ML20004D971
| ML20004D971 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/19/1980 |
| From: | Sassani A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20004D783 | List: |
| References | |
| IEB-79-01B, IEB-79-1B, NUDOCS 8106100422 | |
| Download: ML20004D971 (32) | |
Text
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ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT
({}
IEB 79-01B TECHNICAL EVALUATION REPORT l
VERMONT YANKEE ATOMIC POWER STATION
, DOCKET NO. 50-271 DATED:
i i
Licensee: Vermont Yankee Atomic Power Company Reactor:
BWR, General Electric Rating:
1593 MW Thermal I
Prepared by:
t A. D. Sassani, Jr.
Engineering Support Section II Reactor Construction and Engineering Support Branch, RI i
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1 Rec'd 056 fr Reg I JD 11/19/80 THOMA3(F)
c Contents Page 1.
Introduction....................................................
1 1.1 General....................................................
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2 Background and Discussion.......................................
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1 2.1 General...................................................
1 2.2 On-Site Verification Inspections...........................
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2.3 Evaluation of Licensee's Report............................
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3.
General Information.............................................
2 3.1 Identification of Class 1E Electrical Equipment............
2 3.2 Service Conditions.........................................
2 3.3 Qualification Documentation................................
2 4.
Technical Evaluation............................................
2 4.1 Identification of Safety Related Equipment.................
3 4.2 Master List................................................
5 4.3 Service Conditions.........................................
5 I
i 4.3.1 Inside Containment LOCA.............................
5 4.3.1.1 Radiation....................................
6 4.3.1.2 Submergence..................................
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4.3.1.3 Chemical Spray...............................
6 4.4 High Energy Line Breaks (HELB).............................
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4.4.1 HELB Inside Containment.............................
6 4.4.2 HELB Outside Containment............................
6 4.4.3 Recirculated Fluids.................................
7 4.5 Margins....................................................
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4.6 Aging......................................................
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4.7 Documentation.............................................
9 4.8 SitC 4.9 Equipment Data Review.....................................
9 4.10 Conc 19sions..............................................
10 5.
Licensee Event Reports (LERs)..................................
10
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6.
References.....................................................
10 Appendix A, Test and Rer. orts and Analysis References...........
12 Appendix B, Equipment Status Table.............................
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INTRODUCTION 1.1 General The NRC Office of Inspection and Enforcement (I/E) issued Bulletin 79-018, " Environmental Qualification of Class IE Equipment" in January 1980.
This bulletin required the licensee to perform a detailed evaluation of the environmental qualification on Class 1E electrical equipment required to function under postulated accident conditions and to submit a report on this action.
This document is a report on the evaluation of the licensee's response to this bulletin.
2.
BACKGROUND AND DISCUSSION 2.1 General The evaluation of the licensee's response was accomplished by perform-ing an on-site inspection of selected class 1E equipment and by examin-ing the licensee's report for completeness and technical accuracy.
The licensee's report used in this evaluation is dated October, 1980, and therefore, does not include the response to the bulletin supplement which was issued on 9/30/80 in the form of Generic Questions and r
Answers.
2.2 On-Site Verification Inspections The on-site inspection, made on selected IE equipment, verified proper installation of equipment, overall interface integrity, and manufacturers nameplate data.
The manufacturer and model number from i
the nameplate data was compared to information given in the Environ-i mental Qualification Worksheets of the licensee's report.
If any discrepancies were noted between the installed equipment and I
the correspondence equipment addressed in the licensee's report, they are referenced in Section 4.8 of this report.
The site inspection is documented by Report Number IE 50-271/80-13.
- 2. 3 Evaluation of Licensee's Report i
Each component as addressed on the Environmental Qualification Work-sheets of the licensee's report was examined for completeness and accuracy to the criteria given in the bulletin.
This examination assumed qualification documents (analysis, test reports, etc.) refe-renced by the licensee in their submittal are acceptable.
The results of this examination are documented in Appendix B.
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3.
GENERAL INFORMATION 3.1 Identification of Class 1E Electrical Equipment The licensee's list of systems was compared to the :ystems list t
issued by the Equipment Qualification Branch (EQB) and discussed in Section 4.1 of this report.
i It is recognized that there are differences in nomenclature of systems because of plant vintage and engineering design, therefore, many of these systems may not exist or have dif ferent titles.
These differences will be addressed in the Safety Evaluation Report (SER)
I that will be prepared for this site.
3.2 Service Conditions The service condition accident environment, HELB/LOCA inside contain-ment and HELB outside containment are indicated or discussed in the licensee's report and are based on the FSAR accident analysis and Section 4.3 of this report.
i 3.3 Qualification Documentation i
Appendix A is a list of documents (test reports, analysis, letters, etc.) used by the licensee in dett.rmining the environmental qualifi-cation of plant equipment for Vermont Yankee Atomic Power Station.
These references have been tabulated by the licensee and are indi-cated on the applicable Environmental Qualification Worksheets of their report.
4.
TECHNICAL EVALUATION The basis for.the technical evaluation is the information provided by Yankee Atomic Electric Company (YAEC) in their submittal 3 YAEC-1228, dated October 1980, for the Vermont Yankee Atomic Powar Company and the verifi-l cation inspection of the as-installed equipment.f the Main Steam System and High Pressure Coolant Injection System.
The installation verification consisted of an inspection of components located inside primary containment that could be exposed to a harsh environment and documented by IE Inspection keport 50-271/80-13.
Utilizing the information identified above, the reviewer assessed its ade-quacy in relation to the D0R gusdelines6, NURdG 05887, and supplements 4 to IEB-01B which provides the Commission's requirements and staff positions.
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4.1 Identification of Safety-Related Equipment The licensee reviewed his documentation to establish the systems required to achieve a safe shutdown or provide isolation for the events identified in IEB 79-018.
These systems were then evaluated against the 00R Guidelines 6.
The systems identified by the 1*censee and included in his submittal aru:
1.
Automatic Depressurization System 2.
Circulating, Service and Cooling Water Systems 3.
Containment Atmosphere Dilution System 4.
Core Spray System 5.
Emergency Power System 6.
Heating and Ventilating System 7.
High Pressure Coolant Injection System 8.
Low Pressure Coolant Injection System 9.
Main Steam System 19.
Neutron Monitoring System 3
11.
Nuclear Boiler Vessel Instrumentation 12.
Post-Accident Monitoring System 13.
Post-Accident Sampling System i
14.
Primary Containment and Atmospheric Control System 15.
Process Radiation Monitoring System 16.
Reactor Core Isolation Coolant System 3-17.
Reactor Recirculation System 19.
Reactor Water Cleanup System 20.
Residual Heat Removal System 21, Standby Gas Treatment System 22.
Standby Liquid Control System The iist of systems including those that wtre excluded was provided to the Equipment Qualification Branch (EQB).
The EQB compared the list to a "Q" list developed by the staff and the lists provided by similar facilities to determice the completeness of the licensees response.
Based on the information provided by the licensee and the reviewers comparison, it was determined that the systems identified are within the guidance provided in Section 3.0 and Appendix A of the 00R Guidelines 4 with the exception of the following:
1.
CRD Hydraulic System The licensee did not address the CRD System in his submitta13.
The omission will be evaluated by the EQB and addressed, if l
applicable, in the Safety Evaluation Report (SER) to be written for this facility.
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The above examples identified by the reviewer will be evaluated by the EQB and addressed, if applicable, in the SER to be written
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for this plant.
4.2 Master List The licensee developed a master list based on their system evaluation as required by IEB 79-018. The licensees submittal 3 provided the basis for including specific component / equipment detailed data work sheets as required by IEB 79-018.
We have reviewed the master list for the inclusion of equipment and have the following comment:
1.
The licensees submi*tal3 does not identify terminal lugs, cable splices, splice insulation, instrument and terminal box sealant material, terminal boxes, penetration connection boxes, rigid conduit, flexible conduit, and MSIV position status limit switches.
The above components identified by the reviewer will be evaluated by the EQB and addressed, if applicable, in the SER to be written for this facility.
4.3 Service Conditions 4.3.1 Inside Containment - LOCA The licensee provided temperature and pressure profiles for the Vermont Yankee containment resulting from a LOCA.
The reactor containment temperature and pressure profiles are shown on Figures III.1-1 and III.1-2 of the submittal.
These curves were obtained from Vermont Yankee FSAR Figures 14.6-10 and 14.6-11.
The maximum environments identified are:
0 Temperature:
325#F Pressure:
44 psig Relative Humidity:
100%
Chemical Spray NA Radiation Maximum not stated The analysis of the design basis accident (LOCA) is indicated in FSAR Section 14.6.3.
Figures 14.6-10 and 14.6-11 indicates that the servise conditions in the containment will return to the levels that existed prior to the event in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
The profiles, Figures III.1-1 and III.1-2 of the submittal 3 do not contain units for time along the x-axis.
In discussions with the licensee it was determined that the time for these curves along the x-axis was in hours.
The licensee has indicated that the profiles will be revised and re-submitted.
This item is considered l
Category IV, Unresolved.
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The licensee has provided a Master List of all safety-related elec-trical equipment located within the primary containment, steam tun-nel, and reactor building, and normally available for accident miti-gation and bringing the plant to cold shutdown status.
The Master List is based on the systems included in Table I.1 of his submittal 3 shown above, and identifies the electrical equipment which is requir..
to function under postulated accident conditions.
Equipment is iden '
tified as such by a reference to Appendix II of his submittal 3, envi-ronmental qualification worksheet.
The licensee has identified elec-trical equipment not being required to function under postulated acc dent conditions and has referenced a note instead of an environmenta qualification worksheet. The notes are defined as:
(1) Required to function under non-harsh environmental conditions.
J (2) Not required to function for any accident.
(3) Not required to function for the accident producing the harsh environmental conditions.
The reviewer has identified safety-related electrical equipment that the licensee has indicated are not required to function under postu-lated accident conditions and has referenced note 1 above, " Require to function under non-harsh environmental conditions."
Examples of safety-related electrical equipment that the licensee has identified as not requiring environmental qualification are the following:
1.
Reactor Protection System Appendix I of the submittal 3 identifies 62 devices.
2.
Nuclear Boiler Vessel Instrumentation System Appendix I of the submittal 3 identifies 16 differential pressure switches. Appendix IV.2 describes these devices as steam flow differential pressure switches for sensing high steamline flow from a HELB and automatically shuting MSIV's.
3.
Reactor Core Isolation Cooling System Appendix I of the submittal 3 identifies 13 motor operated valves as not requiring environmental qualification. While Appendix IV.2 indicates the system must be available for up to two hours after post-accident, except for RCIC line breaks.
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4.3.1.1 Radi' tion The licensee does not state a maximum radiation dose in the submittal 3.
The radiation dose values for the equipment to be qualified have been identified in the environmental qualification worksheets, Appendix II of the submittal 3.
In Appendix III of the submittal 3 the licensee indicates "the radiation doses to equipment required to function during and after a LOCA have been calculated in accordance with Supplement 1
- 2 of the IEB 79-01B and the D0R Guidelines 6."
The reviewer will identify equipment, in Section 4.9 of this report, that does not meet the radiation doses required by the guidelines.
(Category IV, Qualification of Equipment Unresolved) 4.3.1.2 Submergence The licensee identified the flood level as 239 feet.
The submittal 3 indicated that all safety-related electrical equipment was located above the flood level.
- 4. 3.1. 3 Chemical Spray The licensee's submittal 3 indicated that chemical spray was not applicable to the facility.
4.4 High Energy Line Breaks (HELB) 4.4.1 HELB Inside Containment The Vermont Yankee facility has a manually initiated containment spray system. The licensee has indicated in his submittal 10 that no credit is taken for operation of containment spray.
The system is manually initiated approximately 1/2 hour after a LOCA/HELB.
The effect of containment spray is to immediately reduce post-accident temperature and pressure on the drywell, as indicated in Figures III.1-1 and III.1-2 of the submittal 3.
A further discussion of environmental conditions in the primary containment is contained in Reference #001 of Appendix A.
The HELB profiles are within the LOCA profile envelopes, therefore, electrical equipment qualified for a LOCA is acceptable for a HELB t
inside containment.
4.4.2 HELB Outside Containment The licensees May 1980, submittal 10 indicated that the results of their review of HELB is contained in " Supplemental Report on Effects of Postulated Break in a High snergy Piping System Outside the Con-i tainment", September 1973.
The submittal 10 indicated that the envi-ronmental conditions had no adverse effect on safe shutdown of the plant in all cases analyzed.
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The licensees October 1980, submittal 3 Appendix III notes:
"Because of prelimin;ry results frcm recent HELB and Heat-UP studies many areas previously believed to be non-harsh have now been deter-mined to be harsh.
The impact on equipment qualification has yet to be fully analyzed and resolved."
Appendix III of the submittal contains the environmental service conditions under which certain safety-related electrical equipment is required to function.
The following Appendix III sections address the various HELB environmental parameters for the:
(1) Steam Tunnel (2) Torus Area (3) Reactor Building (4) HPCI Pump Room (5) RCIC Pump Room Section 4.9 of this report identifies the specific equipment that is Category IV, Qualification of Equipment Unresolved.
4.4.3 Recirculated Fluids Appendix III of the submittal 3 contains the environmental service conditions under which certain safety-related electrical equipment i
is required to function.
The following Appendix III sections address the various recirculated fluids environmental parameters for the:
(1) RHR Corner Rooms (2) Torus Area (3) Reactor Building (4) HPCI Pump Room (5) RCIC Pump Room t
Section 4.9 of this report identifies the specific equipment that is Category IV, Qualification of Equipment Unresolved.
4.5 Margins P
The D0R Guidelines indicate that special consideration was given to the time required to remain functional when establishing the criteria in Section 5.2 of the guidelines.
[
NUREG-0588, Section 3(4), requires that a type test be for a minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in duration when the functional requirement is within the l
first seconds or minutes of an event ard the 00R guidelines, Section 5.2, requires that the test duration be at least as long as the period from initiation until the service conditions return to the level that existed prior to the event.
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Therefore, any type test that exceeds the functional operability
-time by 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longer meets the requirements defined in NUREG-0588 and the DOR guidelines for margin in relation to test duration for this facility.
The other consideration identified in the D0R guidelines in relation to the methods of qualification, other than identified specifically in this report will be addressed in the Safety Evaluation Report (SER) which will incorporate an audit of selected analysis and test reports identified in Appendix A.
L The considerations of margins is included in Section 4.9 of this report and identified as Category IV, Qualification of Equipment Unresolved.
4.6 Aging The licensee indicated that a study of the components subjected to harsh environments is still an outstanding item.
Details of the licensee's effort is included in their final submittal 3.
The licensee has identified the components which are still listed as requiring data.
The 00R guidelines, Section 7, does not require a qualified life to be established for all safety-related electrical eauipment, however, the following actions are required:
1.
Detailed comparison of existing equipment to the materials iden-tified in Appendix C of the 00R guidelines 6.
The first supple-ment 4 ta IEB 79-01B requires the licensees to utilize the table and identify any additional materials as the result of their effort.
2.
Establish an ongoing program to review surveillance and mainte-nance records to identify potential age related degradations.
3.
Establish component maintenaacs and replacement schedules which include considerations of aging characteristics of the installed components.
We, therefcce, require that the licensee provide the details of a program which will include a continuing effort to obtain data on existing materials and address the actions identified above.
In addition, we require the licensee provide a schedule for implementa-tion of the program that identifies problem componer.ts.
The considerations of aging is included in 3ection 4.9 of this report.
Equipment with questions is classified Category IV, Qualification of Equipment Unresolved.
4.7 Documentation The second supplement 4 to IEB 79-01B and the order 5, No. CLI-80-21, requires the licensee have the documentation and data identified in the detailed worksheets which supp'rts the qualification of the safety related electrical equipment available for NRC audit.
The second supplement 4 identifies the type of information required and the loca-tions where the records are to be maintained.
The licensees response in the area of documentation appears to be acceptable. A central file containing all the available documenta-tion for environmental qualification is located at the engineering offices of the Yankee Atomic Electric Company.
4.8 Site Verification Inspection An inspection of the installed components associated with the Main Steam Syr, tem and High Pressure Coolant Injection System was conducted on Sept < mber 29-October 2,1980, at the Vermont Yankee facility.
The details of this inspection are included in IE Inspection Report 50-271/
80-13.
The detailed identification of the components and the observations recorded will be addressed in the SER which will incorporate an audit of selected analysis and tc:t reports identified in Appendix A.
4.9 Equipment Data Review The equipment listed in Appendix B was submitted by the licensee in their response to IEB 79-01B.
This list contains equipmant with unresolved items. Appendix B identifies the licensees data in a for-mat that allows the reviewer to quickly scan the unresolved items.
The component column describes the component and references the system from the equipment qualification worksheet from Appendix II of the submitta13. The next three cclumns are self explanatory, the follow-ing three columns are defined i :
Environment - This column identifies the environmental parameter that appears to be unresolved.
Category - This column addresses the equipment status as follows:
I Qualified for Plant Life II Qualified with Restriction
III Exempted from Qualification IV Qualification of Equipment Unresolved V Equipment not Qualified Remarks - This column describes the environmental parameter or other miscellaneous comments.
4.10 Conclusion This evaluation is based on the ons-site inspection, the information supplied by the licensee in their submitta13, their FSAR, and the assumption that the Qualification Documentation (Test Reports, Analy-sis, Letters, etc.) are acceptable.
The Region I reviewer using the guidance 6,7 and instructions for the evaluation of licensee's data submittals and the site verification inspections that were performed to verify the IE Bulletin 79-01B, January 1980 data, submittal information, finds the licansee to be in accordance with the NRC direction 4,5 except as listed in Appendix B and the body of this report.
The results of this evaluation does not necessarily imply that the equipment is unreliable, unsafe or represents a significant safety issue; it does imply that additional information is required and that the unresolved items will be evaluated by the Equipment Qualification Branch (EQB) and addressed in the Safety Evaluation Report (SER) to be written for this facility.
5.
Licensee Event Reports (LERs)
No licensee event reports were submitted by the licensee, associated with their evaluation of IEB 79-018, as of November 10, 1980.
6.
References 1.
IEB 79-018, Memo to V. Thomas (NRC) from A. Finkel (NRC) dated August 18, 1980.
2.
EQ Branch Comparison of syst. ems and parameters.
(Systems List GE-BWR) 3.
Yankee Atomic Electric Company, Revised and Updated Response to IF'79-01B, dated October 31, 1980, YAEC-1228.
F 4.
Supplement Information to IEB 79-01B, dated February 29, 1980, and September 30, 1980 and October 24, 1980.
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-Order requiring licensees implement requirements of Commission Memorandum and Order of May 23, 1980 (CLI-80-21).
6.
Ofvision of Operating Reactors (DOR), " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors", Enclosure 4 to IEB 79-018.
4 i
7.
NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment", dated December 1979.
8.
Inspection Requirements for Verifying Reactor Licensee Responses to IE Bulletin No.79-01B, dated April 25, 1980.
9.
IE Support and Review of Environmental Qualification of Electrical Equipment at Operating Reactors, dated October 10, 1980.
10.
Yankee Atomic Electric Company, Responses to IEB 79-01B dated Llay i
1980, March 1980 and August 1980.
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APPENDIX A Test Reports and Analysis Lists F
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APPENDIX A Test Reports and Analysis Lists 001 Memo, 8. C. Slifer to S. F. Urbanowski, " Post-Accident Containment Pressure /
Temperatures, Vermont Yankee", NED 80-341, dated April 18, 1980.
002 Not assigned.
003 Not assigned.
004 Vermont Yankee FSAR Section 10.12.
005 Not assigned.
006 Engineering Analysis for Equipment Qualification VY.
007 Encineering Analysis #VY-#, " Radiological Dose Calculation".
008 Action Report No.15566-#, Report of " Thermal Aging Analysis of (Specific Equipment Title) for Class 1E Service at Vermont Yankee Nuclear Power Gene-rating Station".
009 EDS Report 02-0570-1057, " Environmental Qualification of Class 1E Electri-cal Equipment", July 1980, Revision 0.
010 EDS Report 02-0570-1068, " Environmental Analysis of HELB Outside Contain-ment at Vermont Yankee Atomic Power Station", November 1980.
011 Nuclear Power Station Qualification Type Test Report, Limitorque Valve Actuators for BWR Service, Limitorque Project No. 600376A, Reissued May 13, 1976.
012 Correspondence between Rome Cable and Vermont Yankee Nuclear Power Station documenting environmental testing.
013 Engineering Analysis #VY-302, " Qualification for Radiation Environment".
0.14 Letter dated August 17, 1978, Cyprus Wire and Cable Company to Yankee Atomic Electric Company transmitting Environmental Certification.
015 Ebasco Specification, VYNP-IV-C-28, "600V Auxiliary Power Cable" -
Revision 5.
016 Ebasco Specification, VYNP-IV-C2C, "600 Control Cable" - Revision 5.
017 Ebasco Specification, VhsP-IV-C
_F, " Thermocouple Cable" - Revision 4.
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018 Letter dated October 12, 1978, from Galite, Inc., to Franklin Institute Research Laboratories requesting FIRL Test Report #F-C3781.
019 Acme-Cleveland Development Company Test Plan dated 8/31/77, " Qualification of Series EA-180 and EA-740 Switches for Class 1E Use in Nuclear Power Plants in Compliance with IEEE Standard 382-1972".
020 Test Report No. QAS21678/TR, " Qualification Tests of Solenoid Valves by Environmental Exposure to Elevated Temperature, Radiation, Wear Aging, Seismic Simulation, Vibration Endurance, Accident Radiation and Loss-of-i Coolant Accident (LOCA) Simulation", dated March 1978.
021 Test Report Document #770831, " Qualification Tests of Thermocouples and RTD Assemblies by Environmental Exposures to Elevated Temperature, Radia-tion, Seismic Simulatian, Vibration Endurance, and Loss-of-Coolant (LOCA)
Simulatiso", dated Aur,ust, 1977.
022 General Electric Quilification Test Report for Electrical Penetration Assemblies.
023 Letter dated September 6, 1978, General Electric to Vermont Yankee Nuclear Power Station discussing electrical penetration assembly test report.
024 Extract from the Vermont Yankee FSAR, providing a further clarification of penetration qualification with respect to radiation.
025 Letter dated February 2,1978, General Electric to Vermont Yankee Nuclear Power Station discussing electrical terminal block testing and terminal block materials.
026 Letter dated January 27, 1978, the States Company to Yankee Atomic Electric Company with enclosures describing terminal block materials.
027 Letter dated October 16, 1978, Limitorque Corporation to Yankee Atomic i
Electric Company discussing environmental testing of Buchanan Terminal Block
- 524 (see FIRL Test Report #F-C3441).
028 Letter dated October 11, 1978, Yankee Atomic Electric Company to Amerace Corporation requesting identification and comparison of Buchanan Terminal Block used at Vermont Yankee to that tested by Limotorque Corporation.
029 Letter dated October 25, 1978, Amerace Corporation to Yankee Atomic Elec-i tric Company identifying and comparing terminal blocks.
030 Rockbestos Company Report - Qualification of Firewall SR Class 1E Electric Cables.
031_ Rockbestos Company SR Power and Control Cable Specification.
032 Kerite letter dated September 16, 1970 from F. N. Rowland to E. A. Cederborg describing testing of 2/0 AWG stranded, High Temperature Kerite insulated and FR jacketed cable.
033 Kerite Test Report NPC-4-701, " Report on the Effects of Gamma Radiation and Autoclaving on Kerite Power and Control Cables", April 30, 1970.
034 Boston Edison Memo dated February 23, 1973, D. L. Pepi to G. Hierzer releasing surplus Kerite cable from Pilgrim #1 to Vermont Yankee.
035 FIRL Test Report #F-C2781, " Test of Electrical Cables Under Simulated Post-Accident Reactor Containment Service", prepared for Lewis Engineering Company, April 1970.
036 Engineering Analysis #511, " Evaluation of Environmental Qualifications".
037 Limitorque Test Report 600198.
038 Limotorque Test Report B0003.
039 Limitorque Letter to Acton, dated September 29, 1980.
040 Limitorque Test Report F-C3271.
041 Not assigned.
042 Limitorque Letter to Acton, dated July 24, 1980.
043 September 18, 1980, Southern Comp. Letter to EDS Nuclear on States Terminal Block.
044 May 22, 1980, Westinghouse Corporation Letter to EDS Nuclear with WCAP 8754, Revision 1 Test Report.
045 May 19, 1980, Letter Siemens-Allis to YAEC.
046 Letter Siemens-Allis to YAEC, Subject - Radiation Tolerance.
047 Target Rock Corp. Report No. 23758, dated September 26, 1979.
048 Target Rock Corp. Report No. 2005C, dated December 13, 1977.
049 Qualification Tests for Rosemount Pressure Transmitter Model 1152, RMT Report No. 117415, Rev. B.
i 050 Radiation Qualification Test Report for Pressure Transmitters Model 1152DP4E22T0280, RMT Report No. 10763.
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l 051 BWR Equipment Qualification Summary.
Report No. QSR 018-A-02.
052 AETL Report #596-0398.
053 Environmental Qualification Test Summary GE-NSE80036.
054 Suntac Nuclear Corporation, Spec. #34980-1500-201.
055 ITT Barton Qualification Test Procedure for Barton Models 288A and 289A, Number 9999.1217.2.
056 IEEE 344-1975 Seismic and Radiation Qualification Tests for ITT Barton Differential Pressure Indicating Switches, Models 288A and 289A, Report No. R3-288A-1.
i 057 Viking Laboratories, Test Report 30203-2.
J58 Fenwal Inc., Engineering Laboratory, Data Report No. 6350, Qualification of Fenwal 17023-6 Unit to GE Orawing No. 145C3004.
059 Viking Laboratories, Report No. 30203-1.
060 Wyle Laboratories, Report No. 43854-1.
061 Not assigned.
062 Rosemount Product, Bulletin 1011.
063 Shaffer to Moody, SEG 321/79, December 11, 1979.
064 Product Specification Trip / Calibration System, Rosemount Model 5100U.
065 Control Products Division, Document No. EGP, Revision E.
066 Control Products Division, Document ETR, Revist i
067 Lockheed Electronics Co., Test Report No. 3232-3155.
068 Not assigned.
069 Not assigned.
070 Not assigned.
I 071 YAEC Calculation VT-ADH-80-4.
4 072 BWR Equipment Summary Report 015-A-01.
i 073 Radiation Effects Information Center, " Report on Viton", Battelle Columbus Laboratories, April 9,1975.
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074 General Electric Purchase Specifications 21A1079AC, Revision 3, Auxiliary Steam Turbine Drives (HPCI).
075~ General Electric Purchase Specifications 21A5840AJ, Revision 1, Auxiliary Steam Turbine Drives (RCIC).
076 UE&C Memorandum H-338-11, October 23, 1980, W. Majkowski to S. Ruben/
R. N. Brey.
077 VY Calculation #VY-ADH-80-5, Reactor Building LOCA Doses.
078 Letter, Target Rock Corp. to Yankee Atomic Electric Company, September 26, 1980, Target Rock Test Report Number 2302C and 23758.
079 " Radiation Effects on Electrical Insulation by P. H. Ware.
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080 Letter with Enclosure, Collyer Engineering Corporation to Connecticut Yankee Atomic Power Company dated January 13, 1967.
081 Cerro Wire Co. Certificate of Conformance for Suntac P.O. 34980-1601-11.252, May 22, 1975.
082 Boston Insulated Wire & Cable Co. Certificate of Compliance for Suntac, P.O. 34980-1601-11.1, April 28, 1975.
083 Letter, Gerald Tucker, Collyer Insulated Wire Co. to Robert McCoy, Yankee Atomic Electric Company, dated May 23, 1979.
084 Letter, Clyde Hatch, Collyer Insulated Wire Co. to George Tsouderos, Yankee Atomic Electric Company, October 4, 1968.
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h APPENDIX B EQUIPMENT STATUS TABLES 4
f APPENDIX B NOTES 1.
The licensee is presently working with Limitorque to obtain the qualification data.
2.
An aging analysis is being performed by Acton Environmental Testing Laboru-tory. This information, which is incomplete at the present time, will be available for review when completed.
3.
A radiological evaluation is being performed by EDS Nuclear.
This evalua-tion, which is incomplete at the present time, will be available for review when completed.
4.
Because of preliminary results from recent HELB and Heat-Up studies, many areas previously believed to be non-harsh have now been determined to be harsh.
The impact on equipment qualification has yet to be fully analyzed and resolved; and therefore, the conclusions reached herein are to be con-sidered preliminary.
t 5.
Evaluation of the component for long-term operability cannot be completed until the component's qualifications have been determined for all environ-mental parameters.
6.
The qualified life is less than the specified life, a maintenance program has been established to rework or replace these valves every 4.4 years.
7.
The solenoid operated valves currently installed are ASCO 8320 and 8311A31F.
1 Although they have been assured from the vendor that these solenoid operated valves will perform adequately, it is planned to replace them as a precau-tionary measure, with ASCO NP-1, Series 500's.
The ASCO NP-1 Series sole-t ncid operated valves will be installed when available.
8.
General Electric is in the process of qualifying this equipment.
The qualification will be completed in January 1981.
9.
Tbes equipment is included because of NUREG-0578.
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Serlai Contain-Component Manuf.
No.
eent Environment Category Remarks IN 0U7 Salenoid Operated ASCO 206-X Aging II Note No. 6.
Requires a schedule from the licensee.
Valve 381-6F ADS-1 Motor Operator Limitorvjue SMB-1 X
Time IV Note No. 1.
Temperature CSCW-1 Pressure HPSI-2 RH Radiation Aging H Analyzer Delphi KI X
Aging lV hote No. 2.
2 CAD-5 Solenold Operated Athomatic 15840-X RH IV Worksheet shows cosponent qualiffed for 95% RH while Appen-Valve UPI-dix 111.4-1 Indicates 1007 RH for a HELO. Requires licensee MOD to submit a schedule to provide shielding.
CAD-10 Radiation Motor GE FEJ604010.
I Time IV Note No. 2.
i FCJ604011 Radiation Note No. 3.
CS-1 Aging Note No. 5.
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2
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Serial Contala-Component Manuf.
flo.
ment Environment Category Remarts Hotor Operator Limitorque $le-1, X
Time IV isote leo. 3, 5MI-2 Radiation 05-2 Isote flo. 5.
4 Solenoid operated larget 75E002 X
RH IV Worksheet shows component quellfled for 905 RH while Valve Rock Appendix 111.4-1 Indicates 2001 RH for a IKLB.
CAD-1 Solenold Operated Attomatic 15840-I RH IV Worksheet shows component quellfled for 901 151 while Valve UPI-Appendia Ill.5-1 Indicates 1001 RH for a HELB.
CAO-2 Ottor Control Center Westing-Type X
line IV llote 163. 2.
house W
Temperature
[PS-1 EPS-8 Pressure IInte lio. 3.
RH (PS-6 Radiation flote leo. 4.
Aging (PS-7 1
Motor Control Center ITF 5600 X
line IV llote No. 2.
Temperature
[PS-2 Pressure foote lio. 3.
All Radiation flote lio. 4.
Aging I
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Part/
Serial Contain-r y wnt m auf.
W.
asnt Environment Category R m rks W uus UPS Emide 250 I
ilme IV lbte 110. 7.
KVA leaperature E PS-3 Pressure h te lie. 3.
NI Radiation flote Ih. 4.
A 3n9 9
Power Panel GE Ilot I
ilme IV llote 110. 2.
Shown Temperature' EPS-4 Pressure isote Ib. 3.
Al Radiation flote Blo. 4.
Aging Ntor Generator Set GE Not I
ilme IV Isote No. 2.
Shown Temperature EPS-5 Pressure loote Iso. 3.
NI EPS-9 Iladfation leote No. 4.
A IA9 9
Ntor Allis-I-5101 I
Aging IV Note No. 2.
Chalmers 39529-liVAC-1 2-1, 2-2.
2-3 Ntor Allis-51-308-I Aging IV leote No. 2.
Chalers $78 HVAC-2
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Part/
5erial Centain-Component Manuf.
11 9 sent Environment Category Remarks IFoul Motor Operator Limitorque 585-1 I
I!ue IV Note 11o. 2.
Radiation HPCI-I Aging Ilote llo. 3.
Note blo. 5.
leaperature Switch Fenwall 17023-6 I
Aging IV lete llo. 2.
HPCI-3 RCIC-4 Pressure Switch Barksdale B2T-X Time IV Note No. 2.
M12SS Radiation RCIC-6 Aging Note No. 3.
IIPCI-10 Note Ib. 5.
Ilow Transaltter GE/MAC 555 X
Time IV Note No. 4.
RH llPCI-8 Radiation Aging RCIC-7 Pressure Switch Barksdale D2H-X Time IV Note No. 2.
M1255 Radiation llPCI-9 Aging tbte No. 3.
Note No. 5.
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Serial Contain-l Component knof.
No.
aent Enylronment Category Remarks j
i IN OUT Local Controls Terry CCS X
Time IV Note No. 4.
Turbine Temperature itPCI-Il Pressure RH RCIC-8 Radiation A fa9 9
Motor Operator Limitorque SMB-0 X
Ilme
,lV Note No. 1.
596-l.
Tescerature HPCI-12 SM8-3.
?ressure SMB-4 RH Radiation Aging Solenoid Valve ASCO 8311A31F X
Time IV Note No. 7.
Requires the licensee to provide a schedule.
and Temperature
$8GT-4 8320 Pressure l
RH l#PCI-13 Radiation A 1a9 9
e PAS-2 PCA-5 Solenold Valve ASCO NP832-X X Aging IV Note No. 6.
Requires the licensee to provide a schedule.
3A36V M5-1 MS-2 Ntor Operator Limitorque SMB-00G I
Time IV Note No. 1.
Radiation MS-3 Note No. 5.
RCIC-1 RWCU-l
- Y O
- NMO
- 69g-,
M64 WN@
p 9
g, 9
l
e Pert /
Serial Contaln-Component knuf.
16 0 sent Enyjrosusent Categor'y Remark 5 I
F%T Pressure Switch Barksdale B2T-X Radletion IV isote 110. 2.
A1255 Aging felVI-2 Ilote 100. 3.
Pressure Switch Barton 288 X
Aging IV hote 100. 2.
- 8VI-3 PCAC-7 RCIC-5 Safety Relief Valve GE NA X
Ilme IV loote 100. 8.
Position Monitoring Temperature Pressure leote No. 9.
NBVI-5 RH Radiation Aging i
Fressure Switch 5tatic-6N-E 3-I Time IV llote Ilo. 8.
0-CIA-leaperature h8VI-6 Ring GG94R.
Pressure liste 110. 9.
HX10 lut Radiation Aging 1hersoccuple lhervo-Type I
Time IV hote No. 2.
electric i Temperature PAM-1 Pressure flote Ilo. 3.
154 Radiation isole No. 5.
Aging
- Q W
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- M N
MM _' _ _
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i.
Part/
Serial Contain-I Component Manuf.
100.
sent Enytronment Category Remarks IN OUT Acoustic Transmitter BW NA X
Time IV Quellfication tests are in progress and are scheduled Tesperature for completion during the summer of 1981.
PAM-2 Pressure itH lente 100, 9.
Radiation Aging Acoustic Accelerometer BW 10A I
Time IV Qualification tests are la progress and are scheduled leaperature for completion during the summer of 1981.
l PAM-3 Pressure RH Note 100. 9.
Radiation Aging Radiation Detector Vic toreen 877 X
Time IV Qualification tests are in progress and are scheduled Temperature for completion before January 1,1981.
PAM-4 Pressure itH lhte 100. 9.
Radiation A 1ng 9
Level Transmitter GE/MAC 555 I Time IV lete le. 4.
Tenperature FAM-5 Pressure RH Radiation Aging Temperature Element Thenso-Type X Time IV llote 100. 4.
Electric T 1esperature PAM-7 Pressure RH PAM-12 Radiation A 1ng 9
PAM-14 s.,e==
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==
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= * * = *
- 8
Part/
Serial Contain-Cosqsonent, hrsf.
Ilo.
ment Environment Category Remarks Ill WT Mutor (lperator Limitorslue SMB-0, I X ilme IV h te leo. 1.
SMB-00, Radiation RHR-1 SMB-2,
- 10te leo. 5.
58 8 - 3, RHR-2 99-4, 588-4T RHR-7 Ntor GE BEJ.
I ilme IV llote Ilo. 2.
205002 Radtatton RHR-3
- 212003, Aging Note leo. 3.
- 219001, 219003 Biote 110. 5.
N tor Operator titaltorque 585-2 I
Time IV loote leo. 1.
Temperatunt StR-6 Pressure Rif Radiation Aging t
Fan Motor Allis-RG, X
Time IV leote flo. 4.
Chalmers Frame legerature SEGT 1 215T Pressure RH Radiation Aging l
Limit Switch NAHCo EA-740 X X Aging IV Ilote feo. 2.
86700 ELEC-1 Rev. D ELEC-2(Inside) 4 a
s r.
2 Port /
l Serlai Contain-ltemrks l
Cog onert
& nuf.
No.
arnt Cnylroement - Category IR OUT Sensor E
194X9-X Time IV lbte No. 2.
L, 2M12 Iceperature PHM-1 Pressure flote No. 3.
RH Radiation Isole No. 4.
A 1n9 9
Motor Operator Limitorque SMB-000 X
Time IV m te Ib. 1.
Teeperature RCIC-2 Pressure Radiation Aging Pressure Switch Barksdale D2H-A15055 X
Time IV Mote No. 2.
Radiation RCIC-3 Aging Note No. 3.
Note No. 5.
Motor Operator iimitorque 5 % 00, X
Time IV Ilote 100. 5.54-000 Radiation RR$-1 SM-1.
Note No. 1.
$16-2 Motor Operator Limitoque SMB-000 X Radiation IV Note No. 1.
RWCb-2
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Serial Contain-Component k nuf.
80s.
sent Environment Category Remarks e
lN ouT Pressure Transmitter GE/MAC 551 I
Tise IV llote No. 2.
Radiation PM8 Aging hote 110. 3.
Ei Note flo. 4.
Level Transmitter stosemount 1152 X
Ilme IV flote No. 4.
Radiation PE9 Ilote No. 9.
I LITS Yarway 4418Cf X
Aging IV Note No. 2.
l PAM-10 i
Pressure Transmitter GE 552 X
Time IV Note No. 4 Teeperature PAM-Il Pressure Illi P h l3 Radiation Aging Solenoid Valve A5fD 206-X Aging IV Note No. 6.
Requires the licensee to provide a schedule.
832-3 PAS-1 l
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