ML20071H030
| ML20071H030 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/20/1983 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| LIC-83-129, NUDOCS 8305240474 | |
| Download: ML20071H030 (36) | |
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Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 402/536-4000 May 20, 1983 LIC-83-129 Mr. Robert A. Clark, Chief U. S. Nuclear Regulatory Commission Of fice of Nuclear Reactor Regulation Division of Licensing l
Operating Reactors Branch No. 3 Washington, D.C. 20555
Reference:
Docke t No. 50-285
Dear Mr. Clark:
l Environmental Qualification of Safety Related Electrical Equipment l The Omaha Public Power District received the Commission's Safety Evaluation Report ( ER) dated January 11, 1983, which included the Technical Evaluation Report (TER) prepared by Franklin Research Center. Subsequent to the SER, 10 CFR 50.49 was revised to provide regulations related to environ-mental qualification of safety related electrical equipment.
The new 10 CFR 50.49(g) required all licensees to identify open issues related to their environmental qualification programs by May 20, 1983. The District also received the Commission's letter dated April 4, 1983, which provided clarification on the new rule and the SER requirements.
Accordingly, this letter provides the 90-day response to the Commission's SER, the response required by 10 CFR 50.49(g),
and the additional information requested by the April 4, 1983 clarification letter.
Attachment 1 to this letter provides the District's position on outstanding items identified in the TER. All equipment within the scope of the rule is identified in Enclosure 4 of the District's letter to the Commission dated November 5, 1982. As requested by the Commission's April 4, 1983 3305240474 830520 o#
I 45 5124 Employment with Equal Opportunity Male / Female
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Gw e Mr. Robert A. Clark LIC-83-129 Page Two letter, Attachment 2 contains a description of the methods used to identify equipment within the-scope of 10 CFR 50.49(b)(2). Attachment 3 -identifies the remaining open items and provides a schedule for their resolution, where appropriate.
The' District also wishes to note that all SCEW sheets are referred to by electrical equipment qualification page number. (This number consists of the enclosure number, followed by an alpha-numeric page number; i.e., 6-29A.) It would be helpful to the District if the Commission would use this system in future correspondence.
Since ely,
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W. C. / ones Divist6n Manager Production Operations WCJ/TLP:jmm Attachments cc: LeBoeu f , Lamb, Leiby & MacRae 1333 New-Hampshire Avenue, N.W.
Washington, D.C. 20036 Mr. L. A. Yandell, Senior Resident Inspector
l Attachment 1 TER Equipment Item Nos. : 1, 2, 3, and 4 NRC Qualification Category: II.a (qualification not established)
Subject:
Foxboro Transmitters Pages: 6-61, 6-116, 6-109, 6-113, 6-1A, 6-123 Deficiency: Multiple (see TER)
The District contracted Wyle Laboratories to evaluate the Foxboro transmitter qualification status in accordance with the DOR Guidelines of IE Bulletin 79-01B. Wyle Laboratories concluded the subject transmitters lacked sufficient docu-mentation to support qualification.
In an attempt to preclude an expensive change-out of very reliable instrumentation, the Foxboro transmitter serial numbers were forwarded to Foxboro for evaluation. Based on Foxboro's reply, Modification Requests (MR's) were issued to upgrade the existing transmitters.
As of this date, all reported Foxboro transmitters have been upgraded to the NE-10 series transmitters which have a demon-strated accuracy of 0.5%. The Foxboro Company has notified the District that the NE-10 qualification is complete and the test specimens are certified to IEEE-323-1974. An Engi-neering Evaluation Assistance Request (EEAR) has been issued to gather all transmitter serial numbers and purchase inform-ation to be forwarded to the Foxboro Company, who will deter-mine if adequate similarity between Fort Calhoun equipment and the test specimens can be established. Once es-tablished, the pertinent reports will be forwarded to the District. It is expected that this qualification will be established prior to the completion of the 1984 refueling outage.
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TER Equipment Item No.: 5 NRC Qualification Category: I.b (qualification pending modification)
Subject:
ASCO Pressure Switch - PC-1849 p Page: 5-10 1 Deficiency: Inadequate Qualification Pending Modification PC-1849 functions to isolate instrument air to containment on loss of air header' pressure in conjunction with a CIAS signal. LER 79-007 identified the switch as lacking proper environmental qualification and indicated that the switch would either be upgraded or moved from the containment j- atmosphere. The switch was removed from containment and i
placed between valve PCV-1849 and penetration M-73 in Room
-6 9. At the same time, the switch was changed from the Barksdale Model D2T-M1055 to an ASCO Model SBllAR/T10A32.
.This modification was completed in November, 1981.
At the time of installation, a fully qualified pressure switch that. met the District's specifications could not be located. The ASCO switch was selected because ASCO had already commenced qualification and supplied the District with a test specification (AQS-21623). This specification was later replaced by AQS-02882, Appendix A. Test.results, although not completed as of this date, are expected in June, 1983.
Thetestspcificationrequiresradiation qualification.to 5 x 10 RAD. As radiation is the only harsh environmental condition in Room 69, it is expected that the switch will meet the Room 69 requirement of 2.5 x 105 RAD. Aging requirements obtained from the type test will be incorporated into the Fort Calhoun qualified life program to preclude premature. failure of the switch.
Please note'that the SCEW sheet, Page 5-10, was inadvertent-ly omitted from the District's November 5, 1982 submittal; therefore, it is attached for the Commission's reference.
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- Facility: Fort Calhoun Docket No.: 50-235 SYSTEM COMPONENT EVALUATION WORK SHEET -
l LHv 1 Huidata i uGLuntniatiON MLE. QUALifi- UiST5i1U-CATION ING
'! EQUIPMENT DESCRIPTION Specifi- Qualif- Specifi- Qualifi- METHOD ITEMS Parameter cation cation cation cation System: Instrument Air Operating Time Continous Note 1 Note 1 Note 1 Note 1 Item No.: PC-1849 Tempera- N/A Note 1 N/A Note 1 Note 1 Note 1 ture *F Component: Pressure Switch Pressure PSIg N/A Note 1 N/A Note 1 Note 1 Note 1 Manufacturer: ASCO Relative Model No. : SB11AR/T10A32 Humidity % N/A Note 1 N/A Note 1 Note 1 Note 1 m Function: Loss of Pressure Chemical N/A Note 1 N/A Note 1 Note 1 Note 1 g Isolation Spray Accuracy - Spec: Note 1 Demon: Radiation 2.5x10 5R Note 1 N/A Note 1 Note 1 Note 1 Service:
Location: Room 69 Aging N/A Note 1 N/A Note 1 Note 1 Note 1 Flood Level Elev: N/A Submer- N/A NA N/A N/A N/A N/A Above Flood Level: gence Documentation
References:
Notes:
l 1) Appendix A 1) Pending Qualification per ASCO Test Spec. AQS-02882 Appendix A 43 -
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TER Equipment Item Nos.: 8 through 13 NRC Qualification Category: II.a (qualification not established)
Subject:
Fisher 304 Electrical Valve Stem Position Switches Pages: 6-127, 6-125, 6-91, 6-77B, 6-llA, 6-3, 6-106, 6-8, 6-104, 6-10, 6-14, 6-98, 6-42, 6-31, 6-23, 6-18, 6-45, 6-81, 6-100 Deficiency: Radiation Qualification The material analysis used to support the 106 RAD rating was accomplished using Fisher Controls Bulletin 62.3.304 and Fisher Form 2007B (May, 1969). The above documents list the Model 304 constituent materials as follows:
Base 7 Cover -
Aluminum Cams and Cam Rod -
Stainless Steel Operating Arm -
Cadmium-Pla ted Steel Cam Gears -
Cadmium-Plated Steel O-Rings -
Nitrile Electrical Switch -
Phenolic Only the Nitrile and Phenolic materials are susceptible to radiation degradation (reference Table C-1 of Appendix C to IE Bulletin 79-01B). The threshold (the radiation exposure required to change at least one physical property) is listed as 106 RAD for both materials. TER Item Nos. 8, 9, and 10 (switches in Rooms 13, 21, and 22) are the only Fisher 304 limit switches in an environment where accident radiation levels could exceed 106 RAD. The normal 40-year dose does not exceed the 106 limit. Therefore, the District is unable to use preventive maintenance to preclude radiation aging.
At this time, the District has issued a purchase order to Franklin Institute for radiation testing, which is expected to be completed prior to the end of the 1984 refueling outage.
e 5 TER Equipment Item No.: 14 NRC Qualification Category:- II.a (qualification not established)
Subject:
Fisher 304 Electrical Tosition Indication Switch Pages: 6-5, 6-95, 6-10' De ficiency: Temperature, Pressure, and flumidity Requirements Franklin Research Center found te Type 304 switch installed in Room 81 to be deficient with espect to temperature, pres-sure, and humidity requirements. The District replaced the Type 304 with NAMCO Type EA-180 limit switches during the 1982 refueling outage. These limit switches are fully quali-fled (see TER Item No. 7). SCEW sheets were revised and for-warded to the NRC as part of the District's November 5, 1982 submittal.
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TER Equipment Item No.: 20 NRC Qualification Category: II.a (qualification not established)
Subject:
NAMCO Model D2400X Limit Switch Page: 6-65 Deficiencies (1) Documented Evidence of Qualification Not Adequate (2) Criteria Regarding Radiation Not Satisfied Adequate qualification for NAMCO D2400X limit switches could not be established. Therefore, during the 1981 ref ueling outage, these items were replaced with NAMCO Model EA-180 limit switches. NAMCO Type EA-180 limit switches are classified as NRC Qualification Category I.a (equipment qualified, see TER Item No. 7). The SCEW sheet was revised and submitted to the NRC as part of the District's submittal dated November 5, 1982.
In Note 1 to TER Item No. 7, it was stated that the District made no attempt to qualify EA-180 electrical connection points. SCEW Page 6-59A reports the qualification status of Conax electrical conductor seal assemblies. These seal assemblies are used to seal the electrical connection of NAMCO limit switches. The Conax connector precludes steam from entering the limit switch during an actual LOCA.
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TER Equipment Item No.: 28, 29, 31 through 40 NRC Qualification Category: II.a (qualification not established)
Subject:
Limitorque Valves Pages: 6-118, 6-117, 6-96A, 6-75, 6-92, 6-63, 6-62A, 6-G2, 6-84, 6-ll3A, 6-85, 6-4 Deficiencies: (1) Adequate Similarity Not Established (2) Aging Degradation Not Evaluated Adequately (3) Qualified Life Not Established TER Item Nos. 28, 29, and 31 through 40 note that equipment qualification was not established for Limitorque motor oper-ated valves. The TER cites lack of documentation from the manufacturer which establishes similarity between installed equipment and the actual test specimen. Table I compiles the information obtained from Limitorque, denotes appro-priate test report, provides motor manufactarer, and indi-cates insulation class for each tag number identified.
The District wishes to reiterate its position concerning the necessity of qualifying the PORV block valves (TER Item No.
38). It was and is the District's contention that these block valves do not require qualification. The Franklin Research Center rejected this position claiming " defense-in-depth" requires qualification of these valves. Even in the unlikely event that the PORV's stick open and the block valves fail (a double failure), the resulting condition is still within the bounds of the existing safety analysis and wottld be handled as a small break LOCA.
No further qualification effort is to be made on Item No.
- 38. However, the reference to the equipment in the plant emergency procedures will not be eliminated, thus providing the operator with the maximum flexibility in accident mitigation.
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TASLE I .,;
Valve Limi torque Report No. Insulation SCLV ROOM Mic ro f i l m Model Motor Motor Brake Vo l t s/ Plea se
- icg y O rde r No . Class Number Ma nn fac tu re r Manufacture r Cycles IICV-311 337597 600198 11 6-62A Cont. 17068/1958s7 SMB-00 Rollance NA 880/3/60 1
HCV-312 337597 600198 H 6-62A Cont. 17068/195fi7 SMB-00 Rollance NA 480/3/60 ifCV-313 337597 600198 11 6-62A Cont. 17068/195887 SMB-00 Rollanco NA 480/3/60 llCV-315 337597 600198 11 6-62A cont, 17068/1958:7 SMB-00 Rollance NA- is80/3/60 ItCV-317 337597 600198 il 6-62A Cont. 17068/1958s7 SMB-00 Roilance NA 880/3/60 4
.IICV-318 337597 600198 H 6-62A Cont. 17068/195887 SMll-00 llollance NA 8880/3/60 llCV-320 337597 600198 11 6-62A Cont, 17068/195f7 4 SMB-00 Rolianco NA 880/3/60 6
HCV-319 337591 600198 H 6-62A Cont. 17068/19547. SMB-00 Relianco NA 480/3/60 l !
HCV-327 337988A 600198 il 6 -88: Cont. 1708:2/19546 SMB-0 Rollanco NA 8880/3/60 IICV-329 337988A 600198 il 6-84 Cont. 1708a2/195816 SMB-0 Reliance NA a80/3/60 IICV-331 337988A 600198 ll 6 - 8 88 Cont. 1708:2/19586 4 SMll-0 Rollanco NA 8680/3/60 !
ItCV-333 337988A 600198 H 6-886 Cont. 17042/19586 4 SMB-0 Rollanco NA 480/3/60 alCV-388 1 337637A 600198 il 6-85 Cont. 1708:2/19546 SMB-O Rollanco NA 880/3/60 4
llCV-383-3 380217A 4 B0003 8 6-63 Cont. /15160 Reliance NA 180/3/60 4
llCV-383-4 3fO217A i B0003 8 6-63 Cont. /15160 Roilance llCV-2914 337063 600198 11 6-62 Cont. 17071/19545 SMB-00 Rollanco NA 4fs0/3/60
' ilCV-29384 331063 600198 H 6-62 Cont. 17071/19585 6 SMB-00 Rollanco NA 480/3/60 o, llCV-29586 337063 600198 11 6-62 Cont. 17071/19585 SMll-00 Rollanco NA is80/3/60
> llCV-297ta 337063 600198 il 6-62 Cont. 17071/1958a5 SMB-00 Retlanco NA la80/3/60 LCV-218-3 38:37148 00003 8 6-4 .7 17052/15158 SMB-00 Roilanco NA 480/3/60 llCV- 3 fs 7 337637B 600198 11 6-92 13 1701:4/19584 4 SMB-2 Rolianco NA 480/3/60 llCV-1941C 3ta5763 A B0003 8 6-96A 81 17055/15180 SMB- MO Roliance NA 480/3/60 llCV-1042C 3fS763A i B0003 0 6-96A 81 17055/15180 SMB-000 Rollanco NA 480/3/60 IICV-1384 352016A B0003 0 6-117 81 170',7/15181 SMB-000 Retianco NA 380/3/60 1
llCV-1385 385255A 4 f-C3271 0 6-118 81 HA SMU 4 I Elec. Appa ra tus Dings 180/3/60 6
IICV-1386 315255A 4 F-C3271 B 6-118 81 NA SMU 41 Elec. Appa ra tus Din 9s 480/3/60 flCV-150 38e9804A
- B0003 B 6-113A Cont. 17058s/17034 SMB-00 Peorioss NA 880/3/60 4
IICV- l '11 319401s 4 A B0003 B 6-113A Cont. 17054/1703fs SMB-00 Poo rl es s NA 480/3/60 llCV-303 3ri6081 B0003 8 6-75 13 170$6/15160 SMB-000 Peorioss HA #80/3/60 4
(for District uso)
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TER Equipment Item Nos.: 41, 42, 43 NRC Qualification Category: II.a (qualification not established)
Subject:
GE Motors Pages: 6-83, 6-60, 6-36 Deficiencies: (1) Aging Degradation Not Evaluated Adequately (2) Qualified Life or Replacement Schedule Not Established (3) Criteria Regarding Aging Simulation Not Satisfied (4) Criteria Regarding Radiation Not Satisfied The District contracted Wyle Laboratories to perform a thermal aging analysis of GE Series 8000 electric motors.
The Wyle report #26333-27 concluded that the HPSI motor and the containment spray pump motors have useful lives in ex-cess of 120 years.
The LPSI pump motors, due to their increased operating time, were identified as being susceptible to aging degradation.
The two components in question are Permifil Rope and Vulkaflex Power Cables. All remaining materials have useful lives calculated in excess of 120 years. At the present time, the District plans to use preventive maintenance to preclude the aging of the components in question. However, the District will continue analysis of the items, since GE has indicated that the time / temperature condition specified in the aging analysis may have been excessive. GE has also stated that they know of no failures of the materials cited by Wyle.
The TER also requested additional information as to the radi-ation qualification of the motor to lead splices. The splices are the standard 4160 and 480 volt bolted splice used during the construction of Fort Calhoun. The District will inspect the splices in question and perform a radiation qualification analysis. This analysis and corrective action, if necessary, will be completed by the end of the 1984 refue'.ing outage.
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TER Equipment Item Nos.: 44, 45 NRC Qualification Category: II.a (qualification not established)
Subject:
Containment Cooling Fan Motors (VA-7C, 7D)
Containment Vent Fan Motors (VA-3A, 3B)
Pages: 6-25, 6-24 Deficiencies : (1) Adequate Similarity Between Equip-ment and Test Specimen Not Es-tablished (2) Aging Degradation Not Evaluated Adequately (3) Qualified Life or Replacement Schedule Not Established (4) Temperature and Pressure Duration Not Adequate (5) Required Profile Not Developed Adequately (6) Criteria Regarding Spray Not Satisfied (7) Criteria Regarding Radiation Not Satisfied The District contends that the fan tested in report X-377A was similar to the Port Calhoun fans. In fact, the tested fan was one of four fans purchased by Waldinger Corporation for installation at Fort Calhoun. Joy Manufacturing Com-pany's letter dated April 5, 1983 is submitted to sub-stantiate equipment similarity. As X-377A was a destructive test, only three of the four fans were initially shipped to Fort Calhoun. The fourth ( the tested fan) was refurbished and then shipped to the station where it is presently in-
- stalled as VA-3B.
Wyle Laboratories performed an aging analysis on containment ventilation fans. Their report #26333-28 concluded that there are no materials susceptible to aging degradation for a 40-year qualified life.
It is the District's engineering judgement, based on Joy test reports X-377 and X-377A and Wyle Laboratories report
- 26333-28, that containment vent fan and containment cooling fan motors are f ully qualified for their environment and further qualification will not be undertaken.
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@April 5,1983 JOY INDUsTRI AL EQUIPMENT COMPANY NEW PHiLADELPHI A OlVISION 338 SOUTH BROADWAY P. O. 80 X 431 C New PaitAOstewi A. OwiO 44ss3 Tolophone: (216) 3391111 Stone & Webster Engineering Corporation P.O. Box 5406 Denver, Colorado 80217 Attention: Mr. Mike Watson
Subject:
Omaha Public Power P.O. #7391-CN481-108, Section 21 Joy Fan Test Report Gentlemen:
Please find attached one partial copy of Report X-377 which covers the four fans that were supplied under P.O. #7391-CN481-108 to American Air Filter Company for Omaha Public Power. The partial outlines the fan serial numbers to the motor serial numbers to assure that the motors were manufactured at the same time as the test conducted on the fan indicated in Report X-377A.
I am including a copy of our Qualification Test of the fan and motor design for service in Nuclear Containment. This report covers the fan and motor as it was tested to simulate a LOCA condition.
- iery truly yours ,
J0Y MANUFACTURING. COMPANY By:
R.D. Schilz, Proposal Engineer New Philadelphia Division Joy Industrial Equipment Company RDS:pm Attachment.
l ? A J O Y M A N U F A C I U R I N C C O M P A N Y 8A - S
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TER Equipment Item No.: 48
% ,[ g , NRC Qualification Catagory: II.a (qualification not
- established)
~' ' ' ' f subject: Component Cooling Water Pump Motors (Allis-s- s' Chalmers)
\f5 Page: 6-15 0
De ficiency: Documented Evidence of Qualification Inadequate
\-(
Wyle Laboratories report #26333-23 (Rev. A) cited six com-ponents of these motors that may be of concern under the si specified environmental parameters. Three components were susceptible to radiation-induced damage at the specified 2.5 x 105 RAD. Threshold radiation levels for these three com-ponents are as listed below:
Thermister Lead Wire Insulation -
1.7 x 104 RADS (Teflon) 4 Polyester Varnish -
1 x 105 RADS Vellumoid Gasket -
1x 105 RADS
, The specified 2.5 x 105 radiation parameter was determined
' from a design review of plant shielding and environmental i
,V qualification for equipment operating spaces and systems which may be used in post accident operations. This review wasjdone by Combustion Engineering for the District. The CE l report states that Room 69 radiation levels vary widely
\i throughout the large area. The highest readings are near the containment apray pipes. As the component cooling pumps f'
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are more than 30 feet from the major radiation source, GSE Nuclear performed component-specific radiation calculations t for each pump. The calculations show that the dose levels
,/ expected are below the threshold levels of both the varnish and gasket materials. Therefore, further discussion of these items will not be necessary.
The thermister lead wire insulation is subject to embrittle-ment at the specified radiation levels. However, as the lead wire is inside the motor case and not electrically con-nected at Fort Calhoun, it is the District's engineering judgement that failure of the teflon lead wire insulation will not impair the operation of the motor.
The Wyle Laboratories report also cites 3 components of the motors that have qualified lives of less than 40 years. The
< three components are:
- 1. Vellumoid Gasket
- 2. Corprene Gasket
- 3. Phase Insulation
Items 1 and 2 are not necessary for motor operation. The District will use preventive maintenance to rewind the motor prior to the specified life. This will preclude all aging degradation.
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TER Equipment Item Nos.: 53, 63, 72, 73, 75, 76, 81, 108 NRC Qualification Category: II.a or I.b
Subject:
Solenoid Valves Deficiencies : (1) Documented Evidence of Qualification Inadequate (2) Adequate Similarity Between Equip-ment _ and Test Specimen Not Established TER. Item Nos. 53 and 108 both refer to the same solenoid pilot for valves HCV-883A and 884A. As such, Item No. 53 should be disregarded. The corrected SCEW Page 6-78A for Item No. 108 was provided in the District's November 5, 1982 submittal.
TER items listed in Table 'I concern solenoid valves in-stalled as pilots for various system valves. The solenoid valves were replaced by ASCO solenoid valves Model NP-1 except as noted in Table II.
The District's November 5, 1982' submittal utilizes the latest ASCO test repot-t (AQR67368/REV 0). The information contained in the new report was used to establish a 40-year:
qualified life for AFOO NP-1 solenoids. The 40-year established life requires that ASCO-recommended' maintenance be accomplished. This maintenance is now being accomplished on a year-to-year basis until the qualified life program is instituted at Fort Calhoun. For the above reasons, it is.
the District's engineering judgement that all NP-1 solenoids are fully qualified.
Th'e five solenoids listed as "not accomplished" will be replaced prior to the completion of the 1984 outage. The
, reason for the delay in replacing.these is a problem associated with the high dif ferential pressure required across the valve seat. The District is engineering a
- i. control system that utilizes qualified NP-1 solenoids at a pressure below' the 200 psig rating of the NP-1 solenoids.
These five solenoids are de-energized during normal and accident operation and-are not expected to fail.
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TABLE II TER TAG ORIGINAL REVISED PAGE ITEM # NUMBER EQUIPMENT EQUIPMENT NUMBER ASCO 63 HCV-349 WPHT831429 NP8320A185E 6-77A 63 HCV-350 WPHT831429 NP8320A185E 6-17A ASCO 72 HCV-2917 LB8316C44 NP8316A75E 6-73 72 HCV-2927 LB8316C44 NP8316A75E 6-73 72 HCV-2947 LB8316C44 Not accomplished 6-86 72 HCV-2948 LB8316C44 NP8316A75E , 6-88A 72 HCV-2957 LB8316C44 NP8316A75E 6-37 72 HCV-2958 LB8316C44 NP8316A75E 6-37 ASCO 75 HCV-2918 HT8321AS Not accomplished 6-68 75 HCV-2928 HT8321AS Not accomplished 6-68 . ;
ASCO
, 76 HCV-2908 HT8321A6 Not accomplished 6-71 VALCOR 81 HCV-438A V70900213 NP8320A185E 6-19 81 HCV-438C V70900213 NP8320A185E 6-19 81 HCV-864 V70900213 NP8321A185E 6-36A 81 HCV-865 V70900213 NP8321A185E 6-36A 81 HCV-881 V70900213 SP8320A185E 6-79B 81 HCV-882 V70900213 NP8320A185E 6-793 81 HCV-1107A V70900213 NP823DA175E 6-125D 81 HCV-1108B V70900213 NP823DA175E 6-125D 108 HCV-884A V70900213 NP8320A185E 6-78A 108 HCV-883A V70900213 NP8320A185E 6-78A ASCO 73 HCV-2907 LB8316B24 NP8316A75E 6-70 73 HCV-2937 LB8316B24 Noe acccmplished 6-88 73 HCV-2937 LB8316B24 NP8316A75E 6-88A 73 HCV-2967 LB8316C44 NP8316A75E 6-39 73 HCV-2968 LB8316C44 NP8316A75E 6-39 73 HCV-2977 LB8316C44 NP8316A75E 6-39 73 JCV-2978 LB8316C44 NP8316A75E 6-39
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TER Equipment Item No.: 82 NRC Qualification Category: II.a (qualification not established)
Subject:
Target Rock Solenoids Model 80B0017 Page: 6-113B Deficiency: Adequate Similarity Between Equipment and Test Specimen Not Established The District has enclosed a letter (Reference 1) from Target Rock Corporation to Combustion Engineering dated May 6, 1981 for use in establishing similarity. It is the District's engineering judgement that the 808-001 valves are indeed similar to the test specimens of the 2375 and 2804 Target Rock reports and are therefore fully qualified.
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e REFERENCE 1 I. g G-GO741- /O 7/
saa m na m Target Rock Corporation,1966E Broadhollow Road. East Farmingdale. New York 11735/ Phone: (516) 293-3800 May 6, 1981 SUBSID*av CURTISS.WAGet CCMORATC',
/
f Please refer to:
Combustion Engineering, Inc.
1000 Prospect Hill Road Windsor, Conn. 06095 Attention Mr. W. Blowers Dept. 9487-403
Subject:
9070109-NSSS P.O. No.
PCGVS " Solenoid Valves" Contract Spec. No. NSSS-STD Target Rock Project 80B Gentlemen:
This letter does hereby certify that the Qualification to IEEE 382-1972 demonstrated by Report No. 2375, Revision C extends to all the valves and valve operators and accessories used on the ASME Section III, 80B-001 valves supplied to Combustion Engineering Inc. on the subject contract. The basis for this extention is that the supplied valves, operators and accessories use the same materials, details of construction and design features that have been qualified by the previously submitted Report No. 2375, Revision C.
One copy of Target Rock Report No. 2804, Revision B, 7
" Qualification Analysis Report, Aging Seismic & Accident Conditions, Models 79Q-001 thru 79Q-009 Solenoid Operated Globe Valves", is also
~~ce included with the letter. This report is submitted as substantiation of qualification for the models 79Q Solenoid Valves. In particular valve models 79Q-005 and 79Q-009 are both represented by the 1032110-4, Revision G, Assembly Drawing. (See Paragraphs 3.13, 3.21, and 3.22).
Only minor differences exist between the two valves, (i.e . : The size "
of the 79Q-005 valve is 3/8" while the size of the 79Q-009 valve is 3/4".
Both valves are made from the same 1" Basic Body Forging.), which are not considered signigicant for the purpose of IEEE Qualification.
The 80B-001 valves supplied to Combustion Engineering Inc., are represented by the 1032110-4, Revision F, Assembly Drawing.
There are no significant differences between the 80B-001 valves and the 79Q-005 and 79Q-009 valves discussed above. Therefore, the substantiation of IEEE Qualification Presented in Target Rock Report No. 2804, Revision B applies equally as well to the 808-001 valves.
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TAnarr Roca CamroRArsow Page 2 of 2 CN 9025 l May 6, 1981 We hope the inforration supplied above satisfies your requirements,and please do not hesitate to call this office if any questions arise.
Very truly yours, TARGET ROCK CORPORATION ec -~ /h Steven Karidas Project Engineer Thomas D. Crowley Contracts Manager Power Products sx/xx -
Enclosures (2) cc: 5. Hughes S
0
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TER Equipment Item No.: 91 NRC Qualification Category: II.a (qualification not established)
Subject:
Cable Splices at Solenoids and Transmitters Page: 6-50 De ficiency: Documented Evidence of Qualification Inadequate The District contracted Wyle Laboratories to evaluate six types of cable splices. Wyle completed its evaluation in accordance with the requirements of IE Bulletin 79-OlB (DOR Guidelines) in June, 1982. Both the transmitter and sole-noid splices lacked suf ficient accident test data to support full qualification in strict accordance with DOR Guidelines.
As a result of this Wyle report (#26333-26), the District reviewed all solenoid valve control circuits. Ten valves were identified as having to operate after a LOCA. Of the ten valves, all but two, as a result of the 79-018 solenoid upgrade, had been replaced including installation of quali-fled Raychem inline splices (WCSP-N). A modification pack-age was issued to upgrade the splices at the remaining two valves. This modification will be completed prior to or during the 1984 refueling outage.
The transmitters in question have been upgraded to 1974 standards. The new Fonboro transmitters contain f ully qualified terminal block- inat replace the splices. No further action is required.
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TER Equipment Item No.: 92 NRC Qualification Category: II.a (qualification not established)
Subject:
Splices at Containment Penetrations for Solenoid Valves and Instrumentation Page: 6-49 Deficiencies: (1) Aging Degradation Not Evaluated Adequately (2) Qualified Life or Replacement Schedulo Not Established (3) Criteria Regarding Radiation Not Satisfied (4) Criteria Regarding Test Sequence Not Satisfied With regard to deficiency (4), the District was aware of the test sequence requirement for materials with known sensiti-vity to radiation and aging. The District contracted Wyle Laboratories to complete an aging and radiation test of electrical penetration and associated splices. The testing is presently in progress at the Wyle Laboratories NORCO facility. The results and schedule of modifications, if required, will be forwarded to the Commission as soon as they become available.
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TER Equipment Item Nos.: 95, 96 NRC Qualification Category: I.b
Subject:
Containment Vent Fan Cable Splice at Motor and at Penetration Pages: 6-51, 6-52 Deficiency: Documented Evidence of Qualification Not Adequate It is the District's engineering judgement that the 480V, 3 phase, AC splices at the motor leads and both sides of con-tainment electrical penetrations are adequately qualified for the environmental conditions in which they must operate in both LOCA and post LOCA situations. A description of the motor lead splices can be found in Enclosure 9 of the Dis-trict's submittal on IE Bulletin 79-01B, dated November 5, 1982. A description of splices at the electrical pene-trations can be found on Page 22 of the Wyle Laboratories report #26333-26, " Environmental Qualification Evaluation of Cable Splices Inside Containment".
The District believes the key to qualification is viewing the splices as a system in which the inner layers of tape provide for the electrical integrity of the circuit. The outer layer of RTV (Dow Corning 3144 or 3145 clear) provides protection for the electrical insulation from those para-meters of harsh environment which could be detrimental to the splice's electrical integrity.
To substantiate the District's engineering j udgement, each of the environmental stress parameters (pressure, humidity, steam, temperature, chemical spray, radiation, and aging) were investigated to determine any adverse impact on the splice systems. The results of the investigation are as follows:
(1) Pressure: In general, the splices are viewed to be a mechanically passive system which provides electrical insulation for the connection. The insulation is placed over the connector and wire jacket in such a fashion as to minimize voids. The RTV is placed over all. In addition, the pressure would tend to compress the splice, making it watertight. It is also felt that since the splices are so "over-built", any small compression could not cause damage that would lead to splice failure. A pressure transient should not cause splice failure. This is also substantiated by the fact that the splices have remained functional through-out four (to date) containment ILRT's in which the f ans are operated at accident pressure.
(2) Humidity: This was not judged to be a factor. The RTV coating provides an adequate seal over the splice, thus causing the ef fects of humidity to be insigni-ficant.
(3) Steam: The judgement for steam is the same as for humidity. The RTV should provide an adequate barrier to steam. (Reference Dow Corning's letter dated March 24, 1980 regarding Dow Corning 3145 RTV and adhesive / sealant integrity.)
(4) Temperature: A review of the Wyle Laboratories report
- 26333-26 indicates all materials in these splices are capable of operating in an accident environment.
Temperature is not judged to be a problem.
(5) Chemical Spray: The only splices which are exposed to chemical spray are containment-side electrical pene-tration splices. (The motor leads are protected by a junction box, and the splices in Room 81 are not ex-posed to chemical spray.) RTV is not affected by mild basic solutions and should prevent any damage to the splices.
(6) Radia tion : There are three types of splice install-ations to be considered. The first type is the electrical penetration splice in Room 81. These are outside the containment and are not expected to see radiation. No further consideration is required on these splices. The second type of splice is the motor lead splice. These are protected by a junction box which eliminates the need to consider beta radiation.
A revised calculation indicates VA-3A and VA-3B could see an accident gamma dose of 8.64 x 106 RADS, while VA-7C and VA-7D would see 1.92 x 107 RADS. Both are below the 1.0 x 108 RADS value as summarized in Wyle Laboratories' report.
The third type of splice is the electrical penetration in containment. Under accident conditions, this splice could be exposed to 1.44 x 107 RADS gamma and 2.0 x 108 RADS beta (IE Bulletin 79-01B supplied number). It should be noted that the total integrated dose of 1.12 x 106 RADS used in Wyle Laboratories' report #26333-26 (as supplied by the District) was in error. The normal 40-year operating dose should be given as 3.5 x 105 RADS (which is a very conservative lR/HR for 4 0 year life) . This would bring the gamma dose to 1.475 x 107 RADS which is not considered a significant change.
The beta dose is dealt with in two ways. For the in-sulation inside the RTV, Wyle Laboratories estimates the beta dose would be reduced to 2 x 105 RADS due to the RTV shielding. This would indicate a total inte-grated dose of 1.495 x 107 RADS to the insulation.
For the RTV, the outer surface would be exposed to 2 x 108 RADS beta and the inner surface to 2 x 105 RADS beta (in both numbers, 1.475 x 107 RADS gamma must be added as the qualification radiation dose). In review-ing the radiation test information supplied by Dow Corning, both the RTV 3144 and its equivalent 3145 clear become brittle with radiation exposure. How-aver, 3144 did not fail at 1.94 x 108 RADS and failure occurred only after mechanical stress at 4.55 x 108 RADS. Based on this information and the fact that the inner approximately 85 mills should be exposed to approximately 3.475 x 107 RADS, which is well below test values, the "operabilitj" of the RTV should be ensured.
In reviewing the electrical insulation, a system ap-proach is necessary. The insulation is performed by the inner layers of Irrathene SPT tape. The Trrasil tape and Scotch 33 tape are used for protection and to hold the Irrathene SPT tape. From Table 3, Page 29, Item 3, of Wyle report #26333-26, it can be seen that all materials except the Irrasil tape show adequate radiation resistance. The Irrasil could be expected to degrade af ter approximately one hour of accident operation. However, since it performs as a holder for the qualified Irrathene tape and is itself supported by Scotch 33 tape, the District feels the Irrasil tape should not contribute to a failure of the splice system.
(7) Aging: The Wyle Laboratories report #26333-26 indi-cates that no aging related failures should be on-countered.
Conclusion It is the District's engineering judgement that adequate qualification documentation exists. It is recognized that the Wyle Laboratories report #26333-26 states that the splices do not meet the DOR Guidelines for testing. How-ever, it is the District's judgement that the key is the testing done on the RTV and the fact that it should protect the electrical insulation. Based on this analysis and the supporting documents, no further action will be taken.
TER Equipment Item No. : 97 NRC Qualification Category: II.a (qualification not established)
Subject:
States Type NT Terminal Blocks Page: 2-53 De ficiency: Lack of Adequate Similarity Qualification of these blocks was deemed lacking by the Franklin Research Center because similarity between the test specimen and the actual equipment in use was not demon-strated. This problem arose from the District's submittal of qualification documentation which identified the terminal blocks in use as States Models M25014, M25016, M25018, and M25112. Because the States terminal blocks corresponding to these model numbers are, in fact, Type NT, this concern is unwarranted. (Verification of this fact can be found by referring to the vendor's catalog.)
States Type NT terminal blocks were subjected to LOCA condi-tions - and performed satisf actorily ( Plant-Specific Reference
- 10 of the TER). A materials analysis was performed to determine radiation and chemical spray qualification. This was submitted as Enclosure 10 of the District's November 5, 1982 submittal. Results of this analysis showed spray and radiation effects would not prevent the terminal blocks from performing their design function. Aging analysis was also performed for the District by Wyle Laboratories. The Wyle Laboratories report #26333-29 demonstrates that the quali-fled life of the equipment exceeds 40 years. In addition to these qualifications, the Type NT terminal blocks installed are enclosed in junction boxes of NEMA 12 rating and are covered with Dow Corning 3144 or 3145 RTV.
It is the District's j udgement that the qualification docu-mentation plus the additional conservatism supplied by the junction box and RTV make States Type NT terminal blocks NRC Qualification Category I.a.
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TER Equipment Item No.: 99 NRC Qualification Category: II.a (qualification not established)
Subject:
Conex Electrical Penetrations Page 6-46 De ficiencies : (1) Aging Degradation Not Evaluated Adequately (2) Qualified Life Not Established (3) Criteria Regarding Radiation Not Satisfied (4) Criteria Regarding Test Sequence Not Satisfied The District is in agreement that the sequence testing of Conax penetrations was not in strict accordance with DOR Guidelines. The Conax penetrations are undergoing testing by Wyle Laboratories. The District has provided the Com-mission with status updates on this testing in our letters dated December 30, 1982 and March 8, 1983. The results and final resolution are expected prior to the 1984 refueling outage.
TER Equipnient Item No.: 100 NRC Qualification Category: II.a (qualification not established)
Subject:
Conax Electrical Conductor Seal Assembly Page: 6-59A De ficiency: Adequate Similarity Not Established Conax electrical conductor seal assemblies were reported in the District's November 5, 1982 submittal with the model number listed as N/A. The seal assemblies in use at Fort Calhoun are Model No. N-11006-64, which are indeed similar to the test specimens of IPS-409. Enclosed for your infor-mation is the Conax Certificate of Conformance to IPS-409, which establishes qualification.
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CONAX CORPORATION 2300 WALDEN AVENUE BUFFALO, NEW YORK 14225 QUALITY CONTROL DEPARTMENT (NAHA PUBLIC POWER DI5rRICT FT. CAIROUN STATION Fr. CAIROU, NE 68023 P
e' CERME OF CONFORMANCE 57526 ,
DATE October 6, 1981 Conex W.O. F 7-20150 item No.(s) 1
~
Quontity 15 Port No. (s) N-11006-64
- Serial No. (s) N/A Conox certifies thott
- 1. The materials, processes, and equipment furnished on this purchase order were produced In conformance with all contractually applicable specifications as referenced in, or fur-nished with the above purchase order,
, 2. N materials and equipment furnished under this purchase order were produced either i
from materials furnished by the customer for the production of suph parts or from materials for which Conox has ovelloble for examination, chemical or4/or physical test reports or other evidence of conformance to opplicable specifications.
N I
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i SIGN O '
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_b CO X CORPO IATIOb By: 0 -
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luniel A. Pellow ,
form 80002A
Title:
Quality Control Engineer 20A
CERTIFICATE OF CONFORMANCE CUSTOMER: OMAHA PUBLIC POWER DISTRICT CUSTOMER P.O. NO. DATE August 7, 1981 CONAX W.O. NO. 7-20100 ITEM NO (S)- 1 QUANTITY (11 15 PART NO.(S) (1) N-11006-64 CONAX CERTIFIES THAT:
- 1. ELECTRIC CONDUCTOR SEAL ASSEMBLIES
~
FURNISHED ON THIS ORDER ARE DESIGNED TO PROVIDE ELECTRIC CONDUCTOR SEALS FOR SAFETY RELATED INSTRUMENTS, SWITCHES,
- n MOTORS, VALVES, AND OTHER EQUIPMENT LOCATED INSIDE CON-
> TAINMENT OF A NUCLEAR POWER GENERATING STATION.
I
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- 2. ELECTRIC CONDUCTOR SEAL ASSEMBLIES
" FURNISHED ON THIS ORDER ARE QUALIFIED TO THE REQUIREMENTS 0F IEEE &TANDARDS 323-1974, 344-1975 AND 383-1974, SECTION 2.5, AS DOClMENTED IN CONAX IPS-409 " QUALIFICATION REPORT FOR CONDUCTOR MODULES FOR ARKANSAS NUCLEAR ONE, UNIT 2".
- 3. ELECTRIC CONDUCTOR SEAL ASSEMBLIES
, FURNISHED ON THIS ORDER WERE FABRICATED IN ACCORDANCE WITH THE REQUIREMENTS OF THE CONAX NUCLEAR QUALITY ASSURANCE PROCEDURES MANUAL FOR MANUFACTURE OF NUCLEAR POWER PLANT COMPONENTS TO ANSI-N45.2 AND 10CFR50, APPENDIX B.
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Q.A. REV1 D /OWo-V/
, SIGN b #
G. M. RHODES MANAGER OF ENGINEERING NUCLEAR PRODUCTS DIVISION CONAX CORPORATION 2300 WALDEN AVENUE hBUFFALO,NEWYORK14225
- 208 -
f TER Equipment Item No.: 103 NRC Qualification Category: II.a (qualification not established)
Subject:
Raychem Inline Splices Pages: 6-598, 6-59C, 6-59D Deficiency: Lack of Adequate Simiarity The District inadvertently reported these splices as Raychem a breakout kits, when in fact the splices are Raychem Model WCSF-N inline splices. The discrepancy was discovered in 3 August, 1982 and correctly reported in the District's November 5, 1982 submittal. The PRC's TER states that there
- is no question of the qualification of the inline splices (reference TER Item No. 103, Page 5).
9 Therefore, the
- District concludes that the correctly reported splices are NRC Qualification Category I.a (equipment qualified).
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TER Equipment Item No. : 106 NRC (,ualification Category: II.a (qualification not established)
Subject:
High Range Containment Radiation Area Monitor Pages: 6-100A, 6-100B Deficiency: Lack of Adequate Similarity The District inadvertently reported the model number of the radiation detectors as 878-1. The SCEW sheet was revised and issued as Page 6-100A of the District's November 5, 1982 submittal.
The Victoreen test report cited a degradation of exposed signal and high voltage cable during LOCA conditions. To preclude this degradation, radiation detector cables inside containment at Fort Calhoun are totally enclosed in stain-less steel tubing (see SCEW Page 6-100B) . It is the Dis-trict's engineering judgement that the reasons above correct all outstanding discrepancies and this equipment should be classified as NRC Qualification Category I.a.
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Attachment 2 10 CPR 50. 49(b)( 2) EQUIPMENT 10 CFR 50.49(b)(2) equipment was identified as that equip-ment contained in the following three categories:
Category 1: Non-safety grade electrical equipment whose failure may cause equipment actuation'which would result in a situation outside the bounds of the safety analysis.
Category 2: Non-safety grade electrical equipment whose failure may cause COE electrical bus inter-action.
Category 3: Non-safety grade or unqualified CQE equipment whose failure may provide misleading displays to the control room operator.
Category 1 items were previously addressed in the District's response to IE Bulletin 79-22. Any concerns were addressed and corrected at that time. The District is unaware of any systems installed since then (most of those being TMI-related) which could result in a Category 1 failure. The District will conduct an investigation to verify this.
Category 2 items may be divided into two classifications:
(1) Control power buses.
( 2) Power buses.
IE Bulletin 79-27 investigated the control power buses.
Based on this, the District feels adequate isolation devices (fuses or breakers) are installed to ensure proper oper-ation. As in the Category 1 items, the control power as- -
pects of the modifications installed since 1979 will be in-vestigated. No problems are anticipated.
The power buses consist of 480V and 4160V, 3 phase, distri-bution systems in which each item is equipped with a circuit breaker and/or a fuse which would act as an isolation device in the event of a fault. The District will review this system in more detail, but again anticipates no problem.
Category 3 equipment was previously addressed in response to the NRC's SER of the District's electrical equipment quali-fication. The SER required assurance that failure of non-safety related display equipment would not mislead the oper-ator. An orange dot placed on the control panel next to the l
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. . . - _ = ._
, e nameplate of all qualified electrical equipment was selected as the means to identify the qualified device. Fort Cal-houn's operating staff is trained in the use of orange dots to mitigate the consequence of misleading display inform-ation during accident conditions.
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Attachment 3 OPEN ITEMS The following items have been identified as requiring follow-up action:
(1) The Foxboro and GEMS transmitter test reports have not been recected as yet. These reporto include the test data for the Foxboro transmitter change-out and the containment sump level detection system. Appropriate SCEWS and preventive maintenance must be completed.
(2) Conax penetration testing must be completed.
(3). MR-FC-80-88 involves 79-OlB equipment. When this MR is completed, appropriate reporting must be accom-plished. Reference the AR package and memorandum TS-FC-82-325.
(4) When testing of Fisher 304 limit switches is complete, the appropriate SCEWS must be updated. In addition, maintenance to maintain qualified life must be incor-porated into the Fort Calhoun preventive maintenance system.
(5) MR-82-15 concerns modifications to mitigate the con-sequences of an FW-10 steam line break. This area should be reviewed to identify any possible harsh environment equipment.
(6) The following valves require site survey to determine nameplate data for the components indicated:
1 HCV-238 -
Solenoid HCV-240 -
Solenoid HCV-383-1 -
Solenoid HCV-383-2 -
Solenoid HCV-400A,B,C,D -
Solenoid and Fisher 546 HCV-401A,B,C,D -
Solenoid and Fisher 546
- HCV-402A,B,C,D -
Solenoid and Fisher 546 HCV-403A,B,C,D -
Solenoid and Fisher 546 HCV-4 38 B , D -
Fisher 304 HCV-467B,D -
Fisher 304 HCV-2927** -
Solenoid and NAMCO HCV-724A,B** -
Solenoid and NAMCO HCV-725A,B** -
Solenoid and NAMCO l 1 i
i
HCV-742A,C** -
Solenoid and NAMCO HCV-8 20 A, B -
Solenoid HCV-8209C,D,E,F,G,H -
Solenoid HCV-883B,,C,D,E,F,G,H -
Solenoid HCV-884** -
NAMCO and Solenoid HCV-1107A** -
NAMCO and Solenoid HCV-1108A** -
NAMCO and Solenoid HCV-1107B -
Fisher 304 and 546 and Solenoid HCV-1108B -
Fisher 304 and 546 and Solenoid HCV-1849 -
Fisher 304 and Solenoid HCV-2506A** -
NAMCO and Solenoid HCV-2507A** -
NAMCC and Solenoid HCV-2604A -
Solenoid HCV-2929** -
NAMCO and Solenoid HCV-2969** -
NAMCO and Solenoid
- Require additional PM item forms for NAMCO limit switch (7) High energy line break consideration should be given to the new steam generator blowdown system.
(8) Inspect SI pump motor and AC motor lead splices.
(9) Review control system failure / safety analysis inter-action for modifications completed since 1979.
(10) Review control power bus isolation for modifications completed since 1979.
(11) Review power bus isolation.
(12) The qualified life maintenance program will be imple-mented formally by December 1,198 3. This was origi-nally scheduled for full implementation by July 1, 1983; however, because of a main turbine failure, the 1983 refueling outage caused delays in program imple-mentation due to the unanticipated increased burden on the maintenance staff. All qualified life maintenance for the current year was completed during the 1983 outage.
(13) The justification for deletion of TE866 and 861 from the qualified equipment list will be completed by the end of the 1984 refueling outage.
(14) Complete qualification of Pyrotrol III cable.
(15) Complete qualification of main feedwater isolation valve operator brakes.
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