IR 05000348/1987030
| ML20148S415 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/22/1988 |
| From: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Mcdonald R ALABAMA POWER CO. |
| References | |
| NUDOCS 8802030037 | |
| Download: ML20148S415 (32) | |
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f()fby JAN 2 21988 Docket Nos. 50-348, 50-364 License Nos. NPF-2, NPF-8
, AJ1ama Power Company W TN: Mr. R..P. Mcdonald
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Senior Vice President P. O. Box 2641 Birmingham, AL 35291-0400 Gentlemen:
SUBJECT: ELECTRIC EQUIPMENT ENVIRONMENTAL QUALIFICATION ISSUES (NRC INSPECTION REPORT NOS. 50-348/87-30 AND 50-364/87-30)
This letter refers to the Management Meeting held in the Region II office, Atlanta, Georgia on November 25, 1987. The issues discussed at this conference related to environmental qualification (EQ) of electrical equipment. A meeting summary, a list of attendees, and a copy of the handout are enclose It is our opinion that this meeting was beneficial in helping to bring a speedy resolution to a very complicated EQ operational issue. It also provided for a better understanding of the inspection findings and the status of your corrective action In accordance with Section 2.790 of the NRC's "Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and its enclosures will be placed in the NRC Public Document Roo Should you have any questions concerning this matter, please contact u
Sincerely,
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- J. Ne' son Grace
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Regional Administrator
Enclosures:
Meeting Summary List of Attendees Handout:
(a) Justification for Continued Operation (JC0) Unit 1 -
Technical Blocks Used in Instrument Circuits (b) Raychem/ Chico Environmental Seal Qualification
REGION 11 y"j;"L 1
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101 MARIETTA STRE ET, ATL ANTA, GEORGI A 30323
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%,,,,,# ATTACHMENT A DEC 2 1937 Docket No. 50-348 License No. NPF-2 Alabama Power Company ATTN: Mr. R. P. Mcdonald Senior Vice President P. O. Box 2641 Birmingham, AL 35291-0400 Gentlemen:
SUBJECT: CONFIRMATION OF ACTION - DOCKET NO 50-348 This refers to the Management Meeting held in the Region II Office, Atlanta, Georgia on November 25, 198 This meeting was held to discuss the issues of equipment qualification (EQ) that derived from the recent inspection and any ramifications of these issues on continued operation of Unit 1 and ccrrective actions taken before restart of Unit The licensee stated at the outset of the meeting that, except for justifi-cations for continued operation (JC0s) on grease and lubricants, all EQ discrepancies identified on Unit 2 affecting operability will be fixed prior to Unit 2 startu Since the meeting, the Region issued a letter (November 30, 1987) permitting Unit 2 startup with one additional outstanding EQ issue regarding Chico /Raychem seals, which is to be resolved by December 2, 198 The licensee's position in the meeting regarding Unit 1 focused principally on terminal blocks installed inside containment in various instrument loop The licensee's previous position had been that these terminal blocks were qualified, but the staff disagreed. In the meeting, the licensee then took the position that the terminal blocks were qualifiable, and also presented a JC0 to justify continued operation based on a combination of EQ data and a semi-quantitative assessment of containment temperatures under realistically bounding accident conditions. The licensee also stated that if safety systems initiated early in the accident sequences of concern were allowed to operate, design conditions would not be exceeded. However, the licensee also stated that, using current emergency operating procedures, operator action response to erroneous signals could result in inappropriate actions. Such erroneous signals could occur, if the containment temperature were to exceed that for which the terminal blocks are qualified. After review of the data presented by the licensee, the staff acknowledged that there is disparity in EQ test data for like and different terminal blocks and differences in interpretation of the EQ test data to be applied to Farley. The staff also acknowledged that while the licensee's operability argument had some merit, it was largely based on qualitative assumptions and contained some elements of nonconservatis OMOTDTLP
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, Alabama Power Company 2 In view of the staff's position, Alabama Power Company (APC) made the following connitments during the meeting, as clarified by telephone calls between C. W. Hehl of the Region II staff and W. G. Hairston of APC on November 30: Once Unit 2 is stabilized at power, but no later than December 9,1987, initiate an orderly shutdown of Unit . In the interim, APC will increase the awareness of shift supervisors and Shift Technical Advisors to the possibility of inaccurata data from instruments located inside containment in the event of a large loss of coolant or steam line break accident. APC will alN thoroughly train Shift Technical Advisors of the need to monitor di,erse parameters in the event of a large loss of coolant or steam line t,reak accident to assure that inappropriate actions are not taken by operators based on potential inaccurate instrument readings. Further, one of the two STAS on shift will maintain presence in the Control Roo . Effect repairs on environmental qualification deficiencies associated with Instrument Terminal Boards and Head Vent prior to restart of Unit . Walkdown Unit 1 containment during the spring 1988 refueling outage in a timely manner to identify deficier.cies between as-found and as-designed splices and o+.her types of E0 deficiencies found during the recent Unit 2 walkdowns. This walkdown will include all V-splices and a representative sample of other systems and components with field wiring connections to detennine if other deficiencies exis . Deficiencies identified shall be repaired on Unit I at least to the same extent that repairs had been made to Unit 2 EQ deficiencies. Such repairs shall be made prior to plant startup following the spring 1988 outag . Evaluate operability issues on Unit 1 on any unrepaired deficiencies prior to startu . Plant startup of Unit 1 from the 1988 refueling outage shall not occur without prior concurrence by NR It is the staff's judgment that the Farley terminal blocks might possibly pass a qualification test and that the temperatures at the terminal blocks during a large loss of coolant or steam line break accident might not exceed the temperatures for which the blocks could be qualifie Also, there is a small ;
likelihood of such accidents, and the licensee's compensating actions should i enhance proper operator action should instrument inaccuracy occur during such a design basis event. Therefore, we believe that APCS commitments present a reasonable and timely resolution of the issues of environmental qualification of equipment for Unit 1 and will provide reasonable assurance of continued safe operation of the Farley Nuclear Plant Unit 1 for the interim perio l
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Alabama Power Company 3 This confinnation of Action (CAL) letter supercedes our CAL of October 6, 198 If your understariding of these matters differs from the above,-please advise us promptl
Sincerely,
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J. Nelson Grace I Regional Administrator CAL 50-348-87-02 cc: W. O. Whitt, Executive Vice President J. D. Woodard, General Manager -
Nuclear Plant W. G. Hairston, III, Vice President - Nuclear Support J. W. McGowan, Mananer - Safety Audit and Engineering Review J. K. Osterholtz, Supervisor - Safety Audit and Engineering Review
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ENCLOSURE 2-List of Attendees Licensee: Alabama Power Company Date: November 25, 1987 Facility: Farley Nuclear Plant Units I and 2 IR No: 50-348, 364/87-30 Location: NRC Region II Office Atlanta, GA Alabama Power Company ( APC0)
R. P. Mcdonald, Senior Vice President, APC0 W. G. Hairston, III, Vice President Nuclear Support, APC0 J. D. Woodard, General Manager, Nuclear Plant APC0 W. B. Shipman, Assistant Plant Manager, APC0 R. Berryhill, Systems Performance Manager, APC0 J. McGowan, Manager, Safety Audit and Engineering Review, APC0 J. E. Garlington, Manager, Engineering and Licensing (NEL), APC0 D. H. Jones, Supervisor, Design Support, APC0 D. McKinney, Supervisor, Licensing, APC0 B. S. Monty, Manager, Operational Safeguards, Westinghouse R. W. Trozzo, Senior Engineer - Nuclear Safety, Westinghouse P. Dibenedetto, EQ Consultant, DBA J. Love, Project Engineer, Bechtel US NRC Region II M. L. Ernst, Deputy Regional Administrator A. F. Gibson, Director, Division of Reactor Safety (DRS)
E. W. Merschoff, Deputy Director, DRS C. W. Hehl, Deputy Director, Division of Reactor Projects (DRP) l
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A. R. Herdt, Chief, Engineering Branch, DRS D. M. Verrelli, Chief, Projects Branch 1, DRP H. C. Dance, Chief, Project Section IB, DRP )
T. E. Conlon, Chief, Plant Systems Section, DRS P. Fredrickson, Chief, Project Section 1A, DRP R. J. Goddard, Regional Counsel .
L. P. Modenos, Project Engineer, DRP l
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N. Merriweather, Reactor Inspector, DRS C. Smith, Reactor Inspector, DRS ,
A. B. Ruff, Reactor Inspector, DRS P. A. Taylor, Reactor Inspector, DRS l S. J. Vias, Project Engineer, DRP W. S. Little, Acting Deputy Director Regional Inspection, TVA M. D. Hunt, Reactor Inspector, DRS l
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Enclosure 2 2
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NRC Headquarters _
B. Grimes, Caputy Director, Division of Reactor Inspection and Safeguards,'NRR G. Lainas, Assistant Director for Region II Reactors U. Potapovs, Chief, Special Projects Inspection Section
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M. Jacobus, Engineer, Sandia National Laboratories
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Correspondance ENCLOSURE 3.(a)
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Alabama Power ut st7_n'K1
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Justification For Continued Operation (JCO) Unit 1-Terminal Blocks Used In Subject Instrument Circuit Da NOV 2 4 W From W. G. Hairston, III To Mr. J. D. Woodard At Vice pregjdent, Nuclear Generation l
Enclosed is a justification to allow continued operation of Farley Unit 1 with terminal blocks installed in various instrument loops. A copy of this JC0 should be placed in the EQ Central File under States, GE and Foxboro j terminal block If you have any questions, please advis hl Y1W W. G. Hairston, Ill WGH,III/BHW: dst-072 l
cc: File: A-3028-JC0 i A-5001 IEB 79-01B l
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Justification for Contint.ed Operation J . M. Fa rl ey - Uni t 1 Tenninal Blocks Used In Instrument Circuits 1. BACKGROUND The qualification of the Farley Nuclear Plant Terminal Blocks used in instrument circuits was based on type test information for the States ZWM Terminal Blocks, the GE CR 151B Terminal Blocks, and the Foxboro Terminal Blocks. Each terminal block tested was identical to that installed in the Farley Nuclear Plants. The terminal blocks were tested under simulated LOCA conditions in a configuration similar to that installed at FNP. Each test resulted in the terminal block successfully performing the intended function. However, although these tests substantiate the acceptability of using terminal blocks under LOCA conditions, the performance parameters that would additionally support their acceptability for use in FNP instrument circuits were not measured. On the basis of the 10CFR50.49 provision that permits type test plus analysis for establishing qualification, an analysis was performed to demonstrate that the FNP terminal blocks could have performed as intended for the instrument application. The analysis demonstrated similarity by size, shape, and function to a terminal block that was type tested under similar FNP LOCA conditions where insulation resistence (IR) was measured to determine leakage current. The analysis further assumed, based on review of the Sandia NUREG/CR-3814 report that the input or change in insulation resistance was attributable to a surface film mechanism and not material dependant. The corresponding values recorded during the test of the similar terminal block (Conax Test Report IPS-107, Connectron Terminal Block) provided a worst case IR value of 3 x 107 ohms. Allowance of further margin was provided by accepting a lower value of insulation resistance (i.e., 1 x 107 ohms) for use and input into the FNP setpoint analysis for loop accuracy. (Reference WCAP-11658, Evaluation of the Impact of Cable and Terminal Block Leakage on RPS/ESFAS and ERG Setpoints November 13,1987). The 1 x 107 ohms insulation resistance was provided to Westinghouse for all terminal blocks used in FNP instrument circuit _- _ _ .- . . . - . . .
_ _ . _ . . _ _ _ . . ' , Justification for Continued Operation
. . J. M. Farley - Unit 1 -'
Terminal Blocks Used In Instrument Ci'rcuits Page 2 .
A review conducted by the NRC during the week of November 16 through 20th indicated that the technical analysis approach used to justify the 1 x 10 7 ohms insulation resistance value.was not acceptable to the NRC Staff. APCo believes that the methodology employed for the analysis along with the resulting values are technically-sound and justified. However, to further exemplify the amount of conservation built into the setpoint analysis, additional reviews and studies were performe II. EVALUATION A thorough review of the Sandia NUREG report was performed which resulted in confirmation of basic assumptions such as the insensitivity of the tenninal blocks to chemical spray, the lack of surf ace film dependancy on roughness, and the recovery of IR's as temperature is diminished. Additional discussion is provided in Attachment 1 to this report. As explained in Attachment 1, correlation of the Sandia test results to the post accident performance of terminal blocks at FNP can not be made in a quantitative manne The previous evaluation of the impact of cable and terminal . block leakage on RPS/ESFAS and ERP setpoints (Ref. WCAP-11658, November 13,1987) considered a conservative value of 1 x 107 ohms for terminal block IR and, combined with other contributors to channel inaccuracy, confirmed that the RPS/ESFAS functions will occur as required in the plant safety analysis. Furthermore, the use of existing ERP setpoints (without revision) was confirmed to not impact plant safety. At the time of reactor trip and during post accident monitoring, there were no uncertainty increases which could cause the operator to be mislead into performing inappropriate actions. In view of the centinuing concerns raised by the NRC regarding the terminal block insulation resistance values currently demonstrated in the FNP EQ documentation and used in instrument inaccuracy studies, an evaluation has been performed to assess the impact of reduced IR values on the ability to achieve and maintain safe shutdown following design basis events. The results of this evaluation are described in Attachment . . ~ . -. . . ,
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Justification for Continued Operation J. M. Farley - Uni t 1 Terminal Blocks Used In Instrument Circuits
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The evaluation described in Attachment 2 considered the po;tulisted design basis events of large and small break LOCA and secondary pipe breaks. A ..
minimum set of safe shutdown instruments and their functions, potentially exposed to a harsh environment were identified. The evaluation determined that if a terminal block IR value of 5 x 105 ohms were conservatively assumed as the worst case value for that minimum set of instruments, the resulting instrument inaccuracy will allow the current ERP values to be used -
without chang Terminal block testing performed by Sandia National Laboratories (SNL) is documented in NUREG/CR-3416. As discussed in Attachment 1, correlation of the Sandia test results to the post accident performance of terminal blocks at FNP can not be directly made. However, in recognition of the concerns that the Sandia tests have introduced, an evaluation was made of design basis LOCA and secondary pipe break using IR values derived from the Sandia results. Figure 1 represents a correlation between temperature and IR conservatively assuming a logarithmic relationship between temperature and IR. Inis data is based on IR values for GE EB25 terminal blocks measured at 175 C and 95 C. Additional discussion on the relationship of IR to temperature is contained in Attachment 3. The methodology enployed by Attachment 2 was to determine the containment temperature at which the IR
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value would decrease below the value of 5 x 105 ohms. At values of 5 x 106 ohms and above the operator can use his instruments with confidence under the existing ERP's and setpoints. Having determined this containment temperature, the FNP temperature profile is used to define the periods of time when IR is below this threshold value, thereby defining the periods during DBE's when inaccuracy would be postulated to be greater than that accounted for in the ERP's. The results are shown in Figure 3.0-1 of Attachment 2. This period of interest occurs at a time when no operator action is required based on instruments exposed to the postulated harsh
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. Justification for Continued Operation
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. . J. M. Farley - Unit l'
Terminal Blocks Used In Instrument Circuits Page 4 s -
environmen For large and small LOCA, no mitigative or recovery operator actions are required using instrumentation in a harsh environment.. For secondary breaks, safety injection termination (the' required manual operator recovery action) will occur after the instrument accuracy returns to an acceptable value. The onset of excessive instrument _ inaccuracies as shawn in Figure 3.0-1 is not expected during a DBA since the following conservative assumptions were considered: The test profile shown in Figure 1 of Attachment 3, used to obtain the IR values assumed in Figure 1 greatly exceed.the maximum calculated design basis LOCA/MSLB temperature profile for FN . The physical configuration of Phase I specimens in the Sandia test produced more severe conditions than would occur at FNP. The conduit was routed up the exterior of the enclosure and terminated in the test chamber approximately 12 inches below the steam inlet port and the spray header. Neither end of the conduit was sealed. (See Attachment 1.) Sufficient test data exists to indicate that #12 AWG conductors will exhibit lower IR values than smaller #16 AWG conductors with the same insulation system. The Sandia testing used #12 AWG cables whereas #16
, AWG is used in FNP field cables for RTO and transmitter application (See Attachment 1.) The containment temperature profile assumed is derived from worst case assumptions described in FSAR Chapter 6.2 including 102% power, minimum ESF, and only one containment cooler. The profile which would result from more realistic assumptions would be significantly lowe l l
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. ' Justification for Continued Operation
, , J. M. Farley - Unit 1 Terminal Blocks Used In Instrument Circuits Page 5 The minimum values of IR and corresponding high leakage currents recorded in the referenced SNL test results are conservative, and are not representative of values that would be expected at FNP during LOCA/MSLB design basis events. Minimum values ~of terminal block IR
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values higher than those recorded in the SNL report are supported by CONAX Text Report IPS-107, and Wyle Report Nos. 17775-1 and 17733-1 for MSLB/LOCA temperatures relevant to FNP. (See Attachment 3.)
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111. CONCLUSION Based on the above, Alabama Power Company concludes that there is reasonable assurance that the instrument loops will perfonn their safety function when called upon to mitigate the accident for which they are needed. However, to further remove the point of contention regarding terminal block performance and thereby increase the margin of the Westinghouse setpoint analysis, APCo will replace the terminal blocks of concern in Unit 2, during the fifth refueling outage, with qualified splices not relying on terminal blocks and APCo will take the sane measure for Unit 1 prior to startup from the eighth refueling outage, currently scheduled for March 198 i l
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. . - ATTACHMENT 1
Additienc1 Clarifications Regarding the , Qualification of States NT/ZWM and l CR151B Terminal Blocks at l Farley Nuclear Plant (FNP) Units 1 and 2 in Low Voltage RPS/ESFAS and ERP Transmitter and RTD Circuits I I QUALIFICATION REQUIREMENTS AND STATUS States terminal blocks mounted in NEMA 4 enclosures, and G.E. CR151B terminal blocks provided with the G.E. Series 100 electrical penetration assembly terminal boxes were installed in containment safety related instrumentation circuits at FNP during construction. As such these blocks including their performance and installed configuration were required to be and are qualified to the DDR Guidelines for FNP Unit #1 and to NUREG-05es, Cat. 2, for FNP Unit # In accordance with 10CFR50.49, Par. K, requalification of this electric equipment is not require j EFFECTS OF LOCA/MSLB ENVIRONMENT ON TERMINAL BLOCK LEAKAGE l CURRENTS AND PERFORMANCE ! IE Information Notice No. 84-47 indicated that as a result of testing performed by Sandia National Laboratories (SNL) for the NRC it was shown that a moisture film will form on the surface of terminal blocks during the simulation of LOCA/MELB events. (Ref NUREG/CR-3418; SAND 83-1617, Printed August 198 Note that this reference was not provided in IEN 84-47). This film will result in the reduction of insulation resistance between terminal points and ground, and thus will allow some leakage currents to flow to groun IEN 84-47 further states that the NRC staff recognizes that leakage currents de exist during LOCA/MSLB simulations and that the leakage currents may be of significance in some application No written response to the notice was required, and it was suggested that licensees: Review their facilities to determine if terminal l blocks are used in low-voltage applications, such j as transmitters and RTD circuits, and Review terminal block qualification documents te ensure that the functional requirements and associated loop decuracy of circuits utilizing terminal blocks will not degrade to an unacceptable level due to the flow of leakage currents that might occur during design basis event The notice further stated that the NRC staff considers this review to be part of the on going activities that licensees j i i
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are currently undertaking to resolve other environmental deficiencies per 10CFR50.49 deadlines and requirement IEN 84-47 indicated that where existing terminal block - qualification testing does not provide supporting data fo , instrumentation leakage currents, the following possible l corrective action could be considered ) i Obtain documentation from valid qualification tests already performed with substantiated. data for leakage currents, and perform appropriate analysis to
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demonstrate that acceptable loop accuracy and . associated response times for instrument circuits utilizing terminal blocks are being maintained throughout various operating condition Two other possible corrective actions were also stated which involved either additional qualification testing of 1 installed terminal blocks with provisions for continuous ) monitoring of leakage currents throughout the test with ; analysis of loop accuracies, or replacement of installed
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terminal blocks with qualified splice FNP EVALUATION OF TERMINAL BLOCK LEAKAGE CURRENTS States terminal blocks in NEMA 4 enclosures were qualified for FNP Instrumentation and Control circuits inside ! containment by Wyle Report No. 44354-1. Post LOCA l simulation of Insulation Resistance (IR) values were recorded, but no leakage current or IR values were recorded during the LOCA test phase to permit quantification of the surface moisture film leakage currents discussed in IEN l l
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84-4 CR151 and States NT terminal blocks installed.in E. Series 100 Low-Voltage Instrumentation and Control Penetration NEMA 4 terminal boxes inside containment were ; qualified for FNP by as stated in G.E. Qualification ! Test Summary Report 994-75-011, dated March 27, 197 This report provides one minimum value for IR associated with i LOCA simulation testing of the CR151 and States blocks, but ) insufficient leakage current or IR values recorded during the LOCA test phase exist to permit quantification of the surface moisture film leakags currents discussed in IEN 84-4 Due to the lack of data recorded in the DDR Guideline and NUREG-0588 Cat. 2 qualification reports for the FNP States - and CR151B terminal blocks installed in NEMA 4 enclosures, a ' documentation' search was conducted to obtain documentation from already performed valid qualification tests of l identical or similar terminal blocks which could provide 1eakage current or IR data recorded during the simulated i LOCA steam condition Of the test report dccuments l evaluated, including the SNL test documentation upon which 1 IEN 84-47 was based, the most representative test of FNP in ; containment terminal block and enclosure configurations l which provided IR readings during simulated LOCA/MSLB steam l conditions was Conax Report No. IPS-107, dated 10/5/7 Minimum IR values contained in this report which were obtained during LOCA simulated steam conditions were reviewed and a conservatively low IR value was provided to Westinghouse for determination of the resulting leakage currents and their affects on RPS/ESFAS and ERP setpoint accuracie WCAP-11658 addresses the results of this evaluation , and response to AFCo E. G. Action Items Ole and 067, addressrs the methodology used f.,r the selection of the terminal block IR value used in the W.4tinghouse evaluatio BASIS FOR NOT USING SNL IR OR LEAKAGE CURRENT VALUE FOR WCAP-11658 EVALUATION All the following comments are based on a review of NUREG/CR-3418, SAND 83-1617 entitled "Screening Tests of Terminal Block Performance in a Simulated LOCA Environment" printed August 1984 and are in reference to sections of that document (Attachment #1A to this clarification report). It is important to note that only Phase I testing was performed on G.E. CR1519 (Manufacturer 1, Model B) and States ZWM (Manufacturer III, Model D) terminal blocks as shown in Table 1, Pg. 1 o Environmental Test Temperature and Pressure Profiles - As shown in Figure 1, Pg. 8, the test temperature and pressure peaks as well as profile durations greatly exceeded the maximum calculated-3-
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- DBE LOCA/MSLB surface temperature conditions for f FNP in containment terminal block applications. As < stated in the last paragraph on Pg. 52 of Sec f
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4.3.4, "Terminal biccks 6,11, and 12 (States ZWM) experienced a temperature effect. Their inter-terminal barrier softened almost to the liquid melt point, and flowed from between the ; terminals. The melted material covered some of the icwer posts of the terminals, encasing the wires and drooping below the temrinal block in , large globules. Surprizingly, as Figure 20 shows, i the terminal-to-terminal11, insulation resistances and 12 were among the for terminal blocks 6, highest measured. We have no reasonable , hypothesis to explain this behavio We can speculate that the phase change of the inter-terminal barrier material prevented in someway the formation of a continuous film between terminals, or that changing geometry somehow affected the process of conduction between adjacent terminals". Geometrical changes of the inter-terminal barrier occured in Wyle Test 44354-1, but con.plete melting did not occu o No chemical spray was introduced in Phase I LO'CA Testing. (However, Section 5.5, Pg. 126 of the conclusion states that little change in the moisture film conductivity may be expected as a result of chemical spray and therefore, chemical spray would appear to not be a significant issue.)
o Physical Configuration of Phase I Specimens - Threc 6-pole CR151B and three 6 pole States ZWM terminal blocks were all mounted vertically in the same NEMA 4 enclosure (Enclosure 2) as shown i., Figure 4, Pg. 11. Cables were brought into the side of the enclosure through 3/4 inch diameter liquid tight metal hose using elbow conduit terminators to penetrate the NEMA 4 enclosure wall The conduit was routed up the exterior of the enclosure, and terminated in the test chamber head approximately 12 inches below the steam inlet port and the spray header. Neither and of the conduit ws seale (See bottom Pg. 16, and top of Pg. 18.)
All cables used to connect the terminal block test circuitry were #12 AWG, either 1-conductor or 3-conducto The direct steam Jetting exposure into the open conduit from the steam inlet port is not representative of installed instrumentation conduit configurations at FNP, and the use of #12 ; AWG single conductor and multi-conductor cable is not representative of the FNP installed _
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ATTACHMENT 3 November 24, 1987 - TO: JOHN GARLINGTON FROM JESSE LOVE
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IR vs TEMPERATURE SUPPORTING INFORMATION FOR JCO As de:umented in numerous valid test reports, conducted by Wyle, SNL and other industrial test organizations, electrical cable and terminal blocks exhibit generic characteristics with regard to insulation resistance (IR) versus temperature during simulated LOCA/MSLB test conditions. The generic characteristic lower is that IR values are inversely proportional to temperature temperature yields higher value of IR. Conversely with regard to leakage current, leakage current is directly proportional to temperatur SNL Report SAND 83-1617 provides numerous data representations, which demonstrate this accepted phenomeno Figure 1 of Westinghouse lette- dated 11/23/87 was made from i plots of SANDO3-1617 (SNL) Phase !! test data for exposure of l l G.E. EB25 terminal blocks to the SNL Phase II simulatedIR LOCA/MSLB rnf41- MAttached Figure 2, Pg. 9 of SANDB3-1617). test data R@re 1- for an EB25 block was used from the SNL report as there were no States ZWM, or CR151B blocks tested by SNL in Phase II, and the EB 25 block is similar to these FNP installed blocks. Phase I data which did record leakage currents and IR values for Staten ZWM and CR151B blocks was not used due to the inaccuracies associated with the SNL electrical test circuitry that measured leakage current values during Phase I testin The minimum values of IR and corresponding high leakage currents L recorded in the referenced SNL test results are extremely conservative, and are not representative of values that would be expected at FNP during LOCA/MSLB design basis events. Minimum
- values of terminal block IR values higher than those recorded in the SNL report are supported by CONAX Text Report IPS-107, and Wyle Report No.s 17775-1 and 17733-1 for MSLB/LOCA temperatures relevant to FN )
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87.11/24 10:50 P06 *
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l instrumentation cabl Installed instrumentation cables at FNP for RTD and transmitter applications are #16 AWG.
Sufficient test data exists which appears to indicate that #12 AWG conductors will exhibit lower IR values than smaller #16 AWG conductors with the same insulation system when exposed to LOCA steam condition As the #12 AWG cable is a
part of the test circuit and its contribution to l IR and leakage currents resulting from steam moisture is included in the terminal block measured data, additional error may have been introduce o Electrical Configuration of Phase I Test-(Bec .4, Pg. 10, Figure 10, Sect. 4.1, Pg. 29 and last paragraph Pg. 94).
A serpentine connection of alternate terminal block (TP) poles was used which did not result in i the measurement of a unique pole-to-pole resistive ! pat As stated in Sect. 4.1 "The serpentine l ' connection of the 6-pole terminal blocks actua.11y i
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provided 5 parrallel resistive paths. Each of these paths, indicated R 1 through R in Figure 16, is in turn a parallel combination oY an infinite number of paths, i.e...a surface.*" "In measuring the leakage currents the equivalent resistance of these 5 surfaces is actually measure Without further data or assumption the individual values of the surface equivalent resistances, R g through R cann t be determined".
Also as stated in Sect. 3.4 "only one ground return path existed for all 12 phase I terminal blocks, 6 blocks per enclosure. For the majority ; of the Phase I test, all blocks were powered I simultaneously, and hence only pole-to pole f leakage current data is relevant".
As stated in Section 4.4.3, Pg. 94, last paragraph, "If the conduction paths were uniformly I distributed over the terminal block surface, the differences in wiring between Phase I (serpentine) and Phase II (straight through), would cause the Phase I 1R's to be less than the Phase II ir This result is a simple consequence of multiple parallel conducting paths. For our experimental configuration there was approximately five times the pretested conducting surface available on the Phase I terminal blocks as compared to the Phase II terminal blocks. Consequently, the insulation resistance for the Phase I terminal blocks could-5-
1 : !. 87.11/24 10:50 P07 * ,. 1
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l reasonably be expected to be one fifth of the Phase II IR's. Except for the A path of Phase II c terminal block 4, the 45Vdc data and the 125 Vdc l support the hypothesis of uniformly distributed ,
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conduction."
The, serpentine test circuitry used to measure the States and CR151B test specimen leakage currents and IR's did not yield direct individual pole-to-pole or pole to ground values of IR during the LOCA steam environment simulation, and are subject to hypothesis in order to arrive at ' required pole-to-pole value o General Applicability of Phase !! Test Data - As stated above, no Phase !! testing was performed on CR151B er States terminal blocks. The.only block
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tested in Phase II based on present available 1 information which appears to be similar to the l CR151B and States blocks with regard to, block l material, pole-to-pole spacing, the presence of a f barrier between poles and a one-piece non channel l mounted block is the G.E. EB25 (Manufacturer I, I Model A). It should be noted that Table, i P , incorrectly states that the States ZWM ble'k c is a sectional block. Six EB25 blocks were tested in Phase II. Although, the electrical test circuitry of the Phase Il test yields more realistic values of leakage currents and IR's than Phase I test, other electrical test anomolies, and the configuration and environmental test profile , are not representative of the installed condition
j at FN It is interesting to note that the only physical j design affects analyzed were related to whether or not the blocks were sectional or one piece as stated in Sect. 4.4.1.3, Pg. 8 No apparent attempt was made to correlate leakage current performance to geometrical considerations such as { the presence of barriwrs and height of blocks with ! barriers between poles or pole-to-pole spacin Perhaps the conclusion stated in Sect. 4.4. that "Figures 34 through 39 show about one to two f orders of magnitude difference between the performance of terminal blocks 5, 6, and 12 and the one piece blocks, the one piece blocks being i i better." is not singularly related to the sectional block design, but to other geometrical consideration For the Phase II tests, the one piece blocks ref erenced here are G.E. EB25 blocks I - which have similar pole spacing to the G.E. CR151B and States ZWM one piece blocks and do possess j barriers between pole . W -
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tov 24 '87 04: 16 WEC-EAST 405A .
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ATTACHMENT 2 i Westinghouse Bectric Corporation Powersystema "" J ,W ; l Bor st$ i Pmeourg1Pemeyhan 162310355 AIA-87-882 ! Mf IB 87-1000
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Noved:er 23, 1987-Mr. W. G. Hairsten, III, Vies Prueldent l Nuclear Generatico Alabama Power ocupany 600 North Eighteenth Strast Birmingha:n, Al 35291-0400 Attnt Mr. J.E. Garlingte ' Joseph M. Farley Nuclear Plant Uhits No. 1 1 2 ERP Infomation Daar Mr. Hairstant Attached is additicnal information on the 6,yw.L Wild was ganarated for Alabama Power ocupany entitled "Evaluation of the Ispact of Cable and Taminal Block Imakage on RPS/ISTAS and ERP Setpoints" dated Noveda 1987. 211s information was ganarated as a result of the NRC Equtignent : Qualification Audit which was held durinrJ the weak of Novuter 16, 198 If have any additional questiens regarding this please ocntact th act offic i Very truly yours, WESTIN2CUSE EIBCIRIC CCEtPCRATION
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c. Eicheldinger, Manager AlabamaProject A7/ast/de
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KOV 24 '87 04: 16 WEC-EAST 405A , .
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ATTACHMENT The attached table contains a listing of Farley Unit 1 Emergency Response Procedure (ERP) harsh environment instruments, significant safety related functions of each instrument, and time usage factors and diverse instruments for each function. The purpose of the table is to list the instruments potentially subject to a harsh environment for the Farley design basis events. These instruments have an environmental allowance in their calculated uncertainties used in the ERPs. The design basis events , are large and small LOCA and secondary system pid e breaks; i.e., steam lina and feed line break l A review of this table results in identification of a minimum met of ' instruments, and their functions, subject to a harsh environment and also necessary for safe shutdown from design basis events. These are RCS Subcooling, Wide Range Pressure, and Narrow Range steam Generator Water Level. Backup instruments have been identifi6d where available. Other instruments necessary for safe shutdown are lecated in a mild environment or are not affected by current leakage. Other instrunents used in the ERPs are not used to base any required actions within the Farley design bacis events or will not cause any actions to be taken detrimental to plant safety if the instument uncertainty exceeds the al'cwance present3y in the Farley ERP For RCS Subcooling, Steam Generator Narrow Range Level und Wide Range Pressure, it is recommended that for Farley Unit 1 that a containment temperature criterion be defingd that is indicative of current leakage resistance of less than 5 X 10 ohms. A value of greater than 5 X 10 ohms results in an instrument inaccuracy that will allow the current ERP values to be used by the operator to take action as specified in the ERP The temperdture or a corresponding containment pressure criterion should be used as guidance to the operator using the ERps on when to consider that additional error above that already accounted for in the ERPs may exist. Under conditions exceeding these criteria no actions which could reduce tae margin of safety, specifically termination of safety injection based on RCS Subcooling or stopping of all auxiliary feedwater based on Steam Generator Narrow Range Level or stopping of kHR pumps based on Wide Range Pressure, should be performed since the errors may exceed thoce accounted for in the ERPs. After containment conditions have returned to below these criteria the operator can safely resume the use of the ERp specified values, provid d that the leakage current resistance will 10 chm The temperature criterion based on 5 X i l ingrease ohms towould abovealso 5 X apply to Pressuriser Level use in conjunction with l 10 RCS Subcooling for Safety Injection terminatten and reinitiation. If tha LTP values for RCS subcooling are changed for bafety Injg etion termination, then a leakage current resistance of 1 4 10 or greater oculd be acceptable for us insed on a review of Figure 1 and Figure 3.0-1, the instrament inaccuracy l
; hat exceeds the value that the operator can utilite with confidence l eccurs at a time when no operator action basso on instrumentation ir; 2 i harsh environment is required for the design basis events described acev l For large and small LOCA, no mitigative or recovery operator actions are l required based on instumentation in a harch environment. For secondary l l . . -.
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. breaks, Safety Injection termination (the required manual operator recovery action) will occur after the instrument accuracy returns to an acceptchls vcluc. Thercforc the operator limiention described in the previous paragraph will not prevent any necessary operator actions from being performe A review of the Reactor Protection System tad Emergency Safeguards Features functions has determined that the significant functions required for harch environtent events (pressurizer pressure - Low SI and steam generator water level - Low-Low) are required only before 5 minutes after the event occurrence for pressurizer pressure - Low SI and 60 seconds for steam generator water level - Iow-Lnw. This early time of use in the event should ensure that the function necessary will be performed before a . significant error from leakage current develop . __ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _
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o , TAB E . TIE DIVERSE PARA E TEM _ (XM90ffS PARN EfER PUNCTION, A. Short Ters ( b1) CTMT Radiation (B311st) m CIMP Sump A. Identify IDCA *
/. erg / < 20 min (A,2) CIMT Pressure (NR or- WR)
B. Tmg Ters nWsr Invel only wrification - f B. CENT Recirculation RMST level p l-c=5- - C. Critical Safety C. Jorg Tern None Beyond Design ! Mmetion Ittal.=t fbr Flood 5 CIMT Pressure A. Identify LOCA A. fhort Teru (A-1') CDfr Radiation CfMr Pressuru k A (A,2) C1Mr Sump not affected 5 20 min by ctrrent 8 B. Iorg Ters NOW leakage [ g B. CfMr Integrity CSF D C. Adverse CIME fbr E. Img Term CDfr Te twre Instrumentation A. Img Ters (A1) P2R level Nerded, Needs RCS 3 Subcooling A. SI Termination and Presstuu + ' De-initiation (b2) HCS Pressure (WR)
(A-3) PZR Prvssure Tcuperature, B. img N NDE margin available B. CSF Monitor Tkip on Adverse
- C. RCP Trip D. Short Term CIMr Pressire Cnft as backup
l 1 15 uiri I 4. MR IG A. SI te minationj A. Img Ters (A-1) PZR leral hd at (A-2) PZR l'ressure, backy Pressur= Neeided or haremp B. DCS Sitcooling B. Larg Term PZR Pressure C. Img Term PZR Pressure Not /DBA * C. CSF
& /M w C ty b. Leag Te m W2 reg = w d $
A. BCS Jihocoling
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k. Img Ters (A-1) Core Crit TC Needed, Unit 1 5 WRT (HDT)
(A 2) WRT (Cbld)
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(A-3) 3G Pressure
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TAIRE (continued) ,, i FURCTIDIE TDE DIVERSE PARAPETER G MENTS - PARAPETER_ A. Imag 1bna (A-1) Core Exit 1C h ekip only 6. ilRT (Q)1d) A. RC3 Siboooling, g Ikic~mup to WRT/(H0f) (b2) WRT (Hot)
(A--3) SG Pressure e. Intem-it.r cSr (prs) B. tag Teru nous Nur ma, f f operator-infbreation $
G W~; M Level A. Backup to NR level A. Img Ters (A-1) NR 33 Level haam only g 7 o (A-2) AMI Floor r! i ! K1 SG Ikvel J. Vedfy heat sink fbr A. Img Tene (A-1) APH Floit Needed, or h CSF, IDCI:/STMM (A-2) WR S'i Level backup , one SG Line Break required fbe heat sink I PZR Level A. SI te, h ion & A. Iorig Ters (A-2) RCS Subcoolirig No actions 9 solely on PZR (A-2) WR R C Pressure , - re-initiat$en , , , - level ,-
(A-3) PZR Pressure Above 1700 psig only l
B. CSF Invenwry B. Imrg Tern NOIE Orily yelloir path
Idiich is not l required A. Imdequate Otxe (boling A. Lorig Teru L1t T(Bot) Not; DBA , i 10. OrIts Unit 2 Orily - B. BCS Sibooelirig B. Imrig Term (B-1) WR (7fDT) ' (B-2) WR T(CulD) ' I (B-3) SG Pressure 011. CTMr A. Identify IDCR A. Shart Ters (1-1) CTNT Sump dm(Iy Mf' i Radiation Bacias only < 20 min (1-2) CTMr Pressia e l B. Adverse CTMr fbr B. Tong Tern Sample CfNT Atamosphere l Instrunnreatiert C. CDft Monitor fbr CSF C. Img Teru Seuspie CTMr Atmosphere Ib unisolate
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012. CDfr Tem A. Backip to CTMT Pressin e A. Imrig Term CfMT Pressure 8.c Q l 1 edr adverse c mr o 1 irestrumentation . m
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8904. addressed in idCAP 11658, listed here only as grtential backtg instrumen l
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TANFo (cortinued) ASST #FrIDIE k N 1) All RK Tkip/HT in WCAP are SORT TEIM. *lhey perfbria their Dr.3 tion befbre they see a significant adverse ". CDff. Even SG 1evel for RK trip and PZR kYessre SI perforin their ibnction befbre they see a significant $ adverse CTM .. 2) sort Thru: 5 minutes Rr 7t1p/ESF Fo 20 minutes other short tern g torg W -- covers entire accidert n A 3) 16 instnments are required (minimum f) for a DBA to reach Safe Shutdou dl
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- 12 are in a harsh environnert - see pages 1 thru 4
- 4 are in a mild environment (not listed on Table)
- AFW flow - SG Pressee
- HMST level - CST level
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4) Any other instnmeerts required fbe post-accident ruonitoring (WCAP) are not required fbr DBA to reach Safb stxtdow ; l
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- .- ENCLOSURE 3.(b)
, i i RAYCHEM/ CHICO ENVIRtWENTAL SEAL QUALIFICATION
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I 11/25 08:35 7207316 *07
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WRC Proposed Violation: The Chico Seal qualification package has not demonstrated that Raychem will bond to condui ,
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_APCo Position: The postulated failure mechanism discussed during the cudit was chemical spray during a LOCA reacting with the zinc coating on the }alventzed steel 4 nipple to fonn a gray powder over the nipple. The result es a path for enough moisture to enter the limit switch between the Reychem and the degraded conduit causing the limit switch to fail. The following paragraphs
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describe in detail the Farley configuration and its configuration relative to the postulated failure described above. In sumary, it should be noted i that Chico A alone provides a pressure tight seal inside the pipe nipple which provides a pressure tight sea To provide additional assurance that moisture will not enter the Ifmit switch, three additional barriers have i been applied to the FNp configuration. They are: i
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1) Raychem breakout boot i' 2) Keeper sleeve 3) Compression adapter clamp The Raychem breakout kit used for the FNP application is environmentally Wyle Testincluding qualified Report No.thermal aging, irradiation, and LOCA testing (Reference 58442-2, dated 4/03/81). The Farley configuration uses a breakout at the end of pipe nipple. Since the breakout had been qualified previously. Farley conducted a test on the RAYCHEM/ CHICO environmental seal configuration shown in Figure 1 for pressure and temperature conditions postulated during a LOCA (Reference Qualification Testing of Raychem Environmental Seals for Alabama Power Co., Joseph M. Farley Nuclear Plant, dated 12/30/81 chemical spray.). The test did not include exposing the test specimen for spray, The following paragraphs address the affect of chemical i l i The environmental seals used with NAMCO EA-180 limit switches are composed of a Raychem WCSF breakout boct that has been shrunk onto a 1" pipe nipple attached to the limit switch (See Figure 1). The individual conductors l l connected to the switch pass through the breakout boot which forms a seal to ' the conductor insulation / jacket. To provide mechanical rigidity to the breakout boot, the nipple and the breakout boot are filled with Crouse-Hinds sealing compound (CHICO A) and allowed to cure. In addition to providing mechanical rigidity to the breakout boot crotch, the CHICO A provides an ;
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additional pressure tight barrier (seal) inside the pipe nipple which is environmentally qualified. CHICO A was qualifted by test conducted by Southwest Research Institute (SWR 1 Project No. 03 4974-001) for use as drywell penetrations for Grand Gulf Nuclear Station. In addition, on the recomendation of Raychem, a keeper sleeve was installed over the breakout boot and the nipple to add rigidity to the boot, and to keep the boot in place during elevated accident temperatures when the adhesive soften /25 08i36 7207316 #28 _ - - - -- - w-wwwww
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l M Co Postt1on: (continued) keeper sleeve to provide support for the flexible cond mechanically clamps the keeper sleeve to the pipe nippl l l The zinc coating or. the galvanized steel nipple may interact with the , chemical spray during LOCA and form a gray powder over the nippl However ! the chemical Wyle Test Reportspray No. does not react with the Raychem 51119 adhesive (Refer i 58442 2, Sutton 3.4, Page S and Section 3.4.2).
addit *on to the duration of spray at Farley is only 87 minutes and the In indiviiual conductors will be effectively shielded from the spra breakout boot, for whatever reason, the seal assembl because adapte of the keeper sleeve and the clamping action of the compression the compression adapter clamp 411 fail, the internals of t switch will still be protected by the approximately 3 inch long CHICO A sea ,
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