- HT8321A6, 8321A6 The "HT" prefix denotes high temperature coils with Class "H" insulation and are rated for continuous use at 165*F ambient temperature.
LOCATION:
RHR ROOM, CORE SPRAY ROOM, AND REACTOR BUILDING TECHNICAL DISCUSSION:
Component materials of the ASCO solenoid valves have been identified and qualification documentation located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques.
Results of this evaluation indicate that all the nonmetallic materials, except Buna-N, have greater than 660 years expected life at the maximum 104*F temperature.
The Buna-N has an expected life of 11.86 years.
In a letter dated 8-3-79, ASCO stated the following:
"The materials used in the construction of these valves are brass bodies, zinc plate steel bonnets, Buna-N (Nitrile) elastomers, copper shading coils, and all additional internal components are 302, 17.7PH, 305, 416, 430F stainless steel and monel. The valves have Class "H" coils and Nema Type 4 solenoid enclosures.
Based on Engineering judgement, test of similar valves, experience, and rubber manufacturer's literature, the elastomeric components utilized in these valves will function satisfactorily under the accident and post-accident conditions specified in the UE&C Specification. The Class 'H' coils utilized in these valves have been designed for satisfactory operation at 165*F ambient.
l~
UNIT 2 j
BSEP JCO No. 5 PAGE 2 0F 2 EFFECTIVE 9/6/85 Valves of similar design utilizing the said Class 'H' coil system and ethylene propylene elastomers have been satisfactorily qualification tested for use inside containment in accordance with the requirements of IEEE 323-1974, 383-1972, and 344-1975.
Part of this test programs was a thermal aging test.during which the valves were exposed to an ambient temperature of 268'F for 12 days. The valves were continuously energized at nominal voltage and de-energized for 5 minutes every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. At the completion of this test, the valves functioned satisfactorily with no internal or external leakage. The results of this. testing are recorded in ASCO test report AQS21678/TR.
Ethylene propylene was chosen as the elastomer in the valves because they are for use inside containment and it is expected that during an accident the temperature could rise to a maximum of 346*F.
Since the coils passed the 12 day exposure at 268'F, and rubber manufacturer's literature recommends Buna-N for use at 200*F continuous, it is our opinion that this is justification for stating that these valves are capable of satisfactory operation during the accident and post-accident conditions stated in the UE&C Specification".
Although ethylene propylene was the elastomer in the tested valves, the actual service condition of total time above 200*F of less that 3-minutes followed by a rapid drop off to approximately 135*F for these solenoid at Brunswick is such that Buna-N is an acceptable material.
There-is also a Rockwell test report (2972-03-02, Rev. 1; dated 12-1-70) which shows that the HTX8320A20 had successfully functioned during and after exposure to 345" and 110 psig steam for about 2-1/2 hours.
Additionally, a Masoneilan test report (1003, dated 4-19-73) shows that WPHT8300B61 valves successfully functioned during and after exposure to 310*F and 65 psig steam for 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />.
In{ormationonradiationdamagevaluesshowsthat the postulated TID of 1 X 10 will not significantly degrade the function of the nonmetallic materials
-except for the actual disc holder. Testing has been performed on acetal 7
retaining washers to 1 X 10 rads with successful results (
Reference:
MCC Powers Report No. 734-79.002, Rev. 3).
This analysis makes use of partial test data in support of qualification and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 N0. 6 PAGE 1 0F 1 TER NO.:
34, 113, 114, 123 EFFECTIVE 9/3/85 COMPONENT I.D. NO.:
VA-TS-936A, D, F VA-ZS-936A VA-SV-936A MFG / MOD. NO.:
SV - JOHNSON SERVICES V24 TS - JOHNSON SERVICES TS-A19AAC9 ZS - ALLEN BRADLEY 802T-A LOCATION:
RHR ROOM TECHNICAL DISCUSSION:
The operation of the RHR Pump Room Cooling Systems has been reviewed.
In the event of room A fan cooling unit failure, the Qualified room B fan cooling unit will supply the post-LOCA cooling requirements of both RHR pump rooms and the HPCI room simulataneously via interconnecting HVAC ductwork.
Because the safety function in question will be performed by a qualified alternative equipment and because the environmentally caused failure of the A room cooling unit will not degrade any safety function or mislead the operator, this analysis meets the criteria of 10CFR50.49, paragraph (i)(1) and (1)(5).
Therefore, continued operation is justified.
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J
UNIT 2 BSEP JC0 NO. 9 PAGE 1 0F 1 TER N0.:
62 EFFECTIVE 9/3/85 COMPONENT I.D. NO.:
E41-PS-N010 E51-PSL-N006 (No TER)
MFG / MOD. NO.:
STATIC 0 RING PRESSURE SWITCH 6N-AA21X9SVTT AND 6N-AA21-X9-ST LOCATION:
REACTOR BUILDING EL. -17' TECHNICAL DISCUSSION:
Component materials of the Static-0-Ring (SOR) pressure switch have been identified and qualification documentation on a similar SOR pressure switch has been obtained. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques.
Results of this evaluation indicate that the lowest expected life of any nonmetallic material used in the pressure switch is 11.86 years.
The pressure switch nonmetallic materials are exposed to the plant postulated accident temperature peak of 288 F for 70 seconds.
The accident temperature then decreases to 205 F at 100 seconds and returns to snbient af ter approximately 20 minutes.
This postulated peak temperature transient has been compared to accident test data obtained (212 F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) for this switch.
Thodgh the testing does not envelop the postulated peak accident temperature, it is judged that no significant detrimental effects to switch operation should occur as a result of the peak temperature transient.
This assessment is based on the severity of the test performed in comparison to the short duration of the temperature transient (
Reference:
Viking Lab Report No.
30203-2).
Additionally, a radiation analysis was performed to determine the threshold of each nonmetallic material used in the pressure switch.
It was determined that each material has a radiation tgreshold greater than the maximum postulated total integrated dose of 2 X 10 rads gamma.
This analysis makes use of partial test data in support of qualification and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 NO. 10 PAGE 1 0F 1 TER NO.:
67 EFFECTIVE 9/3/85 COMP 0NENT I.D. NO.:
CAC-PT-1257-2
. MFG / MOD. NO.:
BAILEY KQ12C LOCATION:
RHR ROOM TECHNIC AL~ DISCUSSION:
The information provided the operator by these transmitters is also provided by an independent, redundant, and fully qualified transmitter (Rosemount). As such the safety function of this equipment can be accomplished by alternative equi pment.
- Since the safety function is performed by alternative qualified equipment, this. analysis meets the criteria of 10CFR50.49, paragraph (i)(1).
Therefore, continued operation is justified.
1 m
UNIT 2 BSEP JC0 NO. 11 PAGE 1 0F 1 TER NO.:
68 EFFECTIVE 9/3/85 COMPONENT I.D. NO.:
C32-PT-N005A, B MFG / MOD. NO.:
GENERAL ELECTRIC MODEL 551032GKZZ2 PRESSURE TRANSMITTER LOC ATION:
REACTOR BUILDING 50' TECHNIC AL DISCUSSION:
Partial qualification documentation has been obtained for a similar pressure transmitter with the same components and of similar application.
The data was evaluated per the D0R guidelines.
The pressure transmitter measures the RPV pressure and gives the operator information regarding plant performance.
Testing has been successfully conducted to show that the device will not fail catastrophically under elevated temperature and humidity conditions (Re ference: General Electric Document NSE80036). The accident simulation included a peak temperature of 180 F during which time a 6 point calibration functional test was performed. This was estimated to take about 5 minutes.
Additionally, a separate test subjected the transmitter to a 68 F to 158 F at 100% RH test. The tests do not envelop the BSEP requirenent of 200 F for 40-50 seconds and the subsequent ramp down to 150 F in 8 minutes.
However, the accident peak tenperature excursion will not cause significant degradation of equipment operation during that period of exposure above the test maximum tenperature (
Reference:
General Electric Report No. 327, File DV145C3007 and General Electric Document No. NSE80036).
Adgitionally, analysis indicates that the plant radiation requirement of 1 X 10 rads gamma is less than the lowest radiation damage threshold of the transmitter canponents.
This analysis makes use of partial test data in support of qualification and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 NO. 12 PAGE 1 0F 1 TER NO.:
69 EFFECTIVE 9/3/85 COMPONENT I.D. NO.:
E11-PDT-N002A, & B MFG / MOD. N0.:
GENERAL ELECTRIC 552032HKZZ2 PRESSURE TRANSMITTER LOCATION:
REACTOR BUILDING RHR ROOM TECHNIC AL DISCUSSION:
These instruments measure the change in pressure across the RHR heat exchanger and provide a signal to the RHR service water outlet valve to regulate service water pressure so it is always greater than RHR systen pressure. This function can be manually overridden if necessary, and the plant can be safely shutdown in the absence of these devices.
Since the manual override capability provides an alternative means of performing the safety function and since failure of these devices will not degrade the safety function nor mislead the operator, this analysis meets the criteria of 10CFR50.49, paragraph (i)(1) and (i)(5).
Therefore, continued operation is justi fied.
i h
UNIT 2 BSEP JC0 NO. 13 PACE 1 OF 2 EFFECTIVE 9/6/85 TER NO.:
71, 72, 73, 74, 76, 79, 80, 81, & 99 COMPONENT I.D. NO.:
Ell-PS-N016A E41-PSH-N012A E51-PS-N020 B32-PS-N018A Ell-PS-N016B E41-PSH-N012B E51-PSH-N009A B32-PS-N018A-1 Ell-PS-N016C E41-PSH-N012C E51-PSH-N009B B32-PS-N018B Ell-PS-N016D E41-PSH-N012D E51-PSH-N012A SW-TSH-1109 Ell-PS-N020A E41-PSH-N017A E51-PSH-N0128 SW-TSH-ll10 Ell-PS-N020B E41-PSH-N017B E51-PSH-N012C SW-TSH-llli Ell-PS-N020C E51-PSH-N012D SW-TSH-1112 Ell-PS-N020D 1A-PSL-3594,3595*
SW-PS-ll75,ll76*
E41-PSH-N027
BARKSDALE B2T-M12SS D2H-M150SS D2T-M18SS D2T-M150SS P1H-M340SS TC9622-1 T2H-M2515-12 D2T-M80SS LOCATION:
REACTOR BUILDING, RHR ROOM, CORE SPRAY ROOM TECHNICAL DISCUSSION:
Component materials of the Barksdale switches have been identified and qualification documentation located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this evaluation indicate that all materials, except for Buna-N rubber, have greater than 261 years expected life at the maximum reactor building temperature of 104*F.
The switch materials are exposed to the plant postulated accident temperature peak of 288'F for only 70 seconds. The accident temperature then decreases to 145'F within one (1) hour of event initiation. This postulated peak temperature transient has been compared to accident test data obtained (212*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, Ref. AETL TR #596-0398) for those switches. Although the testing does not envelop the postulated peak accident temperature, it is judged that no detrimental effects to switch operation should occur as a result of the peak temperature transient. This is based on the severity of the test performed and the short period of switch exposure to the accident peak temperature.
In addition, the Brunswick switches are located in NEMA 3, 4, 12, or 13 enclosures which protect them from the effects of direct steam impingement / humidity during the postulated accident.
1 m. - - _ _ _ _ _ _ _ -. _ _ _ _ _ _ _ _ _ _ _ _ _ ________________.._____________________-.________________________.--__________________________-.__________;
UNIT 2 BSEP JC0 NO. 13 PAGE 2 0F 2 EFFECTIVE 9/3/85 V
Also, the component nonmetallic materials have been sucessfully radiation aged during qualification testipg (while being used in similar applications) to levels greater'than 1 X 10 rads gansna, the postulated accident TID for BSEP.
This analysis makes use of partial test data in support of qualification and therefore meets the criteria of 10CFR50.49(1)(2).
Therefore, continued operation is justif'ied.
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L-
UNIT 2 BSEP JC0 N0. 18 PAGE 1 0F 1 TER N0.:
93 EFFECTIVE 9/3/85 COMPONENT I.D. NO.:
VA-FT-2577 MFG / MOD. NO.:
BAILEY B013221 LOCATION:
REACTOR BUILDING ELEVATION 50' TECHNICAL DISCUSSION:
Component materials of the Bailey transmitters have been identified and compared to qualification documentation located for transmitters similar in design, construction, and operation. The qualification data has been evaluated per DDR guidelines and by Arrhenius techniques.
Results of this evaluation indicate that these transmitters consist of essentially the same materials and components as Rosemount 1153 transmitters.
The Bailey transmitter includes Teflon and Viton o-rings. These 0-rings are used as static seals between the flange adapter and process flange (Teflon), the process flange and sensor module (Viton), and the electrical housing and cover (Viton). These materials were evaluated at the normal and peak accident conditions and will not experience significant degradation of performance.
The Rosemount transmitters were tested to parameters which envelop the BSEP reactor building conditions (
Reference:
Rosemount Reports 3788, 108025, and 08300040). Based on the similarity of the Bailey transmitters to the Rosemount transmitters, the testing levels, and the environment at this logation (104 F normal, < 200 F for less than 10 minutes peak accident,1 X 10 rads TID) use of the Bailey transmitters is justifed.
This analysis makes use of similarity and partial test data and therefore meets the criteria of 10CFR50.49(1)(2).
Therefore, continued operation is justified.
--a
i UNIT 2 BSEP JC0 N0. 19 PAGE 1 0F 2 EFFECTIVE 9/3/85 TER NO.:
94, 122 COMPONENT I.D. NO.: CAC-PV-1218C CAC-P V-1219B CAC-P V-1219C E41-P V-1218D E41-P V-1219D E41-P V-122 00 E41-PV-1221D and including items identified as various on Rev. 0 which are:
B21-F014A-H B21-F014J-N B21-F014P B21-F014R&S B21-F042AaB B21-F044A&B B21-F046A&B B21-F048A&B B21-F050A&B B21-F056 B32-F0398&C B32-F041C B32-F056EaF B32-F058A&B CAC-PV-12090* CAC-PV-1225C*
E11-F037A-D E11-F043A-D E41-F023A-D E51-F043A-D IA-P V-1201A
CHERRY E2360H LOC ATION:
REACTOR BUILDING 20' AND 50'; RHR ROOM TECHNICAL DISCUSSION:
Component materials of the Cherry switch have been identified and qualification documentation on a switch of similar materials and application has been located. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques.
Results of this evaluation indicated that the nonmetallic components have greater than 66 years expected life at the maximum reactor building temperature of 104 F.
The Cherry switch nonmetallic materials are exposed to the plant postulated accident temperature peak of 295 F for 70 seconds.
The accident temperature then decreases rapidly to 205 F at 100 seconds after accident initiation. The recommended maximum service temperature for continued use of these non-metallic materials exceed the peak temperature of 295*F with the exception of DELRIN.
This plastic material has a 220 F maximun service temperature but has heat deflection temperature of 316F at 66 psi and, therefore, will not be affected during a short exposure to 295 F.
Since CP&L has successfully qualified a switch of similar configuration (Environmental Qualification Test Report for Electrical Devices.
Brunswick Steam Electric Plant, Units 1 & 2, CC&L Project #84-1857) to conditions enveloping the BSEP worst case, it is judged that no detrimental effects to switch operation will occur as a result of the peak tenperature transient.
This assessnent is based on the severity of the test performed and the short exposure time at the postulated accident peak temperature.
E UNIT 2 BSEP JC0 N0.19 l
PAGE 2 0F 2 l
EFFECTIVE 9/3/85 Additionally, radiation testing on switches of the samg material and application supports a qualification level of 3.6 X 10 radsgamma,althougy the testing does not envelop the postulated total integrated dose of 1 X 10 rads gamma, a radiation threshold analysis shows that the radiatjon threshold analysis for each material used in switch is greater that 1 X 10 rads gamma except for the Delrin button. For the Delrin button there is testing to support the use f this material in a mechanical application to a radiation 7
level of 1 X 10 rads gamma (
Reference:
MCC Powers Report No. 734-79.002, Rev. 3).
l This analysis makes use of similarity and partial test data and therefore meets the criteria of 10CFR50.49(i)(2).
I Therefore, continued operation is justified.
l 1
4 UNIT 2 BSEP JC0 NO. 20 PAGE 1 0F 1 TER N0.:
95 EFFECTIVE 9/3/85 COMP 0NENT I.D. NO.:
E41-FT-N008 l
MFG / MOD. NO. :
GENERAL ELECTRIC 50-555111BDAA3PDH FLOW TRANSHITTER LOCATION:
RHR ROOM TECHNICAL DISCUSSION:
l i
This flow transmitter provides control of the HPCI Turbine Control Valve position to maintain design rated HPCI flow.
It also provides the control i
room with an indication of HPCI pump flow.
t Partial qualification test data has been obtained and evaluated for the flow i
transmitter.
Testing has been successfully conducted to show that the device will function under elevated temperature and humidity conditions (
Reference:
l G.E. Document No. NSE80036).
l l
The accident simulation included a peak temperature of 180*F for a time sufficiently long enough to perform a 6 point calibration estimated to take about 5 minutes.
Additionally, a separate test subjected the transmitter to a 68*F to 158'F at 100%RH test.
The tests do not envelop the BSEP requirement l
of 199 F (3" RCIC line break) for 30 minutes and the subsequent ramp down to l
l 150'F in 8 minutes.
However, the accident peak temperature excursion will not l
cause significant degradation of equipment operation during that period of exposure above the test maximum temperature (
Reference:
General Electric Report 327, File DV145C3007 and General Electric Document No. NSE80036).
l l
In addition, an operational analysis was performed to address the effects of the postulated accident radiation environments on the operability requirements of the transmitter.
In the event of a large break LOCA for which the HPCI system cannot maintain RPV level, the transmitter may be subject to high radiation.
However in this Case, the HPCI system is not required since the RPV will be depressurized by the break and/or actuation of the ADS system. Adequate core cooling is then provided by the low pressure ECCS systems.
Therefore, operation of this s
device is not required for safe shutdown.
In the event of a small break LOCA l
for which the HPCI system can maintain RPV level, the core never uncovers.
l hence cooling is maintained and the harsh radiation evironment is not present.
i i
This analysis makes use of partial test data and therefore meets the criteria l
of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
l
I UNIT 2 BSEP JC0 NO. 22 1
PME 10F 1 TER NO.:
NONE EFFECTIVE 9/3/85 i
l COMPONENT !.D. NO.:
E51-FS-N002 MFG / MOD. NO.:
BARTON 289 LOCATION:
REACTOR BUILDING RHR ROOM TECHNICAL DISCUSSION:
i Component materials of the Barton differential pressure switches have been i
i identified and qualification documentation located.
The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques.
Results of this evaluation indicate that the nonmetallic components have greater that 266 years of expected life at the maximum reactor building l
temperature of 104*F.
I The pressure switch nonmetallic materials are exposed to the plant postulated l
accident temperature peak of 288*F for 70 seconds. The accident temperature l
then decreases to 20S*F at 100 seconds and returns to anbient after i
approximately 20 minutes. This postulated peak temperature transient has been compared to accident test data obtained (212*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) for this switch.
Though the testing does not envelop the postulated peak accident temperature, it is judged that no significant detrimental effects to switch operation l
should occur as a result of the peak temperature transient. This assessment is based on the severity of the test performed and the short time for heat transfer through the heavy metal casing.
Additionally, radiation testing n the subject switches supports a 6
qualification level of 3.6 X 10 rads gamma.
Thoughtgetestingdoesnot envelop the postulated total integrated dose of 1 X 10 rads gamma, a radiation threshold analysis shows that the radiatjon threshold for each material used in the switch is greater than 1 X 10 rads gamma except for the Viton 0-Ring.
For the Viton 0-Ring there is testing to support the us9 of this material in an 0-ring application up to radiation level of 2 X 10 rads gamma (
Reference:
ASCO Report No. A0R 67368, Rev.0, paragraph 4.1.4).
This analysis makes use of partial test data and therefore meets the criteria of 10CFR50.49(1)(2).
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 N0. 23 PAGE 1 0F 1 TER N0.:
100 EFFECTIVE 9/3/85 COMP 0NENT I.D. NO.:
CAC-TE-1258-1 TO 14 CAC-TE-1258-17 TO 24 MFG / MOD. NO.:
PYC0100 OHM PLATINUM RTD LOC ATION:
DRYWELL TECHNICAL DISCUSSION:
These temperature elements monitor drywell air space-temperature for recording on a multipoint recorder located in the control room.
Pyco has performed qualification testing on similar RTD enveloping BSEP normal and accident service conditions (
Reference:
Pyco Qualification Test Report No. 16436-82N, Rev. 5, dated 5/18/84).
The similarity of the installed equipment has been confirmed by Pyco.
This analysis makes use of partial test data and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
l i
1
UNIT 2 BSEP i
JC0 NO. 26 PAGE 1 0F 1 TER NO.:
115 EFFECTIVE 9/3/85 COMP 0NENT I.D. N0.:
2(A-D)-BFIV-RB-L MFG / MOD. NO.:
NAMC0 D2400X-R l
LOCATION:
REACTOR BUILDIfG 80' TECHNICAL DISCUSSION:
Component materials of the NAMC0 D2400X-R position switch have been identi fled.
The materials have been evaluated per D0R guidelines and by applying Arrhenius techniques.
Results of this analysis indicate that all materials, except for Buna-N rubber (used as a binder in the asbestos gasket),
have greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature of 104 F.
The gasket, which is comprised of 20%
Buna-N and 80% asbestos, is judged acceptable for continued operation since the Buna-N is used as a binder and once the gasket is properly installed and left undisturbed, no significant degradation would occur.
The analysis performed on the D2400X-R switch is based on testing conducted on NAMC0 series SL3 switches (generically similar in materials, construction, and operation).
These switches were exposed to a 310'F and 65 psig steam environment (
Reference:
Masoneilan Test Report 1003, dated 4-19-73) which exceeds the BSEP requirement.
A radiation analysis indicates thgt the lowest damage threshold for the nonmetallic materials is 8.6 X 10 radggamma.
This damage threshold value envelops the BSEP requiremnt of 1 X 10 rads gamma.
This analysis makes use of partial test data and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continued operation is justifled.
l l
l UNIT 2 BSEP JCO NO. 30 PACE 1 0F 1 EFFECTIVE 9/6/85 TER NO.:
142, 144, 145, 146, & 147 COMPONENT I.D. NO.:
MCC-2XA, MCC-2XA-2, MCC-2XB, MCC-2XB-2, MCC-2XC, MCC-2XD, MCC-2XDA, MCC-2XDB, MCC-2XE, MCC-2XF, MCC-2XH, MCC-2XM*
CENERAL ELECTRIC IC-7700 MOTOR CONTROL CENTER AND RELAYS LOCATION:
REACTOR BUILDING TECHNICAL DISCUSSION:
CE test data applicable to the environmental qualification of the General Electric Series IC 7700 motor control center has been identified and obtained..The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques.
C.E. has'provided BSEP with a test report which contains detailed Engineering Change Notice (ECN) reviews, Product Analysis Reports, and Similarity Analysis Reports on specific components contained in the motor control centers (THED circuit breakers, CR109 magnetic starts, and a control power transformer).
This report also indicates that the test data obtained demonstrated fqualification of the IC 7700 motor control center to the BSEP normal and postulated accident conditions, subject to the certain configuration changes.
CE has indicated that it will be necessary to close the conduit entries to the e
BSEP MCC's in order.to meet the assumptions of the Report / Analysis; a program to accomplish this work is in progress. These changes will reduce the calculated peak component temperature to within the tested parameters. At this time ~the calculated peak component temperatures exceed the CE test parameters by approximately 10*F (approx. 140*F vs. 131*F).
Typical components of the type contained within a motor control center would not be adversely affected by this short time increase in temperature, CP&L is in possession of a type test of another manufacturer's MCC, with a much higher peak temperature (250*F for 6.2 hrs.) which tends to support this conclusion.
This analysis makes use of partial test data and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
'w
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t UNIT 2 j
BSEU i
JC0 NO. 32 PACE 1 0F 1 EFFECTIVE 9/6/85 I
TER NO.:-
141, 155 COMPONENT I.D. NO.:
E41-C002 MFC/ MOD. NO.:
TERRY STEAM TURBINE MODEL CCS HPCI PUMP SYSTEM LOCATION:
REACTOR BUILDING EL. -17' i
TECHNICAL DISCUSSION:
An operational analysis has been performed on the Terry Steam Turbine Model CCS HPCI Pump System. The following postulated BSEP accidents were considered
~
in this evaluation:
l.
HPCI Steamline Break 2.
Large Break LOCA 3.
Small Break in RCIC Steamline 4.
Small Break LOCA In all cases alternate qualified ECCS systems in conjunction with the ADS system (auto or manual mode) are available to maintain core cooling for a safe shutdown. Operator response is covered in the Emergency Operating Procedures.
Since the safety function is performed by alternative qualified equipment, this analysis meets the criteria of 10CFR50.49(i)(1).
Therefore, doatinued operation is justified.
b d
4 9
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UNIT 2 BESU JC0 NO. 33 PAGE 1 0F 1 TER N0.:
143 EFFECTIVE 9/3/85 COMPONENT I.D. NO.:
080-74-17 MFG / MOD. NO.:
AGASTAT 7022 AC TIME DELAY RELAY LOCATION:
REACTOR BUILDING RHR ROOM TECHNIC AL DISCUSSION:
BSEP has one Agastat time delay relay (model 7022AC) installed in the control circuit of RHR pump room cooler fan A-FCU-RB. An automatic start signal to RHR pump roan cooler fan A-FCV-RB de-energizes the coil of the time delay relay which initiates the tine delay function.
If, after the timer delay setting has elapsed, the fan motor contactor has not closed, an annunciator alarm is sounded in the control room indicating that fan A-FCU-RB has failed to start.
It is important to note that this relay does not perfonn any control function to start or stop the fan; it only gives indication.
The result of the failure of this relay would possibily be:
(1)
Loss of control power to the fan A-FCU-RB and (2)
Loss of alarm to the control room that fan A-FCU-RB has failed to start.
If control power is not lost, the fan would start as designed.
However, should the first fan fail to start the RHR pump rooms are provided with another 100% capacity fan B-FCU-RB that is fully quali fied.
This fan will automatically start as soon as RHR pump room tenperature reaches 145*F or above. There is no time delay relay involved in the control circuit of fan B-FCU-RB.
Since the safety function is performed by qualified alternative equipment, this analysis meets the criteria of 10CFR50.49, paragraph (1)(1)
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 No. 34 PACE 1 0F 1 EFFECTIVE 9/6/85 TER NO.:
148 COMPONENT I.D. NO.:
D12-RE-N010A, B MFC/ MOD. NO.:
C. E. MODEL 194X927Cll RADIATION DETECTORS LOCATION:
REACTOR BUILDING EXHAUST AIR PLENUM EL. 80' TECHNICAL DISCUSSION:
Partial qualification documentation has been obtained for the Cencral Electric radiation detectors. The test data was evaluated per the DOR guidelines and using Arrhenius techniques. The results of this evaluation indicate that the radiation detectors were tested at 212*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and performed satisfactorily before, during and after the test exposure. The test parameters envelop the BSEP requirement of 200*F accident peak temperature
(
Reference:
General Electric Report no. 248A9178).
The reactor building flVAC exhaust air plenum radiation levels are continuously monitored by two redundant radiation detector sensors. The detectors provide output signals which initiate the automatic start of the Standby Cas Treatment System and secondary Containment Isolation when the radiation levels exceed 11 MR/HR.
Duringnormaloperation,thegotalintegratedradiationexposureforthe detectors will be only 3 X 10 rads which is well below the damage threshold level of the detector nonmetallics. The detectors activate at 11 MR/ilR and complete their function before damage due to higher levels of radiation is experienced as a result of the accident.
Since the detectors performed their mitigation function immediately upon accident detection, failure would not prevent ECCS actuation or prevent the mitigation of a HELB.
Failure to automatically start the SBCT system and isolate the secondary containment during a HELB will not result in an off-site radiation dose in excess of the 10CFR100 limitations. The resultant radiation release is less than a main steam line break in the turbine building.
SBCT and reactor building isolation may be manually intiated from the control room and/or automatically initiated in response to other sensed parameters which occur during a LOCA.
Additionally, the detectors are periodically tested once every 18 months by physically removing them from their mounting and performing a complete functional test.
Since these redundant detectors perform their actions immediately upon l
accident detection, and since partial test data is used in this analysist this analysis meets the criteria of 10CFR50.49(i)(2) and (i)(4).
Therefore, continued operation is justified.
l
UNIT 2 BSEP JC0 N0. 35 PAGE 10F 1 TER NO.:
151 EFFECTIVE 9/3/85 COMP 0NENT I.D. NO.:
RING AND TONGUE TERMINATION LUGS MFG / MOD. NO.:
LOCATION:
DRYWELL TECHNICAL DISCUSSION:
The nylon insulated lugs are used to terminate Class 1E cables inside the drywell at the Penetration Termination Boxes.
Field inspections were made of these terminals to verify that the lugs were properly aligned and the insulation sleeves were physically separated between adjacent terminals.
Thi s spacing is sufficient to prevent shorting of adjacent conductors at the maximum voltage levels without taking credit for the insulating sleeves.
In the unlikely event of failure of these items no significant degradation of safety function or misleading information to the operator will result, i
therefore, this analysis meets the criteria of 10CFR50.49, paragraph (1) (5).
Therefore, continued operation is justified.
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UNIT 2 BSEP JC0 NO. 36 PAGE 1 0F 2 EFFECTIVE 9/3/85 TER NO.:
156 COMPONENT I.D. NO.:
SGT-FILT-2A-RB SGT-FILT-2B-RB MFG / MOD. NO.:
FARR MODEL NUMBER 051423 LOC ATION:
REACTOR BUILDING 50' TECHNICAL DISCUSSION:
The SBGT is not assuned to remain operable in the most severe postulated HELB environment, but as discussed below, its operation is not necessary for this event.
The radioactive release from a HELB in the reactor building is substantially less than that assumed for the main steam line break which is released directly to the atmosphere and results in much less site boundary dose than that permitted by 10CFR100.
Since the inventory loss prior to isolation for a HELB is less than the main stean line break, the offsite HELB dose is also correspondingly low even if the SBGT is not immediately operable. The HELB analyses for BSEP have shown that no fuel damage is expected as a result of the event.
Therefore, there will be no excessive radiation levels in the reactor coolant when long term recovery from the event is underway.
Thus, there is no need for the SBGT systen to maintain a negative pressure in the reactor building during recove ry.
This iten is located on the 50-foot elevation of the reactor building. The post-LOCA temperature profile in this area is a gradual increase from normal (maximum 104*F) to equilibrian gt 133*F in approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The total integrated radiation dose is 10 rads for the 40 year li fe plus the accident.
Qualification documentation was obtained for the SBGT systen and analyzed per 00R Guidelines.
The testing was performed on identical and/or similar conponents (
Reference:
Farr Test Report No. L-71167). For those safety-related components not tested specifically by The Farr Company, supplemental qualification data was obtained and analyzed. These conponents include, but are not limited to, the following major components.
1.
Blower Motor This is an enclosed General Electric blower motor with a Class F insulation system.
This insulation systen has been analyzed and found to be superior to the G.E. Class B insulation systen which has been successfully tested tg a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 212*F peak temperature,100% relative humidity and 5.5 x 10 rads gamma.
This testing envelops the BSEP postulated accident transient and through analysis, the post-accident period.
UNIT 2 BSEP JC0 NO. 36 PAGE 2 0F 2 EFFECTIVE 9/3/85 9
2.
ITE Molded Case Circuit Breaker These breakers have been tested separately by ITE at a'tenperature and radiation dose more severe than the BSEP postulated accident conditions
(
Reference:
ITE-Gould Report No. CC 323.74-57 Rev. 2 dated October 6, 1980).
3.
Allen-Bradley Push Button Control and Selector Switches These devices are manufactured basically from phenolic and metallic materials. Similar switches have been tested by Honeywell te paraneters I
which envelop the BSEP postulated accident conditions (
Reference:
Honeywell Test Report No. LTR-24407).
4 Allen-Bradley Series 700 Contactor l
These contactors have been successfully tested to 2 x 108 rads gamma and 1
248*F which envelops the BSEP requirements (
Reference:
ANCO letter for IEEE 323-1974 Qualified Components).
i Since partial test data was used in this analysis and since no degradation of l
safety function will occur nor will misleading information be given to the operator; this analysis meets the criteria of 10CFR50.49(i)(2) and (1)(5).
Therefore, continued operation is justified.
l l
t I
i o
l l
- - -. - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - - - - - ' - ^ " ^ - - - - ^ - - ' - - - - - ^ ^ ^ - ' ^ - - ^ - - ~ ~ ' ^ ^ ' ~ ~ ~ ^ " "
UNIT 2 BSEP JC0 No. 40 PACE 1 0F 1 EFFECTIVE 9/6/85
.TER NO.:
179, 181 COMPONENT I.D. NO.: TERMINAL BLOCKS MFC/ MOD. NO.:
CENERAL ELECTRIC, CR-151 LOCATION:
REACTOR BUILDING - ABOVE 20', RHR ROOM TECHNICAL DISCUSSION:
Component materials of the General Electric terminal blocks have been identified and qualification documentation on similar thermal blocks has been located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this evaluation indicates that the 8
nonmetallic components have greater than 5 X 10 years of expected life at the maximum reactor building temperature of 104*F.
The test data shows that-similar terminal blocks were exposed to test conditions, including radiation, significant more severe than the postulated accident conditions at BSEP.
Leakage current was monitored during that portion of the test program with conditions at BSEP. The average leakage current per terminal block was less than 1 ma at 120VAC which was less than the acceptance criteria established for this test. 'The results of this test coupled with the facts that:
1.
All terminal blocks are in an enclosure and therefore not subjected to direct impingement of steam or water.
-2.
There-is a redundancy of all safety related systems as well as a physical separation.
3.
All systems are periodically tested which would detect any random failure.
-further substantiate the use of these terminal blocks in the Reactor Building
(
Reference:
Amerace Report F-C5143).
This analysin makes use of partial test data and therefore meets the criteria of 10CFR50.49(i)(2).
.Therefore, continued operation is justified.
UNIT 2 BSEP JC0 N0. 42 PAGE 1 0F 1 TER N0.:
NONE EFFECTIVE 9/3/85 COMP 0NENT I.D. NO.:
C12-F010-L E51-C002-L54 MFG / MOD. N0.:
NAMC0 D1200G, D-200G-ST-2 LIMIT SWITCHES LOC ATION:
REACTOR BUILDING 50', RHR ROOM TECHNICAL DISCUSSION:
Component materials of the Namco D1200G and D-200G-ST-2 limit switches have been identified and qualification documentation on sim51ar equipment located. The qualification data has been evaluated per 00R guidelines and by applying Arrhenius techniques.
Results of this evaluadon indcates dat de nonmetallic components have greater than 9 X 103 years at the maximum reactor building temperature of 104 F except for Buna-N.
The Buna-N components have an expected life of greater than 11.8 years.
The test data shows that the switch was exposed to test conditions more severe than the BSEP postulated accident conditions for temperature, pressure, and relative humidity (
Reference:
Masoneilan International Report No.1003).
Additionally, a ' radiation analysis performed on the component materials showg that the radiation threshold Buna-N which is the weaklink material is 1 X 10 rads. The switches complete their safety function in less than one hour and
'10g maximum postulated. total integrated radiation dose during this timg is '. X th rads which is much lower that the Buna-N threshold value of 1 X 10 rays.
Since these switches complete their action. prior to being affected by the accident environment and there is partial test data to support their use, this analysis meets the criteria of 10CFR50.49(i)(2) and (i)(4).
Therefore, continued operation is justified.,
UNIT 2 BSEP JC0 NO. 43 PAGE 1 0F 1 TER NO.:
NONE EFFECTIVE 9/3/85 COMP 0NENT I.D. N0.:
2-NP6-MOT-M1, M2; 2A-RX 2-NP7-MOT-M1, M2; 2B-RX MFG /M0D. NO.:
DOERR MOTORS, A421; LINCOLN MOTORS, 213T; AND ITE CONTROL PANELS, CDP-7 LOCATION:
REACTOR BUILDING EL. 20' TECHNIC AL DISCUSSION:
The above electrical components are associated with the compressors for the standby air supply for the Non-Interruptible Air System.
Non-Interruptible Air is normally supplied from the plant service air system, these compressors are a standby source to be used upon loss of the normal supply. Non-interruptible instrument air is supplied to the following control systems:
1.
Main steam isolation valves 2.
Scram valves 3.
Scram volume vent and drain valves 4.
Reactor instrument penetration (RIP) system valves Items '1, 2 and 3 above will fail to a safe position on loss of air.
The Safety Relief valves are supplied with air accumulators of sufficient size to provide valve actuation air in the event of total instrument air supply failure.
The RIP valve control system contains provisions to isolate the air-operated valves on loss of air system pressure. This will trap air in the header and operating lines maintaining the RIP isolation valve in its required position, components necessary to perform this function are qualified.
Additionally, nonnal station service power to the service air systen can be restored within a reasonable time (< 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />), and BSEP has the capability to cross connect the air systems between the two units which would allow the non-affected unit to supply air for these loads.
The air. compressors do not directly control any indications.
Since the safety function is au.cnplished by qualified alternative equipment and no degradation of safety function nor misleading indication to the operator will occur, this analysis meets the criteria of 10CFR50.49(i)(1) and (1)(5).
Therefore, continued operation is justified.
1
UNIT 2 BSEP JC0 NO. 44 PAGE 1 0F 1 TER NO.:
NONE EFFECTIVE 9/3/85 COMP 0NENT I.D. NO.:
E51-C002-H MFG / MOD. NO.:
SOUARE D 9038-AG-1-S4 FLOAT SWITCH LOCATION:
RHR ROOM TECHNICAL DISCUSSION:
This item was supplied as part of the RCIC turbine assembly, and is used to perform a non-safety-related control function. This component is passive and must not fail so as to jeopardize the isolation logic it is powered from.
Testing has been successfully performed on a HPCI turbine that contained this component (Ref:
Wyle Lab / Terry Turbine Report No. 20458, R14-21-80). The testing wgs performed at 150 F for an undetermined time and radiation testing to 1 X 10 rads. During the HELB accident condition, the temperature gradually rises from 104 F and peaks at 295 F in 60 seconds at which time steam leak isolation is completed. The temperature returns below 150 F in 15 mi nutes. The accident rgdiation dose in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident will be less than 1 X 10 rads.
Since the switch terminations are enclosed in a NEMA metal enclosure it is safe to assume that the switch will maintain its electrical integrity for the required duration.
This analysis makes use of partial test data and therefore meets the criteria of 10CFR50.49(1)(2).
Therefore, continued operation is justifled.
L
v UNIT 2 BSEP JC0 NO. 45 PAGE 1 0F 1 TER NO.:
NONE EFFECTIVE 9/3/85 COMP 0NENT I.D. NO.:
832-CS-F019 B32-C S-F020 MFG / MOD. NO.:
SENTRY MODEL F3N1R1 SWITCH LOC ATION:
REACTOR BUILDING EL. 50' TECHNICAL DISCUSSION:
The Sentry F3N1R1 switch utilizes a Series 2 Honeywell Microswitch as the internal switching mechanism.
Honeywell Series 2 switches have been tested at 149 F for more than 30 days
(
Reference:
Honeywell Microswitch Test Response No. LTR-24407). This test envelops the BSEP accident duration but does not envelop the 70 second BSEP peak temperature transient of 200 F.
A material analysis indicates that the swi tch will not be significantly degraded by the short exposure to the postulated accident peak.
Additionally, the switch has been tested to 1 X 107 Rads (
Reference:
Hogeywell Report No. LTR-15027-1) which envelops the BSEP requirement of 1 x 10 Rads gamma.
Th.is analysis makes use of partial test data and therefore meets the criteria
-of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 NO. 46 PAGE 1 OF 1 EFFECTIVE 9/6/85 TER NO.:
NONE COMPONENT I.D. NO.:
B21-FT-4157 B21-FT-4163 B21-FT-4158 B21-FT-4164 B21-FT-4159 B21-FT-4165 B21-FT-4160 B21-FT-4166 B21-FT-4161 B21-FT-4167 B21-FT-4162 MFC/ MOD. NO.:
NDT INTERNATIONAL 78IN/S ACCELEROMETER LOCATION:
DRYWELL EL. 38' TECHNICAL DISCUSSION:
NDT International accelerometers, Model No. 78IN/S, are qualified on the basis of similarity with the NDT International accelerometer, Model No. 838-1, (Reference Wyle. Qualification Report No. 45638-1). Model 838-1 was fully qualified to meet or exceed all BSEP service conditions inside the drywell.
Similarity
~
Model No. 78IN/S and 838-1 are similar. The only difference is in the accelerometer cable interface connections. These connections are vendor supplied splices between a mineral insulated and an elastometer insulated cable, they are covered with Raychem nuclear grade heatshrink which has not been tested in this application, but is otherwise qualified.
Should the interface connections fail, there is a possibility of faulty indication of safety relief valve position in the control room.
- However, other independent indications of safety relief valve position are provided (vessel pressure, suppression pool temperature, vessel level and suppression pool level). Additionally, the non qualified SRV tailpipe temperature instrumentation may be available. Therefore, safety relief valve position indication would not be lost in the event of accelerometer failure.
Since the safety function is performed by an alternative qualified equipment, and since partial test data is utilized in the analysis, and since no degradation of safety function will occur; this analysis meets the criteria of 10CFR50.49(i)(1), (i)(2) and (i)(5).
Therefore, continued operation is justified.
I~
UNIT 2 BSEP JC0 NO. 47 PAGE 1 0F 1 TER NO.:
180 EFFECTIVE 9/3/85 COMP 0NENT I.D NO.:
TERMINAL BLOCKS MFG / MOD. N0.:
G. E. EB-5 LOC ATION:
DRYWELL TECHNIC AL DISCUSSION:
EB-5 terminal blocks are used inside the drywell as tenninal points for 120V/250V/480V Class IE control and power circuits only and no low voltage signal circuits are landed on these blocks.
The terminal blocks are mounted in Nema 4 enclosures and are not subject to direct steam or water impingement.
Various industry reports indicate that only low voltage signal circuits might be in jeopardy during a DBA.
Limitorque Report No. B0119 supports EB-5 terminal block qualification for the DBA at BSEP.
Upon receipt and successful analysis of this report, these terminal blocks will be considered fully qualified for this application.
This analysis makes use of partial test data and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continued operation is justified.
UNIT 2 BSEP JC0 NO. 48 PACE 1 0F 1 EFFECTIVE 9/6/85
.TER NO.: NONE COMPONENT I.D. NO.: CONNECTOR MFC/ MOD. NO.: AMPHENOL COAX CONNECTOR PLUG 34500 RECEPTACLE 36500 LOCATION: INSIDE/0UTSIDE TECHNICAL DISCUSSION:
Amphenol coaxial connectors are utilized to carry the signal from sensors for safety relief valve position indication. These connectors are hermeticaly sealed and are provided with another environmental seal by application of Raychem heat shrink tubing.
Raychem heat shrink tubing is fully qualified by Raychem to BSEP service conditions. The connector's organic material is of copolymer styrene and is rated at 390*F.
BSEP maximum temperature is only 340*F during the accident condition.
In the unlikely event of a failure of a connector, there is a possibility of faulty indication of safety relief valve position'in the Control Room.
However, other independent indications of safety relief valve position are provided (vessel pressure, suppression pool temperature, vessel level and suppression pool level). Additionally, the non qualified SRV tailpipe temperature instrumentation may be available. Therefore, safety relief valve position indication would not be lost in the event of connector failure.
Since the safety function is performed by an alternative qualified equipment, and since partial test data is used in the analysis, and since no degradation of safety function will occur; this analysis meets the criteria of 10CFR50.49(i)(1), (i)(2) and (i)(5).
Therefore,. continued operation is justified.
l
=-
UNIT 2 BSEP JC0 NO. 49 PAGE 10F 1 TER N0.
NONE EFFECTIVE 9/3/85
/
COMPONENT ID NO.:
B21-F003L, B21-F004L MEG / MOD NO.:
NAMC0 EA 510-17702 LIMIT SWITCHES LOCATION:
DRYWELL 17' TECHNIC AL DISCUSSION:
Component materials of the NAMCO Limit Switches have been identified. The materials have been evaluated per DDR Guidelines and by applying Arrhenius techniques. Results of this analysis indicate that all materials have an expected life of greater than 5 years at the expected average temperature of 120 F.
At the time these switches are scheduled for replacement, they will have been installed less than 3 years.
A functional analysis based on the NAMC0 EA 740 shows that the EA 510 will remain functional during the accident.
The EA 740 was quali fied to the same conditions to which the EA 510 will potentially be exposed. The EA 510, like the EA 740, has a NEMA 4 Enclosure containing a polyester-glass switch assembly.
A radiation analysis based on the EPRI document " Radiation Data for Design / Qualification of Nuclear Plant Equipment" shows that the lopst threshold value of damage for the non-metallic materials is 6 x 10 rads which exceeds the potential integrated dose (including accident) to which these switches may be exposed prior to their scheduled replacement.
?
This analysis makes use of p%artial test data and therefore meets the criteria of 10CFR50.49(i)(2).
Therefore, continu$d operation is justified.
-