ML19095A232

Response to Concerns Re Using of Cable Manufactured by Continental Wire & Cable Company, Summarizing VEPCO Investigation with Conclusions

ML19095A232
Person / Time
Site:	Surry
Issue date:	07/14/1978
From:	Stallings C Virginia Electric & Power Co (VEPCO)
To:	Harold Denton, Schwencer A Office of Nuclear Reactor Regulation
References
Serial No. 401
Download: ML19095A232 (7)

Text

e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND,VIRGINIA 23261 July 14. 1978 Mr. Harold R. Denton, Director Serial No. 401 Office of Nuclear Reactor.Regulation PO&M/DLB:das Attn: Albert Schwencer Docket Nos. 50-280 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

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Dear Sir:

This in response to concerns expressed by members of your staff regarding the use of a certain type cable manufactured by the Continental Wire and Cable Company at Surry Power Station Unit Nos. 1 and 2. This letter summarizes our investigation of this matter and presents our conclusions.

• Background On June 29, 1978:, Anaconda Cable, who now* own Continental Wire and Cable, notified Vepco that their records indicated that certain cable which had failed environmental testing at another utility might also be in use at Surry Power Station.

In response to this notification, an invest.igation was initiated immediate"-

ly to determine if this type of cable was in use in safety systems, inside con-tainment, at Surry Power Station. Concurrently, Anaconda Cable was to determine the exact specifications of the cable which had failed as compared to cable pur-chased for use at Surry Power Station. The utility which had conducted the cable test was also contacted to determine the conditions under which the cable had failed as compared to our LOCA performance criteria. Our findings are summarized below. Throughout this letter the other utilities Continental cable which failed will be referred to as the "failed cable". The Continental cable in use at Surry will be referred to as the "Surry cable".

Cable Specifications A review of the records of Continental Wire and Cable has determined that the failed cable is different in several respects from the Surry cable. The failed cable is described in test reports which you now have. The Surry cable specifications are briefly as follows (additional information is provided in the attachments):

conductor:* 16 gage, 7 strand, copper insulation: 25 mils cross-linked fire-resistant polyethylene, (compound number CC-2210)

Shield: 100 percent coverage aluminum mylar tape, with 18 gage 7 strand copper drain wire jacket: 45 mils hypalon

e VIRGINIA ELECTRIC AND POWER COMPANY TO Hr. Harold R. Denton Page 2 The major differences between the failed cable and Surry cable are in in-sulation compound number and in insulation and jacket thicknesses.

Test Results - Failed Cable The test in which Continental Cable failed was performed recently for another utility. Since you now have the detailed results of this test, only a brief description will be provided here. The test performed was a corngination LOCA/steam break test including a prior radiation exposure of 1.5 X 10 rads.

The test sequence was as follows.

irradiation of cable sample to 1..5 X 108 rads increase temperature and pressure to 340° and 110 psia. T0 was es-tablished when these conditions were reached 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 340° and 110 psia After 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, temperature was dropped to 250° and maintained for a total test duration of 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> This was an extremely conservative test which combined the worst effects of both the LOCA and steam break. This combination of conditions would never occur under any accident conditions. For example, irradiations on the order of 108 would occur only during a LOCA during which temperature and pressure would be considerably less than 340° and 110°F. S:imilarly, the temperature and pressure in this test are cgaracteristic of a steam break wherein irradia-tion levels of approximately 10 rads would occur. This test was apparently intended to emcompass all conceivable test requirements in order to reduce the number of tests required. For this reason, the test did not establish that the cable would perform unsatisfactorily in either a LOCA or a steam break.

Discussions with personnel involved in this testing indicated that the failed cable was replaced with another make of cable following this test.

Our impression from these discussions was that the cable was replaced not so much due to any concern over its performance, but because replacement of the small number of circuits affected was easier and faster than the running of additional, less conservative tests.

In summary, these test results indicate that certain instrument cable which is similar to cable in use at Surry, will not perform satisfactorily when exposed to test conditions which were far more severe than would occur in the event of a LOCA. There is no evidence that the Surry cable would not perform satisfactorily under more realistic test conditions or under actual LOCA conditions. However, to resolve concerns over this issue we have con-ducted a review of the specifications of Surry's cable and of the test data available relative to its perfonnance during a LOCA.

Use of Continental Cable at Surry A complete review of all cable runs has not been completed. It has been determined that Continental Cable is extensively used in safety related applica-tions at Surry. The cable is used only as instrument cable. The maxbnum vol-tage used in these applications is 50 volts. Hate that the voltage applied in

V~RGI:.IA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton

- Page 3 the failed cable test was 300 volts.

The Surry cable is in the pressurizer pressure and.level and the steam generator level instrumentation on Unit 1. Since the exact extent of its use on both units has not been determined, we have assumed for purpose of this evaluation that the cable has been used in every possible instrument application.

Acceptance Criteria and Test Results -*surry Cable All safety related electrical equipment for Surry Power Station was pur-chased to meet the LOCA performance requirements specified in Section 8 of the FSAR. Section 8 requires operabilitT in an environment of 280°F and 40* psig for a period of 30 minutes.* Purchase specifications for instrument cable*

require the capability of withstanding a total radiation dose of 10 8 rads with-out a significant change in physical and electrical properties, a value well in excess of the 2 X *107 rads exposure estimated for a Surry LOCA.

All Surry cable purchased from Continental Wire and Cable, was subjected to extensive testing and inspection to ensure quality ahd performance. A representative test report for one cable sample is included in Attachment 1.

Test .reports for all Continental Cable are available if desired. These tests included the verification of mechanical design parameters and of basic electri-cal properties of the conductor and insulation. Tests were performed to moni-tor the performance of the cable and insulation under a variety of severe en-vironmental conditions. These included measurements of the effect on tensile strength and elongation of 7 days in an air oven at 150°c. The cable was tested for heat *distortion at .1S0°c and accelerated water absorption at .75°c.

In all cases, cable performance was satisfactory. Additional information in-clud~ng acceptance criteria is shown on the test report form (attachment 1).

The suitability of this cable for operation under high irradiation has been confirmed both in tests performed by the manufacturer and by other test performed independently. The following article, included as attachment 2, provides a concise summary of-the effects of radiation on the electrical properties of various insulation materials.

"Insulation and Jackets for Control and Power Cables in Thermal Reactor Nuclear Generat~ng Stations" by Robert:B. Bl~dgett and Robert G. Fisher IEEE Transactions on Power Apparatus and Systems, Vol. PAS-88, No. 5 May 1969 This article, in addition to discussing radiation effects on the standard measures of insulation performance,* i.e *. tensile strength and elongation, also directly addresses the effects of irradiation on other electrical properties.

Note that on page 2 of the article, the types of cable cove~ings.tested are listed. Covering type No. 4, CB CLPE is of the same general type as the Con-tinental Cable *used at Surry. As shown in Table XI of the article, under

e VIRGINIA EU:CTlHC AND POWER COMPANY TO Hr. Harold R. Denton Page 4 column 4 for CB CLPE, elongation begins to show deterioration prior to other parameters and identifies the theshold of irradiation damage. This confirms the validity of the accepted practice of relying on measurement of elongation and tensile strength to check for insulation deterioration for this type of insulation. In reviewing this article, please note the following.

1)' In Table XI under column 4, 5 X 10 7 rads is identified as the theshold of d~age for the type of cable used at Surry. A dose of 1 X 10 rads represents the end of serviceability.

2) Under "conclusions", cross-linked polyethylene is* identified as among the most suitable insulation materials for nuclear plant service.

We will now discuss test results for the specific type of Continental cable used at Surry. This test was performed by the manufacturer in 1971, on insulated conductor only, with no jacket. Test information is included as attachment 3.

The test sequence and results are listed on page 2 of the attaclnnent. The test sequence was as follows:

120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />, 50 PSIG steam, followed by 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> innnersion in O. 5;~ Boric acid solution at 160°F Sequence repeated at radiation exposures of 0, 1 X 10 7 , 5 X 10 7 , and 1 :~ 10 8 The test results are listed below as Table 1 with the addition of estimated tensile and elongation values for an exposure of 2 X 10 7rads. This has been added because 2 X 107 rads is the maximum calculated irradiation under LOCA conditions at Surry.

TABLE 1 LOCA TEST RESULTS CLPE - COHPOUND /F2210 CONDITIONING TENSILE ELONGATION PSI  %

NONE . 2440 (100) 550 (100)

STEAM/BORIC ACID 2390 ( 98) 450 (82)

RADIATION ONLY 1 X 107 RADS (GAHHA) 2640 (106) 425 (77)

• 2 X 107 ?-ADS (GAMMA) *2538 (104) i~378 (69) 5 X 107 RADS (GAHHA) 2230 (92) 238 (43) 1 :is: 10 8 RADS (GANNA) 1710 (70) 100 (18)

e VIRGINIA ELECTRIC AND PowER CoMPANY To Mr. Harold .R *. Denton P_age 5 RADIATION AFTER STEAM BORIC 'ACID 1 X 10 7 RADS (GAMMA) 2580 (105) 393 (72)

• 2 x*107 RADS (GAMMA) *2385 * * (98) *344 * * (63) 5 X 107 RADS (GAMMA) 2200 (90) 200 . (36) 1 X 108 RADS (GAMMA) 1600 (66) 69 (13)

(%.RETENTION VS ORIGINAL VALUE)

• ESTIMATED BY LINEAR INTERPOLATION Based on*IPCEA standards, an acceptable value for tensile strength and elongation following.this.test is 50 percent of the original value *of each.

The test sequence which most closely approximates the Surry LOCA condition is the 2 X 107 rads exposure following the steam and boric acid exposure. Based on a linear interpolation of actual test data the.tensile strength and elonga-tion following a LOCA would be .95% and 63%. (results underlined) of the original values. This is acceptable. These test results indicate that, under the highest possible irradiation, and in temperature, moisture and pressure conditions *of greater severity and duration than Surry LOCA conditions, the cable will per-form satisfactorily.

The results also confirm the theshold of irradiation damage at 5 X 107 rads. Note also that irradiation is the major contributor to deterioration of* cable properties and is far more s_ignificant than the steam and water ex-posure.

P_age 3 of attachment 3 is a graph of tensil strength and elongation versus irradiation for the polyethylene compound number 2210 as used in the Surry cable. This data*provides additional confirmation of the onset of deterioration at approximately 5 X 107 rads, accelerating rapidly as irradiation approaches 108. This graph also demonstrates the validity of linear interpolation between 1 X 107 and 5 X 107 which was used in Table 1.

• Instrument Requirements for LOCA While we are confident that our Continental instrument cable will perform satisfactorily thr~ughout a LOCA and thereafter, it is.pertinent to note that the safety related instrumentation located inside the containment is only needed for a short time folloW'.i:ng a LOCA. The instrumentation and coincidence logic required for the function of engineered safeguards during a LOCA are discussed in Section 7 of the Surry FSAR.

Pressurizer pressure and level are the only instruments inside containment which are necessary for the initiation of saf_eguards during a LOCA. Except

e vrnmNIA ELECTRIC .-\ND PowER CoMPANY To Mr. Harold R.. Denton

• Page 6 for very small breaks, i.e. less than 1 inch, the initiating function would be completed within 5 minutes.

The containment pressure transmitters which are the most important instru-ments for safeguards initiation are located outside.the containment.

The following instruments, located in containment, while not*. required to initiated safeguards, are of value in establish~ng the nature of the accident and for confirm1:ng the proper initiation of safety functions.

containment sump level containment temperature safety injection flow accumulator*. levels steam line pressure steam flow wide range reactor coolant temperature wide range reactor coolant pressure In response to a LOCA, these instruments are.used by the operator to veri-fy system conditions and safeguards operation. A loss of one or more of these instruments would not affect the operation of safeguards. These instruments are of greatest value for* the first half hour following an accident.

In summary, instrumentation located inside containment.is needed only for a short time following a LOCA for safeguards initiation and for verification of system conditions. Within 30 minutes following a LOCA, this instrumentation is no l~nger essential; its failure would pose no problem to safe post accident operation. Thus these instruments have served their function lo_ng .before signi-ficant irradiation has occurred. Thirty minutes after the worst LOCA, irradia-tion is still less than 10 6 rads, far below the threshold of damage.

Sunnnary and Conclusion The objective of this evaluation has been to determine if certain instru-ment cable in use at Surry Power Station is suitable for its intended purpose.

This concern developed following the failure by similar cable of a LOCA/steam break environmental test at another utility.

We have reviewed the failed cable test results to determine if any new cable performance information was developed which would cast doubt on the bases upon which our original cable selection was made * . We found no such evidence.

In fact, in many respects, the failed cable test confirmed the.test data deve-loped for our cable in 1971. The failed cable test has demonstrated once again that cross linked polyethylene insulation, when irradiated beyond 108 rads, will not perform.

The unrealistic and abusive cable test which initiated this concerned is in no way an indication that such cable would not perform its intended function under accident conditions. Indeed, test results performed by the manufacturer and con-firmed by others indicates satisfactory performance under severe accident condi-tions.

VIRGINIA ELECTRIC AND POWER COMPANY TO Hr. Harold R. Denton e

Page 7 The data presented herein demonstrates that for cross linked polyethylene insulation, irradiation is the major contributor to cable deterioration under LOCA conditions. The data also established 5 x 10 7 rads as the theshold for irradiation damage. This is far above the irradiation which would occur under Surry LOCA conditions.

We are confident that the Surry cable will perform its intended function under LOCA conditions. No further investigation or corrective action is con-sidered necessary.

Very truly yours,

. . .:...,;_-!,;. \

C. M. Stallings Vice President - Power Supply Production Operations cc: Hr. James P. O'Reilly